Repair of Concrete Nuclear Power Plant

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Organisation de Coopération et de Développement Economique s
Organisation for Economic Co-operation and Development

NEA/CSNI/R(2002)7/VOL 1
05-Sep-2002
English - Or. Englis h

NUCLEAR ENERGY AGENCY

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

OECD-NEA WORKSHOP ON THE EVALUATION OF DEFECTS ,
REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETE
STRUCTURES ON NUCLEAR POWER PLANT S
Hosted by GRS at the DIN Institute in Berlin, German y

10th-11th April, 2002

JT00130882

Document complet disponible sur OLIS dans son format d’origine
Complete document available on OLIS in its original format

NEA/CSNI/R(2002)7/VOL 1

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMEN T
Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30th
September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed :
- to achieve the highest sustainable economic growth and employment and a rising standard of living in Membe r
countries, while maintaining financial stability, and thus to contribute to the development of the world economy ;
- to contribute to sound economic expansion in Member as well as non-member countries in the process of economi c
development ; and
- to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance wit h
international obligations .
The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece ,
Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdo m
and the United States . The following countries became Members subsequently through accession at the dates indicated hereafter :
Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18t h
May 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996), Korea (12t h
December 1996) and the Slovak Republic (14th December 2000) . The Commission of the European Communities takes part in th e
work of the OECD (Article 13 of the OECD Convention) .

NUCLEAR ENERGY AGENC Y
The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEE C
European Nuclear Energy Agency . It received its present designation on 20th April 1972, when Japan became its firs t
non-European full Member. NEA membership today consists of 27 OECD Member countries : Australia, Austria, Belgium ,
Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg ,
Mexico, the Netherlands, Norway, Portugal, Republic of Korea, Spain, Sweden, Switzerland, Turkey, the United Kingdom and th e
United States . The Commission of the European Communities also takes part in the work of the Agency .
The mission of the NEA is:
-

to assist its Member countries in maintaining and further developing, through international co-operation, th e
scientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclea r
energy for peaceful purposes, as well as
- to provide authoritative assessments and to forge common understandings on key issues, as input to governmen t
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainabl e
development .
Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive wast e
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law an d
liability, and public information . The NEA Data Bank provides nuclear data and computer program services for participating
countries.
In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency i n
Vienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field .
© OECD 200 2

Permission to reproduce a portion of this work for non-commercial purposes or classroom use should be obtained through th e
Centre français d’exploitation du droit de copie (CCF), 20, rue des Grands-Augustins, 75006 Paris, France, Tel. (33-1) 44 07 47
70, Fax (33-1) 46 34 67 19, for every country except the United States . In the United States permission should be obtained throug h
the Copyright Clearance Center, Customer Service, (508)750-8400, 222 Rosewood Drive, Danvers, MA 01923, USA, or CC C
Online : http://www.copyright.com/. All other applications for permission to reproduce or translate all or part of this book shoul d
be made to OECD Publications, 2, rue André-Pascal, 75775 Paris Cedex 16, France .

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NEA/CSNI/R(2002)7/VOL 1

COMMITTEE ON NUCLEAR REGULATORY ACTIVITIE S
The Committee on Nuclear Regulatory Activities (CNRA) of the OECD Nuclear Energy Agency (NEA) i s
an international committee made up primarily of senior nuclear regulators . It was set up in 1989 as a forum for th e
exchange of information and experience among regulatory organisations and for the review of developments whic h
could affect regulatory requirements.
The Committee is responsible for the programme of the NEA, concerning the regulation, licensing an d
inspection of nuclear installations . The Committee reviews developments which could affect regulatory requirement s
with the objective of providing members with an understanding of the motivation for new regulatory requirements
under consideration and an opportunity to offer suggestions that might improve them or avoid disparities among
Member Countries . In particular, the Committee reviews current practices and operating experience .
The Committee focuses primarily on power reactors and other nuclear installations currently being built
and operated . It also may consider the regulatory implications of new designs of power reactors and other types of
nuclear installations .
In implementing its programme, CNRA establishes co-operative mechanisms with NEA's Committee o n
the Safety of Nuclear Installations (CSNI), responsible for co-ordinating the activities of the Agency concerning th e
technical aspects of design, construction and operation of nuclear installations insofar as they affect the safety of suc h
installations . It also co-operates with NEA's Committee on Radiation Protection and Public Health (CRPPH) an d
NEA's Radioactive Waste Management Committee (RWMC) on matters of common interest.
COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S
The NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made up
of scientists and engineers . It was set up in 1973 to develop and co-ordinate the activities of the Nuclear Energ y
Agency concerning the technical aspects of the design, construction and operation of nuclear installations insofar a s
they affect the safety of such installations . The Committee's purpose is to foster international co-operation in nuclea r
safety amongst the OECD Member countries .
CSNI constitutes a forum for the exchange of technical information and for collaboration betwee n
organisations which can contribute, from their respective backgrounds in research, development, engineering o r
regulation, to these activities and to the definition of its programme of work . It also reviews the state of knowledg e
on selected topics of nuclear safety technology and safety assessment, including operating experience . It initiates an d
conducts programmes identified by these reviews and assessments in order to overcome discrepancies, develop
improvements and reach international consensus in different projects and International Standard Problems, and assist s
in the feedback of the results to participating organisations . Full use is also made of traditional methods of cooperation, such as information exchanges, establishment of working groups and organisation of conferences and
specialist meeting.
The greater part of CSNI's current programme of work is concerned with safety technology of wate r
reactors . The principal areas covered are operating experience and the human factor, reactor coolant syste m
behaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reacto r
accidents and their confinement, containment performance, risk assessment and severe accidents . The Committee
also studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operate s
an international mechanism for exchanging reports on nuclear power plant incidents .
In implementing its programme, CSNI establishes co-operative mechanisms with NEA's Committee o n
Nuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation ,
licensing and inspection of nuclear installations with regard to safety . It also co-operates with NEA's Committee o n
Radiation Protection and Public Health and NEA's Radioactive Waste Management Committee on matters o f
common interest.

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NEA/CSNI/R(2002)7/VOL1

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NEA/CSNI/R(2002)7/VOL 1

Foreword
The Committee on the Safety of Nuclear Installations (CSNI) of the OECD-NEA co-ordinates the NE A
activities concerning the technical aspects of design, construction and operation of nuclear installation s
insofar as they affect the safety of such installations . In 1994, the CSNI approved a proposal to set up a
Task Group under its Principal Working Group 3 (recently re-named as the Working Group on Integrity o f
Components and Structures (IAGE)) to study the need for a programme of international activities in th e
area of concrete structural integrity and ageing and how such a programme could be organised . The task
group reviewed national and international activities in the area of ageing of nuclear power plant concret e
structures and the relevant activities of other international agencies . A proposal for a CSNI programme of
workshops was developed to address specific technical issues which were prioritised by OECD-NEA tas k
group into three levels of priority :

First Priority
• Loss of prestressing force in tendons of post-tensioned concrete structure s
• In-service inspection techniques for reinforced concrete structures having thick sections and area s
not directly accessible for inspectio n
Second Priority
• Viability of development of a performance based databas e
• Response of degraded structures (including finite element analysis techniques )
Third Priority
• Instrumentation and monitoring
• Repair methods
• Criteria for condition assessment
The working group has progressively worked through the priority list developed during the preliminar y
study carried out by the Task Group . Currently almost all of the three levels of priority are effectively
complete, although in doing so the committee has identified other specific items worthy of consideration .
By working logically through the list of priorities the committee has maintained a clarity of purpose whic h
has been important in maintaining efficiency and achieving its objectives . The performance of the group
has been enhanced by the involvement of regulators, operators and technical specialists in both the work o f
the committee and its technical workshops and by liaison and co-operation with complementar y
committees of other international organisations . The workshop format that has been adopted (based aroun d
presentation of pre-prepared papers or reports followed by open discussion and round-table development o f
recommendations) has proved to be an efficient mechanism for the identification of best practice, potentia l
shortcomings of current methods and identification of future requirements .

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NEA/CSNI/R(2002)7/VOL1

6

NEA/CSNI/R(2002)7/VOL 1
SUMMARY

OECD-NEA workshop on the evaluation of defects, repair criteria & methods of repair for concret e
structures on nuclear power plant s
OECD-NEA IAGE held an international workshop on the evaluation of defects, repair criteria & method s
of repair for concrete structures on nuclear power plants in Berlin, Germany on April 10-11, 2002 .
Through 2 technical sessions devoted to Operational Experience and State of the Art and Futur e
Developments, a broad picture of the status was given to a large audience composed by 54 participant s
from 17 countries and International Organisations . 21 papers have been presented at the Workshop .
The objectives of the workshop were to examine the current practices and the state of the art with regard t o
the evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe r
Plants with a view to determining the best practices and identification of shortfalls in the current methods,
which are presented in the form of conclusions and recommendations in this report .
This workshop on the evaluation of defects, repair criteria and methods of repair for concrete structures o n
Nuclear Power Plants is the latest in a series of workshops .
The complete list of CSNI reports, and the text of reports from 1993 on, is available o n
http://www.nea .fr/html/nsd/docs/

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NEA/CSNI/R(2002)7/VOL1

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NEA/CSNI/R(2002)7/VOL 1

Acknowledgement
Gratitude is expressed to GRS, Germany for hosting the Workshop at the DIN Institute in Berlin. In
particular, special thanks to Mr . Helmut Schulz and Dr Jurgen Sievers, and also Mrs Brunhilde Laue an d
Mrs Schneider for their help .
Thanks are also expressed to chairmen of the sessions and to the Organizing Committee for their effort an d
co-operation.
Dr Leslie M Smith
Prof Pierre Labbé
M. Jean-Pierre Touret
Herr Rüdiger Danisch
Mr James Costello
Dr Dan Naus
M.Eric Mathet

BEG(UK) Ltd
IAEA

(UK) Chairma n
(International )

EdF

(France)

Framatome ANP GmbH
USNRC

(Germany)
(USA )
(USA )
(International)

ORNL

OECD-NEA

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NEA/CSNI/R(2002)7/VOL1

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NEA/CSNI/R(2002)7/VOL 1

OECD-NEA WORKSHOP
ON THE
EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FO R
CONCRETE STRUCTURES ON NUCLEAR POWER PLANTS
th
10th and 11 April, 2002
Berlin, Germany

A.
B.
C.
D.
E.

CONTENTS
CONCLUSIONS AND RECOMMENDATION S
PROGRAMME
PAPER S
PARTICIPANTS

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NEA/CSNI/R(2002)7/VOL1

12



NEA/CSNI/R(2002)7/VOL 1
A. TABLE OF CONTENTS
PAGE

Volume 1
B.
C.
D.

CONCLUSIONS AND RECOMMENDATION S
PROGRAMME
PAPER S

17
19
23

Introductory Paper
Inspection, Assessment and Repair of Nuclear Power Plan t
Concrete Structures
D. J. Naus, Oak Ridge National Laboratory, U.S.A.
H. L . Graves, J.F. Costello, USNRC, U.S.A .
SESSION A : OPERATIONAL EXPERIENCE
Chairman: Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany)

Repair of the Gentilly-1 Concrete Containment Structur e
A. Popovic, D . Panesar and M . Elgohary, AECL, Canada
The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures an d
Installation of an Automated Cathodic Protection Syste m
L. M . Smith, C . A. Hughes, British Energy Generation UK, Ltd .
G. Jones, Sea-Probe, Ltd.
Feasibility Study of IE-SASW Method for the Non-Destructive Evaluation of Containmen t
Building of Nuclear Power Plant
Mr. Yong-Pyo Suh, KEPRI, Korea

25

37

39

51

59

Field Studies of Effectiveness of Concrete Repair s
N.J.R. Baldwin, Mott MacDonald Ltd.,(UK)

73

Detection and Repair of Defects in the Confinement Structures at Paks NP P
Mr. Nyaradi Csaba, Paks NPP Ltd

83

SESSION A : OPERATIONAL EXPERIENCE (Continued )
Chairman : Dr James Costello, USNRC (USA)

11 5

Steam Generator Replacement at Ringhals 3 Containment, Transport Openin g
Jan Gustavsson, Ringhals Nuclear Power Plant, (Sweden )

11 7

In Service Inspection Programme and Long Time Monitoring of Temelin NP P
Containment Structure s
Jan Maly, Jan Stepan, Energoprojekt Prague, Czech Republi c
Repair Criteria and Methods of Repair for Concrete Structures on Nuclear Power Plants
R. Lasudry, Tractebel Energy Engineering, (Belgium )

13

12 5

13 5



NEA/CSNI/R(2002)7/VOL 1
Post-Fire Damage Assessment Procedures for Nuclear Power Plant Structures
L.M. Smith, British Energy Generation UK, Ltd ., (UK)

157

Volume 2
SESSION B : STATE OF THE ART & FUTURE DEVELOPMENTS
Chairman : Dr Naus, ORNL (US)

17

Various stages to Address Concrete Cracking on NPP s
C. Seni, Mattec Engineering Ltd ., (Canada)

19

Investigation of the Leakage Behaviour of Reinforced Concrete Wall s
Nico Herrmann, Christoph Niklasch, Michael Stegemann, Lothar Stempniewski ,
University of Karlsruhe, (Germany )

31

The Development of a State-of-the-Art Structural Monitoring Instrumentation Syste m
for Nuclear Power Plant Concrete Structure s
L. M . Smith, B . Stafford, M.W. Roberts, British Energy Generation UK ,Ltd .
A. McGown, University of Strathclyde (UK)
Ageing and Static Reliability of Concrete Structures under Temperatur e
and Force Loading
Petr Stepanek,, Stanislav Stastnik Vlastislav Salajka, Technical University of Brno ,
Jaroslav Skolai, Jiri Stastny, Dukovany Power Plant, (Czech Republic )
Efficient Management of Inspection and Monitoring Data for a Bette r
Maintenance of Infrastructur e
Marcel de Wit, Gilles Hovhanessian, Advitam

41

55

67

Aging Process Of A Good Concrete During Forty Year s
Dr. Peter Lenkei, University, College of Engineering (Hungary)

77

SESSION B : State of the Art & Future Developments (Continued)
Chairman : Mr. Jean-Pierre TOURET, EdF, (France)

81

Acoustic Monitorin g
Marcel de Wit, Gilles Hovhanessian, Advitam

83

Concrete Properties Influenced by Radiation Dose During Reactor Operatio n
Takaaki Konno, Secretariat of Nuclear Safety Commission, (Japan )

97

The Use of Composite Materials in the Prevention and Strengthening of Nuclea r
Concrete Structures
D. Chauvel, P .A. Naze, J-P . Touret EdF, Villeurbanne, (France)

14

10 5



NEA/CSNI/R(2002)7/VOL 1

Detection of Reinforcement Corrosion and its Use for Service Life Assessment o f
Concrete Structures
C. Andrade, I . Martinez, J. Munoz, CSIC (SP) Rodriguez, M. Ramirez, GEOCISA (Spain)
Improved Detection of Tendon Ducts and Defects in Concrete Structures Usin g
Ultrasonic Imagin g
W. Müller, V. Schmitz, FIZP, (GE) M . Krause, M . Wiggenhauser, Bundesanstalt fü r
Materialforschung und -Prüfung, (Germany )
Structural Integrity Evaluation of a Steel Containment for the Replacement o f
Steam Generato r
Mr. Yong-Pyo Suh, KEPRI (Korea)

11 5

12 5

13 3

New Methods on Reconstruction of Safety Compartments of Nuclear Power Plants
Z. Kdpper, Kdpper und Partner, Bochum, (Germany)
D. Busch, RWE Solutions AG, Essen, (Germany)

14 1

E.

151

PARTICIPANTS

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NEA/CSNI/R(2002)7/VOL1

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NEA/CSNI/R(2002)7/VOL 1
OECD-NEA Workshop on the Evaluation of Defects, Repair Criteria &
Methods of Repair for Concrete Structures on Nuclear Power Plants ,
GRS, Berlin, Germany April 10-11, 200 2
Conclusions and Recommendations

The objectives of the workshop were to examine the current practices and the state of the art with regard t o
the evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe r
Plants with a view to determining the best practices and identification of shortfalls in the current methods,
which are presented in the form of conclusions and recommendations .
CONCLUSIONS
1. Repairs to concrete NPP structures and their durability will continue to be an issue until final
decommissioning .
2. Experience gained during repair projects, and extensive field studies of repaired structures, show tha t
the effectiveness of the concrete repair is dependent on :
correct diagnosis of the cause of the damage ;
selection of a repair strategy that addresses this cause ;
choice of appropriate repair materials and methods ;
careful management of the process ;
post repair maintenance strategy supported by comprehensive records .
Computerised databases can assist with : recording the detection and diagnosis of damage ; rexcording the
location of, and specification for, repairs ; and management of the subsequent repair .
3. The combination of concrete with composite materials is useful in a repair situation . These materials
now have a track record in structural repairs to a decommissioned prestressed concrete containment
(PCC) . Extensive testing has proved their potential as an alternative to steel as a liner for PCCs . They
are currently being considered for enhancing the leak tightness of unlined containments .
4.

Surface overcoating materials can protect exposed concrete surfaces from deterioration due t o
environmental factors eg carbonation, chlorides etc . Careful design will ensure that the coating syste m
can accommodate structural movement, maximise durability and satisfy aesthetic considerations .

5. Experiences of repairs, supported by field studies of repaired structures, confirm that a principal caus e
of damage to reinforced concrete structures is corrosion of the reinforcement . Impressed current
cathodic protection (CP) has been shown to be effective in improving the durability of a repaire d
structure exposed to a very severe marine environment . Laboratory examination of samples of concret e
removed from a structure protected by CP has shown that long term application of impressed current
CP was not detrimental to the original concrete and did not affect the steel to concrete bond .
6. There was recognition that the nuclear industry might benefit from improved guidance on assessmen t
of defects and the effectiveness of subsequent repairs . However, absolute criteria are difficult to defin e
and may not be universally applicable .
7. Laboratory trials of impact echo and synthetic aperture focusing technique ultrasonic non-destructiv e
testing have shown that they have potential to detect subsurface features in concrete elements but that
significant further development is required for field implementation.
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NEA/CSNI/R(2002)7/VOL 1

8. Laboratory tests are providing important data on the leakage of air and air/steam through cracke d
concrete. These tests will help to inform the assessment of pressure retaining concrete structures o n
NPPs and provide a useful source of validation for numerical models and simulation .
9. Papers on the application of structural monitoring to NPP concrete structures confirmed that th e
practicality of installing instrumentation is of equal importance to its ability to measure the damag e
parameter under investigation . Acoustic monitoring has demonstrated the potential to detect and locate
cracking and may warrant further consideration as a tool for assisting the testing of containment
structures.
10. There is little data available on the effect of irradiation on concrete . Samples of concrete removed fro m
a biological shield structure have provided some information .
11. A pre-prepared structural condition assessment procedure listing nuclear safety related structures ma y
be useful in assessing post-fire damage on NPPs . Materials used in the repair of fire damage must b e
capable of meeting the fire performance criteria required by the original structure .

RECOMMENDATION S
The following recommendations are offered to inform national activities and research programmes for the
inspection, maintenance and repair of concrete NPP structures .
1. The execution and durability of repairs to concrete should be considered as an issue that is relevant t o
the nuclear safety of NPPs throughout the lifetime of the plant and until final decommissioning .
2. Improved guidance is required on the assessment of defects (eg . cracks) and the performance an d
effectiveness of subsequent repairs .
3. Owners/operators of NPP concrete structures should develop procedures for recording : the detectio n
and diagnosis of defects/damage ; the location of and specification for each repair ; and the management
strategy to be applied to the repair.
4. Further development of NDE techniques is required in order to support the assessment and evaluation
of defects and subsequent repairs in concrete structures . The development priorities, conclusions an d
recommendations identified at the OECD-NEA 1997 Risley NDE workshop should be applied .
5. Further work is required on the evaluation of leakage through cracks in concrete structures .
6. Further investigation of the effects of irradiation on concrete is required .
7. Consideration should be given to the development of pre-prepared structural condition assessmen t
procedures listing nuclear safety related structures for the evaluation of post-fire damage on NPPs .

18



NEA/CSNI/R(2002)7/VOL 1

OECD-NEA WORKSHOP
ON TH E
EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FOR
CONCRETE STRUCTURES ON NUCLEAR POWER PLANTS
Hosted by GRS at the DIN Institute in Berlin, German y
10 & 11 APRIL 200 2
C. PROGRAMME

Wednesday April 10 , 2002

9:30 10:00 Welcome

H. Schulz, GR S

E. Mathet - OECD
L . Smith- Chairma n

Introduction
Introductory paper

10:00 10:30 Inspection, Assessment and Repair of Nuclear Power Plant D. J. Naus, Oak Ridge National
Concrete Structures
Laboratory, U.S.A .
H. L. Graves, J . F . Costello,
USNRC, U.S.A.
10:30 10:50 Coffee brea k

SESSION A : Operational Experienc e
Chairman :

Mr. Ruediger.Danisch, FRAMATOME-ANP GmbH (GE )

10:50 11 :10 Repair of the Gentilly-1 Concrete Containment Structure
A. Popovic, D . Panesar and M .
Elgohary, AECL, (CDN )
11 :10 11 :30 The Repair of Nuclear Power Plant Reinforced Concrete
Marine Structures and Installation of an Automated Cathodi c
Protection System
L. M . Smith, C.A. Hughes, British
Energy Generation UK ,Ltd .
G. Jones, Sea-Probe, Ltd.

19



NEA/CSNI/R(2002)7/VOL 1

11 :30 11 :50 Feasibility Study of IE-SASW Method for the Non Destructive Evaluation of Containment Building of Nuclear
Power Plant
Mr. Yong-Pyo Suh, KEPRI (K)
11 :50 12 :10 Field Studies of Effectiveness of Concrete Repairs
N.J.R. Baldwin, Mott MacDonald
Ltd.,(UK)

12:10 12 :45 Detection and repair of defects in the confinement structure s
at Paks NPP
Mr. Nyaradi Csaba, Paks NPP
Ltd

12:45

14:00 Lunch

SESSION A : Operational Experience (Continued)
Chairman : Dr James Costello, USNRC (USA )
14:00

14:20 Steam Generator Replacement at Ringhals 3 Containment ,
Transport Opening
Jan Gustavsson, Ringhals Nuclear
Power Plant, (SW)

14:20

14:40 In Service Inspection Programme and Long
Monitoring of Temelin NPP Containment Structure s

Time

Jan Maly, Jan Stepan ,
Energoprojekt Prague, Czech
Republic
14:40

15:00 Repair Criteria and Methods of Repair for Concret e
Structures on Nuclear Power Plants
R. Lasudry, Tractebel Energ y
Engineering, (BE )

15:00

15:20 Post-Fire Damage Assessment Procedures for Nuclea r
Power Plant Structures
L.M. Smith, British Energy
Generation UK ,Ltd., (UK)

15:20

15:50 Coffee break

20



NEA/CSNI/R(2002)7/VOL 1

SESSION B : State of the Art & Future Developments

15:50

Chairman :
16:10 Various stages to address Concrete Cracking on NPPs

Dr Naus, ORNL (US )
C. Seni, Mattec Engineering Ltd.,
(CDN)

16:10

16:30 Investigation of the Leakage Behaviour of
Concrete Walls

Reinforced
Nico Herrmann, Christoph
Niklasch, Michael Stegemann,
Lothar Stempniewski,
University of Karlsruhe, (GE)

16:30

16:50 The Development of a State-of-the-Art Structural
Monitoring Instrumentation System for Nuclear Power Plan t
Concrete Structures
L. M. Smith, B . Stafford, M .W.
Roberts, British Energ y
Generation UK ,Ltd.
A. McGown, University of
Strathclyde (UK)

16:50

17:10

17:10 Ageing and Static Reliability of Concrete Structures under
Temperature and Force Loading
Paper not presented but in the proceedings

Petr Stepanek,, Stanislav
Stastnik Vlastislav Salajka ,
Technical University of Brno ,
Jaroslav Skolai, Jiri Stastny ,
Dukovany Power Plant, (CZ )

17:30 Efficient management of inspection and monitoring data for
a better maintenance of infrastructure
Marcel de Wit, Gilles
Hovhanessian, Advitam

17:30

17:45 Aging process of a good concrete during forty years
Dr. Peter Lenkei, University,
College of Engineering
(Hungary)
End of the first day

Thursday April 11, 200 2
SESSION B : State of the Art & Future Developments (Continued )
Chairman :
9:00 9 :30

Mr. Jean-Pierre TOURET, EdF, (F)

Acoustic monitoring
Marcel de Wit, Gilles
Hovhanessian, Advitam

21



NEA/CSNI/R(2002)7/VOL 1

9:30 9 :50

Concrete Properties Influenced by Radiation Dose Durin g
Reactor Operation
Takaaki Konno, Secretariat o f
Nuclear Safety Commission, (J)

9:50 10:10 The Use of Composite Materials in the Prevention and
Strengthening Of Nuclear Concrete Structures
D. Chauvel, P .A. Naze, J-P .
Touret EdF, Villeurbanne, (F)
10:10 10:30 Detection of reinforcement corrosion and its use for service
life assessment of concrete structures
C. Andrade, I . Martinez, J.
Munoz, CSIC (SP) Rodriguez ,
M. Ramirez, GEOCISA (SP )
10:30 11 :00 Coffee brea k
11 :00 11 :20 Improved Detection of Tendon Ducts and Defects in
Concrete Structures Using Ultrasonic Imaging
W. Müller, V. Schmitz, FIZP,
(GE) M . Krause, M .
Wiggenhauser, Bundesanstalt
für Materialforschung und Prüfung, (GE)
11 :20 11 :40 Structural Integrity Evaluation of a Steel Containment fo r
the Replacement of Steam Generator
Mr. Yong-Pyo Suh, KEPRI
(KR)
11 :40 12 :00 New Methods on Reconstruction of Safety Compartments o f
Nuclear Power Plants
Z. Kdpper, Kdpper und Partner,
Bochum, (GE)
D. Busch, RWE Solutions AG,
Essen, (GE)
12 :00

14:00 Lunch

SESSION C Chairman:
14:00 16:30 Panel discussion
16:30

Dr L .M. Smith (UK)

Closure

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NEA/CSNI/R(2002)7/VOL 1

C . PAPERS

23

NEA/CSNI/R(2002)7/VOL1

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NEA/CSNI/R(2002)7/VOL 1
INSPECTION, ASSESSMENT, AND REPAIR O F
NUCLEAR POWER PLANT CONCRETE STRUCTURE S
D.J. Naus
Oak Ridge National Laboratory (ORNL )
Oak Ridge, TN
H.L. Graves, III and J .F. Costello
U.S. Nuclear Regulatory Commission (USNRC )
Washington, D .C.

Abstract
Aging of concrete structures occurs with the passage of time and has the potential, if its effects ar e
not controlled, to increase the risk to public health and safety . Activities have been conducted to addres s
factors related to quantifying the effects of age-related degradation on nuclear power plant concret e
structures and components . Environmental effects that can lead to age-related degradation of reinforce d
concrete structures and their manifestations are described. Current regulatory in-service testing and
inspection requirements are reviewed. Techniques commonly used to inspect nuclear power plant concret e
structures to assess and quantify age-related degradation are identified. An approach for conduct o f
condition assessments is presented, as well as criteria, based on visual indications, for use in classification
and assessment of concrete degradation. Materials and techniques for repair of reinforced concret e
structures are noted and guidance provided on repair options available for various forms of concret e
degradation (e .g., cracking, spalling, and steel reinforcement corrosion) . Nuclear power plant degradation
and repair experience is summarized. In-service inspection/repair strategies to maintain the probability o f
failure of a concrete component at or below a target value are discussed .
1.

Introduction

To date, 104 nuclear power reactors are currently licensed for commercial operation in the Unite d
States (US) . Currently 103 of these reactors are in operation, producing about 20% of the nation’ s
electricity supply . The median age of these reactors is over 20 years, with 61 having been in commercia l
operation for 20 or more years. Initial operating licenses for these reactors will start expiring in 2006, wit h
approximately 10% expiring by the year 2010, and more than 40% by the year 2015 . Continuing the
service of existing nuclear power plants (NPPs) through a renewal of their initial operating license s
provides a timely and cost-effective solution to the problem of meeting future electricity demand. In fact,
48 reactor units (as of March 2002) have either completed the license renewal process, submitte d
applications to renew their operating licenses, or announced that they intend to do so . However, the
structures in these plants are susceptible to aging by various processes, depending on the operatin g
environment and service conditions, that can affect the engineering properties, structural
resistance/capacity, failure mode, and location of failure initiation . As a result, the ability of the structure s
to withstand various challenges in service from operating conditions, the natural environment, and
accidents may be impacted . Current aging-related activities in large measure are therefore focusing
technical development and support on condition assessment with the aim of demonstrating that structural
margins of existing plants have not or will not erode during the desired service life due to aging o r
environmental effects . Probabilistic methods can be used to provide the quantitative tools for the
assessment of uncertainty in condition assessment and are an essential ingredient of risk-informe d
management decisions concerning continued service of the NPP structures .

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2.

NPP Concrete Containments

All commercial NPPs contain concrete structures whose performance and function are necessary fo r
protection of the safety of plant operating personnel and the general public . Most of the concrete structures
in NPPs are similar to conventional civil engineering structures; however, certain NPP concrete structure s
can entail thicker sections, increased reinforcement, and limited accessibility and harsher exposur e
conditions at certain locations . Typical safety-related functions that the concrete structures provide include
foundation, support, shielding, and containment. Although a number of concrete structures are importan t
to the overall safety of NPPs (e .g., fuel/storage pools, cooling water intake structures, and foundations) ,
discussions will concentrate on the containment structures because of their unique requirements .
Concrete containments are metal lined, reinforced concrete pressure-retaining structures that in som e
cases may be post-tensioned . The concrete vessel includes the concrete shell and shell components, shel l
metallic liners, and penetration liners that extend the containment liner through the surrounding shell
concrete. The reinforced concrete shell, which generally consists of a cylindrical wall with a hemispherical
or ellipsoidal dome and flat base slab, provides the necessary structural support and resistance to pressureinduced forces . Leak-tightness is provided by a steel liner fabricated from relatively thin plate material
(e.g., 6-mm thick) that is anchored to the concrete shell by studs, structural steel shapes, or other stee l
products . Initially, existing building codes, such as American Concrete Institute (ACI) Standard 318 ,
Building Code Requirements for Reinforced Concrete [1], were used in the nuclear industry as the basi s
for design and construction of concrete structural members . However, because the existing building code s
did not cover the entire spectrum of design requirements and because they were not always considere d
adequate, additional criteria were developed for design of seismic Category 1 (i .e., safety related)
structures (e .g., definitions of load combinations for both operating and accident conditions) . Plants that
used early ACI codes for design were reviewed by the USNRC through the Systematic Evaluation Progra m
to determine if there were any unresolved safety concerns [2] . Current rules for construction of concret e
containments are provided in Section III, Division 2 of the ASME Code [3] . The USNRC has developed
supplemental load combination criteria and provides information related to concrete and steel interna l
structures of steel and concrete containments [4,5] . Rules for design and construction of the metal liner
that forms the pressure boundary for the reinforced concrete containments are found in Section III ,
Division 1, Subsection NE of the ASME Code . Depending on the functional design (e .g., large dry or ic e
condenser), NPP concrete containments can be on the order of 40 to 50 m diameter and 60 to 70 m high ,
with wall and dome thicknesses from 0 .9 to 1 .4 m, and base slab thicknesses from 2 .7 to 4 .1 m. Almost
three-quarters of the NPPs licensed for commercial operation in the US employ either a reinforced concret e
or post-tensioned concrete containment . Boiling-water reactor plants in the US that utilize a stee l
containment have reinforced concrete structures that serve as secondary containments or reactor building s
that provide support and shielding functions for the primary containment .
3.

Potential Degradation Factors

Degradation is considered to be any phenomenon that decreases the load-carrying capacity of the
containment, limits its ability to contain a fluid medium, or reduces its service life . Service-related
degradation can affect the ability of a NPP containment to perform satisfactorily in the unlikely event of a
severe accident . The root cause for component degradation can generally be linked to a design o r
construction problem, inappropriate material application, a base- or weld-metal flaw, maintenance or
inspection activities, or a severe service condition . Primary mechanisms or factors that can produc e
premature deterioration of concrete structures include those that impact either the concrete or reinforcin g
steel materials (i .e., mild steel reinforcement or post-tensioning system) . Degradation of concrete can b e
caused by adverse performance of its cement-paste matrix or aggregate materials under either chemical o r
physical attack. Chemical attack may occur in several forms : efflorescence or leaching, sulfate attac k
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NEA/CSNI/R(2002)7/VOL 1
(including delayed ettringite formation), attack by acids and bases, salt crystallization, and alkali-aggregat e
reactions . Physical attack mechanisms for concrete include freeze/thaw cycling, thermal exposure/therma l
cycling, abrasion/erosion/cavitation, irradiation, and fatigue or vibration . Degradation of mild steel
reinforcing materials can occur as a result of corrosion, irradiation, elevated temperature, or fatigue effects .
Post-tensioning systems are susceptible to the same degradation mechanisms as mild steel reinforcement ,
plus loss of prestressing force, primarily due to tendon relaxation and concrete creep and shrinkage .
4.

Testing and Inspection Requirement s

One of the conditions of all operating licenses for water-cooled power reactors is that the primary
reactor containment shall meet the requirements set forth in Appendix J, "Primary Reactor Containmen t
Leakage Testing for Water-Cooled Power Reactors," to 10 CFR Part 50 [6] . Contained in Appendix J are
requirements pertaining to Type A, B, and C leakage-rate tests that must be performed by each licensee a s
a condition of their operating license. On September 26, 1995, the USNRC amended Appendix J (60 F R
49495) to provide a performance-based option for leakage-rate testing as an alternative to the existing
prescriptive requirements . The amendment is aimed at improving the focus of the body of regulations by
eliminating prescriptive requirements that are marginal to safety and by providing licensees greater
flexibility for cost-effective implementation methods for regulatory safety objectives .
Appendix J to 10 CFR Part 50 requires a general inspection of the accessible interior and exterio r
surfaces of the containment structures and components to uncover any evidence of structural deterioratio n
that may affect either the containment structural integrity or leak-tightness . On August 8, 1996, the
USNRC published an amendment (61 FR 41303) to 10 CFR 50 .55a of its regulations to require that
licensees use portions of the ASME Code for containment in-service inspection . Specifically, the rule
requires that licensees adopt the 1992 Edition with the 1992 Addenda of Subsection IWE, "Requirement s
for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," an d
Subsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Powe r
Plants," of Section XI . In addition, several supplemental requirements with respect to the concrete an d
metal containments were included in the rule . On September 22, 1999 the USNRC again amended 10 CF R
Part 50.55a to endorse use of the 1995 Edition up to and including 1996 Addenda of Section XI ,
Subsections IWE and IWL, of the ASME Code for inspection of containment structures . Subsequently o n
August 3, 2001, the USNRC announced that it intends to amend 10 CFR Part 50.55a to incorporate by
reference the 1997 Addenda, the 1998 Edition, the 1999 Addenda, and the 2000 Addenda of Section XI o f
the ASME Code [7] . Comments on the proposed amendment are presently being addressed .
5.

In-Service Inspection and Condition Assessment

Operating experience has demonstrated that periodic inspection, maintenance, and repair ar e
essential elements of an overall program to maintain an acceptable level of reliability over the service lif e
of a NPP containment, or in fact, of any structural system . Knowledge gained from conduct of an inservice condition assessment can serve as a baseline for evaluating the safety significance of an y
degradation that may be present, and defining subsequent in-service inspection programs and maintenanc e
strategies .
Effective in-service condition assessment of a containment requires knowledge of the expected typ e
of degradation, where it can be expected to occur, and application of appropriate methods for detecting an d
characterizing the degradation. Degradation detection is the first and most important step in the conditio n
assessment process . Routine observation, general visual inspections, leakage-rate tests, and nondestructiv e
examinations are techniques used to identify areas of the containment that have experienced degradation .
Techniques for establishing time-dependent change such as section thinning due to corrosion, or changes i n
component geometry and material properties, involve monitoring or periodic examination and testing .
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Knowing where to inspect and what type of degradation to anticipate often requires information about th e
design features of the containment as well as the materials of construction and environmental factors .
Basic components of the continued service evaluation process for NPP concrete structures include damag e
detection and classification, root-cause determination, and measurement .
5.1

In-service inspection

Nondestructive test methods are used to determine hardened concrete properties and to evaluate the
condition of concrete in structures . 1 Application of these methods for detection of degradation i n
reinforced concrete structures involves either a direct or indirect approach . The direct approach generally
involves a visual inspection of the structure, removal/testing/analysis of material, or a combination of th e
above . Indirect approaches measure some property of concrete (e .g., rebound number or ultrasonic puls e
velocity) and relate it to strength, elastic behavior, or extent of degradation through correlations that hav e
been established previously. Many of the nondestructive test methods are based on the indirect approach ,
in which a small number of destructive and nondestructive tests are conducted in tandem at noncritica l
locations in a structure to develop the required correlation curve(s) . However, destructive tests may not b e
possible in many areas of a NPP structure to develop the required curves so assessment of in-place strengt h
must be based on published relations . Environment-specific methods are used where surfaces of structure s
are not accessible for direct inspection due to the presence of soils, protective coatings, or portions o f
adjacent structures . These methods provide an indirect assessment of the physical condition of th e
structure (i.e., potential for degradation) by quantifying the aggressiveness of the environment adjacent to
the structure (e .g., air, soil, and groundwater) . If results of these tests indicate that the environmen t
adjacent to the structure is not aggressive, one might conclude that the structure is not deteriorating .
However, when conditions indicate that the environment is potentially conducive to degradation, additiona l
assessments are required that may include exposure of the structure for visual or limited destructive testing .
5.2 Condition assessment
Determining the existing performance characteristics and extent and causes of any observed distres s
is accomplished through a condition assessment . Common in the condition assessment approaches is th e
conduct of a field survey, involving visual examination and application of nondestructive and destructive
testing techniques, followed by laboratory and office studies . Guidelines and direction on conduct o f
surveys of existing general civil engineering buildings are available [10,11] . The condition survey usuall y
begins with a review of the "as-built" drawings and other information pertaining to the original design an d
construction so that information, such as accessibility and the position and orientation of embedded stee l
reinforcing and plates in the concrete, is known prior to the site visit. Next is a detailed visual examinatio n
of the structure to document easily obtained information on instances that can result from or lead t o
structural distress. Visual inspections are one of the most valuable of the condition survey method s
because many of the manifestations of concrete deterioration appear as visible indications o r
discontinuities on exposed concrete surfaces . Visual inspections encompass a variety of techniques (e .g.,
direct and indirect inspection of exposed surfaces, crack and discontinuity mapping, physical
dimensioning, environmental surveying, and protective coatings review) . To be most effective, the visual
inspection should include all exposed surfaces of the structure ; joints and joint materials ; interfacing
structures and materials (e .g., abutting soil) ; embedments ; and attached components (e.g., base plates and
anchor bolts) . Degraded areas of significance are measured . The condition of the surrounding structure s
should also be examined to detect occurrence of differential settlement or note aggressiveness of the loca l
operating environment . Results obtained should be documented and photographs or video images taken of
1

Descriptions and principles of operation, as well as applications, for nondestructive test methods most commonl y
used to determine material properties of hardened concrete in existing construction and to determin e
structural properties and assess conditions of concrete are available [8,9] .

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any discontinuities and pertinent findings . A crack survey is usually done by drawing the locations an d
widths of cracks on copies of project plans . Cracking patterns may appear that suggest weaknesses in th e
original design, construction deficiencies, unanticipated thermal movements, chemical reactivity ,
detrimental environmental exposure, restrained drying shrinkage, or overloading . Distress associated with
cracks such as efflorescence, rust stains, or spalling is noted . After the visual survey has been completed ,
the need for additional surveys such as delamination plane, corrosion, or pachometer is determined. The
delamination plane survey is used to identify internal delaminations that are usually caused by corrosion o f
embedded metals or internal vapor pressure. Results of the visual and delamination surveys are used to
select portions of the structure that will be studied in greater detail. To locate areas of corrosion activit y
within reinforced concrete, copper-copper sulfate half-cell studies can be performed . By taking readings at
multiple locations on the concrete surface, an evaluation of the probability of corrosion activity o f
embedded reinforcing steel (or other metals) can be made . Where significant chloride penetration i s
suspected, concrete powder samples or cores should be removed from several depths extending to an d
beyond the embedded outer layer of reinforcing steel . Also, a pachometer survey may be performed a s
part of the detailed study to confirm the location of steel reinforcement. Where there is evidence of severe
corrosion, the steel bar should be uncovered to allow visual inspection and measurement of cross-sectiona l
area loss . Upon return to the office, results of the field survey are evaluated in detail . A crack survey map
is prepared and studied for meaningful patterns . Half-cell data are studied and isopotential lines are drawn
to assist in determining active corrosion sites . Samples of concrete and steel obtained from area s
exhibiting distress are tested in the laboratory . Chloride ion results are plotted versus depth to determin e
the profile and the chloride content at the level of the steel . Any elements that appear to be structurall y
marginal, due either to unconservative design or deterioration effects, are identified and appropriat e
calculation checks made . These analyses may identify distress in the structure that has been caused b y
structural overload and indicate safety factors . If the calculations are inconclusive, suitable load testin g
may be indicated (if feasible). After all of the field and laboratory results have been collated and studied
and all calculations have been completed, a report is prepared .
Cracking is a very common damage by-product from a large number of concrete degradatio n
mechanisms. Active concrete cracking is difficult to assess in terms of impact on structural behavior and is
difficult to repair . Thus, inspection methods that support the early identification, sizing, and cause o f
cracking in concrete structures are of primary interest for future inspections . Also, the primary concern fo r
all metallic constituents of concrete structures is corrosion and corrosion-related damage . Inspections that
identify early signs of corrosion cell initiation and indicate the rate of propagation are similarly valuable .
A visual-based approach based primarily on the results of visual inspections has been developed fo r
assistance in the classification and treatment of conditions or findings that might emanate from in-servic e
inspections of reinforced concrete structures . '
The visual-based approach uses a "three-tiered" hierarchy” so that through use of different levels o f
acceptance, minor discontinuities can be accepted and more significant degradation in the form of defect s
can be evaluated in more detail [13,14] . The three acceptance levels include acceptance without furthe r
evaluation, acceptance after review, and additional evaluation required . Criteria associated with these
acceptance levels are presented elsewhere [13] . Evaluations under these acceptance levels can involve
extensive application of both nondestructive and destructive testing methods and detailed analytica l
evaluations frequently may be required to better characterize the current condition of the structure an d
provide the basis for formulation of a repair strategy (if needed) . Even if the analysis results indicate tha t
the component is acceptable at present, additional assessments should be conducted to demonstrate that th e
component will continue to meet its functional and performance requirements during the desired servic e
Information is also available on a damage-based approach that is founded on the concept that degradation of a
component in service is manifested in physical evidence (i .e ., measurable values) that can be categorized or
classified into distinct stages or conditions in accordance with their impact on performance [12] .

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life (i .e., take into account the current structural condition and use service life models to estimate the futur e
impact of pertinent degradation factors on performance) .
6.

Repair of Reinforced Concret e

Reinforced concrete structures can start to deteriorate due to exposure to the environment (e .g.,
temperature, moisture, and cyclic loading) almost from the time of construction [15]. The rate of
deterioration is dependent on the component’s structural design, materials selection, quality of construction ,
curing, and aggressiveness of its environmental exposure . Termination of a component’s service life occur s
when it can no longer meet its functional and structural requirements . Results provided through periodi c
application of in-service inspection techniques as part of a condition assessment program can be used to
develop and implement a remedial action prior to the structure achieving an unacceptable level o f
performance. Depending on the degree of deterioration and the residual strength of the structure, th e
function of a remedial measures activity may be structural, protective, cosmetic, or any combination of
these .
6.1 Repair considerations
The first step in any repair activity is a thorough assessment of the damaged structure or componen t
including evaluation of the (1) cause of deterioration, (2) extent of deterioration, and (3) effect of
deterioration on the functional and performance requirements of the structure or component . From this
information a remedial measures strategy is developed based on the consequence of damage (e.g., affect o f
degradation on structural safety), time requirements for implementation (e .g., shutdown requirements,
immediate or future safety concern), economic aspects (e .g., partial or complete repair), and residua l
service life requirements (e.g., desired residual service life will influence action taken) [16] . Basic
remedial measures options include (1) no active intervention ; (2) more frequent inspections or conduct of
specific studies ; (3) if safety margins are presently acceptable, take action to prevent deterioration fro m
getting worse; (4) carry out repairs to restore deteriorated or damaged parts of structure to a satisfactory
condition ; and (5) demolish and rebuild all or part of structure . Quite often options (3) and (4) ar e
considered jointly .
6.2 Repair materials and techniques
Deterioration of reinforced concrete generally will result in cracking, spalling, or delamination o f
the cover concrete . Corrosion resulting from either carbonation or the presence of chlorides is th e
dominant type of distress that impacts reinforced concrete structures . More detailed information to that
provided below on typical remedial measures for NPP concrete structures is available [16-18] .
After identifying that a crack is of sufficient size to require repair, it is important to determine i f
the crack is dormant or active (i .e., mechanism still operating) . Dormant cracks can be resin injected using
epoxy or high molecular weight methacrylate (HMWM) . Active cracks must be treated as if they ar e
control joints and require special treatment, especially if fluid leakage is involved . Surface preparation is
critical to a successful spall repair . The concrete substrate must be sound and the exposed surface dry an d
free of oil, grease, and loose particles . The most appropriate materials for patching are those that ar e
closest in composition to the material to be patched . Usually this means portland cement concrete for larg e
patches or portland cement mortar for small ones ; however, non-portland cement binders have been use d
successfully. By patching with a cementitious material, the final thermal and structural properties of th e
repair will be similar to the base concrete. Where the repairs are exposed to aggressive fluids the chemica l
composition of the fluids should be known and the repair materials must be compatible . Delamination s
can be repaired by removal and replacement of the delaminated concrete . In areas where removal o f
concrete is not required, the delaminated area can be repaired by injection of epoxy or HMWM . Proper
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surface preparation, batching, mixing, placing, and curing are all important for long-term durability o f
concrete repairs. Basic repair solutions for corrosion-damaged reinforced concrete include :
(1) realkalization by either direct replacement of contaminated concrete with new concrete, use of a
cementitious material overlay, or application of electrochemical means to accelerate diffusion of alkalis
into carbonated concrete ; (2) limiting the corrosion rate by changing the environment (e .g., drying) to
reduce the electrolytic conductivity; (3) steel reinforcement coating (e .g., epoxy) ; (4) chloride extraction by
passing an electric current (DC) from an anode attached to the concrete surface through the concrete to th e
reinforcement (chloride ions migrate to anode) ; and (5) cathodic protection .
6.3 NPP repair experienc e
A survey was distributed to solicit information on the locations and types of concrete distres s
commonly found in US NPPs [17] . Twenty-nine plants representing forty-two reactor units responded .
The results of this survey are summarized below :
Concrete structure evaluations are usually limited to an assessment of prestressing systems of posttensioned concrete containments and a general visual survey of exposed concrete surfaces ;
Twenty-six of the plants reported concrete damage or deterioration with cracking and spallin g
being most common ;
Most common locations of deterioration in BWR plants were in the containment dome and in the
walls and slabs of auxiliary structures, and in PWR plants the locations were in slabs, walls an d
equipment supports of reactor buildings and auxiliary structures ;
Twenty-seven of the plants have repaired damaged concrete with epoxy injection, grout injection ,
and flexible sealing of cracks being the most common methods utilized ; and
Follow-up evaluation of concrete repairs were not commonly performed .
In general, many of the reported degradation instances associated with the NPP concrete structure s
occurred early in the life of the structures and have been attributed to construction/design deficiencies ,
improper material selection, or environmental effects . Examples of some of the specific problems that
have occurred due to age-related degradation include concrete containment liner corrosion, leaching o f
tendon gallery concrete, corrosion of steel reinforcement in water-intake structures, failure of prestressing
tendon wires due to corrosion, and freeze-thaw damage to containment dome . Although the vast majority
of the structures will continue to meet their functional and performance requirements during their servic e
period, it is reasonable to assume that there will be isolated incidents where the structures may not exhibit
the desired durability without some form of intervention .
7.

Time-Dependent Reliability

Evaluation of structures for continued service should provide quantitative evidence that thei r
capacity is sufficient to withstand future demands within the proposed service period with a level o f
reliability sufficient for public safety . Structural aging will cause the integrity of structures to evolve ove r
time (e .g., a hostile service environment may cause structural strength and stiffness to degrade) .
Uncertainties that complicate the evaluation of aging effects arise from a number of sources : inherent
This survey was conducted prior to the ammendment to 10 CFR 50.55a requiring licensees to use Subsections IWE
and IWL of the ASME Code for containment in-service inspections, and conduct of inspections of selecte d
plants by the USNRC [19] .

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randomness in structural loads, initial strength, and degradation mechanisms ; lack of in-service inspection
measurements and records ; limitations in available databases and models for quantifying time-dependen t
material changes and their contribution to structural capacity ; inadequacies in non-destructive evaluation ;
and shortcomings in existing methods to account for repair. Any evaluation of the reliability of a
reinforced concrete structure during its service life must take into account these effects, plus any previou s
challenges to the integrity that may have occurred .
Time-dependent reliability analysis methods provide a framework for performing conditio n
assessments of existing structures and for determining whether in-service inspection and maintenance ar e
required to maintain reliability and performance at the desired level . The duration of structural loads that
arise from rare operating or environmental events, such as accidental impact, earthquakes, and tornadoes ,
is short and such events occupy a negligible fraction of a structure’s service life . Such loads can be
modeled as a sequence of short-duration load pulses occurring randomly in time . The occurrence in time
of such loads is described by a Poisson process, with the mean (stationary) rate of occurrence, 1 random
intensity, Sj , and duration, ,r The number of events, N(t), to occur during service life, t, is described by the
probability mass function,
P

(1 )

The intensity of each load is a random variable, described by the cumulative distribution functio n
(CDF) Fi(x). In general, the load process is intermittent and the duration of each load pulse has a n
exponential distribution ,
F Td = 1 - exp[-t/ ti] ; t > 0

(2)

in which i = average duration of the load pulse . The probability that the load process is nonzero at an y
arbitrary time is p = 1ti . Loads due to normal facility operation or climatic variations may be modeled b y
continuous load processes . A Poisson process with rate 1 may be used to model changes in load intensity
if the loads are relatively constant for extended periods of time .
The strength, R, of a structural component is described b y
R = B Rm(X 1 , X2, …, Xm)

(3)

in which X 1 , X2 . . . are basic random variables that describe yield strength of steel, compressive or tensil e
strength of concrete, and structural component dimensions or section properties . The function R m(… )
describes the strength based on principles of structural mechanics . Modeling assumptions invariably must
be made in deriving Rm(…) and the factor B describes errors introduced by modeling and scaling effects .
The probability distribution of B describes bias and uncertainty that are not explained by the model R m(… )
when values of all variables Xi are known . The probability distribution of B can be assumed to be normal.
A more accurate behavioral model leads to a decrease in the mean and variability in B and thus in R .
Probabilistic models for R usually must be determined from the statistics of the basic variables, X i, since it
seldom is feasible to test a sufficient sample of structural components to determine the cumulativ e
distribution function (CDF) of R directly .

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The failure probability of a structural component can be evaluated as a function of (or an interval of )
time if the stochastic processes defining the residual strength and the probabilistic characteristics of th e
loads at any time are known . The strength, R(t), of the structure and applied loads, S(t), are both rando m
functions of time . Assuming that degradation is independent of load history, at any time t the margin o f
safety, M(t), is
(4 )

M(t) = R(t) - S(t) .

Making the customary assumption that R and S are statistically independent random variables, th e
(instantaneous) probability of failure is ,
Pf t ) = P M t ) < 0

(5 )

in which FR(x) and fS(x) are the CDF of R and probability density function (PDF) of S . Equation (5 )
provides an instantaneous quantitative measure of structural reliability, provided that Pf(t) can be estimated
and/or validated [20] . It does not convey information on how future performance can be inferred from past
performance.
For service life prediction and reliability assessment, one is more interested in the probability o f
satisfactory performance over some period of time, say (0,t), than in the snapshot of the reliability of th e
structure at a particular time provided by Eq . (5) . Indeed, it is difficult to use reliability analysis fo r
engineering decision analysis without having some time period in mind (e .g., an in-service maintenance
interval) . The probability that a structure survives during interval of time (0,t) is defined by a reliability
function, L(0,t). If, for example, n discrete loads S 1 , S 2 , ..., Sn occur at times t 1 , t2, . . ., tn during (0,t), the
reliability function becomes ,
L(t) = P[R(t 1 ) > S 1 , … , R(tn) > Sn]

(6)

in which R(t i) = strength at time of loading S i .
Taking into account the randomness in the number of loads and the times at which they occur as wel l
as initial strength, the reliability function becomes [21 ]
L (t )= f

exp4- tl 1 - t -1

f

ôFs(g ir ~dt l fR (r ) dt
0

(7)

in which fR0 = PDF of the initial strength R 0 and gi = fraction of initial strength remaining at time of load
S i . The probability of failure during (0,t) i s
F(t) = 1 - L(t).

(8)

The conditional probability of failure within time interval (t, t+ At), given that the component has survive d
up to t, is defined by the hazard function which can be expressed a s
h(t) = -d [ln L(t)]/dt .

(9)

The reliability and hazard functions are integrally relate d

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L(t) = exp

-f ô h (x )dx .

(10)

The hazard function is especially useful in analyzing structural failures due to aging or deterioration . Fo r
example, if the structure has survived during the interval (0, t 1 ), it may be of interest in scheduling inservice inspections to determine the probability that it will fail before t 2 . Such an assessment can b e
performed if h(t) is known . If the time-to-failure is T f, this probability can be expressed as
P

=

1 - exp (f t2 h(xdx .

(11 )

In turn, the structural reliability for a succession of inspection periods i s
t

L (0, t)_

L (ti-1, 4- )exp1~ -f h (x )dx

(12)

ti

t

in which ti-1 = 0 when i = 1 .
Intervals of inspection and maintenance that may be required as a condition for continued operatio n
can be determined from the time-dependent reliability analysis . Forecasts of reliability enable the analys t
to determine the time period beyond which the desired reliability of the structure cannot be assured . At
such a time, the structure should be inspected. The density function of strength, based on prior knowledg e
of the materials in the structure, construction, and standard methods of analysis, is indicated by fR(r). The
information gained during scheduled inspection, maintenance and repair causes the characteristics o f
strength to change ; this is denoted by the (conditional) density fR(r B), in which B is an event dependent o n
in-service inspection . Information gained from the inspection usually involves several structural variable s
including dimensions, defects, and perhaps an indirect measure of strength or stiffness . If these variables
can be related through event B, then the updated density of R following in-service inspection is ,
fR r B ) = P r < R

:9

r + dr, B P B ] = c K r ) fR r

(13)

in which fR(r) is termed the prior density of strength, K(r) is denoted the likelihood function, and c is a
normalizing constant . The time-dependent reliability analysis then is re-initialized following in-servic e
inspection/repair using the updated f R(r B) in place of fR(r). The updating causes the hazard function to b e
discontinuous .
Optimal intervals of inspection and repair for maintaining a desired level of reliability can b e
determined based on minimum life cycle expected cost considerations . Preliminary investigations of such
policies have found that they are sensitive to relative costs of inspection, maintenance, and failure [22] . If
the cost of failure is an order (or more) of magnitude larger than inspection and maintenance costs, th e
optimal policy is to inspect at nearly uniform intervals of time . However, additional research is required
before such policies can be finalized as part of an aging management plan . Applications of the timedependent reliability methodology to concrete components are available [22-24] .
8. Conclusions
The performance of reinforced concrete structures in NPPs has been good, reflecting the initial qualit y
control, the young age, and the generally benign environment within the a plant . However, as thes e
structures age incidences of degradation are likely to increase and if not controlled, degradation has th e
potential to reduce the margins that the structures have to withstand various challenges in service from
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NEA/CSNI/R(2002)7/VOL 1
operating conditions, the natural environment, and accidents . The most common form of degradation
observed in NPPs has been concrete cracking. When properly used and applied, in-service inspectio n
techniques are effective in detecting aging effects and providing vital input for assessing the condition o f
structures. Methods for conduct of condition assessments of reinforced concrete structures are fairly wel l
established and generally start with a visual examination of the structure's accessible surfaces . Some
guidance has been developed to aid in interpreting results of the condition assessment, but more definitiv e
criteria are required to assist in interpreting the data provided . Repair methods for general civil engineering
reinforced concrete structures are fairly well established and effective when properly implemented, however ,
the long-term effectiveness (or durability) of remedial measures require development. Time-dependent
reliability analysis methods provide a framework for performing condition assessments of existing structures
and determining whether in-service inspection and maintenance are required to maintain reliability an d
performance at the desired level, however, quantitative data for input into the methodology are limited and
the reliability models for condition assessment have not been validated .
9.

Acknowledgements

Research sponsored by the Office of Nuclear Regulatory Research, U .S . Nuclear Regulatory
Commission under Interagency Agreement 1886-N604-3J with the U .S. Department of Energy unde r
Contract DE-AC05- 96OR22725 . The submitted paper has been authored in part by a contractor of th e
U.S. Government under Contract No . DE-AC05-96OR22725 . This paper has also been prepared in part by
an employee of the USNRC and presents information that does not currently represent an agreed-upo n
Staff position. The USNRC has neither approved nor disapproved its technical content . The U .S.
Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of thi s
contribution, or allow others to do so, for U .S . Government purposes .
10.

References

1. American Concrete Institute, 1964 . Building Code Requirements for Reinforced Concrete, ACI
Standard 318-71, Detroit, Michigan .
2. Lo, T. et al ., 1984 . "Containment Integrity of SEP Plants Under Combined Loads," in Proceedings of
the ASCE Conference on Structural Engineering in Nuclear Facilities, J . Ucciferro (ed.), American
Society of Civil Engineers, New York, New York .
3. American Society of Mechanical Engineers, 2001 . "Rules for Construction of Nuclear Power Plan t
Components," ASME Boiler and Pressure Vessel Code, Sect. III, New York, New York .
4. U.S. Nuclear Regulatory Commission, 1981 . "Concrete Containment," Sect. 3 .8.1 in Regulatory
Standard Review Plan, NUREG-0800, Directorate of Licensing, Washington, DC .
5. U.S. Nuclear Regulatory Commission, 1981 . "Concrete and Steel Internal Structures of Steel and
Concrete Containments," Sect . 3 .8 .3 in Regulatory Standard Review Plan, NUREG-0800, Directorate
of Licensing, Washington, DC .
6. Office of the Federal Register, 1995 . "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors," Appendix J in Code of Federal Regulations, 10 CFR Part 50, Office of
Federal Register, Washington, DC .
7. Office of theFederal Register, 2001 . "Proposed Rules," Vol . 66, No . 150, pp. 40626-40640, Office of
Federal Register, Washington, DC .
8. ACI, 1995 . "In-Place Methods for Determination of Strength of Concrete," ACI 228 .1R, American
Concrete Institute, Farmington Hills, Michigan .
9. ACI, 1998 . "Nondestructive Test Methods for Evaluation of Concrete Structures," ACI 228 .2R,
American Concrete Institute, Farmington Hills, Michigan .
10. American Society of Civil Engineers, 1991 . Guidelines for Structural Condition Assessment of
Existing Buildings, ANSI/ASCE 11-90, New York, New York .
11. Perenchio, W. F., 1989. "The Condition Survey," Concrete International, 11(1), pp . 59-62, American
35

NEA/CSNI/R(2002)7/VOL 1
Concrete Institute, Detroit, Michigan .
12. Naus, D . J., Braverman, J .I., Miller, C .A ., Ellingwood, B .R., Hofmayer, C .H., 2000. “Factors Related
to Degradation of Nuclear Power Plant Concrete Structures,” Proc. of International RILEM Workshop
on Aging Management and Life Prediction of Concrete Structures, Cannes, France .
13. Hookham, C.J., 1995 . “In-Service Inspection Guidelines for Concrete Structures in Nuclear Powe r
Plants,” ORNL/NRC/LTR-95/14, Oak Ridge National Laboratory, Oak Ridge, Tennesee .
14. ACI, 1996 . Evaluation of Existing Nuclear Safety-Related Concrete Structures, ACI 349.3R-96,
American Concrete Institute, Farmington Hills, Michigan .
15. Browne, R.D., 1989 . "Durability of Reinforced Concrete Structures," New Zealand Concrete
Construction, Parts 1 and 2 .
16. Price, W.F. et al., 1993 . Review of European Repair Practice for Corrosion Damaged Reinforce d
Concrete, Report No . 1303/91/5823, Taywood Engineering Ltd ., R & D Division, London, England .
17. Krauss, P .D., 1994 . Repair Materials and Techniques for Concrete Structures in Nuclear Powe r
Plants, ORNL/NRC/LTR-93/28, Martin Marietta Energy Systems, Inc ., Oak Ridge National
Laboratory, Oak Ridge, Tennessee .
18. Emmons, P .H ., 1993 . Concrete Repair and Maintenance Illustrated, R . S. Means Company, Inc .,
Kingston, Massachusetts .
19. Ashar, H. and Bagchi, G ., 1995 . “Assessment of Inservice Conditions of Safety-Related Nuclea r
Power Plant Structures,” NUREG-1522, U.S. Nuclear Regulatory Commission, Washington, DC .
20. Ellingwood, B .R., 1992. “Probabilistic Risk Assessment,” Eng. Safety, McGraw-Hill Book Co ., Ltd. ,
London, England, pp. 89-116 .
21. Ellingwood, B.R. and Mori, Y., 1993 . “Probabilistic Methods for Condition Assessment and Life
Prediction of Concrete Structures in Nuclear Power Plants,” Nuc. Eng. and Des ., Elsevier Science
S.A., North-Holland, Amsterdam, The Netherlands, Vol 142, pp . 155-166.
22. Mori, Y. and Ellingwood, B.R., 1994 . “Maintaining Reliability of Concrete Structures II : Optimum
Inspection/Repair Strategies,” Ibid, pp . 846-862.
23. Mori, Y. and Ellingwood, B .R., 1994 . “Maintaining Reliability of Concrete Structures I : Role of
Inspec-tion/Repair,” J. of Struct. Eng., American Society of Civil Engineers New York, New York ,
120(3), pp . 824-845 .
24. Ellingwood, B .R. and Song, J., 1996. “Impact of Structural Aging in Probabilistic Risk Assessment o f
Reinforced Concrete Structures in Nuclear Power Plants,” NUREG/CR-6425, U.S. Nuclear Regulatory
Commission, Washington, DC .

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SESSION A : OPERATIONAL EXPERIENCE
Chairman : Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany )

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REPAIR OF THE GENTILLY- 1
CONCRETE CONTAINMENT STRUCTURE
A Popovic, D Panesar and M . Elgohary
Atomic Energy of Canada Limite d
Mississauga, Ontario, Canada

Abstract
Gentilly-1 CANDU Nuclear Power Plant is located in Gentilly, Quebec, south of the St . Lawrence River .
Gentilly-1 was designed in the late 1960'ies and constructed in the mid 1970'ies . The reactor building ring
beam has suffered concrete degradation for more than fifteen years and has been repaired in 1985 and i n
1993 . These repairs were ineffective since extensive deterioration had continued to occur . The ring beam
contains and protects the prestressing anchorages and the horizontal ring beam tendons, in addition to th e
pre-stressing anchorages for the dome prestressing system . In August 1998, an assessment program for the
concrete containment building was initiated . The investigation showed that the structural concrete forming
the dome and the perimeter wall is expected to endure without ageing problems for at least fifty years .
However, portions of the ring beam concrete `secondary concrete' needed repair . Various design options
for the ring beam repair were considered. Atomic Energy of Canada Ltd . (AECL) and Intelligent Sensing
for Innovating Structures (ISIS) Canada developed a design for the remedial work and a subsequent
monitoring system . The innovative Ring Beam Repair Program was implemented and successfull y
performed in 2000/2001 . The purpose of this paper is to describe the design and field implementation o f
the repair program .
Introduction
The Plant was designed by AECL in the late 1960'ies and became operational in 1972 . In 1980, AECL
decided to place the station in a lay-up state and to generate a plan for its final disposition . It wa s
concluded in 1984 that returning the site to a condition of completely unrestricted access and usage was no t
immediately necessary and, for technical and financial considerations, the attainment of this objectiv e
should be delayed for the next fifty to eighty years . AECL, the owner of the facility, was required t o
maintain the facility in a "static state" . In general, the purpose of the maintaining of the facility was t o
provide interim storage for all conventional and radiological hazards until the facility is finall y
decommissioned and demolished . Therefore, it was decided that Gentilly-1 containment structure would
be required to function for more than fifty years beyond its originally expected design life . However, the
structure was visibly aging in the ring beam secondary concrete area that was protecting the prestres s
anchorages . The poor appearance of the structure was significantly influencing perception of safety an d
proper functioning of the nuclear installations . The ring beam deterioration is shown in Figures 1a and 1b .

39

NEA/CSNI/R(2002)7/VOL1

Figure 1a, b: G-1 Ring Beam, South Side

Figure 2 : Gentilly-1 Containment Building – Installation of Formwork (April 2000 )
A detailed study was undertaken to perform a condition assessment of the structure and the action s
necessary to ensure satisfactory performance of the structure over the remaining period of required servic e
life . Structural condition assessment studies have been completed and it has been concluded that, excep t
the part of the secondary concrete in the ring beam, the concrete containment structure has the potential t o
be serviceable possibly for fifty years or more, as required by AECL .
Based on the findings from the condition assessment, recommendations were developed for the repai r
design and implementation performed on the ring beam to ensure the satisfactory performance of all part s
of the concrete containment structure .

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Condition Assessment and Design Requirement s
As a result of the condition assessment completed in the year 2000, and all previous investigations ,
valuable information was gathered based on : visual inspections, in-situ measurements and laborator y
testing and analysis . A comprehensive analysis which included testing of concrete, reinforcing steel and
prestressing steel materials, chloride ion content, water soluble alkali content, air voids, water absorption ,
compressive strength, modulus of elasticity, Poisson’s ratio, etc ., concluded that the containment structur e
was in good structural condition. It was also concluded that the reactor building could be expected t o
remain serviceable for the next fifty years if the ring beam repair is performed .
The repair of the ring beam of the Gentilly-1 containment structure was developed by AECL and ISI S
Canada . The objective of the concrete repair was to : to remove all unsound and unbonded concrete ; to
restore the concrete of the ring beam ; to protect pre-stressing anchorages; and to improve aesthetics of the
building and increase public confidence . The Project had challenges relating to the technical component o f
the repair, as well as to the logistic construction challenges due to the work at the top of the reacto r
building, as shown in Figure 2 .
The technical part of the concrete repair project had three main tasks :
1. Concrete demolition (removal of unsound concrete) and surface preparation .
2. Concrete repair .
3. Fiber Reinforced Polymers (FRP) protection of the repaired area .
Concrete demolition, surface preparation and concrete repair are part of the conventional engineering task s
and adequate experience and expertise can yield a high standard of repair design and installation .
On the other hand, protection of the repaired concrete using FRP is a newer area in the nuclear industr y
and does not have an extensive track record . Therefore, a team of specialists was assembled to study an d
propose the technical solution for the application of FRP .
FRP composites have been used for nearly thirty years in aerospace and manufacturing applications wher e
low weight, high tensile strength and non-corrosive structural properties are required . In civil engineering ,
applications of different types of FRPs are finding their role in fabric roof structures, internal concret e
reinforcement, deck grating and as externally bonded reinforcement or protection . The FRP system ha s
proven benefits in some applications . The technique, known as a wet lay-up, provides flexibility ,
constructability and short installation times, resulting in lower overall cost . The system may use different
type of fibers : carbon, E-glass and Aramid, depending on the particular requirements of an application .
The fiber fabric is installed using epoxy resin formulated for substrate adhesion, durability an d
constructability.

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Figure 3 details the application concept of FRP sheets .

Figure 3 : FRP - Installation Detai l

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Figure 4 : Ring beam FRP Installation Pattern
PROJECT EXECUTION
Concrete Demolitio n
As shown in Figure 5, using the sounding method, all of the unsound and unbonded concrete was marked,
measured and prepared for demolition .

Figure 5 : Sounding the Concrete Surface
The unsound concrete was removed using saw cuts and jack hammers, as presented in Figure 6 . Fo r
shallow demolished sections (less than 50 mm deep), the saw cut around the perimeter of repair was at
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NEA/CSNI/R(2002)7/VOL 1
least 12 mm. For deep demolished sections (more than 50 mm deep), the saw cut was at least 25 mm . The
maximum demolition depth into the ring beam was 0 .66 m.

Figure 6 : Concrete Demolitio n
After the demolition, and prior to the grouting/concreting, skin reinforcement was installed, Figure 7 .
Surface preparation by hydro-jetting or sand blasting was employed to open the pore structure of th e
concrete surface and to remove dirt and other debris material, Figure 8 .

Figure 7: Installation of Skin Reinforcemen t

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Figure 8 : Surface Preparation –Hydro-Jetting

Concrete Repair - Materials and Application Method s
The concrete repair materials and repair techniques differ for shallow repairs and deep repairs . The
shallow repair patches have an average thickness of less than 50 mm . The deep repair patches have a n
average thickness of greater than 50 mm .
The shallow patches are repaired with Sika Repair 225 . The repair material is a prepackaged ready to use ,
cementitious, high strength, shrinkage compensated mortar, which includes silica fume and fibr e
reinforcement. The mix is used with the amount of water specified in the technical data sheet . The
material was applied by hand troweling .
The deep patches are repaired with Sika Grout 212, SikaCem 810, PeaGravel (5-9 mm) and water . The
patches are formed, the concrete mix is poured and vibrated while being placed in the forms as shown i n
Figures 9 and 10 .

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NEA/CSNI/R(2002)7/VOL1

Figure 9 : Deep Repair - Concrete Pour

Figure 10 : Deep Repair – Complete d

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Glass Fiber Reinforcing Polymer (GFRP) Installatio n
For this application, the MBrace (Master Builder Technologies) Composite Strengthening System wa s
chosen. Its function is primarily to protect and ensure durability of the concrete repairs . The structural
strength and high modulus of elasticity were not critical requirements . Based on the material properties, Eglass fiber (EG-900) fabric was the selected material since it exhibited the best flexibility and elasticity o f
the system for this application .
The steps for the GFRP application are as follows :
Step 1 : MBrace Primer (low viscosity to penetrate the concrete pore structure) .
Step 2 : MBrace Putty (high viscosity epoxy paste used for surface leveling) .
1st
Step 3 : The Resin Coating, MBrace Saturant (low sag epoxy that encapsulates the fibers) .
Step 4 : MBrace GFRP EG 900 E-Glass Fiber Fabric (instead of C-Fiber shown in Figure 3) .
Step 5 : The 2nd Resin Coating, MBrace Saturant (low sag epoxy that encapsulates the fibers) .
Step 6 : Protective Coating: Sonocoat Topcoat Super Colorcoat VOC Top Coat was used .
Site application of the GFRP to the ring beam is shown in Figure 11 .

Quality Control
Quality control of the freshly mixed repair material included: water/cement ratio, slump of the mix,
sampling the mix for the compression tests, compression test results . The 28 day compression test result s
ranged from fc ’=39 to 51 MPa, which exceeded the specified requirement of fc ’=30 MPa .
The quality workmanship and the concrete properties of the placed repairs was verified by drilling 101 . 6
mm diameter cores ‘in situ’ and performing pull-out tests as shown in Figure 12 . The pullout tests gave
average bond strength (repair material to old concrete) from 1 .2 MPa to 2 .1 MPa, which is above the
revised value of 1 .0 MPa required by the Specifications .

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Figure 12 : Pull-out Test
The quality of the installation of the GFRP and compliance with the Specification was also tested b y
performing pull-out tests on four 101 .6 mm diameter cores . The pull-out tests gave average bond strength
of the GFRP to substrata (repair material or original concrete) from 1 .9 MPa to 3 .4 MPa . That was als o
above the value required by the Specification . The failure mode was never through the GFRP/substrat a
contact joint, but ra

Figure 13 : Completed Repair South-West View

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Conclusions
Based on the results of the assessment studies and the repair work performed, it is concluded that :
1) The repair design and implementation of the Gentilly-1 ring beam has been successfull y
completed .
2) The repaired structure is in good condition and is satisfying current functional, safety, design an d
aesthetic requirements . The life of the structure has been extended for at least the next fifty years .
3) The structure should be closely monitored. The behavior of the repair materials will be compared
with the original design intent, and the required maintenance effort will be performed to ensure
that the design requirements characterized by service life are met .
Acknowledgement s
The authors would like to thank Vector Construction Group for completing the Project on time an d
satisfying high quality requirements, and ISIS Canada for their technical support and expert opinion wit h
the development of the technical specification .

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The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures and Installation of an
Automated Cathodic Protection System .
LM Smith British Energy Generation (UK) Lt d
CA Hughes British Energy Generation (UK) Lt d
G Jones
C-Probe Ltd

Abstract
This paper reports on a project which carried out repairs to the headworks and associated jetty and marine
structures at Hunterston B nuclear power station . Although alternative sources of cooling are available, the
headworks and associated structures have important availability functions with regard to the provision o f
cooling water for the nuclear power plant . These marine structures are situated in an exposed coasta l
location with consequentially aggressive environmental conditions . The jetty and headworks were
originally designed to codes that have now been superseded and in order to ensure that the repaired
headworks structure would have sufficient future service life an automated impressed current cathodi c
protection system was installed at the time of the repair works .
As a structural material concrete is strong in compression and weak in tension . In order that concrete may
be utilised in practical structures, reinforced concrete is provided with steel reinforcement that carries an y
tensile loads or stresses by composite action . It is therefore important that the steel reinforcement i n
reinforced concrete structures is maintained in good condition otherwise the capacity of the structures wil l
be degraded and reduced. The major threat to concrete structures is corrosion of the steel reinforcement
and this is particularly the case in marine structures due to chloride contamination . The extent of corrosion
may be monitored by measurement of the corrosion potential of the steel reinforcement relative to a
reference electrode and cathodic protection may be employed .
The remedial works carried out to the jetty at Hunterston B are a typical example of the use of this type o f
repair and monitoring . The Hunterston jetty is the only means of access to the station’s Cooling Water
intake headwork structure . The intake headwork is essential as it provides the sea water that is used as th e
cooling water for the steam condensers within the power station . The approach jetty and the headwor k
were constructed using conventional reinforced concrete in the late 1950s and the early 1970s respectively .
During routine structural inspections it was identified that the structures required remedial works primaril y
due to the high degree of chloride contamination as a direct result of the environmental exposur e
conditions found at the site .
To ensure structural integrity over the remaining life span of the structures, an impressed current Cathodi c
Protection (CP) system was selected for the headwork structure above Mean Low Water Springs (MLWS )
level and a sacrificial anode system below this level . The impressed current CP (ICCP) system on th e
structure above MLWS is divided into a number of anode zones and each zone is independently powere d
and monitored. The anode for all the zones comprises a mixed metal oxide coated titanium mesh fixed
directly to the repaired and prepared concrete substrate with proprietary fixing pins which hold the mes h
rigidly against the concrete.
This paper describes the repair works carried out to the headworks and the function, technical details ,
installation and performance of the cathodic protection system .

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The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures and Installation of a n
Automated Cathodic Protection System .
Introduction
This paper reports on a project which carried out repairs to the headworks and associated jetty and marine
structures at Hunterston B nuclear power station . Although alternative sources of cooling are available, the
headworks and associated structures have important availability functions with regard to the provision o f
cooling water for the nuclear power plant . These marine structures are situated in an exposed coasta l
location with consequentially aggressive environmental conditions . The jetty and headworks were
originally designed to codes that have now been superseded and in order to ensure that the repaired
headworks structure would have sufficient future service life an automated impressed current cathodi c
protection system was installed at the time of the repair works . This paper describes the repair works
carried out to the headworks and the function, technical details, installation and performance of th e
cathodic protection system .

History
The remedial works carried out to the jetty at Hunterston B are a typical example of the use of this type o f
repair and monitoring . The Hunterston jetty is the only means of access to the station’s Cooling Water
intake headwork structure . The intake headwork is essential as it provides the sea water that is used as th e
cooling water for the steam condensers within the power station . The approach jetty and the headwor k
were constructed using conventional reinforced concrete in the late 1950s and the early 1970s respectively .
The water intake head works structure is of reinforced concrete construction as shown in Figures 1 and 2 .
During the mid-1980s repairs were carried out to the water intake headworks to remedy deterioratio n
related to chloride ingress and reinforcement corrosion . These repairs were largely patch repairs but als o
included:
Application of sprayed Gunite with an embedded secondary steel reinforcement mesh typically 125m m
in depth to the entire soffit of the upper deck slab .



Application of black pitch extended epoxy resin to the upper deck slab soffit .



Application of an epoxy paint system to the columns and intake shaft above high water level only .

No cathodic protection was included in the repair scheme .
By the mid-1990s the repaired areas and some additional areas of the structure had begun to deteriorat e
once more and, during routine structural inspections, it was identified that the structures again required
remedial works . This was primarily due to the high degree of chloride contamination as a direct result o f
the environmental exposure conditions found at the site which was aggravated by the dynamic effect o f
wave impact and storm damage on the .existing repair s

Repair s
As a structural material concrete is strong in compression and weak in tension . In order that concrete may
be utilised in practical structures, reinforced concrete is provided with steel reinforcement that carries any
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NEA/CSNI/R(2002)7/VOL 1
tensile loads or stresses by composite action . It is therefore important that the steel reinforcement i n
reinforced concrete structures is maintained in good condition otherwise the capacity of the structures wil l
be degraded and reduced. The major threat to concrete structures is corrosion of the steel reinforcement
and this is particularly the case in marine structures due to chloride contamination . The extent of corrosion
may be monitored by measurement of the corrosion potential of the steel reinforcement relative to a
reference electrode and cathodic protection may be employed .
To ensure structural integrity over the remaining life span of the structures, an impressed current Cathodi c
Protection (CP) system was selected for the headwork structure above Mean Low Water Springs (MLWS )
level and a sacrificial anode system below this level . The combined system therefore comprises :
Sacrificial anode system to all exposed surfaces between +0 .60m OD and –1 .325m OD, and
Titanium mesh impressed current anode and cementitious overlay to all surfaces above +0 .60m OD
The impressed current CP (ICCP) system on the structure above MLWS is divided into a number of anod e
zones and each zone is independently powered and monitored . The anode for all the zones comprises a
mixed metal oxide coated titanium mesh fixed directly to the repaired and prepared concrete substrate with
proprietary fixing pins which hold the mesh rigidly against the concrete .
The structure was divided into six Remedial Work Areas (RWAs) . A 40mm gap was left between th e
installed anode meshes in adjacent RWAs except in areas with double layer anode meshes where 50 mm o f
the top layer mesh was removed. All metallic objects within the protected areas were made electricall y
continuous with the steel reinforcement . A 25 mm sprayed concrete overlay was generally applied over the
anode mesh, which was increased to 75 mm at the top deck . In some areas existing concrete was cut back
to maintain both external clearances and the overlay thickness .
The impressed current cathodic protection system chosen was the C-Probe Achilles system (Figure 3 )
which allows the structure to be divided into a number of zones for both cathodic protection and corrosio n
monitoring by means of embedded electrodes . A zoned ICCP and monitoring system has superio r
performance to a single global system as the impressed current can be tailored to meet the requirements o f
separate areas of the structure .
Existing previous repairs to the deck slab soffit, including secondary steel mesh reinforcement, wer e
removed and the surface was profiled to maintain a minimum of 30 mm cover to the main reinforcement ,
which was repaired as required. The sequence of removal and reinstatement operations (A through E) is
shown in Figure 4 . High pressure water jetting was used to remove the concrete and patch repairs (Plates 1
& 2) . Where required, existing reinforcing bars that were excessively corroded were supplemented and/o r
replaced with new steel suitably anchored into the structure . Once the entire deck slab soffit had been
reprofiled and prepared a titanium anode mesh and anode ribbons were fixed and a sprayed concret e
overlay applied (Figures 5 & 6) .
Performance
The latest repairs have performed very well without deterioration over a period of six years and have not
shown the rapid deterioration that occurred with previous repairs . This is considered to be due to the
inclusion of the zoned ICCP and monitoring system and it is recommended that future repairs to importan t
structures in this type of location include such a system.

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Figure 1

F ig u re 2

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Figure 3

55

Plate 1 Water jetting

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Plate 2 Exposed steel reinforcement after removal of cove r

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FEASIBILITY STUDY OF IE-SASW METHOD FOR THE NON-DESTRUCTIVE EVALUATIO N
OF CONTAINMENT BUILDING OF NUCLEAR POWER PLAN T
Yong-Pyo Suh, Korea Electric Power Research Institute, Kore a
Jong-Rim Lee, Korea Electric Power Research Institute, Kore a
Jeong-Moon Seo, Korea Atomic Energy Research Institute, Kore a
ABSTRAC T
The IE-SASW method that combines Impact-Echo (IE) method with spectral analysis of surfac e
waves (SASW) is proposed as a newly developed nondestructive testing method in concrete structures .
This method is based upon the idea that IE method uses the elastic p-wave velocity measured from SAS W
method on the concrete member, and applied to specimens to evaluate its feasibility . It was shown that the
thickness of the concrete structure member and the depth of the defects such as voids could be identifie d
by IE-SASW method with good reliability . Additionally, the GPR (Ground Penetrating Radar) technique s
have been applied to the same specimens in order to establish the performance and reliability of th e
proposed method, and compared with IE-SASW method . The experimental studies show that it is mor e
preferable to use the IE-SASW than GPR to detect the voids just beneath the steel reinforcing bars that
may exist in concrete structures.
Keywords: Non-destructive Testing, Concrete, Impact-Echo, SAS W
1. INTRODUCTIO N
The construction quality of the containment building in the nuclear power plant is carefull y
controlled and thoroughly inspected to prevent from unexpected flaws . In general, the concrete of
containment building is deteriorating as time passes by, so the periodic safety assessments using non destructive tests are also crucially required. Until now, the non-destructive tests such as radar, impactecho, and ultrasonic methods have been developed for the concrete structure and compared thei r
characteristics by various test specimens (M . Krause . et al., 1997). Even though each non-destructive
testing method has its own advantage and capability, it is necessary for the user to understand the
limitations of the methods exactly before applying them . Also, the best combinations of testing method s
can be selected only after comparative studies are performed .
In this study, IE-SASW testing method was introduced and the performance was evaluated by
applying to the containment building of nuclear power plant. The IE-SASW method combines IE (impactecho) with SASW (spectral analysis of surface waves) methods, and is applied to detect the flaws as wel l
as the thickness of concrete structures . In the original IE method, the thickness and flaw depth in th e
concrete structure are determined using the predetermined P-wave speed (Sansalone, 1997) . The accuracy
of the result is much dependent on the accurate measurement of representative P-wave speed of the testin g
location. In the IE-SASW method, the representative P-wave speed is determined using SASW method .
In this paper, the performance of applying IE-SASW method to the non-destructive evaluation o f
containment building of nuclear power plant was studied . Two test specimens were constructed and the
defects were included at the known locations. One of the specimens was the prototype of a structura l
member of the containment building typically built in Korea . IE-SASW method was performed to locate
the flaws and to determine the thickness .

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NEA/CSNI/R(2002)7/VOL 1
Additionally, the GPR (Ground Penetrating Radar) method is applied to the same specimens, an d
compared with IE-SASW method . The series of experimental studies were focused to detect the void jus t
beneath the steel reinforcing bars that may happen in concrete structures .
2 . NON-DESTRUCTIVE TESTING METHODS USING ELASTIC WAV E
2.1 . Impact-Echo Method
Impact-echo method is a nondestructive testing method of concrete structures that is based on th e
propagation characteristics of impact-generated stress waves that are reflected by internal flaws and
external surfaces . It can be used to determine the location and extent of flaws such as cracks ,
delaminations, voids, honeycombing, and debonding in reinforced concrete structures (Sansalone, 1997) .
A schematic diagram of impact-echo method is shown in Figure 1 .
The stress pulse generated by an impact on the surface propagates back and forth between th e
internal interface and top surface of a test object. Surface displacements caused by reflections of thes e
waves are recorded by transducer (accelerometer) located adjacent to the impact . The resultin g
displacements versus time signals are transformed into the frequency domain, and plots of amplitud e
versus frequency spectra are obtained. Multiple reflections of stress waves between the impact surface ,
flaws, and/or other external surfaces give rise to transient resonance, which can be identified in the
spectrum, and used to evaluate the integrity of the structure or to determine the location of flaws . The
typical wave forms and amplitude spectrum of signal are illustrated in figure 2 .
The period of reflected waves is equal to the travel path 2T, divided by the compressive wave velocity, Vp .
Since the frequency is the inverse of the period, the resonance frequency f, is :
f = Vp 2T

(1 )

This is a fundamental equation of impact-echo response for solid member, and if the Vp is predetermined, the thickness and location of internal flaws can be identified .
The above analysis is valid for the cases where the reflecting boundary or internal interface ha s
lower acoustic impedance (density × P-wave speed) than the member . When this case occurs, P-wave
incident upon the interface changes sign . For example, the P-wave generated by impact is a compressio n
wave . When this wave is incident upon a solid/air interface, the reflected wave is a tension wave a s
illustrated in figure 3(a) . If, however, a P-wave is incident upon an interface that has higher acousti c
impedance, such as steel reinforcing rebar in the concrete, or enlarged area, the incident P-wave does no t
change sign, and tension wave will be reflected at the interface (higher acoustic impedance) as a tensio n
wave as illustrated in figure 3(b) . Thus equation (1) must be modified, because the period is twice as long .
f = Vp 14T

(2)

2.2. Spectral Analysis Of Surface Waves (SASW) Method
The spectral analysis of surface waves (SASW) method is a method of seismic testing that ha s
developed for determining shear wave velocity profiles at soil and/or pavement sites (Nazarian and Stoko e
ii, 1986) . The SASW method is a nondestructive method in which both the source and receivers ar e
located on the surface as shown in figure 4 .
The source is simply a transient vertical impact that generates a group of surface waves of variou s
frequencies in the medium . Two vertical receivers located on the surface monitor the propagation o f
surface wave energy . By analyzing the phase information of the cross power spectrum determine d
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NEA/CSNI/R(2002)7/VOL 1
between the two receivers, surface wave velocity - wavelength relation is determined. If the stiffness of a
site varies with depth, then the surface wave velocity will vary with wavelength . The variation of surfac e
wave velocity with frequency (wavelength) is called dispersion, and a plot of surface wave velocity versu s
wavelength is called a dispersion curve . The dispersion curve is developed from phase information of the
cross power spectrum . This information provides the relative phase between two signals (two-channe l
recorder) at each frequency in the range of frequencies excited in the SASW test . In a homogeneou s
medium, surface wave velocity, Vr , is constant and independent of the wavelength . Detailed procedure o f
determining dispersion curve in the SASW method is described elsewhere (Joh, 1996, Nazarian and Stoko e
ii, 1986).
In order to apply the SASW technique to the nondestructive test in concrete structures, the variatio n
in surface wave velocity along the whole thickness of a concrete structure should be determined, and the P wave velocity can be converted using Equation (3), assuming the Poisson's ratio of the concrete to be i n
the range between 0.15 and 0 .2 .
Vp _

1 +V

2(1 -v)

VR
(3 )
0.87 + 1 . 12v (1 - 2v)
The probable error in determining P-wave velocity with the assumed Poisson's ratio of concrete betwee n
zero and 0 .2 is about 3%, which is minimal, and with Poisson's ratio of 0 .2, the Equation (3) reduces to :
Vp = 1 .79 VR

(4)

2.3. IE-SASW method

In order to predict the thickness of concrete member or to identify the defect using IE method, P wave velocity of concrete is required. The P-wave velocity can be found as presented in Equations (1) an d
(2), when the boundary condition and the thickness of concrete member are predetermined . In general ,
however, the thickness of member such as slab and wall of building remains unknown . The concrete
specimen should be extracted from the structures using core-boring machine, and substituting the height o f
core specimen into the equations (1) produces the P-wave velocity. But this method has such
disadvantages that the surface of structure is destructed due to core boring, and the P-wave velocit y
calculated may not be representative value of the structures because of the non-homogeneity of th e
concrete material.
Therefore, IE-SASW method that enables to obtain the P-wave velocity from SASW metho d
nondestructively and then to apply the IE method, is proposed in this study . In order to obtain the material
properties in the multi-stratified soil system using SASW method, the iterative inversion processing is
required until the discrepancy between the experimental dispersion curve and theoretical dispersion curv e
is minimal (Joh, 1996) . But in concrete structures, the surface wave velocity (Vr ) can be easily obtained by
experimental dispersion curve without executing the inversion procedure, because concrete material i s
assumed to be composed of a single layer . Using the relationships among surface wave velocity (V r), Pwave velocity (V p), and Poisson's ratio ( v) with assumed Poisson's ratio of 0 .20, P-wave velocity o f
concrete can be determined by Equation (4) . Then, the thickness or the location of defect such as void in
the concrete member can be identified using IE method nondestructively .

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3 . APPLICATION S
3.1 Test Specimens
Two test specimens were made to study the feasibilities of IE-SASW method in the nondestructiv e
evaluation of reinforced concrete structures . One test specimen, named “A” (length : 150cm, width : 50cm,
thickness : 30cm, including two voids) was designed as illustrated in figure 5 . Voids were simulated wit h
styrofoam of which the acoustic impedance is distinguishably smaller than that of concrete . Especially, in
this specimen, one void is located just beneath the rebars and the other is located apart from the rebar. The
cover depth to the void beneath the steel reinforcing bar is 10cm and the other is 15cm. The mix
proportion of the concrete is shown in table 1 .
In order to obtain the P-wave velocity, three concrete test molds were made and cured in a wate r
bath at 20 c for 3 days . After casting, the P-wave velocity was measured using IE method respectively ,
and averaged to provide the P-wave of about 3300 m/sec .
Another test specimen, named “B”, is prototype structural member of the containment building of a
nuclear power plant with three tendon sheathing pipes (diameter of 150mm) that are unfilled and som e
dozens of rebars (diameter of 55mm) as shown in figure 6 . In order to compare the feasibilities of both IESASW and GPR methods, the cubic styrofoam (10cm x 10cm x 10cm) and water container (width : 20cm,
height: 40cm, thickness : 40cm) filled with water, and a PVC pipe which is 15cm in diameter were inserte d
intentionally during the construction at the depth of 30cm from the surface of the specimen . The
photograph 1 shows the prototype member of the containment building of a nuclear power plant .
3.2 Results of IE-SASW metho d
3.2 .1 . Test specimen- A
The SASW test was performed on the test specimen-a to get the p-wave velocity before applying th e
IE test. The interval between source and the first receiver is set to 20cm that is the same distance betwee n
receivers . Figure 7 shows the dispersion curve produced by the phase difference between the two
receivers, and the phase velocity is approximately chosen out to be 1870m/sec . The phase velocity
between two receivers is equal to surface wave velocity without inversion process, because it can b e
assumed that the surface wave velocity is constant and independent of wavelength in a concrete layer.
Thus, the P-wave velocity can be calculated as 3347m/sec by the equation (4), assuming the Poisson’s rati o
of concrete to be 0 .2. It is resulted that the P-wave velocity of 3,300 m/sec obtained from the SASW tes t
shows good agreement with the velocities that are previously obtained from the concrete test molds and
from IE test using known thickness . Therefore, it is revealed that the p-wave velocity can be obtained
from the SASW test reliably.
The IE test was performed at the three positions : (1) on the void caged behind steel reinforcing bars ,
(2) on the void where there is no rebar, and (3) on the surface where there is no defect through tes t
specimen. Several steel balls were used as impact sources, and accelerometer (PCB 353b15) was used as a
receiver. Signals were recorded and analyzed using a dynamic signal analyzer (HP 35665a) . The
amplitude spectrums of acceleration at these three positions are presented in figures 8 (a), (b), and (c) ,
respectively .

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NEA/CSNI/R(2002)7/VOL 1
The location of the void and the thickness of specimen can be determined by amplitud e
spectrums using P-wave velocity determined by SASW method, and substituting these resonanc e
frequencies in figures 8 (a), (b), and (c) into equation (1) produces 10 .6cm, 15 .1cm, and 30 .1cm,
respectively. Comparing these results with true values of 10 .0cm, 15 .0cm, and 30 .0cm, it can b e
concluded that the IE-SASW method shows a good potential to identify the defect and unknown thicknes s
of concrete member reliably .
3.2 .2 . Test Specimen- B
The P-wave velocity of test specimen was determined by the SASW test . A test was carried out
along the line where there were no defects and no rebars as designated in figure 6 . SASW test wa s
performed twice, changing the interval between source and the first receiver from 20cm to 40cm . The
dispersion curves are presented in figures 9 (a) and (b) . The P-wave velocities calculated from the surfac e
wave velocities were 4200m/sec and 4100m/sec, respectively, and average value of 4150m/sec was used i n
IE test.
A series of IE test were performed at the position (E1 to E13) designated in figure 6, and the
amplitude spectrums of acceleration are shown in figure 10 . At position of E1, the resonance pea k
frequency of 1824hz was observed, and the thickness of wall was calculated as 1 .14m, which is in goo d
agreement with the true thickness of 1 .2m . At the position of E5, E6, E11, and E12 where the metal sheat h
pipes are located, the clear and large resonance peak frequencies were observed in the range from 4700h z
to 4900hz . These frequencies produce the depth to the metal sheath pipe as 0 .45~0 .47 m, which fits wel l
with the actual depth of 0 .45m, and it is revealed that the metal sheath pipe have lighter impedance tha n
concrete and acts as a free boundary . However, the impact-echo response from the rebar of position E 2
shows the dominant peak at 4700hz, which does not give the actual depth to the steel reinforcing bar . The
acoustic impedance of rebar is about five times that of concrete, but the area of rebar is much less than that
of concrete . In this case, the reflected and refracted P-waves from the interfaces of concrete/steel o r
steel/concrete created by the rebar makes the signal to be complicated, because the location of rebar i s
shallow but the depth to the bar is relatively deep . This fact makes it difficult to locate the rebar by IE test .
the amplitude spectrums corresponding to styrofoam, water container, and PVC pipe are shown in figures
11 respectively . The impact-echo responses from these objects show large amplitude resonance betwee n
6khz and 7khz . At the position of p1 in PVC pipe, the highest amplitude peak occurs at 6 .976khz: this
peak is used to calculate a depth of 0.297m, in good agreement with the known depth of 0 .3m . At the
position of s4 in styrofoam, the highest amplitude peak occurs at 6 .336khz : this peak is used to calculate a
depth of 0.327m, which is close to the known depth of 0 .3m. Finally, at the position of w4 in wate r
container, the highest amplitude peak occurs at 6016khz : this peak is used to calculate a depth of 0 .345m,
which is a little larger than the actual depth of 0 .3m. These results show that it is feasible to use the IESASW method to detect the voids (filled with either water or not) and PVc pipe .
2.4. Results of GPR method

2.4.1 . Test specimen- A
A ground penetrating radar (GPR) was employed and tested on both specimens A and B for th e
comparison purpose. Figure 12 shows the typical processed image corresponding to the profile of test
specimen-a, which recorded with 1200mhz antennas . This section shows reflected and diffracted signal s
(hyperbolas designated by arrows) in response to the different objects : (1) six rebars, and (2) the void
where there is no rebar . Contrary to the ie method the void beneath reinforcing steel bars could not b e
detected by GPR method.

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2.4.2 . Test specimen-B
The GPR profile corresponding to the test specimen-b is shown in figure 13 . A series of rebar
spacing evenly is illustrated as hyperbolas . However, it is difficult to identify the tendon sheath locate d
behind the rebar, because most of the electromagnetic wave is reflected when encountered with rebar .
Pulse type dipole antenna of 1,200mhz (central frequency) failed to detect the objects such as PVC pipe ,
styrofoam, and water container in the specimen-b .
3.

Conclusions

The comparative studies presented here illustrate the performance and feasibility of nondestructiv e
test to the concrete structures . The experiments were quite useful to select the most suitable techniques fo r
specific applications . In this study, the IE-SASW method and GPR method are employed to evaluate thei r
feasibilities. The conclusions obtained are as follows :
(1)
(2)
(3)

(4)

The IE-SASW method can be applied to identify the location of defect and the thickness o f
concrete structure with good accuracy.
Especially, the void just beneath the rebar in test specimen-a could be easily detected by the IE SASW method . On the other hand, the rebar itself could not identified by this method .
The location of tendon sheathing and thickness of the structural member (test specimen-B) coul d
be identified by the IE-SASW method . Also the location of the objects such as styrofoam, wate r
container, and PVC pipe that were intentionally included in the test specimen-B can be detecte d
by this method .
The GPR method provides an objective and reliable image corresponding to the rebars an d
defects such as voids. But it was difficult to identify the void just beneath the reinforced stee l
bars . Therefore, the detection potential can be improved by the combined utilization of the IESASW method and the GPR method .

REFERENCE S
[1] Carino, N. J., and Sansalone, M ., 1992, "Detection of Void in Grouted Ducts Using the Impact-Ech o
Method", ACI Material Journal Vol . 89, No . 3, pp.296-303.
[2] Carino, N . J., Sansalone, M ., and Hsu, N. N., 1986, "Point Source-Point Receiver Technique for Flaw
Detection in Concrete", ACI JOURNAL, Proceedings Vol . 83, No . 2, pp .199-208.
[3] Joh, S . H ., 1996, "Advances in Interpretation and Analysis Techniques for Spectral-Analysis-o f
Surface-Waves (SASW) Measurement," Ph . D . Dissetation, The University of Texas at Austin .
[4] Lin, Y. and Sansalone, M ., 1992, "Detecting Flaws in Concrete Beams and Columns Using the Impact Echo Method", ACI Materials Journal, Vol . 89, No . 4, pp .394-405.
[5] Malhorta, V. M ., and Carino, N . J ., 1991, "HANDBOOK on NONDESTRUCTIVE TESTING of
CONCRETE", CRC Press, New York, 343p .
[6] M . Krause, M . Bormann, R. Frielinghaus, F . Kretzschmar, O . Kroggel, K . J. Langenberg, C.
Maierhofer, W. Müller, J . Neisecke, M . Schickert, V . Schmitz, H . Wiggenhauser and F . Wollbold,
1997, "Comparison of pulse-echo methods for testing concrete", NDT&E International, Vol . 30, No .
4, pp . 195-204 .
[7] Nazarian, S ., and K.H . Stoke, II, 1986, In Situ Determination of Elastic Moduli of Pavements System s
by Spectral Analysis of Surface Waves Method (Theoretical Aspects), Research Report Number 437 2, U .S. Department of Transportation, Federal Highway Administration, pp .2-6 .
[8] Richart, F . E. Jr., Hall, J . R . Jr., and Woods, R . D., 1970, "Vibrations of Soil and Foundation", Prentic e
Hall, Inc ., Englewood Cliffs, New Jersey, 414p .
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[9] Sansalone, M ., and Streett, W . B ., 1997, Impact-Echo Nondestructive Evaluation of Concrete an d
Masonry, BULLBRIER PRESS, ITHACA, N .Y ., pp .9-320.
[10] Sansalone, M ., and Carino, N . J., 1989, “Detecting Delaminations in Concrete Slabs with and without
Overlays Using the Impact Echo Method”, ACI Material Journal, Vol . 83, No . 2, pp .175-184 .

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Figure 1 . Simplified diagram of impact-echo metho d

(b) amplitude spectrum
Figure 2. Typical result of impact-echo test

f1 = VP / 2 T

f1

Freq.

f1

Freq.

void

(a) lower acoustic impedance
(b) higher acoustic impedanc e
Figure 3 . Comparisons of resonance frequency between lower and higher acoustic impedance

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NEA/CSNI/R(2002)7/VOL 1

Figure 4 . General configuration of SASW testing

Figure 5 . Drawing of test specimen-A

11

SPA . @ 35 .56 cm
rebar (diameter=55mm,

Styrofoam(depth=

Figure 6 . Drawing of test specimen-B that is the prototype structural member
of containment building of nuclear power plan t

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NEA/CSNI/R(2002)7/VOL1

0

4000

Figure 7. SASW test result to determine the P-wave velocity of test specimen-A

0

10000

20000

30000

frequency(Hz)

(a) amplitude spectrum for the void behind steel reinforcing bar s
3 .0E-0 5
2 .0E-0 5
1 .0E-0 5
0 .0E+00
0

10000

20000

3000 0

frequency(Hz)

(b) amplitude spectrum for the void where there is no steel reinforcing ba r
1 .0E-05

5 .0E-06

0 .0E+00
0

10000

20000

3000 0

frequency(Hz)

(a) on the surface where there is no defect through test specimen
Figure 8 . Amplitude spectrums of acceleration for test specimen-A at the three positions :
(a) on the void behind steel reinforcing bars, (b) on the void where there is no stee l
reinforcing bar, and (c) on the surface where there is no defect through test specimen .

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Phase Velocity (m/s)
0.00

1000

0

VR=2350 m/s
Poisson’ Ratio =0.2
Vp = 4200

2000

3000

Phase Velocity (m/s)
4000

0.00

Distance : 20 cm

0

1000

VR=2300 m/s
Poisson’ Ratio =0.2

m/s

2000

3000

4000

D istance : 40 cm

Vp = 4100 m/s

0.60

0.30

0.80

(a) Distance between receivers : 20cm

(b) Distance between receivers : 40cm

Figure 9 . SASW test results to determine the P-wave velocity of test specimen- B

rebar

E4
E5
E6

E1 0
E1 1
E1 2

Figure 10. Amplitude spectrums of acceleration for test specimen-B at different positions (Theoretica l
resonance frequencies corresponding to wall thickness and sheath location were plotted a s
dotted lines respectively.)

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NEA/CSNI/R(2002)7/VOL 1

10000

20000

30000

0

10000

Frequency (Hz)

20000

Frequency (Hz)

(a) Amplitude spectrum corresponding to PVC pipe (B) Amplitude spectrum corresponding to Styrofoa m

10000

20000

Frequency (Hz)

(c) Amplitude spectrum corresponding to water containe r
Figure 11 . Amplitude spectrums of acceleration for PVC pipe, Styrofoam, and water container in tes t
specimen-B at different positions

(a) Schematic plan showing the location of reinforcing bar and void s

(b) Radar image of the reinforcing steel bars and voi d
Figure 12. GPR image at test specimen-A (scanned by 1200MHz antenna)

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NEA/CSNI/R(2002)7/VOL 1

profile
Reinforcing steel bar
Metal sheath

(a) Schematic plan showing the location of profile with reinforcing steel bars and metal sheath s

Figure 13 . GPR profile at test specimen-B (scanned by 1200MHz antenna )

Water
185

Table 1 . Mixture proportion of test s pecimen concrete
Aggregate
Air Content
o
Cement
S/A(/o)
Coarse
Fine
(%)
320
1026
713
41
5

W/C
(%)
58

Photograph 1 . Prototype Structural Member of Containment Building Of Nuclear Power Plan t

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NEA/CSNI/R(2002)7/VOL 1
Field Studies of Effectiveness of Concrete Repairs
NJR Baldwin, Mott MacDonald, UK
Abstract
This paper presents a summary of the work carried out under an HSE research study entitled `Field Studie s
of Effectiveness of Concrete Repairs' .
There is little published information describing or comparing the long-term performance of different repai r
types. This project examines 45 sites where repairs have been carried out and have remained in service fo r
several years and evaluates the effectiveness of a range of concrete repair systems as applied in practice .
The sites include a range of structure types, ages and service environments, and include bridges, tunnels ,
building frames, and car parks in industrial, public, highway and nuclear environments . The repairs
include hand and trowel applied materials, sprayed materials, and cathodic protection techniques, and als o
some sites with coating and crack injection systems .
The objective is to improve practices for maintaining and improving the integrity of operational structure s
and so achieve higher standards of structural safety and reliability and better whole-life structural
management.
The project employed a range of visual, non-destructive and destructive investigation techniques at
repaired sites and compared the condition found with records of the repair procedures and objectives . The
site investigations involved detailed visual inspection, and surveys using hammer tapping, covermeter ,
half-cell and carbonation depth, pull-off testing and core sampling for petrographic analysis of the repai r
material, repair layer interfaces and repair/substrate interface .
Examination of records of the repairs allowed assessments to be made of the level of understanding of the
original cause of deterioration and the need and objectives for the subsequent repairs . Comparisons have
been made between the specification and evidence from the repair sites . The owner's repair objectives an d
constraints, and the quality and effectiveness of the repairs have been considered .
The site investigations have provided detailed information on the performance, structure and effectivenes s
of repairs . Full records of the repairs were characteristically difficult to obtain . Most repairs were to
corroded reinforcement in structures affected by chloride-induced corrosion . There was often evidence of
post-repair corrosion with the exception of those structures with a cathodic protection system . Fine cracks
and surface crazing were common in the repair materials and at the perimeter of patch repairs . These can
penetrate to the reinforcement and represent performance limiting features .
The project output includes production of guidelines covering the decision making process of concret e
repair. These guidelines will be disseminated to industry via the HSE and through new guidance notes
within industry documents and a proposed ICE publication .
1.

Introduction

1.1

Origin

Mott MacDonald Ltd (MM) was commissioned by the UK's Health and Safety Executive (HSE) in June
2000 to carry out a research study entitled `Field Studies of Effectiveness of Concrete Repairs' . The project
follows on from the project conducted by the UK's nuclear industry entitled `Concrete repair materials an d
protective coating systems' that produced a compendium of repair materials and systems 1 .

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In 1999 the effectiveness of in situ repairs was added to the HSE’s Nuclear Research Index . The issue
questioned the potential performance, maintainability and longevity of nuclear safety related structure s
with extended service and safe storage lives using current concrete repair practices . HSE’s Nuclear
Industry Inspectorate (NII) proposed a programme of field studies to examine typical repair sites . NII
recognised the difficulties in gaining access to repair sites within nuclear facilities and in 2000 secure d
funding from the HSE to conduct field studies on the repairs to a variety of concrete structures away from
nuclear sites .
The scope and objectives have been developed between the HSE, MM and other organisations whos e
interests are represented in an Expert Group associated with the project . Funding has also been received
from the Highways Agency (HA) and the Institution of Civil Engineers (ICE) Research and Development
Enabling Fund . The project receives substantial additional support from collaborating organisation and
individuals, as well as the co-operation of owners of repaired structures .
1.2

Objectives

The aim of this project is to evaluate the effectiveness of a range of concrete repair systems as applied i n
practice, in order to improve practices for maintaining and improving the integrity of operational structure s
and so achieve higher standards of structural safety and reliability and better whole-life structural
management. This includes assessment of the whole process whereby repair is carried out, and i n
particular what parts of the process lead to success or failure. The project also investigates the effects o f
ageing of repairs and identifies the most effective means of providing enhanced durability . Guideline s
covering the decision making process of concrete repair will be disseminated to industry .
The project has focussed primarily on non-structural patch repairs to reinforced concrete intended to arres t
deterioration resulting from the ageing processes prior to any significant effect on structural integrity suc h
that structural intervention is avoided . Structural repairs, in which the load paths pass through the repair,
have not been specifically targeted .
Investigations have also been carried out into the long-term performance of practical cathodic protectio n
(CP) schemes . The performance of the systems has been assessed, and the effects of CP on the bond of
reinforcement and the distribution of ions throughout the concrete has been investigated .
1 .3 Definition s
The terms used within the study have been the subject of intense debate . The problems lie in finding
definitions which are agreeable for civil, structural and materials engineers, the repair industry an d
operators of structures, for terms such as defect, repair, effectiveness, performance, non-structural repai r
and structural repair. Examples of definitions from ENV 1504-9 2 and EU FP5 research project
“LIFECON” are presented in Table 2 .

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NEA/CSNI/R(2002)7/VOL 1

Table 1 Definitions
ENV 1504-9
Defect

An unacceptable condition whic h
may be in-built or may be the resul t
of deterioration or damage .

Repair

A measure which corrects defects.

Protection

A measure which prevents or reduce s
the development of defects .

EU FP 5

Return of a structure to an acceptabl e
condition by the renewal, replacement
or mending of worn, damaged or
degraded parts .

The definition and objective of repair works are also variable and the following definitions are found in the
literature :
Etebar3 defines the objective of repair as being to restore or enhance one property such as durability ,
structural strength, function or appearance .
Walker4 states that rehabilitation involves controlling degradation to enable a structure to continue t o
serve its intended purpose, either through repair to a state similar to the original, or using methods to
arrest deterioration processes .
Emmons and Vaysburd5, state that the object of any repair project is to “produce a repair at relatively
low cost with a limited and predictable degree of change over time and without deterioration and/or
distress throughout its intended life and purpose”.
These demonstrate the different elements and concepts of the objectives of repair and intended repai r
performance. Measurement of repair effectiveness is more complex and subjective, and is best addresse d
by comparing condition over time relative to the original objectives .
1.4 Effectiveness
A repair may be effective if it has achieved the performance that it was originally intended to . However,
there may be several aspects to the original intention, such as cost, longevity and cosmetic issues . There
may also have been requirements or restrictions for preparation and application of the repair, that form par t
of it’s effectiveness . This project has used a detailed definition of effectiveness, adopting the principles of
the ‘SPALL’ criteria, defined by King and Ecob6 and listed below .
Structural: Possesses the required structural properties .
Protection : Provides protection for the reinforcement.
Application : Can be applied effectively within the given constraints .
L ongevity : Once applied it remains in place .

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NEA/CSNI/R(2002)7/VOL 1
L ooks:

Has an appropriate surface finish .

The ‘SPALL’ parameters, in combination with assessment of the quality of the processes of selection an d
specification of the repair, have been reviewed at individual sites . Site observations and testing have been
used to gather information on the effectiveness of the technical, site activity aspects of the repair, i .e. the
physical actions involved in execution of the repair . The contract information and records have been use d
to evaluate the process effectiveness, i.e. the planning and management of the repair.
1.5

Structure

The project was carried out in four main stages . The first was a data gathering phase in which the literatur e
available for the types, methods, standards, investigation techniques and performance of repairs wa s
reviewed7 . This confirmed there is little independent data on the long-term in situ performance of most
concrete repair systems . The second stage involved identifying the sites to be visited, contacting th e
owners/operators, sourcing record documents, and executing risk assessments and designing th e
investigation procedures . The third stage involved visiting 45 sites to form a database of defects 8.
Specialist techniques were used to investigate a large number of repairs and CP systems at one major site 9.
The final and current stage of the project involves the analysis and presentation of the data 10. All of the
reports will be available via the HSE in 2002 .
2.

Site investigation

2.1 Number and type of sites investigated
Patch repairs were targeted where the depth of repair was less than the full depth of the element, typicall y
less than 100mm deep, and the areas repaired were typically less than 1m x 1m. A wide variety of repairs
and repaired structures were included to provide a representative population for study . In total, 45 site s
were visited, and have been described in detail in relation to age, natural and service environment, repai r
history and condition. However, the sites remain anonymous . At certain sites, there was more than on e
type, generation or condition of repair, and in total 65 locations were examined .
The sites were located in throughout England and included infrastructure, public buildings and nuclea r
facilities in coastal, estuarine, river and inland locations . Repaired elements included beams, columns ,
slabs and walls from structures including bridges, tunnels, power stations and other reinforced concret e
frame buildings . The type and environment of structures visited is summarised in Table 2 and Table 3 .
Table 2 Types of repair examine d
Structure
type and
location
Total
Total

Concrete frame structure
Public

Industrial

Inland

Inland

Estuarine

Coasta l

2

2

4

6

14

Car
park

Road bridge/viaduct

Road
tunnel

Others

Inland

Inland

River

Estuarine

Estuary

14

2

4

5

Inlan d
underpass
1

5

1

5
5

20

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NEA/CSNI/R(2002)7/VOL 1

Table 3 Types of repair examine d

Total

Hand/trowel applied
proprietary
conventional
dense high buil d
25
8
7

Total

40

Type of
repair

Sprayed
proprietary

Flowa
ble

CP

conventional

Crac k
Sealin g

9

3

9

5

3

9

5

3

12

At 48 locations the repairs were investigated through visual inspection, hammer surveying, non-destructive
testing, and intrusive sampling, and 60 core samples were examined in the laboratory . At 13 locations the
repairs were examined by visual inspection and hammer surveying . At 3 further locations visual
inspection was supplemented with non-destructive testing (NDT) .
The sites were selected where safe access could be achieved to repairs where some knowledge o r
documentation of repair age, locations, type and method existed . The repair locations were selected to
represent a range of sizes, types and application methods . The repairs were applied by hand or trowel ,
spray, or flowable methods . Mostly the repairs were of cement-based conventional materials or proprietar y
repair systems, and up to 12 years old . Seven different CP systems were also examined at four different
structures. Each was associated with repair with cementitious materials prior to installation of the C P
system.
2.2

Records of repai r

For each site, an attempt was made to discover the maintenance history, the cause of deterioration, and th e
parties involved in the repair process. This involved consultation with the structure owner/operator and
recovery of documents related to condition prior to repair, the cause of defects, the repair contract ,
specification and method statement, and detailed description and/or data sheets for the materials used in th e
repair. Photographs of the repair contract were also invaluable.
The level of documentation available varied greatly. The older the repairs, the more difficult it was t o
recover all of the information. There was also a difficulty in recovering documents where the ownership o f
the structure had changed . Where more than one repair generation occurred, there was often n o
information relating to the earlier repair phases .
For a number of sites no records were available, or records of repair did not appear in existing engineerin g
and maintenance files . Records from Health and Safety files did not contribute significantly to th e
information recovered for the sites, despite a requirement under CDM regulations 11 for retention of the
Health and Safety file for the life of the structure .
2.3 Road Tunnel Deck Stud y
The research also took advantage of replacement of the reinforced concrete deck of a major UK road
tunnel during 2000-2001 . Sections of the deck containing repairs and CP systems were inspected an d
sampled. A total of 26 cores and 18 sections measuring approximately 600mm x 500mm, were saw n
through the full depth of the deck . The sections were water jetted to expose a 250mm length of each bar i n
one face, and pull-out testing was carried out on 57 bars at the University of Birmingham .

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3.

Site investigation methods

The performance of the repairs was assessed through a combination of visual inspection, non-destructiv e
testing e .g. hammer tapping, half-cell potentials, cover meter survey, carbonation depth . Destructive
sampling and testing was carried out where permitted by the structure owner, and included pull-off
resistance, extraction of a core for petrographic examination and electron microscopy, assessment o f
carbonation depth, and sampling for chloride and alkali contents . Other investigation techniques, such a s
chloride and alkali determinations were used at specific sites where deemed necessary . The sample hole s
were repaired with a proprietary high-build repair system and the site cleared and vacated .
Where possible, more than one repair was sampled, identifying contrasts in condition, appearance, age ,
location, and exposure conditions .
3.1

Effectiveness of investigations

Visual inspection and hammer tapping were of greatest value on site, particularly when combined wit h
core sampling or break out . The visual inspection provided a rapid assessment of the overall condition o f
the structure, its environment, and condition of repairs and coatings . Inspection of defects, particularly
when combined with documentation of the condition and maintenance history of the structure, provided a n
insight into the causes, severity and timing of deterioration . In comparison, the value of the half-cell survey
and pull-off resistance tests were limited .
3.2

Half-cell surveys

Half-cell surveys were carried out at many locations and most required careful interpretation . In general,
there was a marked difference in potential values for the repairs and for the substrate concrete . Typically
the repairs had a less negative value, and was often positive, indicating a dry substrate . A common trend
found in columns and walls, in both repairs and substrate, was of increasing negative potential toward s
ground level, and particularly within 0 .5m of ground level . This is associated with the increase in moisture
content of the substrate and not necessarily associated with corrosion activity . However, in the road and
car park structures, chloride concentrations may be higher in the lower portions of elements exposed to sal t
spray . This can result in higher corrosion potentials . The portions within 1 .5m of ground level exposed to
spray were observed to have a high incidence of deterioration and repair .
Where reinforcement was exposed at the surface, either through low cover or through spalling of the cove r
concrete, an increase in negative potential was typically found only by detailed mapping and often over a
very limited area. This was probably related to the general lack of moisture within the dense repai r
materials . Consistently high negative potentials were found only in wet substrates, for example beneat h
leaking joints .
Previous research 12 has found the surface zone of cementitious systems to have significant effects on
oxygen diffusion and resistivity . These ‘skin effects’ can interfere with half-cell surveys and provide a n
inaccurate impression of corrosion potential at depth .
3.3

Pull-off resistance

The pull-off testing was characteristically problematic and time consuming particularly on overhea d
locations and on rough substrates . Failure typically occurred at the interface between repair material an d
the substrate (36%), within the concrete substrate (21%) or in the adhesive (25%) . The recorded failure
values typically ranged between 0 .1 and 0.9N/mm2, with a mean value was 0 .48 N/mm2 and standard
deviation of 0 .22 N/mm2.
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No difference was found between pull-off resistance in repairs with substrates prepared by waterjetting
compared to those prepared by mechanical break out . However, other research has demonstrated
significantly higher bond strength on overlays applied to water jetted substrates compared with san d
blasted and mechanically broken substrates 13 .
3.4 Core sampling and petrographic examinatio n
Examination of the core samples and core holes provided information on macroscopic feature such as voi d
content and distribution, discontinuities, cracks, condition of reinforcement, and configuration of repairs
and interfaces. The petrographic examination of the repair and substrate provided detailed information o n
the microstructure and quality of the repair material, substrate and the interface between them . Thi s
identified differences and subtleties in repair composition and structure that were not appreciated on site,
such as the extent and depth of fine cracking and microcracking from the external surface, the depth of
carbonation along these features, layering in repairs and subtle differences in composition of the binder .
4.

Preliminary findings

4 .1 Reason for repai r
The majority of the repairs examined related to corrosion of the reinforcement embedded in the original
concrete substrate . Overall, approximately 60% of the locations examined had deteriorated, or wer e
repaired, as a result of chloride induced corrosion or the potential for it. Approximately 25% had
deteriorated as a result of carbonation induced corrosion, and the remainder by drying shrinkage or othe r
mechanisms. The structures built between the 1930’s and 1950’s had mostly required repair as a result o f
carbonation induced corrosion . These were almost exclusively in locations devoid of an external source o f
chlorides . In most structures there was evidence of some post-repair deterioration. This was often
associated with the original cause of deterioration, such as chloride ingress . Performance of the repair s
was not always a measure of the effectiveness of the management strategy .
4 .2 Repair structure
Many features of the repairs could be identified, such as sawn perimeters, feathered edges, different
methods and depths of break-out, presence of bond coats, reinforcement primers, levelling coats, external
coatings, and the occurrence of layers, partings, cracks and voids . The components and procedures
described in the repair records did not always match those found in the site investigation . The thickness of
applied layers in hand or trowel applied materials was often greater than that recommended in the materials
data sheets. There was little evidence that this had compromised repair effectiveness .
Many of the repairs were finished flush with the surrounding surfaces, resulting in reinstatement of lo w
cover depths, sometimes less than 10mm. Repairs at several locations had an intentional local increase in
cover to overcome this . At other sites, the cover was enhanced by application of cementitious render or
additional protection was afforded by a paint or high performance coating at the external surface .
4.3 Cracking in repairs
Cracking occurred at the external surface in approximately 60% of repair locations examined.
Approximately 45% of repair locations contained surface crazing and/or cracks resulting from dryin g
shrinkage . The cracks commonly passed from the external surface of the repair to the embedded
reinforcing steel, and can pass through the full thickness of the repair . The drying shrinkage cracks are
typically less than 0.1mm wide, and the fine cracks grade into microcracks, clearly visible in the repair s
under petrographic examination, and remain readily resolved at 0 .005mm width .
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NEA/CSNI/R(2002)7/VOL 1
There is evidence of penetration of water into cracks and carbonation of the binder adjacent to cracks . At
approximately 20% of repair locations the depth of carbonation along the cracks was greater than the
minimum cover depth at the site or the depth of reinforcement intersected in the core sample . Fine cracks
and microcracks in repairs less than a year old were carbonated to depths of up to 21mm . The depth of
carbonation from the crack surfaces was very limited and typically no more than 2-5 times the width of the
crack .
Cracks were also found at the margins of repairs . The extent of carbonation in the repair material wa s
mostly insignificant, but the substrate concrete was commonly more porous than the repair, and carbonate d
to greater depths . This is significant where reinforcement bars pass from the substrate into the repair an d
local depassivation at the steel could result in corrosion . The ingress of chlorides into cracks is also known
to result in corrosion of reinforcement at the perimeter of repairs 14 .
4.4 Deterioration of repairs
Deterioration had occurred to some ‘holding’ repairs in aggressive environments . A small proportion of
‘long-term’ repairs were failing prematurely. The cause of deterioration was mostly site-specific and
included reoccurrence of the original cause of deterioration, ingress of water and chloride, shortcomings i n
the specification, crazing or cracking in the repair material, carbonation of poor quality material, vibratio n
in the structure, reinstatement of low cover and the presence of a cavity at the repair/substrate interface.
Repairs were ineffective mostly in the ‘Protection’, ‘Longevity’ and ‘Looks’ aspects of the SPALL criteria .
Evidence for incipient anode formation was found at several sites with cracking or spalling of the repair
and/or surrounding concrete, typically at or close to the repair perimeter . All of these locations had been
repaired because of chloride-induced corrosion . There were also locations where corrosion had continued
at the repair site, not necessarily through incipient anode formation, but mostly through ongoing exposur e
to saline water, or the presence of carbonated concrete adjacent to the repair . The repair planning and
management processes for these sites were not fully effective .
4 .5 Performance of C P
Each of the sites where CP had been operational showed very low levels of deterioration to the substrat e
concrete and patch repairs. Some deterioration of the external coating systems was noted . The mean bond
stress determined through pull-out testing on 57 bars was 3N/mm 2. The pull-out testing found no evidenc e
that CP had affected the bond strength for plain round bars in original deck concrete or the sprayed
concrete repairs in areas with CP and in areas without CP .
Examination of core samples confirmed that there was no evidence of significant corrosion of the bar s
embedded in concrete protected with CP systems . Ionic mapping of the binder in repairs and origina l
concrete confirmed trends of ionic concentrations at the surface anodes and embedded cathodic stee l
reinforcement. No evidence was found for significant deterioration resulting from these concentrations .

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5.

Preliminary Conclusion s

This project aims to define and assess the effectiveness of in situ patch repairs and CP systems. A large
population of repairs of up to 12 years age have been investigated .
Records for the repair contracts have been examined but are not always complete or available .
The majority of structures had repairs to corroding reinforcement, caused predominantly by chlorid e
contamination . Many of the repaired structures showed evidence of post-repair deterioration, particularl y
where the cause of deterioration was not effectively treated. However, the structures with CP systems
were in notably good condition .
Many of the patch repairs contain shrinkage cracks which represent a performance limiting factor .
6.

Acknowledgements

The author acknowledges the following funders of the research :
Health and Safety Executive .
Highways Agency .
Industry Management Committee .
Institution of Civil Engineers Research and Development Enabling Fund .
MM Group Research and Development Fund.

7.

References

1 Sheffield University Centre for Cement and Concrete, Mott MacDonald Special Services Division, DEWPitchmastic PLC (1999), "Concrete repair materials and protective coating systems", Nuclear repair contract
BL/G/31221/S, IMC Reference CE/GNSR/5020 .
2 ENV 1504 (1997), "Products and Systems for the Protection and Repair of Concrete Structures - Definitions,
requirements, quality control, evaluation of conformity", Part 9, "General Principles for the Use of Products an d

Systems "
3 Etebar, K., "Integrity of repaired concrete under repeated loading conditions" pp 493-502 Proceedings of th e
international conference on concrete repair , rehabilitation and protection, University of Dundee, Scotland, UK ,
27-28 June 199 6
4 Walker, M., "An overview of rehabilitation methods and selection of an appropriate system" . Pp169-180.
Proceedings of the international seminar on controlling concrete degradation, University of Dundee, Scotland ,
UK, 7 September 199 9
5 Emmons, P .H . & Vaysburd, A.M . (1993), "Factors affecting durability of concrete repair - The contractor' s
viewpoint", Proceedings of 5th international conference on structural faults and repair, Vol.2 Extending the Life of
Civil & Building Structures, University of Edinburgh .
6 King, E .S . & Ecob, C.R . (1993), "Review and Specification of Concrete Repair Materials" .

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NEA/CSNI/R(2002)7/VOL 1

7

Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R109 9
‘Phase 1 Report’, April 2001 .

8 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R112 0
‘Phase 3 Report : Records of site investigation’, 2002 .
9 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R112 1
‘Phase 3a Report: Investigations at UK Tunnel’, 2002.
10

Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R113 4
‘Phase 4 Report’, 2002

11

Construction (design and management) Regulations, 1994 (amended 2000), UK Legislation, Stationary Office Ltd,
London .

12 M B Leeming, Concrete in the Oceans – Phase II – Co-ordinating report on the whole programme, Fina l
Report to contributors, July 1986 .
13

J Silfebrand ‘Improving Concrete Bond in Repaired Bridge Decks’, Concrete International, September 199 0

14

TRRL Contractor Report 209, J G Keer and J R Chadwick, ‘Corrosion of reinforcement embedded in concrete
repair materials exposed to de-icing salts’, 1990.

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OECD-NEA WG 10-11 .04.2002 BERLIN

Detect and repair of defects on th e
confinement structure at Paks NPP

CSABA NY ARADI

system technologist engineer
Paks Nuclear Power Plant Ltd .
P.O.B 71 . 7031 Paks,
Hungary
[email protected]

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ABSTRACT
Paks Nuclear Power Plant is the only commercial nuclea r
facility in Hungary, which has been operational since 1982 .
Like other N-plants, Paks NPP is also exposed to majo r
cha llenges due to plant aging and changes in
circumstances that affect the operation . T he defense in
depth concept is a corner of nuclear safety, therefore th e
confinement concrete structure and its aging phenomena is
on the focus of the utilities . Declaring the lifetime extension
program, Paks NPP pays distinguished attention on
confinement integrity and achieved significant results on this
field .

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Content of presentatio n
1. Introduction
2. The construction and structure of hermetic spac e
(confinement.)
3. Methods of examination of leakage and strengt h
test
4. The results of leakage tests .
5. Repairing of defects found during examinations .
6. Short review of repair of the liner and concret e
structure at the Unit 1 . of NPP Paks .
7. Conclusions.

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1 . Introduction
The third element of the defence in depth concept is th e
confinement which provides the last physical barrie r
against activity release to the environment .
Therefore, the leakage of the confinement is strictl y
limited, which is one of the fundamental conditions o f
plant operation .
To assess the integrity of the confinement, the utility ha s
to determine the leak rate of the confinement, even there
has been a locally uncontrollable hole opened on the liner .
Outage for maintenance and refuelling is a typical case .

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2. The construction and structure of confinement

Net volume of confinement is 50 000 ÷ 54 000 m3 .
Lower stage is –6 .5 m-, higher stage is +40 .5 m.
Carbon steel liner is on the inner side of confinement’s rooms ,
and outer side of reactor compartment walls .
Steel liner’s thickness is 8 mm.
Thickness of concrete is 800÷1500 mm .

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Pressure transient at a LB LOCA DBA at VVER-440 confinement.
A [16 . s] - HPSI started
B [30 . s] - Water spill
back from the tray s
started
C [48 . s] - Water spill
back from the tray s
finished (90 % of water
volume spilled back)
D [86 . s] - spray injectio n
started

Containment pressure
(492 mm - double ended cold leg

50

100

150 200

250

300

Time [s]

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Confinement
inent

n building

i with the

89

bl Bubble Condenser

NEA/CSNI/R(2002)7/VOL1

3-D modell of confinemet

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Comfinement’s 3-D modell

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3. Methods of examination of leakage and strength test
We had carried out a leakage test in full designed presssure 250 kPa - at first start-up phase. We had measured the leakage
of confinement at three gaugh level ;
120, 170 and 250 kPa .
We had measured deformation of walls at same time .
Repeted leakage tests are achieved on 120 kPa .
Leakage rate is converted to 24 hours and 250 kPa .
We had managed a leakage test at 170 kPa on each unit a t
1994÷1997 .

We use the two point method, while pressure drops .

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Leakage rate of NPP units 1÷4.

1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1994 1995 1995 1996 1996 1997 1997 1998 1999 2000 2001 200 1

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5. Repairing and checking defects found during examination s
Defects of kiner
- welding --> local test
- injection plastic material --> integrated leakage tes t
Defects of hermetical closure s
- exchanging gasket material --> local tes t
- exchange sealing equipments --> integrated leakage tes t
- making new gasket construction --> integrated leakage test
- resealing of door’honges --> integrated leakage tes t

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6. Short review of repair of the liner and concret e
structure at the Unit 1 . of NPP Paks
We have made a program improve hermetization of unit 1 .
from 1998 co-operating with company V UEZ a.s . Slovakia.
Mean activities :
- searching defaults during ILRT and depression tes t
- injecting the space between liner and concret e
- measuring alteration of pressure gaugh under liner while
carried out an ILRT
- decomposition of concrete to liner at blowdown air corrido r
and injecting
- decompositioning of concrete at a hermetic roo m
explorating a connection to liner and repairing i t

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Injektion point No 1 ./19 deconstructed concrete in A201/ 1

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NEA/CSNI/R(2002)7/VOL 1

Injektion point No 1 ./19 with injection cap

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Injektion point No 1 ./19 with local control cap

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NEA/CSNI/R(2002)7/VOL 1

Excaveted connection to liner in room A306/ 1

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Excaveted connection to liner in room A306/ 1

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NEA/CSNI/R(2002)7/VOL 1

Repaired connection in room A306/ 1

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NEA/CSNI/R(2002)7/VOL 1

Excaveted connection in room A306/ 1

102

NEA/CSNI/R(2002)7/VOL 1

Repaired connection in room A306/1

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Hole in structural concrete discovered by drilling for injektio n

104

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Inside of hole in structural concrete
discovered by drilling for injektio n

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Checking hole No 1 .for borical leakage and concrete at unit 2

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Checking hole No2 . for borical leakage and concrete at unit 2

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Checking hole No3 . for borical leakage and concrete at unit 2

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Checking hole No1 . for borical leakage and concrete at unit 3

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Checking hole No2 . for borical leakage and concrete at unit 3

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Checking hole No 3 . - unit 3

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Other activities to repair estate of concrete :
- gap of concrete between the spent fuel storage pool an d
No 1 . poo l
- destructive and chemical disquisition of concrete sample s
- model supported mechanical and chemical experiments have
been performed
- long term program to check the status of concrete structur e
- evaluation of concrete structure for lifetime extensio n

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7. Conclusions
7.1 . Notable degradation of concrete hardness has not bee n

found.
7.2. Surface corrosion has been found on the metal structure i n
the concrete
7.3 . The confinement leak rate is within the limits of Techn.
Spec.
7.4. Leak tightness enhancement program on the unit 1
confinement is in progress, significant results are achieved .

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SESSION A: OPERATIONAL EXPERIENCE (continued )
Chairman : Dr . James Costello, USNRC (USA )

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Steam Generator Replacement At Ringhals 3
Containment, Transport Opening
Jan Gustavsson, Ringhals Nuclear Power Plant, Swede n

Abstracts
At the steam generator replacement at Ringhals 3 1995 an opening 6 x 8 m was taken in the cylindrical
containment wall about 12 m over the ground level. The wall is build up of an outer prestressed concret e
wall, steel liner and an inner concrete wall . The prestressing reinforcement in the outer concrete wal l
consists of horizontal and vertical tendons . Each tendon was prestressed to about 5000 kN from the
beginning . The tendon ducts were filled with grease.
Before the opening was taken a temporary wall with a gate was build on the inside of the containment wall .
When starting making the opening, the tendons within the opening were de-tensioned and pulled out . A
additional number of horizontal tendons under and over the opening were de-tensioned . The opening was
cut through drilling 300 mm holes along the side of the opening into the steel plate from both outside an d
inside . Then the steel plate was cut and the wall plug removed .
During the transport of the steam generators the work with the restoring of the wall started . The concrete
surface was prepared and reinforcements bars were drilled into the old concrete . When the transport wa s
finished the steel plate was restored, tendon ducts joined and the reinforcement put in place. To prevent
early cracking in the concrete tubes for cooling water were mounted in the form . The inner form was
casted at first and then the outer one .
After that the concrete has reached sufficient strength the tendons were mounted in the ducts and tensione d
to the same force as before de-tensioning. After the tensioning the ducts were filled up with grease again .
A special inspection program was performed to see if there were any degradation in the concrete, the
tendons or the steel plate . The result was that we could see no degradation on the steel plate or the tendons .
The measured tendon forces are higher than calculated. The strength of the concrete had reached a mediu m
level of 91 MPa (Original level 50 MPa) . The carbonisation had reached 8 – 10 mm into the concrete . Our
conclusions are that we have not seen any degradation which is a threat to the containment in th e
foreseeable time.

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1.

Introductio n

1.1 Backgroun d
In 1992 Ringhals decided to change steam generators at Ringhals 3. The degradation of the tubes had gone
far and the capacity was reduced to 88 % since 1988 . A steam generator replacement had been done a t
Ringhals 2 during 1989 and the experiences from the replacement were good .
The planing of the replacement started during 1992, and KWU was chosen to manufacture the stea m
generators . For the replacement at Ringhals a consortium consisting of Siemens and Framatome wa s
engaged. The civil works were carried out of a subcontractor, NCC, to the consortium . NCC is one of the
biggest contractors in the Swedish building industry . The replacement work started in the beginning o f
June 1995 and was finished within 90 days .
1.2 Transport procedure
The steam generators were transported to Ringhals by boat to the harbour at Videbergshamn near Ringhals
Site . From the harbour the steam generators were transported by a special vehicle to a storing place withi n
the Ringhals area . Some preparation works were done on the steam generators . The replacement procedure
started with making the opening. As soon as the opening were ready the old steam generators were taken
out from the containment one at a time. The steam generator was lifted up from its position by the polar
crane and laid down on trolleys on girders at level +115 . Then the steam generator was transported throug h
the containment opening and lifted down to ground level by a lifting device . The vehicle then took the
steam generator to a storing place . When the three old steam generators were taken out it was time for the
new ones to be transported into the containment. The transport procedure was the same in the opposit e
direction.
1.3 Description of the wall
The cylindrical containment wall consists of one outer concrete wall, a steel liner and an inner concret e
wall. The inner diameter of the containment is 35 .4 m and the height is 52 m .
The inner concrete wall is 0 .33 m thick concrete reinforced with a grid of 016 mm rebars distance 200 m m
in the inside. The compressive strength of the concrete in the wall is K50 .
The steel liner is 7 mm thick . On the outside of each joint between plates in the steel liner there is U-profil e
welded in order to make it possible to see if there is any leakage through the joint. The canal inside the Uprofile is connected to each other in special sections . From each section there is a pipe drawn to a galler y
where it is possible to test each section . To the steel liner there are profiles welded to which the tendon
ducts are fixed .
The outer concrete wall is 0 .77 m thick concrete reinforced with a grid of 0 25 mm rebars distance 20 0
mm in the outside . The covering concrete layer is 50 mm. The compressive strength of the concrete is K50 .
About 0 .3 m in the wall there are horizontal tendons and about 0 .5 m in the wall there are vertical tendons .
The distance between the vertical tendons is 0 .75 m and the average distance between the horizonta l
tendons 0.4 m.
Both vertical and horizontal tendons consists of 139 06 mm wires . The tension load is just below 5 MN.
Each horizontal tendon stretches half a turn round the containment. The total number of horizontal tendons
is 245 and vertical tendons 153 .

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1.4 Preparation work
In order to prepare the staff that were going to do the civil some practice was done on the wall block from
Ringhals 2 . The containment wall in Ringhals 2 has the same construction as it has in Ringhals 3 . On the
wall block the staff trained seem drilling and other work procedures in the same manner as it was planne d
to be done at Ringhals 3 .
2.

The transfer opening

2.1 Protective wall inside containmen t
The work inside the containment started as soon as the plant was shut down .
After erecting scaffolding and cleaning the inside of the containment wall and the floor next to the transfe r
opening, a protective wall was installed . The floor between the protective wall and the containment wal l
was covered with steel plates . The protective wall had a steel structure and covered with steel panels . There
also was a ceiling of steel panels built between the protective wall and the containment wall .
In the protective wall there was a MEGA-door installed through which the steam generators were
transported.
2.2 Detensioning and removal of the tendon s
Within the opening there were nine vertical tendons and 20 horizontal tendons which were detensioned an d
dismounted. Another 56 horizontal tendons were detensioned . Theese tendons are situated below, abov e
and on the other side of the containment than the opening . The work with the tendons started as soon as th e
plant was shut down .
The detensioning work started with erecting of scaffolding outside the containment at the pilasters . Also
for preparation the jacks and x/y-writers were calibrated .
For the tendons the grease cans were dismounted and the anchor parts were cleaned from grease . The jack
was coupled to tendon top end for the vertical tendons and at both ends for the horizontal . The present
tension load was checked. Then the tendon was tensioned to a load of 5 MN and the extension recorded.
The detensioning of the tendon was done in two steps and the contraction recorded . The jack was removed.
Then the tendons were dismounted through special procedures, one for the vertical tendon and one for the
horizontal . The tendons were cleaned from unnecessary grease, inspected and winded up on a hydrauli c
winch .
The tendons and grease cans were stored indoors during steam generator replacement .
2.3

Cutting of the opening

The cutting of the opening 8 m x 6 .6 m started as soon as the tendons in one side of the opening wa s
removed . The first operation was to drill holes with diameter 0 300 c 200 from both outside and inside a t
the same time . Totally six drilling units worked parallel . The holes were drilled almost into the steel liner .
The concrete near the steel liner was chipped away . In the bottom of the opening a hydraulic jack wer e
mounted and special holes were made in the corners for the sliding beams which should be used fo r

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removal of the wall block. Some drilling for rebars and bolts were done for the supports for the slidin g
beams and to fix the inner concrete plug to the outer .
When the drilling was done and the fuel elements were removed from containment the steel liner was cut .
In the bottom corners of the block holes were cut in order to make it possible to place the sliding beams . A
sleigh was mounted on the sliding beams and the wall block was lowered down on the sleigh . And before
the last pieces of the liner was cut the wall block was secured to the sleigh .
The block which weighed about 125 tons was drawn out from the containment wall, then lifted down to th e
ground level and transported to a storage area .
2.4 Preparing surfaces
After removal of the wall block the work with preparing the concrete surfaces took place . Shear recesse s
were cut in the concrete surfaces in order to transfer shear forces in the joint . In the surface at the top of the
opening there were ventilation channels cut out to make it possible to fill up the form with concret e
properly .
The tendon ducts were cut at the level of the concrete surface and the new tendon ducts were joint wit h
inserts.
The concrete around the test channels for the steel liner was cut out .
To replace the rebars that were cut, new rebars were drilled in the concrete near the outside respectively th e
inside . A new layer of rebars was installed between the vertical and the horizontal tendons . The drilled
holes were up to 2500 mm deep . Rebars with joint nuts were grouted in the drilled holes according to a
special procedure .
This work were done during the period when the transportation of the steam generators went on throug h
the opening .
2.5

Restoring of the steel plat e

The work with preparing the edge of the steel liner went on during the period of transportation . A new
prefabricated steel liner was placed in the opening. The new steel liner was connected to the old steel liner
with two small steel plates one on the inside and one on the outside . The channel between the small stee l
plates was connected to the test channels for the old steel liner. The test channels that were cut off, when
drilling the opening, were restored . Attachments for the tendon ducts had been welded to the steel liner at
the prefabrication .
2 .6 Restoring the inner part of the containment wal l
The first work sequence to be done was to mount all the connecting rebars . Then the vertical and the
horizontal rebars up to 1,5 m were mounted . The first part of the formwork was mounted and the first par t
of the casting could start up to 10 cm below the edge of the formwork . The working procedure wa s
repeated with mounting the horizontal rebar, mounting next part of the formwork and casting for each 8 0
cm.

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In the last section there were some special arrangements done to secure that the opening was properly fille d
with concrete. A pump was connected to the formwork and the concrete was pumped into the formwork
until it came out in the ventilation opening at the top of the opening . The formwork was made of stee l
plates and left after the work was finished .
2.7

Restoring the outer part of the containment wall

In order to secure a low temperature in the concrete during the hardening pipes for cooling water wer e
mounted at three depths in the concrete . The first row of cooling pipes was mounted close to the steel liner ,
the second at approximately the same depth as the vertical tendons and the third just outside the horizonta l
tendons .
The work procedure had the following parts :
Mounting the first row of cooling pipes .
Mounting vertical tendons ducts to the attachments and connecting with the existing ends of th e
tendons ducts .
Connecting and mounting the first layer of vertical rebars .
Connecting and mounting the first layer of horizontal rebars .
Mounting the second row of cooling pipes .
Mounting horizontal tendons ducts to the attachments and connecting with the existing ends of th e
tendons ducts .
Mounting the third row of cooling pipes .
Connecting all the vertical rebars and horizontal up to 1,5 m in the second layer .
Mounting all the vertical rebars and the horizontal up to 1,5 m .
The first section of the formwork was erected .
After connecting the cooling system to the cooling pipes the casting took place up to 10 cm below th e
top of the formwork .
The work procedure with connecting and mounting horizontal rebars, erecting formwork and castin g
was repeated each 80 cm .
To the last section of the formwork a pump was connected and the concrete was pumped into th e
formwork until it came out in the ventilation opening at the top of the opening .
The concrete quality used was K50 with the temperature 5 °C. In order to prevent cracks from high
temperature during the hardening the concrete was cooled . The incoming temperature of the cooling wate r
was to 2 °C. The temperature of the cooling water coming out of the pipes was registered and compare d
with calculated values . The supervision of the temperature went on for some time after that the casting wa s
ready and the temperature had decreased to a normal level .
2.8 Tensioning of the tendons
The work with the mounting and tensioning of the tendons started about 12 hours after the casting wa s
finished. The first moment was to insert the tendons that are crossing the opening in the ducts . When the
tendons had got through the duct the anchors were thread on each end and a button head was made at eac h
wire.

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Before the tensioning work started the jacks and x/y-writers were calibrated . When the compressive
strength in the concrete in the opening had reached 36 MPa the tensioning of the tendons outside th e
opening could start . To start with the tensioning of the tendons crossing the opening the compressiv e
strength in the concrete must have reached 40 MPa.
The tensioning was done according to a specified procedure to the specified load and the same shims wer e
placed at both ends as before detensioning. Both the load and the elongation was recorded . When all the
tendon were tensioned the grease cans were mounted the cans and tendon ducts were injected with grease .
One incident happened when the horizontal tendons were inserted in the tendon ducts . One of the duct s
was damaged so that the plate in the duct was pressed to a stop in the duct . It wasn’t possible to insert the
tendon . The place, where the stop occurred, was located . It was situated outside the opening . It wa s
necessary to cut the concrete away to uncover the stop. It was done with water jetting . The stop was found
and the duct repaired and the concrete restored. The tendon was then inserted and tensioned .
2.9 Removing of temporary construction s
A couple of days after the casting of the concrete the removal of the temporary wall on the inside begun .
The area was cleaned and the material was transported out of the containment .
When the tensioning of the tendons was finished the scaffolding was removed .
3.

Inspection program

3.1 Contractors control program
In the scope that the contractor had there was prescribed a control program for the civil work that include d
normal control on concrete works, the drilling, cutting, welding and prefabricating the steel liner and th e
tendon works . A special control plan was drawn up for this purpose .
3.2 Status of the containment wall
At some incidents that had happened at other plants before, questions about the status of the containmen t
wall had been raised . In order to answer some of these questions a control program was established . The
main parts in the program were :
Visual inspection of all uncovered concrete surfaces to look for cracks, cavities from the casting an d
other abnormalities .



Visual inspection of uncovered rebars to look for corrosion and how the concrete has enclosed the
rebars .
Visual inspection of the steel liner to look for corrosion .
Take out concrete cores to examine the compressive strength, the carbonation deep and the deep of th e
chloride penetration.



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3.3

Results

One observation made, was a crack between the steel liner and the concrete at the inside of the steel liner .
This phenomenon is known before from Ringhals 2 at the steam generator replacement . The crack width
was approximately 1 – 2 mm and according to a special investigation the crack will remain in the area next
to the opening but disappear for the rest of the containment wall at the tensioning of the tendons .
There also was one small crack in the outer wall at the top of the opening . The crack width was less than 1
mm and was probably caused by shrinkage of the concrete . The crack closed probably at the tensioning o f
the tendons . No other signs of degradation of the concrete were discovered .
There were no signs of corrosion on the rebars and the concrete had enclosed the rebars quite good .
The uncovered parts of the steel liner showed no sign of corrosion . The steel liner surface had a thin layer
of cement mortar .
The compressive strength was 91 MPa as an average value . The origin compressive strength was set to 5 0
MPa and samples taken by the casting showed values of 60 - 70 MPa after 91 days. There is an increase of
the compressive strength with increasing age of the construction .
A normal value of the carbonation was 8 – 10 mm . It is about 20 years since the containment wall was
poured. Our estimate is that when the construction reach 100 years of age the carbonation have reached a
depth of about 20 mm in the wall .
The result from the test of chloride gave that the concentration of chlorides is very low . The highest value
was 0,2% chloride of the cement weight at the surface and less than 0,1% in depths of 3 – 6 cm .
The conclusion of the inspections and tests mentioned above are that the containment wall is in goo d
condition and there are nothing that indicates a degradation of the containment .

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In Service Inspection Programme and Long-term Monitoring o f
Temelin NPP Containment Structure s
Jan Malÿ, Jan gtepân - Energoprojekt Praha a.s.
Czech Republic, Prague
Abstract
The paper describes monitoring systems of Temelin containment concrete structures and pre-stressing
systems. Reliable information on the actual state of containment structure is necessary for conditio n
assessment as well as for detection of local defects and effects of ageing . In service inspection programme
as a main tool of preventing structural defects and damages will be discussed . Results of Temeli n
containment inspections and tests will be presented .
Containment structure
The Temelin NPP is formed by two VVER 1000 MW units . At present the start-up of the 1st unit at 100%
power output is underway, for the 2nd unit fuel has been loaded and the physical start-up process i s
underway. The units of the VVER 1000 MW type have a PWR (pressurised water reactor) reactor and a
containment of pre-stressed reinforced concrete . Typical cross section of the reactor building is shown i n
Fig. 1 .
The containment consists of a cylindrical and a dome part . Connection between the cylindrical and dom e
parts is made with the help of a rigid ring beam in which the anchoring blocks of pre-stressing cables ar e
placed . The wall thickness of cylindrical shell is 1 .2m, the dome wall thickness is 1 .1m. The containment
structure is placed on the reinforced concrete slab at a level of +13 .20m, thickness being 2 .4m. This slab
contains also supporting blocks of pre-stressing cables of the cylinder which are built-in there . The
containment is made of concrete, grade B40 according to the Czech standards (CSN) . The tightness of the
containment is ensured by the steel liner of a thickness of 8 mm made of carbon steel .
A chart of the pre-stressing method is shown in Fig . 2 . The cylindrical part of the containment is prestressed by 96 cables running in helical direction . The cable anchors are installed in the upper part of th e
ring beam, the bending of the cables takes place in the slab at a level of +13 .20m . The dome part of the
containment is pre-stressed by an orthogonal grid plan of pre-stressing cables formed by 36 cables . Alway s
two cables are conducted against each other, anchors of one cable and bending of the other one are situate d
on one side. The anchoring blocks are installed from the ring beam side . The cables of the cylinder and
dome parts are of the same structure and cross section. Cable preservation was made with grease durin g
production, preservation of anchors was made after pre-stressing . The anchors are protected from climatic
effects by means of the sheet covers installed .
The pre-stressing unbonded cables are conducted in polyethylene tubes . Every cable is formed by 45 0
wires featuring a diameter of 5 mm. Low-relaxation wire was used for production, its yield point being
1620 MPa . The initial pre-stressing force according to the design is 10 MN . On the basis of experienc e
acquired during the pre-stressing of the 1st unit the anchor details were modified . These modifications have
ensured better arrangement of wires on the anchor, and thus also a more even distribution of the prestressing force into the individual wires .

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The containment function was verified during the structure integrity test (SIT) . The test carried out on both
the 1st and 2nd units was implemented as a combined test . The function was tested for strength at a n
overpressure of 460 kPa, the function was tested for tightness at an overpressure of 400 kPa . The SIT o f
the 1st unit was carried out in 1998, the SIT of the 2 nd unit was carried out in 2000 .
Overview of inspection activities during operatio n
For the purpose of ensuring full performance of the containment for the entire period of operation of th e
unit there was created an inspection programme of the containment structure . The inspection of the
containment structure consists of the assessment of measurement of sensors of the permanently installe d
measurement systems and of the inspection of the conditions of concrete, liner and pre-stressing system .
According to the frequency of the work carried out the inspections divide the work into two phases . During
the first phase (the first 4 years of operation) the activities are carried out in full extent . Within the
framework of the following phase a part of activities is carried out on an annual basis, and another par t
once in four years .
The inspection of the containment concrete consists of the following activities :
inspection of the containment surface – to be carried out twice a year . Focused on checking for
damage, corrosion of the reinforcement system and crack development .
non-destructive concrete strength tests – carried out once a year, in the second phase once in fou r
years.
Liner checks are carried out always when it is possible to enter into the containment . The inspection
consists of the following activities :
inspection of the coating for integrity and of the liner for damage .
non-destructive measurement of liner thickness .
check for tightness – carried out within the periodical test framework .
The checks of the pre-stressing system is carried out in the first phase of the inspection work once a year ,
in the second phase once in four years . The inspection consists of the following activities :
inspection for humidity at the place of anchors and bends .
inspection for integration of preservation at the place of anchors and bends and change in chemical
properties of grease .
Inspection of wires and anchors for damage .
checks of the pre-stressing force by lift up tests .
The above listed activities are completed with the inspection of the cable removed . The dismantling of the
cable will be made three times during operation, there will always be removed two cables of the cylinde r
and one cable of the dome. The inspection is focused on the condition of preservation, level of corrosio n
and damage to individual wires and mechanical properties of wires are verified on selected samples .

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A new methodology has been developed for the lift up test . This methodology makes it possible to specify
exactly the pre-stressing forces in the anchor . during the test the pressure in the press is recorded, as well a s
the force measured by Hottinger sensors and anchor lift from the supporting block (measured by the
displacement sensors temporarily installed) . This monitoring enables exact specification of the moment o f
anchor relief and thereby also the exact determination of the pre-stressing force in the cable .
Assessment measuremen t
In order to enable the inspection of the level of the containment pre-stressing, measurement systems ar e
installed permanently on the structure, and these systems measure structure deformations and pre-stressing
force in the cables . The measurement is carried out once a month, once a year the setting of the reacto r
building is measured . Inspection of the values measured is carried out at each measurement, a complet e
assessment is made once a year .
The following measurement systems are installed on the containment :
NDS and SDM systems – these two systems consist of vibrating wire fitted during concret e
pouring into the containment walls . The sensors are of four types and measure concrete
deformation, temperature and horizontal shift in the middle of the height of the cylindrical part o f
the containment . The containment includes more than 240 sensors which are installed in it (246 on
the 1st unit and 256 on the 2nd block).
Hottinger system – this system is formed by strain-wire gauges stuck on the anchors of all cable s
of the cylinder and of the dome, i .e. 264 anchors measured. The sensors measure force in th e
anchor of the pre-stressing cables .
MEM system – the system is formed by the sensors installed on the conduits of the pre-stressin g
cables . These sensors measure force in the cables by means of the magneto-elastic method . The
sensors are placed on two cables of the cylinder and of the dome . The sensors on the dome cable s
are placed under the anchor and the cable bend . On cylinder cables they are placed under th e
1st
anchor and at the middle of the cylinder height, on the unit the sensors are installed in the lowe r
part of the cylinder as well .
HYNI system – the system measures the settlement of the reactor building by means of hydrostati c
level control . Measurement is carried out with regard to the criterion of the reactor inclination .
Since the reactor building is founded on a rocky bedding, the settlement was at a minimum leve l
and there are virtually no changes anymore .
The distribution of sensors on the containment structure is illustrated in Fig . 3. At present it is the concrete
creep that has the largest impact on the change in the state of stress and containment deformation . With
regard to the age of the structure, the effects of the pre-stressing cable creep and of concrete shrinkage is
minimum .
As an example of the values measured by vibrating wire of the NDS system, Fig . 4 illustrates the
development of proportional deformations in the central part of the dome of the containment of the 2 nd unit.
From the graph it is possible to see the deformation in course of pre-stressing, and subsequentl y
deformation due to concrete creep . The values measured also reflect the effects on the values measured in
the case changes in outdoor temperatures during the year . If measurement is made in a period of several
hours, the effects of outdoor temperature oscillation during the day has similar features . For the purpose o f
comparison, Fig . 5 shows the time history of temperatures in the dome during the same period.

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The example of the pre-stressing force course on the cable is shown in Fig . 6. The graph states the value s
of pre-stressing force measured with the MEM system sensors on the cylinder of the 1st unit, cable no . 21a.
From the graph it is possible to see the decrease in the pre-stressing force along the cable length due t o
friction, as well as decrease in the pre-stressing force over time as a result of concrete creep .
The comparison of measurement Hottinger, MEM and lift up tests is in Fig . 7 and 8 . The graphs provide
courses of regression of the average value of the pre-stressing force specified by the systems Hottinger and
MEM (for sensors under the cable anchor) on the cylinder and on the dome of the 2 nd unit. The graphs also
state average values of pre-stressing force determined by the lift up test after 2000 hours from the prestressing and before the SIT .
Conclusio n
The methodology of the in-service containment inspection presented in this paper enables us to inspect an d
monitor the containment of the 1st and 2nd units of the Temelin NPP, and provides a guarantee that the
containment structure is able to perform the function of the last safety barrier .

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Fig . 1 Cross section of the reactor buildin g

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Fig . 2 Scheme of pre-stressing of containment

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5 ensors of system NDS and 5 DM
in the wall of containment

5 ensors of system Hottinger and ME M
on the cables of containmen t

Fig. 3 Distribution of sensors on the containment structure
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~
y
I

N

N

Co

Co

CD

CD

N

Fig. 4 Proportional deformations in the central part of the dome of the containment of the 2nd unit measured
by string strain gauges of the NDS system, sensor type PSAS .

Average of int . sensors

Average of ext . sensors

Fig. 5 The time history of temperatures in the dome during the same period as in Fig . 4.

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NEA/CSNI/R(2002)7/VOL 1

Fig. 6 The example of the pre-stressing force measured with the MEM system sensors - the cylinder of th e
1 st unit, cable no . 21 a

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10

9.5

z
d
v
ô

9

LL

8 .5

8
01 .07.1996

01 .07.1997

01 .07 .1998

01 .07 .1999

01 .07.2000

01 .07.2001

01 .07 .200 2

7LPH

Regresion of Hottinger cyl . -Re gresion of MEM cyl . --O--Lift-up cyl .

Fig. 7 The comparison of measurement Hottinger, MEM and lift up tests - the cylinder of the 2nd uni t

Regresion of Hottiger dome -Re gres ion of ME M donne t Lift-ap dom e

Fig. 8 The comparison of measurement Hottinger, MEM and lift up tests - the dome of the 2nd unit

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REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETE STRUCTURES O F
NUCLEAR POWER PLANTS
PARTICULAR APPLICATION ON NATURAL DRAUGHT COOLING TOWERS IN BELGIUM
R. Lasudry, Principal Enginee r
BELGATOM
GTEC Department
Civil Works Branch
Avenue Ariane 7, B1200 Brussel s
Tel : 32 2 773 81 3 4
Fax : 32 2 773 89 7 0
e-mail : roland.lasudry@tractebel .be
Abstract
A previous paper was presented at the OCDE Workshop held 22 – 23 March 2000 in Brussels explaining
different aspects of the techniques used for “Instrumentation and monitoring of natural draught coolin g
towers in Belgium” .
These monitoring and preventive techniques are now applied in Belgium since already more then 10 year s
by Tractebel on the towers of the nuclear plants .
These huge constructions have to sustain considerable physical, chemical and biological loads . As one can
figure out, and as years go by, these inspections showed deterioration of which type, progress, quantit y
eventually led to the need of repairing these structures .
The present paper goes over 4 main different sorts of defects (beam supports breaking, fast carbonatio n
rate, concrete porosity, and a series of local deteriorations like insufficient concrete cover, cracking, gravel
pockets, corroded reinforcement) encountered on 3 cooling towers situated in Belgium, and affecting th e
shell as well as the inner structures .
The diagnosis, the choice of the appropriate repair techniques and products which will avoid having to fac e
much higher costs in the future are explained .
It also gives an illustration of the works carried on site and points out the uncommon and complex aspect s
the treatment of such a construction implies (planning, both horizontal and vertical curved shape, works a t
great height, …) .

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1.

Introductio n

A cooling tower, whether at a fossil-fired or a nuclear power plant, where the circulation water is coole d
after having left the condenser, is a main component as it provides the cold source in the thermodynami c
cycle of the turbine.
On account of the water flow rate to be treated and the volume of air required, the cooling towers, built of
reinforced concrete, are structures of hyperboloid revolution that are very impressive by their size an d
shape (up to 160 m high for the large installations) . The developed surface area of the thin concrete shell o f
these towers may reach several tens of thousands of square metres . By their nature, the cooling towers have
to sustain considerable physical, chemical and biological loads .
The inner structure is a construction of beams and columns designed in order to provide the support of the
air/water exchange material allowing this construction to perform its cooling function .
While for a long time concrete constructions were considered capable of defying the years withou t
problems, this simplistic approach has since been abandoned, as it is now recognised that, like othe r
materials, concrete is affected by ageing and by various illnesses that need to be treated and kept under
control if the concrete is to reach its optimal operational life . These constructions were generally ordere d
by the operators as a turnkey component and, as such, were not verified (calculations and works) by th e
engineering consultants as Tractebel .
Moreover, in the seventies, several cooling towers even collapsed, like in Ferrybridge (England) o r
Bouchain (France) reminding all structural and electrical engineers that a major failure of this essential
power plant component was always possible .
2.

Technical approach

2.1

General description and diagnosi s

The concrete of a cooling tower, exposed to the aggressions of its environment and to the operatin g
conditions, presents also a number of characteristics that are specific to it as compared to other structures.
Therefore, investigations are first required to identify the causes of the deteriorations, as the causes wil l
dictate the methods and materials required for a successful restoration .
The origins of the deteriorations may be chemical, physical or biological : wind action, the structure’s own
gravity, inside/outside temperature gradient, insulation differential, rain action, air pollution, soil-structur e
interaction, water vapour and proliferation of algae or mosses on the shell leading to cracking, deficiencie s
inherent to the concrete such as the presence of gravel nests, chipping and spalling, deteriorations caused
by carbonation, alkali-silica reaction or by chlorides .
In the case of a cooling tower, all these deteriorations make fragile a structure which is constantl y
surrounded by an atmosphere saturated with water vapour and which is exposed to thermal differentia l
stresses the major part of its existence .
These environmental conditions lead to the slow deteriorations of the tower’s components : reinforcement
corrosion, concrete leaching .

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2.2

Treatment guidelines

Accordingly, treating a component such as the shell of a cooling tower requires many investigations, cor e
sampling at significant locations, laboratory analyses, selection and validation of particular treatmen t
processes and procedures that are suitable, considering also the access constraints and the size of th e
surfaces needing treatment. In the case of serious degradation of the concrete performances, a ne w
calculation of the shell might need to be performed that could lead to strengthening measures bein g
prescribed .
In general and if the chemical deteriorations are limited, treatment involves repair of the physica l
deteriorations, followed by preventive treatment of the concrete against aggressive elements .
Repair of the physical deteriorations comprises the treatment of the reinforcement and the replacement o f
the damaged concrete so as to return the tower as close as possible to its initial condition .
This first action is then followed by surface treatment . Taking into account the geometrical characteristic s
and the deformation of the shell, this treatment usually involves applying to the inner face, which is the on e
exposed to the water vapour, a treatment more impervious than the one applied to the outer face . As a
result, the inner face is sealed off against ingress of water into the concrete, and the concrete is allowed to
"breathe" on the outside . Also, as the two faces are treated, penetration of carbon gases and othe r
aggressive elements is prevented and chemical reactions can be stopped .
In practice the review of the way the treatment is to be implemented is completed by a comparativ e
appraisal of the methods of accessing any point of the inside and outside of the shell . Complex problem s
are taken into consideration in this review, relating to the height of the structure, its both horizontal an d
vertical curved shape and the necessity of supplying to any point the liquids and products under require d
pressure for the treatment . For instance, high pressure water-jet cleaning requires water at 400 bar (or even
more) to be provided at great height. Also, all the means of access have to present sufficient safet y
regarding operators and products, while making possible the rapid progress and quality of the work .
Finally, as the works are only performed during a stop of the tower, usually meaning lowering or stopping
the production, these equipments are to be suited (in number, speed and ease of use) with the shortes t
possible working schedule.
2.3

Encountered problems

The 3 first examples concern degradation affecting the shell of 3 power plant cooling towers (Ruien 5/6 ,
Doel 3 and Tihange 3) .
The last example concerns problems affecting the inner structure of the Tihange 3 cooling tower .

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3.

Ruien power plant units 5/6

3.1

Main features

Cross-flow type cooling tower built in the years 1972-1973 :
- diameter at the ground level ( cold water basin not included)
- diameter at the lintel
- diameter at the throat
- diameter at the top
- height at the lintel
- height at the throat
- total height
- shell thickness at the lintel (varying up to 28 m)
- shell thickness from elev .28 m to the top
The reinforcement is made of only one centred reinforcement layer .

: 60.90 m
: 54.30 m
: 41 .60 m
: 48 .50 m
: 16.15 m
: 62.58 m
: 94.65 m
: 0.70 m
: 0 .12 m

As significant vertical cracks appeared after a few years, the builder placed at his own costs under the 10 year guarantee, an additional external reinforcement to avoid a potential collapse of the structure . This
involved placing 42 cables T13 (each tensioned with 30 kN) every 0 .80 m from the throat to the top .
3.2

Problem description - diagnosi s

The cooling tower of the Ruien power plant was found to be in the following condition :

shell: cracking (partly passing through), local deteriorations, gravel nests (some going right throug h
the shell), corroded reinforcement, uneven concrete surface, though fairly good concrete quality i n
the sound parts of the shell, carbonation depth < 15 mm (pict . 1) .
shell supports : cracking and deteriorations under chemical attack .
The nature, the quantity and the size of the shell cracks were such that the initially monolithic structure ha d
now to be considered as an articulated one .
Also, the concrete needed treatment to stop leaching and protect the reinforcement .
3.3.

Treatment definitio n

In a first phase, core samples were taken from the shell in order to identify the physical and chemica l
properties of the concrete, so that intended treatments could be assessed and their chances of succes s
evaluated.
After this, the phases of the treatment were defined :
complete cleaning of the inner and outer surfaces (pressure jetting with water and/or sandblasting )
(pict. 2) ;
removal of all the portions of poor quality concrete (in and out) ;
treatment of reinforcement (in and out) : cleaning (all around the bar), replacement when necessary
and protection against corrosion with an hydraulic-based product ;

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replacement of the removed concrete with a new hydraulic cement-based compound, filling of th e
grid nests and correcting of the unevenness with the same product (in and out) ;
complete shaping (inside only) with a fine polymer improved hydraulic rendering cement (6 kg/m2) ;
inner face protection : application of impervious coats that would remain flexible enough to follo w
structural deformation . On an impregnation epoxy resin coat, a polyurethane-based multi-laye r
flexible system (2 mm) was laid, protected afterwards with a final coat (40 µm) providing the UV
radiation protection ;
outer face protection: application of protective coats witch allow the concrete to breathe and whic h
are not sensitive to structural deformation . A hydraulic elastic polymer mortar was applied in tw o
phases of 1,5 kg/m2 and 3 kg/m2, the first layer applied with a trowel, the second gun-applied .
These coats provide the concrete with a barrier against attack by various substances . Moreover, the
application of cement-based products brings a complementary quantity of "fresh" lime that contributes t o
stabilize and partly restore the alkaline properties of the existing concrete, improving thus the protection o f
the reinforcement and restoring the intrinsic strength of the concrete .
In a second phase, tests were performed on selected areas in order to determine the corrosion level of the
reinforcement and validate the product application methods that proved the most effective .
3.4.

Works

A particular problem was faced with this cooling tower due to the pre-cast cables being located outside th e
upper half of the concrete shell . This had to be taken into account when carrying out the various operation s
(movement of hung scaffolding, pressure cleaning, application of the coating products) (pict . 3).
Concerning the shell supports, the bases of the columns were injected with a sizing epoxy-based resin .
However the small opening of the cracks of the columns did not make their injection possible . The
columns have been treated with the same product as the inner face, however in two layers only .
The works have been carried out under cover of a quality assurance program and controlled by the expert s
of an insurance company allowing it to be covered by a 10-year full warranty conditional to regula r
inspections .
Representing the treatment of 12,000 m 2 per face plus the supporting columns, they have been carried out
within a 3 months period (of which only one month of real working days due to bad weather conditions )
for the inner face and about one year for the outer face (with an average working ratio of 45%, winter sto p
included) .
4.

DOEL 3 Power plant

4.1.

Main features

Counter-flow type cooling tower build in the years 1981-1983 :
- diameter at the ground level ( cold water basin not included)
- diameter at the lintel
- diameter at the throat
- diameter at the top

139

:
:
:
:

141 .70 m
133 .88 m
76 .64 m
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NEA/CSNI/R(2002)7/VOL 1
- height at the lintel
- height at the throat
- total height
- shell thickness at the lintel (varying along the first 10 m)
- shell thickness from elev . 10 m to 104 m
- shell thickness from elev . 104 m to the top

: 12.70 m
: 108 .70 m
: 167 .28 m
: 0.85 to 0 .25 m
: 0.25 to 0 .18 m
: 0.18 m

The reinforcement is made of two reinforcement layers, the theoretical concrete cover is larger than 4 0
mm .
4.2.

Problem description - diagnosi s

The cooling tower of Doel 3 is affected by an illness that results of its shell outer face incurring a too fas t
carbonation rate (penetration of carbon dioxide gas in the concrete). This phenomenon, which wa s
demonstrated by a series of tests carried out in Belgian laboratories, leads to a modification of th e
alkalinity of the concrete that protects the reinforcement . With time the reinforcements get corroded an d
cause serious deteriorations as the concrete bursts . In our case, the penetration depth was about 30 to 3 5
mm, which is almost equal the concrete cover of the reinforcement .
4.3.

Treatment definitio n

The propagation of carbonation can be arrested by the application of a coat that prevents the carbo n
dioxide gas penetrating into the concrete .
The suitable products are selected following in situ tests and manufacturer’s specifications . The treatment
of the outer face of the wall is however not sufficient . Further investigations resulted in an imperviou s
coating being planned for the inner surface so as to avoid ingress into and diffusion within the concrete o f
water vapour, since this would have induced flaking of the paint on the outer side . As the concrete was not
much cracked, there was no need for flexible coating . An impervious epoxy-based system was chosen for
the inner protection, it is applied in 2 layers (non pigmented impregnation layer of about 500 g/m2 an d
pigmented final coat of about 300 g/m2) .
The outer protection consists of 3 layers of a one component synthetic (plastic mixture based on PV C
copolymers) paint (300, 400 and 300 g/m2 of different shades) .
The latter effectively protects the concrete by preventing carbon dioxide penetration (CO2 diffusion
coefficient µ = 3 .05 106) whilst allowing the concrete to breathe (permeable to air and water vapour) an d
offering a satisfactory resistance against ageing (chiefly UV radiation) . Because of the unpredictabl e
weather in our countries, the paint was chosen (solvent-based) so that it could be applied on wet surface s
and drying time would be short (1 to 2 hours) .
4.4.

Works

Most of the treatment operations are similar to those already described, though a number of particularitie s
exist at Doel 3 .
For instance, as the outer face of the shell had its concrete roughened in alternance by two types of vertical
ribs (the one long and slender and the other short and broad), such surface offering better behaviour under
wind loads. After a few years, the corrosion of the longitudinal reinforcement that had been placed in th e
slender ribs (which had a very thin (5 to 15 mm) concrete cover) induced a general cracking of these rib s
(pict. 5) . A new calculation of the shell showed that the slender ribs could be completely removed from th e
tower. This was in fact less expensive and more feasible than trying to repair them . These works were
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carried out 2 years before the treatment of the complete shell was considered . A new problem occurred
because of the remains of the reinforcement that previously linked the ribs to the shell . This steel had been
cut close to the shell and was corroded due to an insufficient protection . These countless rust point s
required a particular treatment method that was defined after detailed trials and tests (pict . 6) .
Another particularity resides in the fact that the plant of Doel (4 units) is only provided with 2 coolin g
towers, each one serving 2 units (D1 coupled with D4 and D2 with D3) . The stop of two units is thu s
required if work is to be performed inside a tower unless the thermal conditions of the outfall allows the
circulation water of one unit to bypass the tower while the other is stopped . In our case, the entire treatment
inside the cooling tower had to be carried out within the strictly limited period that unit 3 was shut dow n
for steam generator replacement. In this short period had to be covered the assembly/disassembly of the
fixed and mobile scaffolding installations, the automatic remote-controlled and manual cleaning devices ,
and of course all the cleaning, repair and painting operations . Furthermore, the presence of 2 stiffenin g
rings on the inner face of the tower (each protruding by some 1 .25 m) did not make things easier (pict. 4) .
With some precautions, the painting of the inside was done using an airless spray gun . Thanks to good
forward planning, the operation was completed successfully within the six weeks time allocated .
On the contrary, the restoration work on the outer face of the shell took much longer then expected, as the
necessity appeared for corrective treatment of the ribs . Dealing also with the weather and temperatur e
conditions (leading to an average working ratio of 45%, winter stop included), the work could not b e
completed in less than one and a half years despite the simultaneous use of 4 cradles . In order to obtain an
homogeneous final shade, the last coat (representing 48,000 m2 and manufactured in one batch) wa s
nevertheless applied in only one week .
The whole work was also controlled by the experts of an insurance company allowing it to be covered by a
10-year full warranty conditional to regular inspections . The first of these inspections took place after one
year and did not reveal any visible deterioration .
5.

Tihange 3 Power plant

5.1.

Main features

Cross-flow type cooling tower built in the years 1981 - 1983 :
- diameter at the ground level (cold water basin not included)
- diameter at the lintel
- diameter at the throat
- diameter at the crown
- height at the lintel
- height at the throat
- total height
- shell thickness at the lintel (varying along the first 10 m)
- shell thickness from level 10 m to 100 m
- shell thickness from level 100 m to 115 m
- shell thickness from level 115 m to 140 m
- shell thickness from level 140 m to the top

:
:
:
:
:
:
:
:
:
:
:
:

120,5 m
111,30 m
62,9 m
67,8 m
8,7 m
109,3 m
157,5 m
0,86 to 0,21 m
0,21 m
0,21 to 0,17 m
0,17 m
0,17 to 0,50 m

The steel reinforcement is made of two layers, the theoretical concrete cover is larger than or equal to 40
mm .

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NEA/CSNI/R(2002)7/VOL 1
5.2 .

Damage description - diagnosis

Since a few years, the concrete of the external side of the shell of Tihange 3 cooling tower became ochrecoloured. The outer side of the shell had become ochre several years already below the throat, due to thi s
part of the face being exposed to rainfall . More recently this was increased by percolation which formed a
permanent wet patch presenting traces of calcite, several hundreds of m² covering the first 15 casting ring s
(each of 1 .5 m high) .
In addition to the inspection routinely performed in the scope of the monitoring program, several particula r
analysis were carried out to focus on the phenomenon which induced this particular pathology . These
consisted of :
a chemical, petrography and microscopic analysis of core samples taken within and outside the we t
patches;
a chemical analysis (particular for Fe content) and microscopic examination of algae fragments take n
from the internal face of the shell, and for comparison purposes from another cooling tower at the sam e
site;
a corrosion study of the reinforcement by electrical potential measurements in the wet patches and in a
totally opposite area;
The overall analysis of these two sets of investigations permitted drawing coherent conclusions which can
be outlined as follows :
the absence of correlation between the deformation and the cracking hints at the concrete itself being a t
fault;
the reinforcement corrosion was found low except in the wet patches where it can be said medium to
strong ;
the concrete porosity is very high, its density and compressive strength are low ;
the ochre colouring does not reveal reinforcement corrosion, but appears to result from a mineralogica l
transformation of ferrous particles present in the cement, resulting of a drop in pH due to concrete
carbonation . The carbonation front matches exactly the coloured zone .
The percolation observed in the wet area in fact results from the too high porosity of the concrete . The
permeability throughout the mass of the concrete induces the percolation, the engine of this being th e
pressure gradient between the inside and the outside of the tower ..
5.3.

Determination of the treatment

Water percolation jeopardizes the durability of the tower as it, in the longer time, results in reinforcemen t
corrosion and a washing away of the concrete mix constituents .
We therefore found it necessary to fight percolation by impeding the water to penetrate into the concrete ,
by applying an inner waterproofing coating composed of two layers of epoxy, in addition to the primer .
This treatment is adequate because the cracking is low . Several other reasons (limited time, importance o f
permanent cost, technical risks, …) resulted in the decision being taken to treat all the entire inner surfac e
rather than just the affected area.

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No treatment was advised for the outer face as it is important to be able to ascertain the efficiency of the
inner coating . The outer face has to be left as is not to mask a possible evolution. Also no technica l
arguments (since carbonation and corrosion are low) can justify outer treatment which as we know is ver y
costly, and moreover, it allows the concrete to "breathe" .
5.4.

Works

The work was quite similar to that done on the inside of the Doel 3 cooling tower shell, applying the same
methods and essentially requiring a sufficiently long period of stoppage of the plant (pict . 7) .
A steam generator replacement which was planned at Ti3 in the summer `98 offered that suitable window
to treat the cooling tower . However due to other maintenance work to the tower, only 6 weeks wer e
available to achieve this treatment .
The challenge was met : the first site meeting was held on 10 June and the last lick of paint was applied o n
20 July, despite 12 days in all being lost due to bad weather .
The attached tables 1 and 2 illustrate the progress of the repair and coating of the 37 .600 m2 of the inner
face of the structure, which involved applying some 37T of product . Appendix 3 gives an idea of the poo r
weather conditions (essentially rainy days, but wind can be embarrassing as wel), and their impact i s
reflected on the progress curve .
Like at Ruien and Doel the work was covered by Quality Assurance requirements and supervised b y
experts so as to insure the 10 year guarantee .
In addition to the practical aspects of this type of work which are similar to the already mentione d
examples, one point was analyzed particularly : the structural stability of the crown walkway from whic h
the painter's cradles had to be suspended . Depending on the orientation and the suspension points, thes e
cradles induce tilting loads of several ton/m . Indeed this tower inspection walkway is not an integral part o f
the shell . It is composed of a series of prefabricated "U" elements, each fixed individually by means of 2
bolts on the top edge of the shell and 1 bent rebar at each end . These U elements are each 3 .4 m long, ar e
straight and are spaced 12 cm from each other, forming a polygon on top of the crown (pict . 8).
The checking of the stability of these boxes was favourable providing their anchoring, which were alread y
15 years old, had retained their original strength .
This was verified by :



tapping and visual inspection of the seals (absence of cracking, rust, dull sound )



core samples to test the quality of the material
verify the compliance between the reinforced drawings and the execution, through a magnetic "X Ray"
of the elements in order to check the position, the number, the diameter and the cover of th e
reinforcement .

The confirmation of the quality of the concrete and the suitability of the reinforcement compared to the
calculation allowed to go ahead with the treatment with full knowledge of the facts .

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6.

TIHANGE 3

6.1.

Introductio n

Not only the shells of these towers can be source of problems !
As said before, the cooling tower is a heat exchanger the function of which is to cool down the water of th e
third circuit of a nuclear power plant, cooling itself the turbine condenser (circulation water) .
This is obtained by a heat exchange between air (up-draught) and water . The cooling water is sprayed o n
the inner horizontal surface (representing about 10 .000 m², i.e. about 3 football fields!) by a net of channel s
and pipes. It is sprinkled above an exchange material (called packing) across which it trickles (pict . 9) . In
normal conditions, the water flow rate represents an amount of water of 4 l/m²/s, in winter conditions thi s
value can be doubled.
The water is collected at the foot of the tower in a basin and then flows by gravity to the return circuit, to a
discharge in the river or to the pumping station depending on the thermal conditions .
On Thursday 3 January 2002, the break of 2 brackets of the supporting structure resulted in the collapse o f
2 beams and the packing they were supporting (pict . 10) . The flood entrained the debris which was stoppe d
by the protection screens placed at the entry of the returning pipes . The inlet of the pipes became rapidly
clogged and eventually the basin flowed over . Due to the geographical position of the tower, the overflo w
flooded to some of the NPP buildings .
6.2 .

Finding s

The NPP operator ELECTRABEL asked TRACTEBEL Energy Engineering to analyse the accident . The
next day some quick investigations were made (pict . 11) :
- the break affects two brackets embedded at the lower face of principal beams that supported th e
secondary beams, themselves supporting the packing;
- the affected brackets are anchored at the lower face of the principal beams as if they were suspended ;
- the reinforcement steel that became bare and the rupture surface do not show any trace of corrosion o r
dirt;
- the concrete of the rupture surface feels sound ;
- the two fallen beams are situated in the same area, present the same constructive characteristic s
(geometry, supports, …), provide an identical function and incurred the same type of break .
6.3.

Analysi s

As the rupture surface was clean and there is no steel corrosion, one may consider that the failure was
sudden (brittle break, like a lump of sugar) and corresponds to a shear break of the concrete . The concrete
characteristics seemed good, core samples have nevertheless been taken in the wrecked brackets to verif y
this aspect . Results are satisfying .
Other causes were thus to be searched in an inadequate match between the design and the actual loa d
applied to the bracket .

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The consultants have investigated in the following directions while the cooling tower was stopped (in fact ,
the NPP can work with the tower by-pass, the circulation water is not cooled as it is sent to the basin befor e
being sprayed in the sprinkling system) :
- loading condition of the structure
- resistance capacities of the structur e
- visual inspection of the structure
a)

Loading state

An eventual overload of the bracket can only result from an excess of weight applied on the supporte d
beam, originating from the packing . This overweight of the packing may result from its clogging or winte r
service conditions (to minimize freezing problems, the flow is only distributed on the periphery of the
packing, which means that for a same flow rate, the concerned wet surface receives a doubled flow ,
increasing the water weight as well) .
Measurements have been carried out on samples of the packing near the wrecked area . A new problem
arises because the measure is made on a dry packing as the cooling tower is stopped, and therefore th e
water quantity has to be estimated.
The packing itself is made of different elements (part of their weight doesn’t change, while other parts o f
the packing can be fouled, increasing their weight) . Considering a dead load of about 0 .5 kN/m³, the
surprise was great when it was discovered that some packing elements weighed more than 2 kN/m ³ . The
problem was then to evaluate the correspondent weight of water : part of it trickling across the packing ,
part retained by the clog matter . A variation formula has been used to do this, considering that the quantity
of water is a function of the degree of clogging, and that it can reach up to 1,5 times the dry weight of clo g
matter in the worst case.
b)

Structural analysis

Concurrent with this analysis, a complete study of the structural elements was carried out, taking int o
account the initial design calculation codes, to determine the available safety margin for each element on
basis of the as built drawings . This study has enabled us to point out which of the elements wer e
theoretically the most at risk .
c)

Visual inspection

Based on the above mentioned study, a visual inspection was performed on the “most at risk” elements .
The discovery of unexpected damage called for a comprehensive visual inspection involving more than a
thousand spots (pict . 12 & 13) .
d)

Load test of the wrecked beam s

The two fallen beams have been examined and showed, over their entire length, regular cracking at the
inferior face, going up on each side to mid-height . They were taken to a laboratory where they were teste d
to verify if the load they had supported had changed their future behaviour (still elastic or not) .

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6.4

Results and diagnos e

6.4.1 .

Packing support structure (low level)

The failure observed is due to an overload of the packing . This overload has broken 2 brackets of a
series of 24 supporting the longest beams of the tower . Their visual inspection showed a generalize d
cracking .
The tests performed on the 2 beams have shown that they were still in good shape (however they ha d
fallen down!), breaking by flexion at their design load.
The visual inspection of the rest of the structure highlighted the presence of degradations (mostly a t
the extremities of the beams and/or the brackets), some of them being severe enough to represent a
danger for the structure .
6.4.2.

Supports of the spraying system (upper level )

The same kind of degradation was observed randomly distributed on 10% of the supports (pict . 14 &
15). As our theoretical study revealed, the thermal effects due to the temperature variation s
(stop/service conditions, day/night, sun/shadow, winter/summer, …) develop horizontal loads whos e
importance is close to the vertical loads . Originally nothing was planned to limit concrete/concret e
friction between the beam and its support.
To summarize, two main causes were demonstrated :
6.5.

the clogging of the packing, leading to its overweight ;
an underestimation of the horizontal loads caused by a lack of support material .
Treatment and repair works

Twenty-two new steel brackets were designed to support both the initial one and its beam . They have been
ordered, fabricated and placed within 3 weeks on the lower level to replace the brackets wrecked by
overweight of the packing (pict . 16) .
The brackets and beams whose extremities are severely cracked have been supported by scaffolding tha t
will be left in place until the next stop .
At the upper level, the configuration of the brackets allowed us to design another steel system an d
reinforce beam and brackets that were cracked (pict . 17) . These supports were designed, ordered,
fabricated and placed within a week .
6.6.

Conclusion s

This last example shows once more, if necessary, the need of regular inspections and/or monitoring to
avoid the risks of the occurrence of severe and costly damages, possibly even human consequences .

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In this case, we proposed to the operator to monitor the packing weight by two means : continuou s
weighing of selected volumes of packing and by equipping the two new beams (replacing the wrecke d
beams) with strain gauges providing a continuous measure of the actual weight carried by the structure .
The study that was made gave the opportunity to set the adequate levels of alarm .
The water level of the basin shall also be monitored to avoid another flood .

7.

Appendices and pictures

7.1.

Ruien ¾ : repair and painting works

Picture 1

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NEA/CSNI/R(2002)7/VOL 1

Picture 2

7 .2 .

Picture 4

Doel 3 : painting works

NEA/CSNI/R(2002)7/VOL 1

Picture 5

Picture 6

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NEA/CSNI/R(2002)7/VOL 1

7 .3 .

Tihange 3 : painting works

Appendix 1
Appendix 2

NEA/CSNI/R(2002)7/VOL 1

Appendix 3

Picture 7

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Picture 8

7 .4.

Tihange 3 : brackets breakin g

Picture 9

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NEA/CSNI/R(2002)7/VOL 1

Picture 1 0

Picture 11

153

NEA/CSNI/R(2002)7/VOL 1

Picture 1 2

Picture 13

154

NEA/CSNI/R(2002)7/VOL 1

Picture 14

Picture 15

155

NEA/CSNI/R(2002)7/VOL 1

Picture 1 6

Picture 17

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Post-Fire Damage Assessment Procedures for Nuclear Power Plant Structures
L.M. Smith British Energy Generation (UK) Lt d
Abstract
The examination, inspection, maintenance and testing of all plant and structures that may affect nuclea r
safety on Nuclear Power Plants is of prime importance. It is essential that all nuclear safety relate d
structures must be maintained throughout their operational life in such a way that they are always fit fo r
purpose and capable of meeting their nuclear safety role as required . In order to do this they must be
examined, inspected and tested in a manner and at a frequency which is adequate to confirm that th e
structural integrity, performance and reliability claims made in the safety case continue to be me t
throughout the operational life of the station . The performance criteria for nuclear safety related structure s
must be determined on the basis of the duty required of each structure at each of the stages of the lifetim e
of the facility . Consequently, detailed procedures exist for the routine inspection of these structures i n
normal service .
In the event of a fire on a nuclear power plant the standard inspection procedures may no longer b e
applicable and the Licensee would have to demonstrate that any nuclear safety related structures in the are a
of the fire are still fit for purpose and capable of meeting their nuclear safety role .
Following a fire in an area that could potentially affect the condition or performance of a structure, a
specific procedure would be written to cover the examination of that structure, the method and criteria fo r
its assessment and acceptance criteria for its continued use. The specific post-fire inspection procedur e
would be based on the normal in-service inspection procedure amended to take account of the extent of the
damaged area . All areas where structural failure could cause damage to, or failure of, nuclear safety relate d
equipment would be addressed. The resulting inspection report would include lists of the defects found an d
give recommendations on any remedial works required .
This paper considers the factors that would be applicable to specific post-fire inspection procedures for th e
inspection and assessment of nuclear safety related structures .

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Post-Fire Damage Assessment Procedures for Nuclear Power Plant Structures
Introduction
As fire is an identified risk on Nuclear Power Plants (NPPs), it is considered at the design stage an d
preventative measures, segregation and suppression systems are built into the plant design in order to limi t
the effects of fire should it occur . However, although this greatly reduces the likelihood and effects o f
potential fires it does not totally remove the risk of one taking place .
This paper is based on circumstances as they exist in the UK but the general principles are equally
applicable to other locations . Should a fire occur, the post-fire assessment of structures on UK NPPs woul d
be treated as a special investigation not covered by normal standing procedures . This document only
provides a brief over view of the subject. For more detailed information on the effects of fire on th e
structural materials involved the references given at the end of the paper should be consulted [2-10] .
General Background
Each site in the UK is covered by a Nuclear Site Licence which contains 36 standard conditions whic h
must be met by the operator and is issued under the provisions of the Nuclear Installations Act 1965 b y
HM Nuclear Installations Inspectorate . The rules are not prescriptive and the licensee retains absolut e
responsibility for nuclear safety under UK law . [1 ]
The regulatory requirements have major implications for the in-service inspection regime, principall y
through site licence condition 28, which covers the examination, inspection, maintenance and testing of al l
plant and structures that may affect nuclear safety . It is essential that all nuclear safety related structure s
must be maintained throughout their operational life in such a way that they are always fit for purpose an d
capable of meeting their nuclear safety role as required by a detailed safety case, which is identified in the
licence conditions . In order to do this they must be examined, inspected and tested in a manner and at a
frequency which is adequate to confirm that the structural integrity, performance and reliability claim s
made in the safety case continue to be met throughout the operational life of the station [2] . The
performance criteria for nuclear safety related structures must be determined on the basis of the dut y
required of each structure at each of the stages of the lifetime of the facility .
In normal operation, a list of nuclear safety related structures and procedures for the inspection of each o f
them are prepared from examination of the Station Safety Report . The inspections are visual in the firs t
instance and checklists are prepared for each building or area of structure from detail drawings and
preliminary inspections that give guidance to the inspection team .
Under the terms of Site Licence Condition 28, after a fire, the Licensee would have to demonstrate that an y
nuclear safety related structures in the area of the fire are still fit for purpose and capable of meeting thei r
nuclear safety role as required by the safety case . In the event of a fire in an area that could potentially
affect the condition or performance of a structure, a specific procedure would be written to cover th e
examination of that structure, the method and criteria for its assessment and acceptance criteria for it s
continued use . As in the assessment of damaged areas of conventional structures, personnel safety woul d
be of high importance and require a risk assessment to be carried out but on NPPs there is the additional
requirement to ensure radiological safety . This will require survey and monitoring of the area and a written
system of work to keep any potential radiological dose to personnel to a level that is as low as reasonabl y
practicable .

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NEA/CSNI/R(2002)7/VOL 1

The specific post-fire inspection procedure would be based on the normal in-service inspection procedur e
amended to take account of the extent of the damaged area . All areas where structural failure could cause
damage to, or failure of, nuclear safety related equipment would be addressed. The resulting inspection
report would include lists of the defects found and give recommendations on any remedial works required .
IAEA Safety Report No .8 (Preparation of fire hazard analyses for NPPs) [11]covers the need for a fir e
preplan . It may be argued that there is also a need for a fire postplan prepared in advance to cover th e
procedures and methods to be used to assess the immediate safety of nuclear safety related structures (i n
the order of their importance to nuclear safety) following a fire .
On British Energy NPPs, defects are defined in three nuclear safety categories ; Category 1 : Affecting
safety - repair required immediately ; Category 2 : Not affecting safety - repair required as soon as possible
to prevent further deterioration ; and Category 3 : Not affecting safety - repair carried out under norma l
station maintenance programme . The categorisation is based initially on an assessment made by th e
inspection engineer, taking into account the current safety case and overall structural integrity . Identified
defects would be included (with normal inspection defects) in a monitoring database which allows eas y
progress tracking and identification of faults .
In addition to the defect classification given above, a damage classification of the effects of the fire on th e
structural members should be carried out to allow a structural assessment to be carried out . Tabulated
damage classifications are useful in this regard (Table 3 [9]) .
When assessing a fire damaged structure the points given below should be taken into consideration .
Fire Damage Assessment
Where possible detailed information concerning the use and occupancy of a structure should be obtaine d
prior to a post-fire investigation to enable an estimate of the fire loading to be determined. Station Logs ,
Fire brigade reports and eyewitness accounts are useful for determining the duration and course of the fire .
Before the structural investigation is carried out, as-built construction drawings should be obtaine d
showing all the main members of the structure and the form of construction . Care must be exercised to
ensure that the effects of all previous plant modifications and repairs have been included in the informatio n
collated [3] . Previous in-service inspection reports can provide useful information on the pre-fire condition
of the structure and normal in-service inspection procedures identify critical areas of the structure .
A variety of different methods, ranging from visual survey through to non-destructive examination an d
comprehensive instrumentation, is available to the engineer to allow condition monitoring and lifetim e
management of structures to be implemented. Structural investigations should, where possible, utilise non invasive methods such as visual surveys or non-destructive testing although there are limitations to th e
information that may be obtained from such sources . Where it exists, instrumentation may provide usefu l
information on the performance of materials and structures . Invasive investigations involving the remova l
of a sample or specimen for testing should only be used where simpler methods cannot provide th e
required information . Careful attention must be paid to the limitation of damage caused during sampl e
collection and reinstatement of the structure must be given detailed consideration and executed correctly .
The method of investigation that is chosen in any particular post-fire situation for the assessment of fir e
damage to concrete structures will obviously depend on the accuracy of the results required . Each method
has its advantages and disadvantages but every method, if properly used with sufficient care, may give a n
assessment of the fire damage to the structure. No one method is entirely free from error and in most case s
a combination of tests will be used depending on the importance and cost of the reinstatement of the
structure, the time allowed for repair and whether demolition and replacement is a feasible option .
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NEA/CSNI/R(2002)7/VOL 1

The examination of fire debris is a very important stage in any post-fire investigation as it allows an overal l
picture of fire severity to be gained by the investigator . This is especially useful if combined with visua l
damage classification and a sounding survey of the structure . The main drawback with debris surveys i s
that the final position of thermal indicators may bear little relationship to their position at the time of th e
fire if any debris clearance has taken place before examination .
Pre-fire defects must be isolated from fire damage and noted specifically as such along with othe r
influences such as blast or explosion damage (see [3, 9]) .
Photographs are a most important part of any post-fire survey as they provide a visual record and ma y
yield useful information at later stages during investigation or reinstatement should problems b e
encountered .
Concrete Structures
Colour changes in concrete and aggregates may be used as a valuable guide in fire damage assessment i n
cases where significant changes occur due to thermal exposure . In some instances, however, a lack o f
colour change may not be taken as inferring that concrete is unaffected by high fire temperatures and care
should be exercised especially in cases where ingneous aggregates have been used .
Ultrasonic pulse velocity (UPV) measurements may be employed in damage assessment if a sound area o f
concrete is available for comparison and the physical configuration of the member and its reinforcemen t
and the surface condition of the concrete are suitable . In complicated cases, laboratory analysis of th e
results may be necessary to interpret the readings . Any UPV survey must be preceded by a cover meter
survey to identify the position, orientation and size of steel reinforcement .
Cores may be taken from concrete structures to allow the depth of colour change to be determined or fo r
material testing purposes . However, as the damage will be worst at the fire exposed surface, which will b e
likely to be in the area of the core nearest the plattens of the testing machine, care must be taken whe n
interpreting the test results . Coring should be carried out in non-critical areas of structures where stress
levels are low .
Thermoluminescence (TL) measurement provides an excellent indication of the thermal exposure that
concrete has been subjected to and from this the residual compressive strength may be judged . As with
UPV testing, however, it must be realised that concrete strengths may occupy a range of values for an y
thermal exposure and only an approximate figure may be attributed to the residual strength . TL testing i s
particularly valuable in critical areas of a structure where other methods may not be applied. The TL test i s
the only test that can tell definitely if concrete has been heated significantly and in many cases its mos t
useful application is in indicating areas that have not been subjected to damaging thermal exposure s
(especially in cases where a large amount of smoke damage is present) . Post-fire exposure to radiation may
affect the TL signal .
In prestressed concrete structures, such as containments and pressure vessels, the residual level of prestres s
will be of great importance . Installed instrumentation may be used if it has survived the fire and in unbonded prestressing systems load checks may be carried out and tendons withdrawn for inspection an d
testing. In this type of system it is possible to replace and restress tendons that may have been affected b y
high temperatures in order to restore the level of prestress in the structure. The situation with regard to
bonded prestressing systems can be more difficult to assess as withdrawal, replacement and restressing
cannot be carried out.

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NEA/CSNI/R(2002)7/VOL 1
Brickwork and Blockwor k
By virtue of their method of manufacture, clay bricks perform well under fire conditions although th e
mortar used in wall construction and concrete blocks will degrade in a manner similar to that o f
unreinforced concrete. Brick and blockwork panels may also be damaged by restrained thermal forces an d
by global structural movement of steel or concrete frames and by surface damage due to shock cooling
during fire fighting operations .
Steelwork
In general, a structural member remaining in place, with negligible or minor distortions to the web, flange s
or connections should be considered satisfactory for further service . The exception will be for the relatively
small number of structures in cold-worked or tempered steel where there may be permanent loss o f
strength. The change in strength may be assessed using estimates of the maximum temperatures attained o r
on-site tests (such as hardness tests) ; if necessary, the steel should be replaced. Microscopy can be used to
determine changes in microstructure . Since this is a specialised field, the services of a metallurgist are
essential [9] . Bolted connections may require detailed consideration if heated above 360°C .
Supplementary Testin g
Supplementary testing such as load testing or detailed materials testing is normally only required in special
cases where the damage assessment is highly critical or where special problems exist . This may include
tensile and hardness testing of steel, measurement of the modulus of elasticity of the material, durabilit y
testing or load testing of the entire structure or a portion of it . If a fire has occurred inside a containment
structure, a structural overpressure test may be required to confirm the structural behaviour and
leaktightness of the containment .
Once the residual characteristics of the structure have been determined, a design check based on th e
material properties must be carried out to examine the ability of the structure to continue to fulfil its design
role and to identify areas in need of strengthening .
Further Informatio n
Further detailed information on the effects of fire on structures and structural surveys may be obtained b y
consulting the references given at the end of this document [3-10] .
Table 1 gives details of a typical outline procedure for the post-fire inspection of NPP structures and Tabl e
2 gives details of temperature indicators which may be used to assist the investigation .
Recommendatio n
Although detailed inspection procedures should be written for each structure after a fire event, it i s
recommended that an outline higher level generic fire postplan is written which contains a ranking o f
nuclear safety related structures (in order of their importance to nuclear safety) and outlining genera l
procedures for post-fire inspection in order to reduce the response time .

161



NEA/CSNI/R(2002)7/VOL 1

References :
1.

"The Nuclear Installations Act 1965", HMSO, UK, 1965 (et seq. )

2.

McNulty, T, McNair, I, and Bradford, P, "A Regulatory View of Plant Life Management of Civi l
Structures in the Nuclear Industry", Institution of Nuclear Engineers, International Conference on
Nuclear Plant Life Management, 17-19 November 1998, Warrington, U K

3.

Smith, LM, "The Assessment of Fire Damage to Concrete Structures", PhD Thesis, Paisle y
College of Technology, UK, 1983

4.

Lie, TT, "Fire and Buildings", Applied Science Publishers, London, UK, 197 2

5.

Pchelintsev, VA, (ed), "Fire Resistance of Buildings", Amerind Publishing Co . Pvt Ltd, Delhi,
India, 197 8

6.

Malhotra, HL, "Design of Fire-resisting Structures", Surrey University Press, UK, 198 2

7.

Schneider, U, "Properties of Materials at High Temperatures - Concrete", RILEM, France, 198 5

8.

Harmathy, TZ, "Fire Safety Design and Concrete", Longman Scientific & Technical, Harlow, UK ,
199 3

9.

"Appraisal of Existing Structures (Second Edition)", Institution of Structural Engineers, London ,
UK, October 199 6

10.

ACI 349 .3R-96, "Evaluation of Existing Nuclear Safety Related Concrete Structures", American
Concrete Institute, Detroit, USA, 199 6

11.

"Preparation of Fire Hazard Analyses for Nuclear Power Plants", IAEA Safety Reports Series No .
8, IAEA, Vienna, Austria, 1998

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NEA/CSNI/R(2002)7/VOL 1

Table 1 Post-Fire Investigation (modified for NPPs) [3]
Stage
Activity
1

2

3

4

Preparatio n
Obtain (i) Original design calculations and as-built drawing s
(ii) Plant modification details
(iii) Nuclear Safety Cas e
(iv) Normal in-service inspection procedures and previous survey reports
(v) Station log
(vi) Fire Brigade report s
(vii) Eye witness reports
(viii) Any available monitoring instrumentation result s
Write (i) Specific post-fire inspection procedure
(ii) Specific written system of work/radiological documentation & requirement s
Visual Examination
(i) Identify any pre-fire damage (eg settlement )
(ii) Locate seat of fire
(iii) Chart progress of fir e
(iv) Examine debris for maximum temperatur e
(v) Locate special damage (eg explosion damage )
(vi) Locate areas of above average fire damage
(vii) Classify members
(viii) Take photograph s
Accuracy q
CONCRETE
Simple In-situ Testin g
(i) Sounding (hammer survey)
1
(ii) Colour Changes (if applicable)
2-3
More Complex In-situ Testin g
(i) UPV Measurement
(ii) Cores for colour examination
2-4
(iii) Deflection measurement
2-3
4

16 3

Time q

A
A

B-C
B-C
B-C



NEA/CSNI/R(2002)7/VOL1

Table 1 (continued)
5

6

7

Laboratory Testin g
(i) Cores for strength
(ii) Themoluminescence
(a) For thermal exposure
(b) For residual strength
(iii) Materials testing (other than 5(i))
(eg Reinforcement & prestressing tendons)
STEELWORK
In-situ Testin g
(i) Distortion & Deflectio n
(ii) Hardness testing
(iii) Metallographic investigation

C

4
2-4
3-4

C
C
C-C+

1-3
4
4

A-B
A-B
C-C+

-

C+
C+

GENERAL
Supplementary Testing
(i) In-situ load testing (incl . SOPT & Tendon Load)
(ii) Design check

8

3

Final Report

Including photographic record, repair recommendations and update to safety case

Post-Fire Investigation Guide (See Table 1 )
q Accuracy
Time
1
2
3
4

General Guid e
Limited Accuracy
Moderately Accurate
Accurate

A Relatively quick in-situ examination
B Time consuming in-situ examination
C Requires lab analysis and/or time
to recover specimens/analysis

16 4



NEA/CSNI/R(2002)7/VOL 1

Table 2 Post-fire Thermal Indicators [3, 9]
Approximate
Temperature, ° C
100+
120
120-140
130-200
140
150
150
150-180
170
180
200
205
250
250
275
280
300-350
390-400
400-500
420
450-870

Indicator & Condition
Paint deteriorates
Polystyrene items collapse,
polythene items shrivel,
PVC degrades
Polystyrene
softens,
polythene softens and melts
Polymethyl
methacrylate
softens
Polyurethane foam charre d
black
Paint destroyed
PVC fumes
Polystyrene melts and flows
Phenolic resin changes from
yellow to brown
60Sn-40Pb solder melts
PVC browns
Charring & clay like
appearance of acrylic resin
Charring of wood begins
Polymethyl methacrylate
bubbles
Lead base babbitt melts
Copper instrument tubin g
begins to soften ,
recrystallise
Lead, sharp edges rounde d
or drops formed
Zinc die casting alloy melts

Typical Example

Method o f
Observation
Visual
Visual

Class

Plastic items

Visual

I

Handles, covers, glazing

Visual

I

Thermal insulation

Visual

I

Coatings
Cable insulation
Plastic items
Wall linings, roof sheets

Visual
Visual
Visual
Visual

I
I
I
I

Solder joints - electrical
equipment
Cable insulation
Thermal insulation

Visual

I

Visual
Visual

I
I

Doors, floors, furniture
fittings
Handles, covers, glazing

Visual

I

Visual

I

Sliding bearings in pumps
& compressors
Instrument tubing

Visual

I

Hardness test

II

Visual

I

Visual

I

Coatings
Plastic items, cable
insulation

Plumbing, fixtures,
shielding
Plumbing fixtures, small
components
Cable insulation
Galvanised steel
Piping

I, II
I

PVC chars
Visual
I
Zinc coating melts
Visual
I
Austenitic stainless stee l
Metallographic
II
sensitised
480
Asbestos powders/flakes
Column packing
Visual
I
540
High temperature scaling
Carbon steel exposed to
Visual
I
begins on carbon steels
air
595
Bolts tempered to lower
ASTM A193 B7 & B1 6
Hardness test
II
than normal hardness
bolting
Indicator Classes - I Unaffected by time (continued)
II Events of a more complex nature involving functions of time, temperature & coolin g
rate

16 5

NEA/CSNI/R(2002)7/VOL 1

Table 2 (continued) Post-fire Thermal Indicators [3, 9 ]
Approximate
Temperature, ° C
600-650
700-750

Indicator & Condition
Aluminium & alloys melt

Typical Example
Small machine parts ,
brackets, electrical
conduits
Glazing

Method o f
Observation
Visual

Class
I

Sheet glass softened or
Visual
I
adherent
750
Moulded glass rounded
Corrugated window glass
Visual
I
760
Gross deformation of low Structural steel members
Visual
I
carbon steels
760
Inorganic zinc paint
Structural coatings
Visual
I
darkens, spalls off
800
Sheet glass rounded
Glazing
Visual
I
820
Borosilicate glass softens,
Instrument gauges, sight
Visual
I
melts
glasses
Bolting hardened well
Steelwork connections
Hardness test
845
II
above normal range
850
Sheet glass flowing easily
Glazing
Visual
I
900-1000
Leaded red brass/ brass
Plumbing fixtures
Visual
I
melts
905
Zinc coating boils off
Galvanised steel
Visual
I
950
Ni/Au bronze metal melts
Thermocouple wave rings
Visual
I
980
Foamglass insulation melts Thermal insulation
Visual
I
to a black-grey slag
1000-1100
Copper melts
Wiring
Visual
I
I
1100-1200
Cast iron melts
Castings
Visual
1400
Low carbon steel melts
Structural steelwork
Visual
I
Indicator Classes - I Unaffected by time
II Events of a more complex nature involving functions of time, temperature & coolin g
rate

16 6



NEA/CSNI/R(2002)7/VOL 1

Table 3 Classes of Post-fire damage, Characterisation and Description [9 ]
Class

Characterisation

Description

1

Cosmetic damage, surface

Characterised by soot deposits and discolouration. In most cases soot & colour
can be washed off. Uneven distribution of soot deposits may occur . Permanent
discolouration on high-quality surfaces may lead to rejection . Odour is included
in the class ; it may be difficult to remove but chemicals are available fo r
elimination.

2

Technical damage,
surface

Characterised by damage to surface treatments and coatings . Small extent only
of concrete spalling or corrosion on uncovered metals. Painted surfaces can be
repaired. Plastic coated surfaces need replacement or covering . Minor spalling
may remain or can be re-plastered .

3

Structural damage,
surface

Characterised by some concrete cracking and spalling, lightly charred woo d
surfaces, some deformation of metal surfaces or moderate corrosion damage .
This type of damage includes Class 2 damage and an be repaired similarly .

4

Structural damage, cross section (interior)

Characterised by major concrete cracking and spalling, deformed flanges an d
webs of steel beams, partly charred cross sections of timber constructions, an d
degraded plastics . Damage can in many cases be repaired on the existin g
structure. Within the class are also deformations of structures so large that th e
loadbearing capacity is reduced, or dimensional alterations prevent proper fittin g
into building. This applies in particular to metal/steel constructions .

5

Structural damage to
members and components

Characterised by severely damaged structural members and components ,
impaired materials and large deformations . Concrete constructions are
characterised by extensive spalling, exposed reinforcement and impaire d
compression zone. In steel structures extensive permanent deformations hav e
arisen due to diminished loadbearing capacity caused by high temperature
conditions. Timber structures may have almost fully charred cross sections .
Changes in materials may occur after fire, so they may display unfavourable
properties . Class 5 damage will usually lead to rejection .

16 7



Unclassified
Organisation de Coopération et de Développement Economique s
Organisation for Economic Co-operation and Development

NEA/CSNI/R(2002)7/VOL 2
05-Sep-2002
English - Or. Englis h

NUCLEAR ENERGY AGENCY

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

OECD-NEA WORKSHOP ON THE EVALUATION OF DEFECTS ,
REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETE
STRUCTURES ON NUCLEAR POWER PLANT S
Hosted by GRS at the DIN Institute in Berlin, German y

10th-11th April, 2002

JT0013093 2

Document complet disponible sur OLIS dans son format d’origine
Complete document available on OLIS in its original format

NEA/CSNI/R(2002)7/VOL 2

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT
Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30t h
September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed :
- to achieve the highest sustainable economic growth and employment and a rising standard of living in Membe r
countries, while maintaining financial stability, and thus to contribute to the development of the world economy ;
- to contribute to sound economic expansion in Member as well as non-member countries in the process of economi c
development ; and
- to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance wit h
international obligations .
The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece ,
Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdo m
and the United States . The following countries became Members subsequently through accession at the dates indicated hereafter :
Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18t h
May 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996), Korea (12th
December 1996) and the Slovak Republic (14th December 2000) . The Commission of the European Communities takes part in th e
work of the OECD (Article 13 of the OECD Convention) .
NUCLEAR ENERGY AGENC Y
The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC
European Nuclear Energy Agency . It received its present designation on 20th April 1972, when Japan became its first
non-European full Member. NEA membership today consists of 27 OECD Member countries : Australia, Austria, Belgium ,
Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg ,
Mexico, the Netherlands, Norway, Portugal, Republic of Korea, Spain, Sweden, Switzerland, Turkey, the United Kingdom and th e
United States . The Commission of the European Communities also takes part in the work of the Agency .
The mission of the NEA is:
- to assist its Member countries in maintaining and further developing, through international co-operation, th e
scientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclea r
energy for peaceful purposes, as well as
- to provide authoritative assessments and to forge common understandings on key issues, as input to governmen t
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainabl e
development .
Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive wast e
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law an d
liability, and public information . The NEA Data Bank provides nuclear data and computer program services for participatin g
countries.
In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency i n
Vienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.
© OECD 200 2

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be made to OECD Publications, 2, rue André-Pascal, 75775 Paris Cedex 16, France .

2

NEA/CSNI/R(2002)7/VOL 2

COMMITTEE ON NUCLEAR REGULATORY ACTIVITIE S
The Committee on Nuclear Regulatory Activities (CNRA) of the OECD Nuclear Energy Agency (NEA) i s
an international committee made up primarily of senior nuclear regulators . It was set up in 1989 as a forum for th e
exchange of information and experience among regulatory organisations and for the review of developments whic h
could affect regulatory requirements .
The Committee is responsible for the programme of the NEA, concerning the regulation, licensing an d
inspection of nuclear installations . The Committee reviews developments which could affect regulatory requirement s
with the objective of providing members with an understanding of the motivation for new regulatory requirements
under consideration and an opportunity to offer suggestions that might improve them or avoid disparities among
Member Countries . In particular, the Committee reviews current practices and operating experience.
The Committee focuses primarily on power reactors and other nuclear installations currently being built
and operated . It also may consider the regulatory implications of new designs of power reactors and other types o f
nuclear installations .
In implementing its programme, CNRA establishes co-operative mechanisms with NEA's Committee o n
the Safety of Nuclear Installations (CSNI), responsible for co-ordinating the activities of the Agency concerning th e
technical aspects of design, construction and operation of nuclear installations insofar as they affect the safety of suc h
installations . It also co-operates with NEA's Committee on Radiation Protection and Public Health (CRPPH) an d
NEA's Radioactive Waste Management Committee (RWMC) on matters of common interest .

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S
The NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made up
of scientists and engineers . It was set up in 1973 to develop and co-ordinate the activities of the Nuclear Energ y
Agency concerning the technical aspects of the design, construction and operation of nuclear installations insofar a s
they affect the safety of such installations . The Committee's purpose is to foster international co-operation in nuclear
safety amongst the OECD Member countries .
CSNI constitutes a forum for the exchange of technical information and for collaboration betwee n
organisations which can contribute, from their respective backgrounds in research, development, engineering o r
regulation, to these activities and to the definition of its programme of work . It also reviews the state of knowledg e
on selected topics of nuclear safety technology and safety assessment, including operating experience . It initiates and
conducts programmes identified by these reviews and assessments in order to overcome discrepancies, develo p
improvements and reach international consensus in different projects and International Standard Problems, and assist s
in the feedback of the results to participating organisations . Full use is also made of traditional methods of cooperation, such as information exchanges, establishment of working groups and organisation of conferences and
specialist meeting .
The greater part of CSNI's current programme of work is concerned with safety technology of wate r
reactors . The principal areas covered are operating experience and the human factor, reactor coolant syste m
behaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reacto r
accidents and their confinement, containment performance, risk assessment and severe accidents . The Committee
also studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operate s
an international mechanism for exchanging reports on nuclear power plant incidents .
In implementing its programme, CSNI establishes co-operative mechanisms with NEA's Committee on
Nuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation ,
licensing and inspection of nuclear installations with regard to safety . It also co-operates with NEA's Committee o n
Radiation Protection and Public Health and NEA's Radioactive Waste Management Committee on matters o f
common interest.

3

NEA/CSNI/R(2002)7/VOL2

4

NEA/CSNI/R(2002)7/VOL 2

Foreword
The Committee on the Safety of Nuclear Installations (CSNI) of the OECD-NEA co-ordinates the NE A
activities concerning the technical aspects of design, construction and operation of nuclear installation s
insofar as they affect the safety of such installations . In 1994, the CSNI approved a proposal to set up a
Task Group under its Principal Working Group 3 (recently re-named as the Working Group on Integrity o f
Components and Structures (IAGE)) to study the need for a programme of international activities in th e
area of concrete structural integrity and ageing and how such a programme could be organised . The task
group reviewed national and international activities in the area of ageing of nuclear power plant concret e
structures and the relevant activities of other international agencies . A proposal for a CSNI programme of
workshops was developed to address specific technical issues which were prioritised by OECD-NEA tas k
group into three levels of priority :
First Priority
• Loss of prestressing force in tendons of post-tensioned concrete structure s
• In-service inspection techniques for reinforced concrete structures having thick sections and area s
not directly accessible for inspection

Second Priority
• Viability of development of a performance based databas e
• Response of degraded structures (including finite element analysis techniques )
Third Priorit y
• Instrumentation and monitoring
• Repair methods
• Criteria for condition assessment
The working group has progressively worked through the priority list developed during the preliminar y
study carried out by the Task Group . Currently almost all of the three levels of priority are effectively
complete, although in doing so the committee has identified other specific items worthy of consideration .
By working logically through the list of priorities the committee has maintained a clarity of purpose whic h
has been important in maintaining efficiency and achieving its objectives . The performance of the group
has been enhanced by the involvement of regulators, operators and technical specialists in both the work o f
the committee and its technical workshops and by liaison and co-operation with complementar y
committees of other international organisations . The workshop format that has been adopted (based aroun d
presentation of pre-prepared papers or reports followed by open discussion and round-table development o f
recommendations) has proved to be an efficient mechanism for the identification of best practice, potentia l
shortcomings of current methods and identification of future requirements .

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NEA/CSNI/R(2002)7/VOL2

6

NEA/CSNI/R(2002)7/VOL 2
SUMMARY

OECD-NEA workshop on the evaluation of defects, repair criteria & methods of repair for concret e
structures on nuclear power plant s
OECD-NEA IAGE held an international workshop on the evaluation of defects, repair criteria & method s
of repair for concrete structures on nuclear power plants in Berlin, Germany on April 10-11, 2002 .
Through 2 technical sessions devoted to Operational Experience and State of the Art and Futur e
Developments, a broad picture of the status was given to a large audience composed by 54 participant s
from 17 countries and International Organisations . 21 papers have been presented at the Workshop .
The objectives of the workshop were to examine the current practices and the state of the art with regard t o
the evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe r
Plants with a view to determining the best practices and identification of shortfalls in the current methods,
which are presented in the form of conclusions and recommendations in this report .
This workshop on the evaluation of defects, repair criteria and methods of repair for concrete structures o n
Nuclear Power Plants is the latest in a series of workshops .
The complete list of CSNI reports, and the text of reports from 1993 on, is available o n
http://www.nea .fr/html/nsd/docs/

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NEA/CSNI/R(2002)7/VOL 2

Acknowledgement

Gratitude is expressed to GRS, Germany for hosting the Workshop at the DIN Institute in Berlin. In
particular, special thanks to Mr . Helmut Schulz and Dr Jurgen Sievers, and also Mrs Brunhilde Laue an d
Mrs Schneider for their help .
Thanks are also expressed to chairmen of the sessions and to the Organizing Committee for their effort an d
co-operation.
Dr Leslie M Smith
Prof Pierre Labbé
M. Jean-Pierre Touret
Herr Rüdiger Danisch
Mr James Costello
Dr Dan Naus
M.Eric Mathet

BEG(UK) Ltd
IAEA

(UK) Chairma n
(International )

EdF

(France)

Framatome ANP GmbH
USNRC

(Germany)
(USA )
(USA )
(International)

ORNL

OECD-NEA

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NEA/CSNI/R(2002)7/VOL2

10

NEA/CSNI/R(2002)7/VOL 2

OECD-NEA WORKSHOP
ON THE
EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FO R
CONCRETE STRUCTURES ON NUCLEAR POWER PLANTS
10th and 11 th April, 2002
Berlin, Germany

A.

CONTENTS

B.
C.
D.
E.

CONCLUSIONS AND RECOMMENDATION S
PROGRAMME
PAPER S
PARTICIPANTS

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NEA/CSNI/R(2002)7/VOL2

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NEA/CSNI/R(2002)7/VOL 2
A. TABLE OF CONTENTS
PAGE

Volume 1
B.
C.
D.

CONCLUSIONS AND RECOMMENDATION S
PROGRAMME
PAPER S

17
19
23

Introductory Paper
Inspection, Assessment and Repair of Nuclear Power Plan t
Concrete Structures
D. J. Naus, Oak Ridge National Laboratory, U.S.A.
H. L . Graves, J.F. Costello, USNRC, U.S.A .

25

SESSION A : OPERATIONAL EXPERIENCE
Chairman: Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany)

37

Repair of the Gentilly-1 Concrete Containment Structur e
A. Popovic, D . Panesar and M . Elgohary, AECL, Canada

39

The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures an d
Installation of an Automated Cathodic Protection Syste m
L. M . Smith, C . A. Hughes, British Energy Generation UK, Ltd .
G. Jones, Sea-Probe, Ltd.
Feasibility Study of IE-SASW Method for the Non-Destructive Evaluation of Containmen t
Building of Nuclear Power Plant
Mr. Yong-Pyo Suh, KEPRI, Korea

51

59

Field Studies of Effectiveness of Concrete Repair s
N.J.R. Baldwin, Mott MacDonald Ltd.,(UK)

73

Detection and Repair of Defects in the Confinement Structures at Paks NP P
Mr. Nyaradi Csaba, Paks NPP Ltd

83

SESSION A : OPERATIONAL EXPERIENCE (Continued )
Chairman : Dr James Costello, USNRC (USA)

11 5

Steam Generator Replacement at Ringhals 3 Containment, Transport Openin g
Jan Gustavsson, Ringhals Nuclear Power Plant, (Sweden )

11 7

In Service Inspection Programme and Long Time Monitoring of Temelin NP P
Containment Structure s
Jan Maly, Jan Stepan, Energoprojekt Prague, Czech Republi c
Repair Criteria and Methods of Repair for Concrete Structures on Nuclear Power Plants
R. Lasudry, Tractebel Energy Engineering, (Belgium )

13

12 5

13 5



NEA/CSNI/R(2002)7/VOL 2
Post-Fire Damage Assessment Procedures for Nuclear Power Plant Structures
L.M. Smith, British Energy Generation UK, Ltd ., (UK)

157

Volume 2
SESSION B : STATE OF THE ART & FUTURE DEVELOPMENTS
Chairman : Dr Naus, ORNL (US)

17

Various stages to Address Concrete Cracking on NPP s
C. Seni, Mattec Engineering Ltd ., (Canada)

19

Investigation of the Leakage Behaviour of Reinforced Concrete Wall s
Nico Herrmann, Christoph Niklasch, Michael Stegemann, Lothar Stempniewski ,
University of Karlsruhe, (Germany )

31

The Development of a State-of-the-Art Structural Monitoring Instrumentation Syste m
for Nuclear Power Plant Concrete Structure s
L. M . Smith, B . Stafford, M.W. Roberts, British Energy Generation UK ,Ltd.
A . McGown, University of Strathclyde (UK)
Ageing and Static Reliability of Concrete Structures under Temperatur e
and Force Loading
Petr Stepanek,, Stanislav Stastnik Vlastislav Salajka, Technical University of Brno ,
Jaroslav Skolai, Jiri Stastny, Dukovany Power Plant, (Czech Republic )
Efficient Management of Inspection and Monitoring Data for a Bette r
Maintenance of Infrastructur e
Marcel de Wit, Gilles Hovhanessian, Advitam

41

55

67

Aging Process Of A Good Concrete During Forty Year s
Dr. Peter Lenkei, University, College of Engineering (Hungary)

77

SESSION B : State of the Art & Future Developments (Continued)
Chairman : Mr. Jean-Pierre TOURET, EdF, (France)

81

Acoustic Monitorin g
Marcel de Wit, Gilles Hovhanessian, Advitam

83

Concrete Properties Influenced by Radiation Dose During Reactor Operatio n
Takaaki Konno, Secretariat of Nuclear Safety Commission, (Japan )

97

The Use of Composite Materials in the Prevention and Strengthening of Nuclea r
Concrete Structures
D. Chauvel, P .A. Naze, J-P . Touret EdF, Villeurbanne, (France)

14

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NEA/CSNI/R(2002)7/VOL 2

Detection of Reinforcement Corrosion and its Use for Service Life Assessment o f
Concrete Structures
C. Andrade, I . Martinez, J. Munoz, CSIC (SP) Rodriguez, M. Ramirez, GEOCISA (Spain)
Improved Detection of Tendon Ducts and Defects in Concrete Structures Usin g
Ultrasonic Imagin g
W. Müller, V. Schmitz, FIZP, (GE) M . Krause, M . Wiggenhauser, Bundesanstalt fü r
Materialforschung und -Prüfung, (Germany )
Structural Integrity Evaluation of a Steel Containment for the Replacement o f
Steam Generato r
Mr. Yong-Pyo Suh, KEPRI (Korea)

11 5

12 5

13 3

New Methods on Reconstruction of Safety Compartments of Nuclear Power Plants
Z. Kdpper, Kdpper und Partner, Bochum, (Germany)
D. Busch, RWE Solutions AG, Essen, (Germany)

14 1

E.

151

PARTICIPANTS

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NEA/CSNI/R(2002)7/VOL 2

SESSION B : STATE OF THE ART & FUTURE DEVELOPMENT S
Chairman: Dr. Naus, ORNL (US )

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Addressing concrete cracking in NPPs.
By C . Seni ,
Mattec Engineering Ltd. Canad a
Abstract
The phenomenon of concrete cracking is one of the most frequently encountered deterioration at NPPs as i t
has been shown by a wide Survey of NPPs performed by IAEA in 1994-9 5
It can be due to a multitude of causes such as the normal ageing process (shrinkage, creep, prestressin g
force loss) as well as exposure to the environment (temperature variation, moisture, freeze/thaw, etc )
The above mentioned Survey has also shown that in 64% of cases, no action was taken or required . It
became also obvious that there is a lack of guidance as when remedial actions are needed .
The paper describes, with the help of a Flow Chart, the various stages to be considered, from the first step
of identification of cracks, to the definition of causes, evaluation of extent of damage, evaluation o f
effect/implications (safety, reliability) , to the final step of deciding if repair action is required .
Finally, based upon a wide literature survey the paper proposes in a Chart format, Criteria for addressin g
concrete cracks in NPPs ., when taking in considerations all these factors .
General
Reinforced concrete structures deteriorate in various ways due to the normal ageing process ( shrinkage ,
creep, prestressing cables relaxation) and/or impact from aggressive environment (temperature, moisture ,
cyclic loading )
The rate of deterioration will depend on the component’s design, material selection, quality of construction ,
curing, and aggressiveness of the environment .
The experience gained from an international survey on ageing of Nuclear Power Plants initiated by IAE A
in 1994-95 [2] was that concrete cracking was the most frequent form of degradation . At the same time the
Survey has shown that in 64% of cases no action was required .
Various reasons could explain this lack of action which could have unpleasant implications .
In-service inspection techniques are available that can indicate the occurrence and extent of such an agein g
or environment-stressor related deterioration . Periodic application of these techniques as part of a
condition assessment program can indicate the progress of deterioration . The results obtained from thes e
programs can be used to develop and implement remedial actions before the structure attains an
unacceptable level of performance . Depending on the degree of deterioration and the residual strength o f
the concrete component, the remedial measures may be structural, protective, cosmetic, or any combinatio n
of these .

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NEA/CSNI/R(2002)7/VOL 2

A plant service life management program would normally include measures for detection and monitorin g
(inspection, instrumentation, reporting, mapping, etc,), assessment of extent of damage (measurement) and
type of damage (impacting safety, reliability), technicalities (possibility to repair, cost, duration) ,
scheduling (prioritization), and selection of a repair method .
However the world literature as well as codes and standards ,if not confusing, at least are lacking clea r
indications when to proceed with repair, taking in consideration all aspects involved .
This paper discusses the process which should lead to the selection of an effective repair method an d
proposes, based upon worldwide standards and literature, criteria which should lead to the decisio n
whether to repair or not concrete cracks, after the cracks have been identified and evaluated, addressing th e
entire range of aspects involved .
Lead way to the selection of a repair metho d
Definition of Caus e
Once cracks have been identified through periodic inspection or instrumentation, their location and
dimension recorded and evolution monitored, the next step is the definition of causes, since no repai r
should be undertaken before the cause of cracks has been identified .
Concrete ageing, external factors or simply the aspect of the cracks could provide information abou t
causes. In each of these three categories there are a number of specific indicators (Chart#1) .
(i) Cracks due to concrete ageing .
One category of cracks is the result of the ageing phenomenon of the concrete per se, and with time crack s
will multiply or increase in width, depth or length, thus giving an indication of the status of ageing of tha t
concrete.
The reasons for cracking which are time dependent could be :
- shrinkage/creep
- prestressing loss of the cables
- reinforcing bars corrosion .
The creep of the concrete will be present in particular in prestressed concrete structures and the associate d
loss of prestressing force viz cable relaxation should always be considered as a potential cause of cracking,
while this cause will not be present in conventionally reinforced concrete .
(ii) Cracks due to external factors .
Another category of cracks is the result of the various external factors independent of the concrete agein g
but associated with ageing since they have an acceleration effect upon it .
In this category fall cracks resulting from the effect of exposure to the environment :
-freeze/thaw cycle,
-chloride penetration,
-carbonation,
-aggressive environment (e .g.atmospheric pollutants, acid rain, fog, sea/ground water exposure.)
20

NEA/CSNI/R(2002)7/VOL 2
-cycling loading ( e .g. mechanical/thermal),
-construction defects (e .g.low concrete quality, excessive permeability, alkali reactive aggregatesAAR, early formwork removal ) ,
-design criteria (allowing for concrete in plastic/non-linear state, thus with acceptable crack
formation in the tension zone, insufficient knowledge of some long term acting desig n
parameters),
-thermal gradient
-accidents (e .g. loss of coolant-LOCA, fire, earthquake)
-excessive testing, (e.g. containment leak rate test )
(iii) Cracks aspect
The location or pattern (direction) off cracks could also provide indication about the reason of thei r
formation.
To recognize, viz differentiate between all these causes and heir related effect, the world literature and i n
particular the extensive work and publications produced by RILEM and IAEA can provide ampl e
information . [1], [3], [4], [5], [6], [7], [8], [13] .
Extent of Damage.
Two aspects need to be considered in the process of identification of the extent of damage i .e. the cracks
dimension and their evolution .(Chart #1) .
(i)Cracks dimension .
There are three parameters which need consideration, i .e. the width, length and depth of the cracks . Thes e
parameters are usually monitored and IAEA Survey [2] has shown that, world wide ,
- the crack width was recorded by 73 .3% of the stations,
- the length was recorded by 74% of the stations ,
- the depth was recorded by 21 .4% of the stations .
Recording crack width and/or depth is important since the crack size can affect the corrosion of the
reinforcing, which in turn will amplify the effect of the freeze/thaw cycles and the ingress of other
aggressive agents (e .g. chlorides) to the proximity of the reinforcing bars . Various authors have rated the
significance of the crack width based upon these aspects as shown in Table1, [3], [7] . From Table1 it
appears that crack widths between 0 .1mm and 0.4 mm are considered acceptable, depending on th e
environmental conditions .
However it does not address the complexity of the cracking phenomenon with all aspects involved .
(ii)Cracks activity
Recording crack length is important in order to determine their evolution, since this will indicate if crack s
are active, i.e. the process accountable for the cracks is still in effect and will have to be identified an d
addressed, while no change in the crack length ( dormant cracks) could indicate that the process ha s
stopped and may not be age related.
There are one time cracks like those due to an accidental loading or construction defects and which will no t
further develop, and cracks originated by time dependent factors like concrete “ageing per se” (e .g.
carbonation) or environmental factors (e .g. chlorides, freeze/thaw) .

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NEA/CSNI/R(2002)7/VOL 2

To the first category belong the Dormant cracks while to the second category belong the Active crack s
(Table2), [3], [8] .
This classification is important when assessing the urgency for repair or the repair method . Also in the cas e
of Active cracks the cause should be first eliminated, before a repair is undertaken .
Table 2 however is limited to a few cases and can not provide sufficient guidance .
Impact of Damage.
The next step of the process consists of establishing the impact the cracks can have upon the concret e
member in particular and upon the structure in general, since this will affect the decision whether t o
proceed or not with repairs . This step involves consideration of the following ;
Structural member ranking.
The effect of cracks will depend upon the structural member ranking which should take into account th e
importance and function of the member based upon its structural and functional role which will indicate t o
what extent the safety or functional performance are affected . Finally this will lead to the necessity an d
urgency to proceed with repair s
A few methodologies were suggested for ranking nuclear plant components, i .e. in the US according to a
program elaborated at the Oak Ridge National Laborator y
[7],[9] ,or in the UK [10], or in Canada [11], and by RILEM [8] .
These should provide the basis for each NPP to develop their own methodology .
Assessment.
Before selecting a type of repair, the last step in the process is the assessment of the necessity to procee d
with repairs or not . This should be based upon a Cracks Acceptance Criteria
The best attempt so far was made by C .J. Hookham in 1995 taking in consideration the crack dimensio n
and environmental factors (chloride penetration and depth of carbonation), as shown in Chart [6] .
Going with the complexity of the crack phenomenon one step farther is the Cracks Acceptance Criteri a
presented in this paper and as shown in a chart form in Figure 1 and 2 .
The two charts include the following parameters for consideration :
-crack width (range 0 .2 mm to 1 .0 mm) ,
-type of environment (mild or aggressive) ,
-cracks activity (active or dormant), and
-depth of chloride penetration or carbonation (low or high) .
The values set are based upon a review of the world literature and codes and standards referenced in thi s
paper.
The crack width is closely related to the urgency of repair since this represents an open path for th e
aggressive agents to reach the reinforcement .

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NEA/CSNI/R(2002)7/VOL 2
The upper limit is 0 .2mm beyond which repairs are not required under any circumstances while beyond th e
lower limit of 1 .0 mm, repairs are required in all cases . For in-between values indication are given in the
two charts.
The aggressiveness of the environment will also affect the decision . Thus an aggressive
environment ( e .g. sea shore location with high chloride content, proximity to air pollutant industries, hig h
freeze/thaw cycle, etc ) will increase the limit and associated repair urgency factor .
Regarding the cracks activity, for the dormant ones the necessity of repair will be less stringent as
reflected in the two charts .
The depth of chloride penetration and/or carbonation are indicators of how close the initiation o f
reinforcement corrosion is . Maximum permissible chloride contents, as well as minimum recommended
reinforcement cover requirements have been provided in codes and guides . The threshold of acid-solubl e
chloride contents reported by various investigators which could initiate steel corrosion ranges from 0 .15 to
1 .0% by weight of cementitious materials, whereas code limits range from 0 .2 to 0 .4%. [3],[7],[13] . In the
two charts (Fig.1 and 2) the limit selected is 0.4% .
As earlier indicated the cracks have also to be considered in connection with the importance of th e
structural member affected which will come from the Plant Components Ranking . For a high ranking
component (Figure 1) the urgency of repairs will be grater than for a low ranking one (Figure 2) .
Recommendations.
Each NPP should develop a concrete component ranking as well as a database (history) of crack repair s
performed and their effectiveness .
An experienced civil engineer familiar with concrete ageing, should be associated with the entir e
assessment process described.
Repair should proceed only after the cause has been determined and the repair should include th e
elimination of cause .
Concluding remarks
The paper has detailed the various steps to follow when dealing with concrete cracks, from the time of thei r
identification to the decision making of when and how to proceed with their repair, with Chart#1 as a guid e
through the various steps involved .
Figure 1 and 2 are the Acceptance Criteria to follow at the end of the assessment process when all required
crack parameters are known, to provide the answer if and when remedial action is required .
The selection of repair materials and method were not within the scope of this paper .
However such information, when required, can be found in the world literature or from the experience o f
other similar NPPs using inter communication grids like WANO or COG since most Utilities will have a
database of repairs performed and their effectiveness[11] .

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NEA/CSNI/R(2002)7/VOL 2

REFERENCES.
1 . A.M.Neville, Properties of concrete, Pitman Publishing .
2.
,

Summary Results of the Survey on Concrete Containment Ageing, IAEA/NENS Working Materia l
Vienna, 199 5

3.

Assessment and management of ageing of major NPP components important to safety : Concrete
containment buildings, IAEA-TECDOC-1025, June 199 8

4.

L.Granger, Assessment of creep methodologies for predicting prestressing forces losses in nuclea r
power plant containment, -RILEM Report 19/1999 , Considerations for use in managing the agein g
of NPP concrete structures .

5.

Assessment and management of ageing of major NPP components important to safety : Metal
components of BWR containment systems, IAEA-TECDOC-1181, October 200 0

6.

C.J. Hookham, In-service inspection guidelines for concrete structures in NPPs -ORNL/NRC/LTR 90 Lockneed Martin Energy Systems Inc .,ORNL,Oak Ridge,199 5

7.

D .J. Naus/C.J.Hookham, Condition assessment of concretestructures in NPPs, RILEM Repor t
19/1999 , Considerations for use in managing the ageing of NPP concrete structure s

8.

D .J.Naus/ P .D . Krauss/C .Seni, Repair techniques and materials for degraded NPP concret e
structures, -RILEM Report 19/1999 , Considerations for use in managing the ageing of NP P
concrete structures.

9.

C.J. Hookham, Structural ageing assessment methodology for concrete structures in NPPs ,
ORNL/NRC/LTR-90/17 Martin Marietta Energy Systems Inc .,ORNL,Oak Ridge,Tenn .March 199 1

10.

R.Judge, Classification of structural components and degradation mechanisms for containmen t
systems, -Proceedings of the Third International Conference on Containment Design and Operation ,
Toronto, Oct .1994

11.

K.E . Philipose, The structural aging assessment program-Ranking methodology application fo r
Candu NPP concrete components, COG-97-269, May 199 7

12 . J.A. Sato et. al , Effective repairs in concrete containment -Materials and guidelines for using join t
sealant, patching concrete and repairing cracks in CANDU containmen t
structures, COG-96-50 1
13.

C.Seni, Methodology for Assessment of Repair of Concrete Cracks and Acceptance Criteria, COG 97-202, Decmber 1997

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NEA/CSNI/R(2002)7/VOL 2
Table 1 : Permissible crack widths to prevent corrosion of steel reinforcement . [3],[7 ]
Author

Environment factors

Permissible width, m m

Dangerous crack width

1 .0 to 2 . 0
0.3

Abeled

Crack width alowing corrosion within 1/2 yea r
saline environment
Structures not exposed to chemical influences

Tremper

Found no direct relation between crack width and corrosio n

Boscard

Structures exposed to a marine environment

de Bruyn

Found no direct relation between crack width and corrosio n

Rengers

Engel and Leeuwen

Voelmy

0 .3 to 0 . 4

0. 4

Unprotected structures (external)

0. 2

Protected structures (internal)

0.3

Safe crack width

up to 0 . 2

Crack allowing slight corrosion

0.2 to 0 . 5

Dangerous crack width

over 0 . 5

Indoor structures

0 .25 to 0 .3 5

Bertero
Normal outdoor exposure

0 .15 to 0 .2 5

Exposure to sea water

0 .025 to 0 .1 5

Protected structures (interior)

0.3

Exposed structures (exterior)

0.2

Fairly harmless crack width

0. 1

Harmful crack width

0. 2

Very harmful crack width

0. 3

For all structures under normal conditions

0. 2

Structures exposed to humidity or to harmful chemical influences

0. 1

Structures subjected to dead load plus half the live load for which the y
are designed
Structures subject to deal load only

0.4

Exterior (outdoor) structures exposed to attack by sea water and fumes

0.05 to 0 .2 5

Exterior (outdoor) structures under normal conditions

0 .15 to 0 .2 5

Interior (indoor) structures

0 .25 to 0 .3 5

Ordinary structures

0.3

Structures subjected to the action of fumes and sea environment

0.2

Haas

Brice

Salinger

Wastlund

0.3

Efsen

Rusch

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NEA/CSNI/R(2002)7/VOL 2
Table 2 : General guide to repair options for concrete cracking . [3],[8 ]

Description
Dormant pattern
or fine cracking

Perceive d
durability
rating (1-5*)

Repair Options

Commentary

Judicious neglect
Autogenous healing
Penetrating sealers

4
3
2

Coatings

3

HMWM or epoxy treatment

2

Overlay or membrane

2

Epoxy injection
Rout and seal
Flexible sealing
Drilling and plugging
Grout injection or dry
packing
Stitching
Additional reinforcing
Strengthening

1
3
4
3
4
5
4
3

Needs experienced applicato r
Requires maintenanc e
Requires maintenanc e

Active cracks

Penetrating sealer
Flexible sealing
Route and seal
Install expansion joint
Drilling and plugging
Stitching
Additional reinforcing

3
3
3
2
4
4
3

Cracks less than 0 .5 mm
Requires maintenanc e
Use for wide cracks
Expensive
May cause new cracks
May cause new cracks
May cause new cracks

Seepage

Eliminate moisture source
Chemical grouting

2
1

Usually not possible
Several applications may be

Coatings
Hydraulic Cement dry
packaging

4
4

Dormant
isolated
cracking

large

Only for fine cracks
Only on new concrete
Use penetrating sealer for H20,
Cl resistance
Use coating for abrasion and
chemical resistanc e
Topical application, bonds
cracks

For severely cracked areas

necessary

May have continued seepag e
May have continued seepag e

* Scale from 1 to 5, with 1 being most durable .

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NEA/CSNI/R(2002)7/VOL 2

0 .4 < Cl
or

< 1

Carbonation Depth
Approaching Stee l

No visible degradation I
or cracking
0 < w < 0.4 mm

Concrete cracking with or
w/o staining
0 .4 < w < 1 .0 mm

I Concrete cracking ,
staining, and spalling
w > 1.0 mm

FIG. 7.3 . Damage state chart relating environmental exposure, crack width ,
and necessity for additional evaluation or repair [Ref. 7.6] .

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35

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The Development of a State-of-the-Art Structural Monitoring Instrumentation System for Nuclea r
Power Plant Concrete Structure s
LM Smith British Energy Generation(UK)Lt d
B Stafford British Energy Generation(UK)Lt d
MW Roberts British Energy Generation Lt d
A McGown University of Strathclyde
Abstract
This paper describes the development of a state-of-the-art diverse monitoring system for application to
existing concrete structures on Nuclear Power Plants . The system has the capability to monitor surface
behaviour or, in cases where surface effects are not the most critical, sub-surface behaviour can b e
monitored with minor modification to the installation arrangement .
The system uses instrument clusters with both fibre optic and AC-LVDT transducers designed to monito r
small structural displacements accurately . By simultaneously measuring the structural response usin g
transducers of different types, type-based errors may be eliminated and the system reliability enhanced .
In order to determine the type of instrumentation to be used, a research project was undertaken to evaluat e
the performance of available equipment and the practicality of its installation. This paper describes the
work carried out under the research project and the development of the system to the practical installatio n
stage.
Introduction
Historically, emphasis has been placed on the use of instrumentation to validate design and analysi s
assumptions and for initial structural integrity testing and it has been normal for the first structure of a ne w
design or series to be extensively instrumented with subsequent structures receiving less attention.
Although instrumentation has been used for long term monitoring this has often been as a result of the
continued operation of systems installed for commissioning and structural integrity tests, which have the n
been adopted for long term monitoring purposes . It is now generally accepted that the installation o f
structural monitoring systems at the time of construction will provide useful information for the lifetim e
management of nuclear power plant structures and the detection of ageing effects .
The OECD-NEA workshop on the instrumentation and monitoring of concrete structures [1] considered
this in some detail . There is now a perceived need to address the ageing of concrete structures . To this end,
the use and acceptance of instrumentation techniques has increased with time and more reliance is bein g
placed on such techniques .
The workshop recommended that, while techniques are available to monitor the ageing and performance o f
structures and new systems are available which may be retrofitted, improvements be made in retrofitting
instrumentation and in relating it to existing instrumentation . It is extremely difficult to replac e
instrumentation that was installed at the time of construction . Additionally, the usefulness of installed
instrumentation is dependent on the accuracy and reliability of the sensors used . The usefulness o f
instrumentation systems has also been improved by developments with regard to computer managemen t
systems and databases.

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The Current Project
British Energy identified a potential need to be able to measure small changes in structural displacement in
reinforced and, particularly, prestressed concrete structures using retro-fitted instrumentation . This paper
describes the development of a state-of-the-art diverse monitoring system for application to existing
concrete structures on Nuclear Power Plants . The system has the capability to monitor surface behaviou r
or, in cases where surface effects are undesirable, sub-surface behaviour with minor modification to th e
installation arrangement. The specification required the investigation of an instrumentation package chose n
to include established and innovative technologies, and instrumentation diversity and redundancy .
This paper describes the development of a state-of-the-art diverse monitoring system for application to
existing concrete structures on Nuclear Power Plants . The system has the capability to monitor surface
behaviour or, in cases where surface effects are not the most critical, sub-surface behaviour can b e
monitoredwith minor modification to the installation arrangement .
Programme of Investigation s
The project was initiated in April 2000 and was undertaken at the University of Strathclyde in three stage s
with project milestones as follows :
Stage 1
Investigation and definition of the monitoring requirements . Preparation of proposals for the
instrumentation required. This work was completed by mid - May 2000 .
Stage 2
Laboratory investigations at the University of Strathclyde and preparation of an Interim Report o n
the outcome of the investigations . This work was completed in late-December 2000 and an Interi m
Report was issued in January 2001 . A recommendation of the Stage 2 work was that an extension to
Stage 2 be granted to investigate the installation problems associated with the use of Automati c
Crackmeters penetrating down below the surface of the concrete some 50 to 100 mm .
Stage 2 (extension )
The Stage 2 (extension) was undertaken during April and May 2001 . This work involved developing
two types of Automatic Crackmeters, installing them in specially constructed reinforced concret e
beams and testing their measurement efficiency .
Stage 3
The Final Report set out the performance and accuracy of the instrumentation and include d
recommendations for the instrumentation system to be used on NPPs .

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NEA/CSNI/R(2002)7/VOL 2
Results
In Stage 1 of this project the normal operational conditions on Nuclear Power Plants were assessed and th e
most likely types of instrumentation to effectively undertake the monitoring were identified . In Stages 2
and 2(extension), evaluation of these instrument types and the methods of installing these wer e
investigated (Plate 1) . Stage 3 involved the preparation and submission of the Final Report .
The Stage 2 investigations were designed to measure vertical distortion, in-plane deformations / strains an d
temperatures on a centrally loaded, simply supported 3m x 1m x 100mm lightly reinforced concrete beam .
The instrumentation package was chosen to include established and innovative technologies ,
instrumentation diversity and redundancy . The range of instrumentation employed is shown in Table 1 .
The data on the claimed resolution, frequency of measurement, mounting and development stage of each
instrument are shown in Table 2 . The details of the installation methodology, tuning/ commissioning an d
data collection technique associated with these instruments are given in Table 3 . The layout of the
instrumentation employed on the beam is given in Fig .1 and illustrated in Plate 1 . In addition to thes e
instruments, ambient air temperature was measured using PRT’s above, below and at the side of the beam .
The objectives and details of the testing regime are shown in Table 4 . Incremental vertical distortion tes t
data was obtained from two separate tests, involving four stages of short-term loading with a single stage
of unloading . In addition, tests were conducted to compare the analogue and digital outputs from the fibre
optic (Fabry-Pérot interferometer) instruments . The reason for undertaking these tests was to prove that
analogue signals could be obtained from the fibre optic instruments, as this had not been established
previously. The main findings of the testing undertaken under Stage 2 were :
Vertical and in-plane deformations could be measured by the instrumentation employed to an
accuracy of +/- 0 .01mm, (+/- 10 microns) .
In-plane strains could be measured to an accuracy of +/- 2 microstrain .
Temperature could be measured to an accuracy of +/- 0 .1 o C .
AC- LVDT’s and fibre optic deformation gauges were the preferred instruments to measur e
deformations .
Whilst highly accurate and possibly extremely useful in the future, some further development wor k
was required on the TMS laser system before it could be employed for structural monitoring .
VW strain gauges and fibre optic strain gauges were the preferred instruments for directly
measuring strain .
Digital outputs from the instrumentation signal conditioners were preferred as they allowed muc h
greater flexibility in data collection and networking . However, it was shown that analogue output s
may be obtained with similar accuracies to digital outputs .
Instrumentation should not be surface mounted to avoid superficial cracking and de-bonding
problems .
The use of Automatic Crackmeters should be considered to measure in-plane deformations /
strains . The pillars and instrument fixings should be manufactured from a temperature
compensated alloy such as Invar (Fe/Ni alloy) to reduce potential temperature effects .

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NEA/CSNI/R(2002)7/VOL 2

The temperature variations at the location of each instrument should be measured and temperatur e
compensations / corrections applied .
It was recognised that potential installation problems have a major influence on the choice o f
instrumentation and consequently on the measurements to be made . Indeed, it was realized that the need to
create a horizontal reference plane against which out of plane distortion could be measured, was not
practical at many NPP locations . Thus measurement of out of plane distortion was not recommended an d
the choice of instrumentation was limited to those instruments that could measure in-plane deformation s
and strains .
It was identified that superficial cracking of the concrete, free surface effects and de-bonding o f
instruments on the surface of the concrete could result in mis-leading data being collected . Thus it wa s
agreed that the instrumentation packages to be used would require to be capable of measurement at a subsurface level . It was recommended that sub-surface deformations and strains should be measured using
both conventional electrical and optical fibre instruments . It was decided that measurements should be
made at a depth of 50 to 100mm below the top surface of the concrete to avoid surface effects influencing
the data collected .
The study recommended that sub-surface deformations should be measured using Automatic Crackmeter s
consisting of pillars drilled and fixed into the concrete with the change in distance between pairs of pillar s
measured using external AC electrical LVDT’s and fibre optic deformation gauges . The types of
instruments should be similar to those used in the Stage 2 investigations, however, they should b e
individually temperature monitored to check for local temperature variations on the top surface of th e
concrete.
Sub-surface strains should be measured using temperature compensated VW strain gauges and optical fibr e
strain gauges fixed within Automatic Crackmeters .
Ambient air temperatures immediately above the concrete surface and concrete temperatures at 25 to
50mm below the surface should be measured.
Although the correlation between analogue and digital outputs from the signal conditioners was good, it
was recommended that digital outputs should be the preferred form as :
Links between instruments and signal conditioners were simpler .
Site specific problems of mutual interference and external interference were avoided .
Networking of instrumentation using RS422 / RS485 links was possible .
The Stage 2 (extension) investigations were designed to measure in-plane deformations / strains an d
temperatures on a centrally loaded, simply supported 3m x 1m x 150mm reinforced concrete beam. The
instrumentation packages used were designed to provide examples of possible combinations of establishe d
and innovative technologies with instrumentation diversity and redundancy .
The two instrumentation packages used were :
A four instrument package, consisting of a VW strain gauge (subsurface), a fibre optic strain gaug e
(subsurface), an AC LVDT (external) and a fibre optic deformation gauge (external), and ,
A two instrument package, consisting of an AC LVDT (external) and a fibre optic deformatio n
gauge (external) .

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The arrangements of the instrumentation packages are shown in Plates 2 and 3 and the layout of th e
instrumentation was as shown in Figure 2 .
Test results were obtained from a nine stage loading and single stage unloading test undertaken over seve n
days . These indicated that for lightly loaded and unloaded conditions there were identifiable variations o f
the outputs, (“instrumentation noise”), of +/- 0.002 mm (+/- 2 microns) and +/- 1 microstrain . In addition,
creep of the concrete could be identified over time . Thus the main findings of the testing undertaken were :
The two types of instrumentation packages could be installed on the beam efficiently .
Owing to the use of subsurface instruments, the installation of the four-instrument package
involved a great deal more drilling than the two-instrument package where the instruments ar e
external to the concrete surface and drilling is only required for the fixing posts . .
The detailed design of the instrument packages and the adjustment /calibration of the instrument s
required great care to ensure that the instruments were not damaged during installation an d
operated efficiently in place .
To calculate the strains in the concrete from the LVDTs and fibre optic deformation gauge outpu t
data, the gauge length was taken as the distance centre to centre between the vertical pillars .
To calculate the strains in the concrete from the VW and fibre optic strain gauge output data, th e
gauge length was taken as the distance centre to centre between the vertical pillars and the strain s
recalculated on this basis. All data were efficiently recorded using digital outputs, ( i .e. using the
recommended / preferred method of Stage 2) .
In plane deformations could be measured by the LVDTs and fibre optic deformation gauges to a n
accuracy of +/- 0 .01 mm, (+/- 10 microns) .
Background noise in the LVDTs and fibre optic deformation gauges was at a level of +/- 0 .002
mm, (+/- 2 microns) .
In plane strains could be measured by the VW and fibre optic strain gauges to an accuracy of +/- 2
microstrain .
Background noise in the VW and fibre optic strain gauges was at a level of +/- 1 microstrain .
Creep occurring in the concrete over time was reflected in a gradual change in the outputs from th e
instrumentation .
Final Recommendations
Instrument packages consisting of a VW strain gauge, a fibre optic strain gauge, an AC-LVDT and a fibre
optic deformation gauge should be used . The strain gauges should be temperature compensated and b e
embedded into the concrete to a depth of 50 to 100mm using a shrinkage compensated cementitious grout
and an epoxy “jacket” to isolate the instruments from local effects . The AC-LVDT and fibre optic
deformation gauge should be mounted above the surface, PRTs should be embedded at 25 and 50 mm t o
measure concrete temperatures . PRTs should be mounted next to the bodies of the AC-LVDTs and fibr e
optic deformation gauges to measure local temperatures and so allow temperature corrections to be made .

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NEA/CSNI/R(2002)7/VOL 2
It should be noted that the instrumentation package gauge length must be precisely determined to allow th e
effective gauge length of the strain gauges to be calculated and so their strain outputs corrected . The
instrumentation package gauge length is used to calculate strains from the LVDT and deformation gaug e
outputs.
An alternative reduced package may be used where space is restricted. This should consist of an AC
electrical LVDT and a fibre optic deformation gauge placed one above the other between two vertica l
pillars . The instruments should be placed just above the top of the concrete surface . The vertical pillar s
should be embedded into the concrete to a depth of 50 to 100mm . PRTs should be embedded at 25 and 5 0
mm to measure in concrete temperatures and a PRT should be mounted next to the bodies of the ACLVDT and fibre optic deformation gauge to measure local temperatures and so allow temperatur e
corrections to be made. It should be noted that the instrumentation package gauge length must be precisely
determined to allow calculation of the strains from the AC-LVDT and deformation gauge outputs .
Specified types of VW and fibre optic strain gauges, AC-LVDTs, fibre optic deformation gauges and PRT s
which have been proven to be accurate and reliable for the purpose of monitoring deformations and strain s
at the very low levels required should be employed in the instrumentation packages .
The installation method should be very carefully specified to provide holes in the concrete into which th e
instrumentation packages will fit. Thus close tolerances should be specified on hole sizes, spacing an d
orientation.
The instrumentation packages should be designed and constructed to close tolerances to ensure :
The instruments are not damaged during fabrication .
The instruments can be calibrated before installation and, at least for the AC-LVDTs an d
deformation gauges, after installation .
The instruments can be easily set to any specified point in their range prior to installation .
The grouting of the instrumentation packages into the concrete should be very carefully specified an d
controlled . (The size of the holes in relation to the size of the vertical posts must take account of th e
grouting process) .
The instrument package and associated cabling should be very carefully protected from damage after
installation .
Calibration of the instrumentation should be very precisely undertaken at a range of movement and rang e
of temperatures appropriate to the operational conditions expected in service .
The output from the instrument signal conditioners should be digital . This provides much greater flexibilit y
in data collection, allows networking and avoids site-specific interference problems . The use of the
analogue signals is possible if other factors preclude the use of digital signals . The signal conditioners to b e
used should be those specifically designed by the manufacturers of the instrumentation for the variou s
instruments . They should be connected to the instruments with the appropriate cable type and length . The
cabling should have the minimum possible number of joints/connections .
The conditioned digital signals should be collected and processed in a PC using specialist data collectio n
and processing software with the capacity to collect, process and store data at an appropriate rate for a
specified period . The PC thus requires to possess sufficient processing and storage capacity.

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NEA/CSNI/R(2002)7/VOL 2
“Instrumentation noise” and evidence of creep in the concrete over time, are likely to be a feature of the
outputs from all the instruments . Thus it will be necessary to monitor structures in order to identify thei r
“normal engineering behaviour” .
In-service Performanc e
A number of the two-instrument external instrument packages have been installed on actual NPP structure s
and these are giving very good results with levels of noise and accuracy comparable with the laborator y
tests .

References
1. OECD Nuclear Energy Agency, NEA/CSNI/R(2000)15 “Proceedings of the Workshop o n
Instrumentation & Monitoring of Concrete Structures 22-23 March 2000, Brussels, Belgium .”

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NEA/CSNI/R(2002)7/VOL2

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NEA/CSNI/R(2002)7/VOL 2
Ageing and Static Reliability of Concrete Structures under Temperature and Mechanical Loadin g

Petr Stépanek a, Stanislav 9stni d , Vlastislav Salajka, b, Petr Hradil b, Jaroslav Sol , c, Jifi
9fastnÿ c
a Department of Concrete and Masonry Structures, Technical University of Brno, Ùdolni 53, 60200 Brno ,
Czech Republic, e-mail: stepanek .p@fce .vutbr.cz
b Department of Building Mechanics, Technical University of Brno, Ùdolni 53, 60200 Brno, Czec h
Republic, e-mail : salajka [email protected]
c
Power Plant Dukovany, 675 50 Dukovany, Czech Republi c
d
Department of Technology and Building Materials, Technical University of Brno, Ùdolni 53, 6020 0
Brno, Czech Republic, e-mail : [email protected] .cz

Abstract
The contribution presents some aspects of the static reliability of concrete structures under temperatur e
effects and under mechanical loading. The mathematical model of a load-bearing concrete structure wa s
performed using the FEM method . The temperature field and static stress that generated states of stres s
were taken into account. A brief description of some aspects of evaluation of the reliability within th e
primary circuit concrete structures is stated . The knowledge of actual physical and mechanical
characteristics and chemical composition of concrete were necessary for obtaining correct results of
numerical analysis .

Studied problems were divided into a number of fields and worked out in details :
1. Verification of contemporary physical and mechanical characteristics of concrete (input parameters o f
the FEM models) .
2. Checking of the concrete microstructure and verification of the grade and kind of possibl e
microstructure changes .
3. Experimental verification of the boundary conditions from point of view of the temperature field and
the radiation stress .
4. Setting up a mathematical model of the structure for an examination of the interaction of temperatur e
and static stresses (finite element method, software ANSYS) in two alternatives :
a) Macro-model representing the essential part of concrete structures in the proximity of the reactor ,

b) Model of extremely stressed parts of the concrete structure (a part of the macro-model) .
1.

Introduction

The nuclear power plant Dukovany (EDU) has been in use under reliable operation for more than 15 years .
Within the programme Harmonisation whose aim is to ensure high-quality and safe operation of thi s
nuclear power plant at least until the year 2025 so for this reason there has been disposed a great number o f
tasks concerning various areas. Actual static reliability of concrete structures is besides others one of the
problem of the power plant building part . Considering the fact that concrete structures have to b e
functional, safe and reliable for substantial time period after the operation of the nuclear power plant ther e
have been worked out large-scale procedures and models for evaluation of particular - for the nuclea r
power plant reliability - dominantly important structural parts . The problems of evaluation of the concrete
structure reliability are solved from the experimental and theoretical point of view .

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NEA/CSNI/R(2002)7/VOL 2

2.

Experimental par t

Concretes of the load-bearing structures of the primary circuit are effected not only by the mechanica l
stress but also by the moisture stress . Moreover, these concrete structures are also subjected to long-period
influence of high temperatures during their lifetime . Due to the present temperature and moisture stresses ,
new crystalline formations inside the concrete structure can develop (e .g. 11 A-tobermorit) and so one part
of evaluation works was focused on observations of the actual physically- mechanical characteristics o f
concrete. The experimental part of the work was especially concentrated on the following areas :
-

-

Determination of the distribution of temperature field on the surface of the concrete structures whic h
serves as the boundary conditions for temperature field (calculated by the means of mathematical
models )
Determination of the distribution of moisture field inside the concrete structures . It is used as the
constraint condition for moisture field determination by the means of mathematical mode l
Determination of the actual physically-mechanical characteristic s
Determination of different degradation effects on the concrete structur e
of the primary circuit .

2 .1 . Measurements in situ and laboratory test s
There were carried out
-

-

Measurements of concrete moisture (by gravimetric and by neutron method) ,
Determination of the physically mechanical characteristics of concrete by laboratory tests o n
samples obtained by drilling (volume mass, modulus of elasticity, stress-strain diagram under
the temperature stress, characteristics of temperature and moisture expansivity, determinatio n
of the content of soluble boron salt in concrete etc .),
Evaluation of radionuclide level activity of concrete ,
Measurements of thickness within the non-hermetic internal protective lining in the zones o f
the moisture content measurement of concrete by the neutron method .

Concrete samples were drilled from predefined areas and according agreed schedule within the period
1997-2001 (diameter of the samples was 100 mm) . The other samples were fragments of the external
surface of concrete under the protective lining used for determination of the concrete structure moisture .
The range of measured moisture values within the observed period was roughly the same . However,
significant differences were among the moisture values measured on the same areas . This proved
considerable changes of the moisture conditions - Fig. 1 . It is possible to state that the moisture migration
inside concrete occurs during the time period . From more measurements carried out in the identical area s
within one shutdown it is evident that the moisture varies in time dependency .

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NEA/CSNI/R(2002)7/VOL 2

Following physically mechanical concrete characteristics of the primary circuit were found at the analyse s
- Volume mass (density, thickness) under natural conditions p
- Compression strength
fc

- Modulus of elasticity

Ec

2192 to 2330 kg/m 3
48 .1 to 60 .8 MPa,
25 .2 to 33 .4 GPa .

Furthermore, the temperature and moisture expansivity characteristics were measured and the stress-strai n
diagrams of concrete were verified (including the decreasing branch) under the temperature stress .
Considering the fact that there were prescriptions and records available concerning the evident concret e
tests within the time of the nuclear power plant construction, it was possible to compare the origi n
physically mechanical characteristics with the ones after 15 years operation . No significant difference s
were found .
2.2. Conclusions of the experimental par t

From the experimental tests of the steel and concrete samples and from the measurement in situ i t
is evident that
-

-

-

-

3.

o
temperatures of the concrete structure exceed 100 C in some areas ,
the migration of moisture inside concrete demonstrates itself within the time of shut-down, which was
found out by comparison of the moisture in the identical areas of the RC load-bearing structure at th e
block 2 ,
amount of boron was found in concrete by physically chemical tests ,
in the course of the nuclear power plant operation there have not appeared any substantial physically mechanical characteristic changes of concrete within the observed structures of the primary circuit ,
the incidence of pitted corrosion was found in the samples of non-hermetic protective lining (it doe s
not have any static importance but it only forms the protection of concrete against the contaminatio n
at purification during the shut-down),
the appearance of artificial radio-nuclides in the samples of concrete taken from the structure of th e
primary circuit was found . The radionuclide content in samples is so low that it does not influence th e
physically mechanical properties . Also from the point of view of the State Supervision of Nuclea r
Safety (SÙJB) classification there is not dealt with any emitter,
the appearance of CSH-gels and 11-A-tobermorit in the samples of concrete that only demonstrate s
the present existence of the increased moisture and temperature at the time of its origin .
Modelling of the structure behaviou r

Consistent with the observation of the structure behaviour of all four blocks of th e
nuclear power plant it was stated tha t
-

-

-

all EDU blocks can be considered to be identically thermally stressed and that is why it is no t
necessary to distinguish the temperature stresses of the individual blocks . The differences o f
temperature about ±5 C in the extremely stressed areas cannot be considered as substantial ,
in some areas of the concrete structure the temperature of 50 oC is exceeded in long terms. It is the
temperature that causes the decrease of the concrete strength in accordance with the standard • SN 73
1201-86,
locally variable moisture was found in the concrete structures close to the reactor .
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NEA/CSNI/R(2002)7/VOL 2

3.1. Types of loading
3.1.1 . Radiation loading
The detailed analysis of the radiation influence on the concrete structures reliability was carried out detailed theoretical summary can be found in [4], [5] . According to the American standard ANSI/ANS-6 .41985 [6] the radiation influence on the primary shielding is minimal on the condition that the density of th e
temperature flow of energy does not exceed 10 10 MeV . cm-2 . s-1 = 16 W m2. According to our carried out
calculations, the density of the temperature flow of energy at the primary concrete shielding entry is 71 4
times lower than the value considered by the USA standard. The literature states that the radiation flow o f
one mW cm-2 results into the temperature increase of concrete by approximately 1 .5 C. On this assumption
(and at the validity of the linear dependence between the temperature and energy of the falling dow n
radiation), the temperature increase on the surface of the primary concrete shielding of the reactor would
be 0.0036 o C.
On the base of these calculations it is possible to neglect the influence of the internal sources of the
temperature inside concrete initiated by the radiation on the temperature field values when it is solved b y
the FEM method.
3.1.2. Moisture loading
Considering the fact of random moisture migration inside the concrete structure, the influence of th e
moisture expansivity on the state of stress of the structure was neglected at simplifying modelling . Up
today, there have not been available reliable time-dependent changes of the moisture during the tests .
3.1.3. Other types of loading
The limiting importance for the state of stress evaluation, function reliability and durability of the concret e
structure has the distribution the temperature field inside the concrete structure and by this generated th e
state of stress (of course, in the co-action with the mechanical, radiation and moisture stress) . That is why
we solved the problem s
n

n
n

relating with the specification of a mathematical model of the parts of the primary circuit structures ,
which are based on:
1. definition of the boundary conditions of temperature field within the mathematical model of th e
primary circuit structures,
2. theoretical analysis of the determination of the concrete structure referential temperature . The aim
was to formulate the theory needed and to define the referential temperature field within th e
solved structures in dependence on their thickness ,
3. final solution of the problem considering the influence of random temperature sources insid e
concrete initiated by absorption of neutron radiation and by gamma emission ,
4. complementation of some missing data relating to geometry, material characteristics and th e
composition of some structural parts.
application of the specified model for solution of the particular situation defined after the discussio n
with the nuclear power plant workers ,
evaluation of the actual static reliability of the concrete structures within the above stated situations.

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3.2. Model of the Structure
Based on the experimentally obtained data (the boundary conditions of the temperature field), followin g
amendments on the mathematical model of the primary circuit structures were carried out taking into
account the comments issuing from the discussion with some EDU department s
-

Geometry improvements of the modelled structure (more exact model) .
Physically-mechanical characteristics improvements of some materials in accordance with results o f
the carried out experimental tests .
Specification of the boundary conditions of the temperature field according to the measured result s
(non-contact thermometers, thermo-vision, standard and non-standard measurements) .

Following calculations were carried out :
calculation of the temperature field distribution ,
calculation of the state of stress generated by the temperature field in the steady state,
calculation of the mechanical state of stress under the static load,
calculation of the total state of stress (under the temperature and static load) .
Following design states were taken into account :
n STANDARD : the standard operating situation that corresponds to the steady (time-independent )
behaviour of the concrete structures during testing procedures when the influence of the reactor
starting-up operation is not substantial (stationary problem). This situation is defined b y
-

-

the boundary conditions of the temperatures measured (non-contact thermo-meters,
thermo-vision, standard and non-standard measurements performed on the modelled part of th e
structure),
the dead load, the technology load (machinery parts loading the structures) and by the boro n
content of a reservoir ,
consideration of the creep influence on the state of stress generated by the temperature stress (the
temperature loading is assumed as the long-terms one) .
LPT 30 : the hypothetical operating situation that corresponds to the steady behaviour of the
concrete structures during the testing procedures and that is defined by
the dead load, the technology load (machinery parts loading the structure) and by the boro n
content of a reservoir ,
final definition of the temperature boundary conditions that correspond to the temperature fiel d
increasing values on the measured ones . There is assumed : the temperature increase in the reacto r
zone by 300 C, zero temperature increase on the boundary of the solved situation (solved part)
and the linear course of the temperature increment among the boundary values defined by above
mentioned two items ,
the consideration of the creep influence on the state of stress generated by the temperatur e
loading .
KPT 30 : the hypothetical operating situation that corresponds to the steady behaviour of th e
concrete structures during the testing procedures and that is defined by:
the dead load, the technology load (machinery parts loading the structure and by the boro n
content of a reservoir ,
final definition of the temperature boundary conditions that correspond to the temperature fiel d
increasing values on the measured ones . It is assumed: the temperature increase in the reacto r
zone by 300 C, zero temperature increase on the boundary of the solved part and the quadrati c
approximation course of the temperature fields of the concrete structure s

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n

KPT 30 KR : the hypothetical operating situation that corresponds to the steady of behaviour of th e
concrete structures during the testing procedures and that is defined as the same as the situatio n
KPT 30 with the only difference when the short-time temperature stress is assumed (e .i. the creep
influence on the state of stress generated by the temperature stress is neglected) .
LPT 30 KR : the hypothetical operating situation similar to LPT 30 . The short-time temperature
loading is assumed .
For details see [1], [2], [3] .

n

3.3. Conclusions of the experimental part
From the works that were carried out in the years 1998-2001, which dealt with the problems of th e
concrete structures reliability on the primary circuit, can be concluded :
according to the carried out evaluation of the static reliability in accordance with the Czech standard • SN
731201-86 at the steady temperature operating regime can be stated that all the structural parts meet th e
requirements of the calculated stresses for following combinations :
n

4.

situation STANDARD (dead load + technology + temperature in the steady operating regime )
dead load + technology + temperature in the steady operating regime increased by 30 0 C in the
reactor zone (linear course, loading situation LPT 30) ,
dead load + technology + temperature in the steady operating regime increased by 30 0 C in the
reactor zone (quadratic course, loading situation KPT 30) .
satisfactory compliance among the results of the numerical temperature field solution an d
measured values was found .
Conclusion

By solving of the described interactive problem we ca n
â obtain values of the deformations and internal forces of the concrete structur e
â analyse the influence o f
• the radiation, temperature and static stress acting on the structure
• the physical non-linear behaviour of concret e
• the shrinkage and creep of concrete .
The results of analyses can serve as the base for correct computing of time reliability and lifetime
prognosis of the concrete load bearing structures .
It was confirmed that the set-up mathematical models would be possible simply to extend for the cases
using the simulation method and for evaluation of the concrete structure reliability in which some of the
input data are considered as random . At present, input statistic data are gathered for application of these
models . But the substantial complication will be – regarding the extent of the solved task - the tim e
demand of the calculation, but even this problem can be solved – [7], [8] . Application of the non-linear
modelling is another sphere of the result accuracy specification .

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In regard to the extreme loading (mechanical, temperature, moisture and radiation), owing to importance o f
the concrete structures and in view of extremely high demands on the reliability (that must be regularl y
controlled during the nuclear power plant operation and even after its closing as well), there appears as th e
only possibility of the experimental monitoring combination of important properties of a concrete structure
with numerical verifying of the structure actual reliability by the help of mathematical modelling (wit h
regard to the actual physically-mechanical characteristics) in the future . In connection with this method o f
the reliability conclusive evidence there are being prepared :
-

application of the totally reliable access for the reliability evaluation of the concrete structure s
(mathematical modelling, simulation methods – input data considered as random values/fiel d
preparation and completion of a complex of tests for fast models of ageing (degradation processes ,
complex of referential samples) .

Acknowledgements
This contribution has been prepared on the base of the scientific research order „Problems of Thermally
Stressed Structures in the Nuclear Power Plant in Dukovany“ and the research project CEZ 322/98 2 6
100007 „Theory, Reliability and Defects of Statistically and Dynamically Stressed Structures“, Faculty o f
Civil Engineering.

190, 0

PG4PG 5

180, 0
170, 0

PG2PG 3

160, 0
PG6PG I

150, 0
140, 0

PGIPG2

130, 0
120, 0

PRû
MER

110, 0
100, 0
90, 0
80, 0
70, 0
60,0
4 .5

14 .5

24 .5

3 .6

13 .6

23 .6

Fig. 1 : Time-dependence of moisture on a tested are a

61

3 .7

13 .7

23 .7

2 .8

12 .8

dat e

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Literature
1]

Analysis of problem of loading of building structures in power plant with reactor VVER 440 . Final report 1997,
ÙBaZK FAST VUT v Brn , 12/1997

2]-

[3] Temperature and mechanical stress and reliability of concrete structures . Final reports 1998, 1999 an d
2000. ÙBaZK FAST VUT v Brn , 12/199 8

[4] Kaplan M .F. : Concrete Radiation Shielding (Nuclear Physics, Concrete Properties, Design and Construction) ,
Longman Scientific & Technical, England, 198 9
[5] Hubbell J .H.: Photon Mass Attenuation and Energy-absorption Coefficients from 1 keV to 20 MeV, National
Bureau of Standards, Washington, USA, 198 1
[6] ANSI/ ANS – 6 .4 – 1985, American National Standard for Guidelines on the Nuclear Analysis and Design of
Concrete Radiation Shielding for Nuclear Power Plants, American Nuclear Society, Illinois, USA, 198 5
[7] IAEA : Safety Aspects of Nuclear Power Plant Ageing . IAEA-TECDOC-540, 199 0
[8] IAEA : Data Collection and Record Keeping for the Management of Nuclear Power Plant Ageing . Safety Serie s
No. 50-p-3, 1991

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Fig. 2 : Distribution of the temperature field on the concrete structures surfac e

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Fig . 3 : State of stress (principal stress) of the concrete cantilevers initiated by the temperatur e
loading

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Fig. 4 : Total state of stress (principal stress) of the concrete cantilevers (interaction of mechanical an d
temperature loading
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Efficient management of inspection and monitoring data for a better maintenance of infrastructur e

Marcel de Wit 1

Gilles Hovhanessian2

Area Manager

Deputy general manager ,

Advitam, Northern Europe

Advitam, Paris

Keywords : management, monitoring, inspection, maintenanc e

Abstract
In North America, Europe and Japan, government agencies and larg e
private owners are now facing the challenge of maintaining, with limited
resources, large stocks of vital structures like traditional and nuclea r
power plants, Cooling towers but also highways, railways, bridges ,
dams, harbors, industrial facilities etc… These structures are
representing a large amount of money, have not been designed to b e
easily repaired or replaced, and are getting older and more vulnerable .
People involved in structure management have developed extensiv e
technical methods and tools to monitor the condition of the structure and establish the diagnosis . Each
authority has been developing is own inspection maintenance procedures, taking into account thei r
specificity, their different priorities, safety requirements, resources and range of competence .
In most cases visual inspections are used to detect deteriorations, to rank structures, define priorities ,
estimate repair costs, etc… These visual inspections require to record, report, analyze and store for year s
large quantities of data (inspection records, drawings, photos…) and it is easy to get lost in the clerica l
work . Moreover a number of decision steps (inspection record, ranking of defects, long-term analysis) ar e
still highly subjective and can greatly affect the quality of the final diagnosis .
An inspection-based management software system has been developed to optimize this process an d
provide decision-makers with objective information on the condition of the infrastructure . The system is a
comprehensive management system which integrates : database of structural defects, on-site computerize d
record, analysis, maintenance, diagnosis, repair and budgetary functionalities .
This paper describes the basic functions and benefits of the system .

1

[email protected]

2

ghovhanessian@advitam-group .com
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1. LIMITATIONS OF CONVENTIONAL INFRASTRUCTURE MANAGEMENT PROCES S
The characteristics and limitations of, still widely used conventional structure management process ar e
listed below:
-

Design data (drawings), inspection data, detailed investigation data and repair data are not stored
in a single system.

-

Inspection frequency for a given structure is based on the type and age of the structure . It is rare
that the date of inspection is based on the results of the previous inspection .

-

Before inspection, inspectors must prepare inspection drawings - very often original desig n
drawings are not available and inspectors must spend time to make new drawings .

-

During inspection : the inspectors take hand-written notes of the defects . Inspectors usually do not
bring with them the heavy reference manuals .

-

Back in office the inspectors copy the deterioration onto the structural drawings, along with their
dimensions and characteristics . Sometimes these data are stored electronically (excel sheets an d
CAD drawings) .

-

In accordance with the inspection manual, a ranking indicator or a comment is affected to eac h
deterioration . The inspectors then establish reports that are transmitted to the engineers in charg e
of the analysis.

-

The engineers receive several reports from different inspectors . They may have difficulties with
inconsistent data, inhomogeneous ranking systems, unreadable handwriting or confusin g
dimensions . However on the basis of these reports the engineers must estimate the condition of the
structures and recommend actions for maintenance or repairs .

-

When needed, detailed investigations are performed by specialized consultants and specialize d
contractors propose repair solutions .

-

Maintenance and repair costs are then presented to decision makers .
After reviewing the above points, it becomes clear that even with a clear inspection manual and a n
efficient organization, conventional infrastructure management process allows too much room fo r
subjectivity and conventional infrastructure management is expensive .

2. OBJECTIVES OF THE INSPECTION-BASED MANAGEMENT SOFTWAR E
-

The inspection-based management software has been developed with the following objectives :
improve the overall efficiency of the maintenance process ,
reduce the cost of maintenance process at all steps ,
build a comprehensive database system which integrates all steps of the maintenance proces s
(inspection preparation - inspection - reporting - analysis - repair - budget) ,
facilitate the task of inspectors ,
assist engineers in the compilation and analysis of large quantities of data ,
allow easy access to all data at any step of the engineering and decision process ,
provide decision makers with valuable and objective information on the condition of th e
infrastructure, on which they can base and justify their decisions .

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3 . BASIC DESCRIPTION OF THE INSPECTION-BASED MANAGEMENT SOFTWAR E
The inspection-based management software consists of several components specifically designed t o
handle the tasks of each people involved in the maintenance process (Table 1) .
Software components

Infrastructure Management

Inspection Management

Inspection

Tasks
- database of detailed informatio n
on structures (drawings, designconstruction-inspection-repair
data),
- ranking of structures according
to preset rules,
- management of the database,
- budgetary tools ,
- scheduling of tasks ,
- management of repair works
- preparation of inspections ,
- transfer of data from mainfram e
to mobile inspection units
- on site recording of deterioratio n
(on pen-touch light computers)
reference availabl e

Photo-based inspection
Report
Analysis

- time-effective survey of
deterioration based on photos
- reporting of site-records ,
- automated standard report
- advanced analysis functions ,
- detailed investigation ,
- detailed repair definition

Designed fo r

- operators of structure ,
- consultants ,
- specialized inspection companie s

- inspectors ,
- specialized inspection companies ,
- consultants

- engineers,
- consultants ,
- contractors

Table 1 : Components of the inspection-based management software

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Fig.1 : General information tab

3.1 Infrastructure management software
With the infrastructure management software, operators can organize their inventory of structures . The
system allows to build a database including :
-

general information about each structure (design, construction, location, pictures, drawings – se e
Fig .1) ,
detailed check-lists for each structural component (see on Fig .2 an example for expansion joints) ,
damage criteria for ranking of deteriorations (see Fig .2 where limit values for joint opening ar e
defined) ,
catalogue of repair solutions and corresponding costs and durations (see Fig .3) .

The software has been built so that all parameters can be changed and adjusted to the specific usage o f
each industry/administration/operator. For example : the limit criteria for crack opening is smaller in nuclea r
containment vessels than in highway tunnels . Such a limit criteria can be set in accordance with th e
corresponding regulations .
.

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Fig .2: Definition of check-points and damage criteria for expansion joint s

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Fig.3 : definition of tasks, with corresponding costs and scheduling
3.2 Inspection softwar e
The inspection software is designed to simplify the task of inspectors and ensure coherent and regula r
deterioration surveys and maintenance records (Fig .4). Running on pen-touch computers (Fig .5), the
inspector can access at any time :
-

the inspection reference manual ,
the drawing of the inspected structure ,
the history of the deterioration that was recorded in past inspections ,
the maintenance check-list for each structural component .

When the inspector detects a deterioration, he draws with the pen directly on the computer screen the shap e
of the deterioration. He can use a reference of more than 200 deterioration types classified in families . In
accordance with the inspection manual, the software then requests the inspector to measure and record a
certain number of parameters to describe the deterioration (dimensions, color, humidity level…) . He may
also want to take pictures of the deterioration . The software saves all this information in the database :
-

the deterioration type ,
its graphical representation in the CAD drawing,
its dimensions and other specific parameters ,
any picture of the deterioration.

Because these data are linked together in the database, it will be easy to access and sort such information a t
any step of the maintenance process, even years after the inspection .

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Fig .4 : Site inspection software running on pen-touch computer s
3.3 Photo-based inspection softwar e
In some cases when access is difficult, survey of deterioration can be done using high quality digita l
pictures . The photo-based inspection software has the following functions :
-

correction of the lens deformation of the picture,
deformation/scaling of image,
on-scale insertion of image data into CAD drawing ,
highlighting of typical deteriorations (rebar, cracks… )

Photo-based inspection is often a cost-efficient alternative for the inspection of large structures (dams ,
cooling towers, etc… )

Fig.5: Site inspection using pen-touc h
computers

Fig.6: Photo-based deterioration survey

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3.4 Reporting software
Immediately after an inspection has been performed, the inspectors can edit standard reports with the
reporting software :
output of drawings with deteriorations (Fig .7),
tables ,
pictures .
The process is automated and the output format can be adapted .

3.5 Analysis softwar e
Designed for engineers in charge of the analysis of data, the analysis software consists of a set of tools for :
browsing and sorting of data,
evolution of one deterioration or a group of deteriorations ,
comparison of similar structural components ,
evolution of the condition of one structure, evolution of the stock of structures ,
identification of deterioration to be repaired in priority ,
bill of quantities for repairs ,
assistance in the diagnosis .

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4. BENEFITS AND DIFFICULTIE S
The direct benefits are cost reduction and improved efficiency due to :
-

organized database of consistent data (structural, inspection, maintenance and repair data) ,

-

easier access and sharing of information ,

-

time savings for preparation of drawings, reporting and analysis of data ,

-

long-term management.

Whereas the main difficulties in implementing the system are :
-

staff need to be computer literate ,

-

drawings that exists only on paper must be digitalized (scanner) or redrawn with CAD ,

-

older data should be input in the new system .

The system is therefore easier to implement on recent structures because CAD drawings are available and
older data is smaller : the system has been used from the beginning on the Tagus estuary crossing in
Portugal (Vasco de Gama bridge) .
However, the extra work required to input paper-based drawings and older data can be recovered throug h
cost and time savings at all steps of the process . During the inspection and analysis of the Zilwaukee bridg e
(a twin 2 .5 km-long precast segmental viaduct carrying I-75 over the Saginaw River, Michigan, USA), th e
system proved to be very cost-effective. Every day the inspectors sent the data by email for review by th e
Project Manager and the analysis team, located 3,000km from the bridge . Data from previous inspection s
was later inputted electronically so that it can be compared, sorted and visually displayed along with ne w
data.
5. CONCLUSION
Structure management is an increasingly important concept for structure owning authorities or privat e
companies around the world today . Each owner has been developing its specific health condition indicato r
system, allowing to express the structural condition of a structure by some quantitative measure, to monito r
durability, safety and to decide at what point action needs to be taken .
A management software that would integrate all the steps of this maintenance process can dramaticall y
optimize its efficiency through easier management, storage, sharing and analysis of the structura l
information . While such a concept is not new, this paper presented in detail the capabilities and benefits a s
they were observed in actual large scale implementation of the inspection-based management software .
REFERENCES :
Stubler, Domage, Youdan, "New Developments in Structural Monitoring and Management for Bridges",
IABSE Conference, Cairo .
Stubler, Le Diouron, Elliott, "New Tools to Listen and Watch Structures for a Complete Monitoring" ,
Proceedings of The Korea Institute for Structural Maintenance Inspection, Vol .5, No .1, May 2001 .

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AGING PROCESS OF A GOOD CONCRETE DURING FORTY YEAR S
Dr. Peter Lenkei
Pécs University, College of Engineering (Hungary)
A prestressed concrete truss (Fig . 1) was used to cover an uranium ore processing (concentrating) hall in a
nuclear industry plant . The environment of the hall was slightly aggressive, containing sulphuric acid an d
carbon dioxide . Due to the uranium oxid dust all the equipment and the reinforced concrete structures wer e
covered with a special paint coating for easier decontamination .

After 40 years of service the uranium ore mining was terminated and the hall was demolished (Fig . 2) .

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During the life span of the structures several investigations were made . Destructive tests (DT) were made
after construction in 1960 (200*200*200 cubes), in 1990 and in 2000 thorough non-destructive test s
(NDT) were carried out. Finally, after demolition another DT was made, core samples were taken from th e
truss. The results converted to 150*300 cylinder strength are shown on Fig . 3.

Likewise after the demolition parts of the prestressed bottom chord of the truss were surveyed and neither
on the prestressed wires, nor on the other reinforcement traces of corrosion could be find (Fig . 4).
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Fig. 4 Surface of prestressing wires and rebars
The results demonstrated, that the concrete aging process was characterized by a definite increase over the
initial strength and even after 40 years the concrete strength was 3% higher of the initial strength . Most
probable the NDT gave a little higher results over the true strength values . Neither visible cracks o r
carbonization, nor reinforcement corrosion were detected .
CONCLUSIONS
1. Initially good quality concrete and reliable concrete cover, with sufficiently maintained paint
coating could guarantee the long term life span of prestressed concrete structures .
2. Even the correct NDT may slightly overestimate the concrete properties .

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SESSION B: STATE OF THE ART &FUTURE DEVELOPMENTS (Continued )
Chairman : Mr. Jean-Pierre Touret, EdF, (France )

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THE USE OF ACOUSTIC MONITORING TO MANAGE CONCRETE STRUCTURES IN TH E
NUCLEAR INDUSTRY

Marcel de Wit, Gilles Hovhanessia n
Advitam

ABSTRAC T
Concrete and steel are widely used in containment vessels within the nuclear industry . Both are excellent
acoustic transmitters . In many structures tensioned wire elements are used within containment structures .
However, tensioned wire can be vulnerable to corrosion . To reduce the probability of corrosio n
sophisticated protection systems are used . To confirm that the design strength is available through time ,
extensive inspection and maintenance regimes are implemented .
These regimes include tests to confirm the condition of the post-tensioning, and pressure tests (leak tests )
to verify the performance of vessel .
This paper presents an acoustic monitoring technology which uses widely distributed sensors to detect an d
locate wire failures using the energy released at failure . The technology has been used on a range o f
structures including post-tensioned concrete bridges, suspension bridges, buildings, precast concret e
cylinder pipelines (PCCP) and prestressed concrete containment vessels (PCCV), where it has increase d
confidence in structures and reduced maintenance costs .
Where the level of ambient noise is low then SoundPrint® acoustic monitoring can detect concrete
cracking . This has been shown in PCCP pipelines, on laboratory test structures and also in nuclea r
structures. The programme has shown that distributed sensors can locate internal cracking well befor e
there is any external evidence .
Several projects have been completed on nuclear vessels . The first has been completed on an Electricité de
France (EDF) concrete test pressure vessel at Civaux in France . The second at the Sandia PCCV Tes t
Vessel in Albuquerque, New Mexico, USA, which involved the testing of a steel lined concrete vessel .
The third was on a PCCV in Maryland, USA .
Acoustic monitoring is also able to monitor the deterioration of post-tensioned concrete structures as a
result of seismic activity . Summary details of a case history are presented .
1. INTRODUCTION
Continuous acoustic monitoring has been used since 1994 to monitor failures in bonded and unbonde d
tendons in post-tensioned structures, where it has shown major benefits in confirming the performance o f
structures, increasing Client confidence and reducing maintenance costs . To extend the application of thi s
technology to the monitoring of concrete cracking required that the effectiveness of the principles an d
methods was evaluated for each structural type .
For acoustic monitoring technology to function in a particular environment it must be shown that th e
signals generated by cracking can be detected above general noise levels and distinguished from event s
which are not of interest . Furthermore, to assess the structural implication of each event it is generall y
important to be able to locate the source of each emission . Provided with high quality data of this type, th e
engineer can appraise a structure with knowledge of the actual failures in damaged elements, and their
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location, in the entire structure over the monitoring period . The alternative, to base the assessment on a
physical inspection at a sample of locations, leads to uncertainty when for practical and economic reason s
the number of inspection points is limited. Monitoring the entire structure may also reveal failures no t
detectable by a conventional investigation .
In many applications the acoustic data is transmitted over the Internet for processing and analysis . After
processing and quality control checks, the data can be made available on a secure section of th e
SoundPrint® website, allowing owners rapid independent access to their database of results .
SoundPrint® acoustic monitoring systems have also been placed on structures, which are in active seismi c
zones. Rapid status reports on wire failures / structural damage allows Owners and Regulators to assess th e
condition of a structure within a few hours .
The technology is useful in providing cost-effective long-term surveillance of both unbonded and groute d
post-tensioned containment structures. This paper shows how the technology can also monitor the
cracking of concrete structures which are subject to low levels of ambient noise .
2. DEVELOPMENT OF CONTINUOUS ACOUSTIC MONITORIN G
The principle of examining acoustic emissions to identify change in the condition of the structural element s
is not new . However, until recently, continuous, unattended, remote monitoring of large structures was no t
practical or cost-effective. The availability of low-cost data acquisition and computing hardware,
combined with powerful analytical and data management software, resulted in the development of a
continuous acoustic monitoring system called soundprint ®, which has been successfully applied to
unbonded post-tensioned structures in North America since 1994 .
Corrosion of the steel strands in these post-tensioned structures has become a concern for designers and
owners. As with grouted post-tensioned bridges, the extent of corrosion is not known, primarily because o f
the difficulty of identifying corrosion due to the inaccessibility of the corrosion sites, the lack of externa l
evidence and the limited spatial coverage of intrusive inspections .
The SoundPrint 7 system uses the distinctive acoustic characteristics of wire breaks to separate them fro m
other acoustic activity on a structure . Using a combination of instrumentation, data acquisition and data
management, it is possible to identify events, as well to locate the failure and time of failure .
This concept allows the non destructive identification of broken strands, so that these strands can be
replaced periodically as part of a long term cost effective structural health programme . In addition, an
understanding of the condition of the steel wire elements allows the life of the structure to be extended .
A typical system includes an array of sensors (Figure 1) connected to an acquisition system with coaxia l
communication cable . The sensors are broadband piezo-electric accelerometers fixed directly to th e
concrete slab . Sensor locations are chosen so that an event occurring anywhere on the slab can be detecte d
by at least four sensors . Sensor spacings range from 1 per 60 square meters for fully grouted slabs up to 1
per 100 square meters for ungrouted tendons . Multiplexing techniques are able to acquire data from man y
hundreds of channels on 32 acquisition channels .

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Time

Frequency

Figure 1 - Standard sensor for buildings, bridges and Figure 2 - Time domain and frequency spectrum plot s
parking structures
of wire break detected by sensor 10 .0 m. from event

Figure 3 - Time domain plot showing relative arrival Figure 3 - Time domain plot showing relative arriva l
time of signal at different sensors
time of signal at different sensor s
Using several characteristics of the acoustic events including frequency spectrum it is possible to classif y
wire breaks and to reject environmental noise .
By analyzing the time taken by the energy wave caused by the break as it travels through the concrete t o
arrive at different sensors, the software is able to calculate the location of the wire break, usually to withi n
300 - 600 mm of the actual location . Independent testing showed the system to be 100% correct whe n
spontaneous events classified as "probable wire breaks" were investigated . Figure 2 shows a typica l
acoustic response to an unbonded wire break at a sensor 10 .0 m from the break location . Figures 3 and 4
illustrate how the system locates events .
SoundPrint® site systems download all data automatically using the Internet to the Calgary processin g
center. This allows the cost of data transfer to be minimized . All data can be viewed by the owners team
directly on the Pure Technologies secure web site . This allows the owner to review areas of concern in

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parallel with the generation of routine reports . Various levels of alarms can be triggered semiautomatically using e-mail, automatically voice activated phone alarms, etc .
Presently, over 300,000 square meters of unbonded post-tensioned slab in twenty structures, five bridges
and almost 100km of large diameter water pipe are being simultaneously monitored . The analytical
software is capable of automatically generating reports summarizing the time and location of wire break s
and other significant events . The operating efficiency of the system over the monitoring period is also
recorded.
3.

MONITORING OF WIRE BREAKS IN GROUTED POST-TENSIONED BRIDGE S

Acoustic monitoring has been used in a wide range of applications including suspension and cable stay
bridges (reference J .F . Elliott), and pipelines (reference Mark Holley) .
The technology has also been applied on many post-tensioned concrete bridges as described at thi s
conference (reference Carlyle, Adkins, Youdan) .
4.

MONITORING OF CRACKING DEVELOPMENT IN CONCRETE STRUCTURE S

Description of Concrete Projects
During the UK TRL grouted post-tensioned bridge evaluation program, the developers of the acousti c
monitoring system, Pure Technologies Ltd (Pure) and TRL had the opportunity to evaluate the applicatio n
of the method to crack development in a partially hollow reinforced concrete beam specimen. This
specimen was tested with three-point loading .
Dr. Walter Dilger of the University of Calgary provided access to a flat post-tensioned slab specimen . The
slab was 5 m by 10 m by 150 mm thick supported by 3 columns and tested in shear .
Electricité de France allowed access to a large-scale model prestressed concrete containment vessel a t
Maeva being tested with internal pressure . This vessel has part of the perimeter wall instrumented fo r
crack detection . This specimen had previously been tested to the same pressures used in the curren t
experiment.
At the Sandia National Laboratories in Albuquerque, New Mexico, a ¼ scale test vessel was pressured t o
the full ‘failure’ test pressure in a modification of a standard leak rate test . The objective was to monito r
concrete cracking, tearing of the liner, and gas leakage .
Finally in Maryland, US an operating PCCV was tested during an Integrated Leak Rate Test to determin e
if any wire failures were recorded .
Transport Research Laboratory, UK
A partially-voided reinforced concrete bridge beam was loaded as shown in Figure 5 .

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NEA/CSNI/R(2002)7/VOL 2

Microphones / accelerometers were installed at six locations on a voided beam specimen . Stress wa s
applied as three point loading . Emissions were noted at all loads and continued throughout the test .
Figure 5 - Arrangement of Test Specimen VS1 7

Figure 6 – Location of Events from 0 kN to 30 kN (0 to 6,700 lbf)

Figure 7 – Location of Events from 30 kN to 100 kN (6,700 lbf to 22,500 lbf)

Figure 8 – Location of Events from 100 kN to 200 kN (22,500 lbf to 45,000 lbf)
Results
As used here, ‘cracking noise’ means the generation of acoustic events associated with the propagation o f
cracks, some of which were not visible . Amplification through the data acquisition system produced
audible cracking noise throughout most of the test . No sounds were heard or recorded during periods whe n
displacement had stopped . Acoustic events were located as plotted in Figures 6, 7 and 8 .
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NEA/CSNI/R(2002)7/VOL 2

It was noted several times during the test that the locations of cracks could be determined from the acousti c
data before the cracks became visible . On most occasions, the operators of the acoustic equipment wer e
able to direct researchers to the area where cracks had occurred, resulting in the visual confirmation of
cracks at those locations .
Crack Monitoring at University of Calgar y
Procedure
Ten accelerometers were attached to the underside of the test slab . The slab-column arrangement is show n
in Figure 9 and sensor locations are shown in Figure 10 . Lateral motion of the slab was commenced and
the resulting cracking events were heard and recorded . Upon first loading of the specimen, a very large
number of small emissions were heard . This is known as the Kaiser . This effect describes the generatio n
of acoustic events coincident with initial load sharing and redistribution when a concrete specimen is first
loaded to a given level. Subsequent unloading and reloading to the same level will not produce ne w
acoustic events until the previous maximum load is exceeded. The rate of occurrence of these emissions i s
estimated at between 10 and 100 per second in the specimens tested at the rate of loading used . A sampl e
of the time-domain data is shown in Figure 11 . Each graph represents the output of one sensor.

10 m

Direction of Load Applicatio n

Figure 9 - Slab Arrangement – Specimen UCS 1

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NEA/CSNI/R(2002)7/VOL 2

Figure 10 – Sensor Locations

Figure 11 – Time Domain Graphs

Results
Amplification through the data acquisition system produced audible cracking noise throughout most of th e
test . No sounds were heard or recorded during periods when displacement had stopped . The locations o f
these events are shown in Figure 12 .
As was the case with the TRL test, it was possible to direct researchers to the location of cracking befor e
the cracks were visible . The locations of cracks identified by the acoustic system coincided with th e
confirmed locations as determined by the researchers .

Figure 12 – Event Location s
Maeva Model Containment Vesse l
The Maeva vessel was built for other purposes relating to internal pressure testing . The vessel consists of a
cylindrical wall with an internal diameter of 16 .0 m (52 .5 ft.) and an external diameter of 18 .4 m (60 .4 ft.).
The floor and roof of the vessel consist of concrete slabs connected by four columns each containin g
sixteen x 75 mm (3 in .) high-strength steel Macalloy bars . The concrete wall is enclosed by a watertigh t
steel bulkhead. Instrumentation has been installed on two panels of the vessel to confirm the ability of th e
acoustic monitoring system to detect cracking of the concrete as pressures change . The programme
includes a medium pressure test to 5 .66 bars (82 .1 psi) (which has been completed) and a high-pressure tes t
to 10 bars (145 psi) (which is to be completed later) . Because the vessel had previously been pressurize d
to a greater pressure than was used in the medium pressure test, the vessel was not expected to produce th e
Kaiser effect cracking .

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NEA/CSNI/R(2002)7/VOL 2
Sixteen sensors were attached to two contiguous wall sections of the vessel in the pattern shown in Figure
13 . The sensors were attached to the outer perimeter of the vessel in the annular space between the inne r
and outer wall . The sensors are connected to a data acquisition unit located in a building 200 m (656 ft .)
from the structure .

-~ . .

FR1Z1A0 2

FR1Z1AFR1Z1A01

FR1Z1A10

[5

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FR1Z1A13

6

[1 ]

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Z1A14

A

3 .42

[1 [8]

~E

[11 ]

~

2 .28

[9 ]

X

FR1Z1A12

[15]

FR1 Z

[10]

[2]
[14]

[71 RZ1A15
-

FR1Z1A08

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X

++

4 .94
4 .56

FR1Z1A07

1 .1 4

[3 ]
FR1Z1A00

FR1Z1A11

FR1Z1A03

0 .00

14

12

10

80

60

Figure 13 – Sensor Layout and Location of Acoustic Cracking Events o n
Test No. 3, June 1999
First Result s
Sixteen acoustic cracking events were detected and located. Of these, events numbered 2, 6, 14 and 1 6
occurred outside the area monitored and therefore could not be accurately located. Locations of events are
shown on Figure 13 . The largest cracking events were detected at locations 1, 5, 6, 12 and 15 .
Comparison with a known impact from a Schmidt hammer suggests that some of the cracking event s
released approximately 1 Joule of energy . Event #1, the first large event, occurred at an internal pressur e
of 3 .01 bars (42 .66 psi) . Maximum pressure achieved was approximately 5 .66 bars (82 .09 psi) . Events 1 5
and 16 occurred after the pressure began decreasing, at 5 .65 bars (81 .95 psi) and 5 .46 bars (79 .19 psi)
respectively. Time domain and frequency spectrum plots for Event #1 are shown in Figures 14 and 15.
To confirm the relationship between acoustic events and cracks it is common to correlate the visible
surface evidence of cracks and the measured locations of the acoustic cracking events . This process i s
limited by the fact that cracking events recorded by SoundPrint ® may reflect cracks that are internal to the
structure. Also a visual inspection will normally only record cracks that are of the order of 0 .2 mm (0 .008
in.) wide, whereas SoundPrint® will record cracks finer than this . For this test vessel two additional factor s
are relevant . The first is that no crack survey was taken before this latest phase of the testing commenced,
and the vessel had previously been taken to higher pressures than were used in this phase of testing. The
second is that the internal surface of the concrete had been treated with a thin layer of epoxy, which woul d
have the effect of reducing the clarity of the any crack survey . However, a preliminary crack survey was
carried out after the medium pressure tests and this survey will be repeated after the high pressure tests .
The results of the first phase of acoustic monitoring have confirmed that the system is working
satisfactorily, can identify acoustic cracking activity and is able to locate the source of such cracking
events .

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NEA/CSNI/R(2002)7/VOL 2

Figure 14 - Time Domain Plot of Sensors Figure 15 - Frequency Domain Plot of Sensors
Responding to Crack #1 .
Responding to Crack #1 .
Outline Details of the Sandia Test
In a US project sponsored by MITI and the National Research Council (NRC), a one-quarter scale mode l
of a steel lined containment vessel has been constructed at Sandia National Laboratories in Albuquerque ,
New Mexico . The model is approximately 16 .5 m (54 ft .) high and 11 m (36 ft .) in diameter . The
objective of the work was to validate numerical simulation methods by comparing measured to calculate d
responses well into the inelastic regime, up to and including failure . The intention was to map the
development of strains and eventual damage as the pressure in the vessel is brought above the design
pressure of the vessel . Although more than one thousand strain and other gauges were installed on th e
vessel, much of the vessel was not monitored by strain gauges . Acoustic monitoring was being installed on
the entire vessel wall area as part of the programme to detect tendon failures and with the hope o f
monitoring concrete cracking and liner tearing/leakage . The system has been specially configured to
stream data offsite to a back-up computer incase the onsite unit is destroyed during the test . Preliminary
details of the test are reported by Hessheimer 5. Results of this test will be published first by the NRC in
2001 . See Appendix A for update to June 2001 .
PCCV Maryland
This vessel is approximately 42m diameter reinforced with two sets of circumferential tendons . A total of
204 tendons are present ; a normal tendon is comprised of 90No. 6mm wires .
The vessel was monitored with 36 distributed acoustic sensors, which were placed on the external concret e
surface . The vessel was tested during an Integrated Leak Rate Test. Surveillance of the vessel was by onsite monitoring in real time during the test . During the ILRT zero wire breaks were recorded. To confirm
the performance of the acoustic monitoring system events were generated using impacts with a simila r
energy (and a similar acoustic signature) to a wire break . Impacts made at anchorage cans, were correctly
identified and located . This was quite an achievement as some of the cans were less than 1m apart. See
Appendix A for update to June 2001 .

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5.

SEISMIC MONITORING

In numerous applications Owners need to know the effect of a seismic event on their structures . This is
particularly important in seismically active zones and in the nuclear sector where seismic standards are
rigorous. In California where there is a statutory requirement to report on the effect of a seismic even t
within 24 hours, SoundPrint ® acoustic monitoring has been used to provide rapid and relevant details to th e
Owner .
East Bay Municipal District of Southern California - Brookwood Reservoi r
Above ground prestressed water storage tanks are common in many areas of the world, includin g
California . Failure of these tanks due to corrosion or other factors can be catastrophic . To investigate the
usefulness of long term acoustic monitoring as a management tool, the East Bay Municipal District o f
Southern California commissioned the installation of a monitoring system on the Brookwood reservoir, a
10 million litre (264,00 US gal) capacity tank in the Walnut Creek area East of San Francisco .
As part of the commissioning process, two individual wires were corroded to failure and the result s
monitored with the system installed there . Both wire breaks were detected and located successfully withi n
300 mm (1 ft.) of the actual location.
At 18 :06 on 17 August 1999 there was a Mercalli 5 earthquake in the Bay Area of California . At 23 :04
SoundPrint® recorded 1 wire break at the water tank resulting from the earthquake some 30-km (19 mi.)
distant. Details of this minor damage were e-mailed to the owner within hours . With this data the owner
was able to report quickly, and positively, recording only minor damage . With the proven success of the
system the owner is planning to extend the number of structures monitored .
If many sites were monitored, the owner can also use the rapid notification capability of the system t o
direct emergency repair teams to the areas where most damage has occurred in the event of a larg e
earthquake .
6.

SUMMARY

Continuous remote acoustic monitoring has been used successfully to determine the time and location o f
wire breaks in prestressed structures. Testing of the technique as a method of detecting cracking i n
concrete structures subject to loading has been carried out on post-tensioned and reinforced concret e
structures of different configurations . On two of the three structures tested, the locations of crack s
identified by the monitoring system were confirmed by visual inspection . Crack development was detected
by the monitoring system before cracks became visible .
The SoundPrint® acoustic system has been shown to provide reliable continuous remote monitorin g
capability and the ability to determine the times and locations of both prestressing failures, concret e
cracking and liner tearing/leakage . These capabilities are useful in providing cost-effective long-ter m
surveillance of unbonded and grouted post-tensioned containment structures, and of cracking within the
concrete itself where ambient noise levels are low .
SoundPrint® has been shown to be a valuable management tool where rapid monitoring of seismi c
response is required .

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ACKNOWLEDGEMENT S

The authors thank TRL and the UK Highways Agency, Dr. Walter Dilger, University of Calgary ,
and Electricité de France for use of these data . The authors wish to thank the many individual s
and agencies who contributed to the gathering of these data .
REFERENCE S

1. J.F. Elliott. Continuous Acoustic Monitoring of Bridges . International Bridge Conference ,
1999, Pittsburgh, Pennsylvania IBC-99, pp. 70.
2. M.F.Hessheimer, D . W. Pace, E. W. Klamerus, T . Matsumoto, and J . F. Costello.
Instrumentation and Testing of a Prestressed Concrete Containment Vessel Model . 14th
International Conference on Structural Mechanics in Reactor Technology, Lyon, France ,
1997, pp 792-1, 792-9 .
3. M. Holley, Acoustic Monitoring of Prestressed Concrete Cylinder Pipe . American Society o f
Civil Engineers, Pipelines in the Constructed Environment, California, USA, 1998, pg 46 8
4. F. Carlyle, A . Adkins and D .Youdan. The Use of Acoustic Monitoring to Extend the Life o f
Post-Tensioned Overbridges at Huntingdon and Mossband, UK . Structural Faults & Repai r
2001,9thInternational Conference, July 2001, London .

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Concrete Properties Influenced by Radiation Dose During Reactor Operatio n
Takaaki Konn o
Technical Counselor
Secretariat of Nuclear Safety Commission
ABSTRACT
The radiation dose effects on the physical, chemical and mechanical properties to the biological shieldin g
concrete of the Japan Power Demonstration Reactor (JPDR) were studied to obtain useful information fo r
the plant life management strategy of commercial nuclear power plants . The JPDR was passed 25 year s
from the construction and performed 957 days operation and the total reactor operating time 14,230 hours .
The cumulative radiation dose rate on the biological shielding concrete of the JPDR was estimated a s
equivalent with the one that from the current commercial nuclear power plant after operated 40 years . High
radiation dose is a special unique environment for the concrete structures in nuclear facilities . The
evaluation of the radiation dose effect to the concrete structures in the nuclear power plants is importan t
factor for the plant life management strategy considering aging of the concrete properties during plan t
operation . Usually, studies on concrete properties influenced by the radiation were performed under th e
test condition of short term and high irradiation rate . The test results under the condition of long term and
low irradiation rate for the concrete are rarely exist . This study was conducted using the actual concrete
samples from the JPDR biological shielding concrete obtained when the plant was decommissioning . The
maximum fast neutron and gamma ray at the reactor side surface of the biological shielding concrete ar e
1 .11 × 10E+18 n/cm2 and 4 .77 × 10E+18 Gy, respectively, at the level of the reactor core . The test results
showed that the compressive strength of the concrete samples were not decreased by the radiation exposur e
which was rather shown the tendency to increase along with the fast neutron fluencies within the test range
to 10E+17 n/cm2. The test results showed the biological shielding concrete with steel lining have good
durability in the test range of radiation exposure dose rate in spite of affected with the heat generation
within the shielding concrete by the neutron and gamma ray flux .
INTRODUCTION
The evaluation of the radiation dose effect to the concrete structures in nuclear power plants is important
factor for the plant life management strategy considering aging of the concrete mechanical propertie s
during plant operation . Usually, studies on the concrete properties influenced by radiation exposure were
performed under the accelerated test condition of short term and high irradiation rate . Test results under the
condition of long term and low irradiation rate for the concrete using the actual concrete specimen s
extracted from operating nuclear power plants are rarely existing . The study of radiation dose effects on the
physical, mechanical and chemical properties of concrete was performed using the actual biologica l
shielding concrete sampled by coring from the Japan Power Demonstration Reactor when the plant wa s
decommissioned . The Japan Power Demonstration Reactor (JPDR) was the first nuclear power generatio n
reactor of the rated reactor power 45 MW in Japan that was passed 25 years from the construction at th e
1986 and performed 957 days operation that the total reactor operating time was 14,230 hours when the
plant was on the decommissioning . The cumulative radiation dose rate on the biological shielding concrete
of the JPDR was estimated as equivalent with the one from the current commercial nuclear power plan t
after 40 years operation. High radiation dose is a special unique environment for the concrete structures i n
the nuclear facilities . The tests were conducted considering the various environmental state conditions o f
the radioactivity and the heat generation caused by the neutron exposure, gamma-ray exposure and massiv e
concrete hydration in the biological shielding concrete of the JPDR .

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OUTLINE OF THE JPDR CONCRET E
The biological shielding concrete of the JPDR was constructed with the Portland cement concrete as th e
mixing strength 35 MPa from the March to November 1962 . The mixing proportion is shown in Table 1 .
The maximum thickness of the concrete was 3 .0 m and had a lining of 13 mm thick steel plate on th e
reactor side surface and epoxy paint finishing on the outer side surface . Cooling pipes in order to reduce
the thermal heat by radiation exposures and guide tubes of neutron monitoring were installed in the reacto r
side concrete .

Table 1 Mixing proportion

The location of the core samplings were selected from the level of the reactor core as the high irradiation
concrete, and the upper and lower distant level from the reactor core as the low irradiation concrete in
order to clarify the influence of the radiation environment and the concrete placing. Figures 1 and 2 show
the sampling location and the concrete core sample .

Fig. 1 Sampling location
of the concrete core

Fig. 2 Concrete core sampled

EVALUATION OF THE ENVIRONMENTAL CONDITION S
Influencing environmental conditions to the biological shielding concrete are radiation exposure and hea t
of hydration in the hardening of the massive concrete after the placing . Figure 3 shows neutron flux and
exposure dose distribution in biological shielding concrete at the level of reactor core obtained b y
calculation using computer code ANISN-JR . Neutron exposure dose for concrete samples used for th e
strength tests were evaluated by the calculation of the neutron flux and exposure dose distribution to the 2
dimensional R-Z cylindrical column model using the computer code DOT 3.5 . Distribution graph of the
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NEA/CSNI/R(2002)7/VOL 2
Fast neutron flux (E>0 .11 MeV) and thermal neutron flux (E<1 .85 eV) were obtained by the calculatio n
using the computer code DOT 3 .5. Neutron flux at the test specimens were decided by the distributio n
graph and the neutron exposure dose for the concrete samples were calculated multiplied with th e
converted dose for the rated reactor power operation time of the JPDR. Since the error of the calculated
value of the thermal neutron flux from the measured one become larger according with the distance apar t
from the core center to upper or lower directions the calculation was corrected using the radioactivity o f
Eu-152 measured . Figure 4 shows comparison of the radioactivity of Eu-152 by the calculation and th e
measurement .
The maximum neutron irradiation dose rate to the biological shielding concrete were estimated that the fas t
neutron exposure rate was 1 .11 X 10 E+18 n/cm2 and the thermal neutron exposure rate was 4 .75 X
10E+17 n/cm2 at the reactor side concrete of the reactor core level . Figure 5 shows the gamma ray flux
distribution in the biological shielding concrete obtained by calculation at the level of the JPDR reacto r
core. The maximum gamma dose rate obtained 4 .77 X 10 E+8 Gy by the flux converted to the effectiv e
dose and multiplied with the total rated reactor power operation hours . Figure 6 shows the calorific valu e
distribution that was calculated one dimensional transportation analysis using the computer code ANISNJR generated by the total neutron and the total gamma ray in the biological shielding concrete at the leve l
of reactor core . The calorific value in the biological shielding concrete at the reactor side biologica l
shielding concrete was obtained 3 .0 X 10 E-4 w/cm3 by neutron exposure, and 5 .68 X 10 E-3 w/cm3 by the
total of primary and secondary gamma-ray exposure . The contribution to the maximum calorific value in
the biological shielding concrete that was obtained 5 .98 X 10 E-3 w/cm3 by the total neutron and the total
gamma-ray was largely contributed by the gamma-ray .

Fig. 3 Neutron flux and exposure dos e
distribution in the biological shieldin g
concrete at the level of reactor core

Fig . 4 Activity distribution in the
biological shielding concrete at the
level of reactor core (Eu-152 )

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NEA/CSNI/R(2002)7/VOL 2

a o
13-

" •

Case of 45 M

- SHROUD

'' .



-

L PRESSURE
-VESSÉ

CORE

/ LI
i

i

f

0

~

40

I

I

80

n

l ~ ~

120

160

~

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I.

200

1

1

240

1

1

280

1

l

~

320

Fig.5 Distribution of total gammaray flux at the level of reactor core

Fig .6
Calorific value distribution
generated by total neutron and tota l
gamma-rays at the level of reactor core

The distributions of the concrete temperature for both case of the hydration heat during cure after concrete
placing and the radiation exposure heat during the operation were calculated . The results were shown in the
Figures 7 and 8, respectively . As shown in the figures, biological shielding concrete were therma l
influenced by the heat of hydration in the early stage and heat of radiation exposure during the reacto r
operation .

Fig. 7 Temperature distribution of th e
biological shielding concrete by the heat
of hydration at the placing

Fig. 8 Temperature distribution of the
biological shielding concrete in th e
operation

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METHOD AND RESULT OF THE TESTS
Test on the mechanical properties of the irradiated concret e
In the tests on the mechanical properties of the concrete, compressive strength, tensile strength, modulus o f
static elasticity, and Poisson's ratio were investigated . The test specimens were shaped the core samples i n
the size of 8 cm diameter and 16 cm height for the tests of compressive strength, modulus of elasticity an d
Poisson's ratio, and in the size of 8cm diameter and 8cm height for the tensile strength tests . The concrete
specimens were cured in the water 24 hours before the tests . The modulus of static elasticity were
evaluated by the stress-strain ratio at the one third of the maximum stress on the stress-strain curv e
obtained by the compression gauge and the strain gauge in the compressive strength tests . The Poisson' s
ratio was evaluated from the strain ratio of lateral to longitudinal in the linear strain range .
Figure 9 shows the distribution of the compressive strength along with the depths from the reactor side t o
outer side concrete . The compressive strength of the concrete core samples was distributed from the rang e
of 29 .4 to 53 MPa (average 44 MPa) and the average was 20 % larger than the strength of mixin g
proportion. The compressive strength showed the tendency to increase along with the fast neutron fluence
increased in the range from 1X 10E+13 n/cm2 to 1X 10E+17 n/cm2 when it was looked in the relationshi p
between the compressive strength and the fast neutron fluence calculated as shown in Figure 10 . Since the
compressive strength was influenced by the over burden of the placing height of the flesh concrete at th e
construction, the compression strength were converted to the concrete placing height at 0 cm .
In the previous study, the compressive strength did not decrease in the range of the radiation dose rate
2X 10 E+18 to 2X 10E+19 n/cm2 but when the irradiation dose rate is increased to over the 5X 10E+1 9
n/cm2 the compressive strength is decreased significantly, and also the compressive strength of th e
concrete have been said to decrease by the Gamma irradiation accumulated approximately over 10E+9 G y
(Hilsdorf, H.K. et al .) . In our tests, however, the maximum fast neutron and gamma ray dose rates at th e
reactor side concrete are 1 .11X 10E+18 n/cm2 and 4 .77X 10E+18 Gy, respectively and it could not confir m
the tendency .
Figure 11 shows the relationships of the modulus of static elasticity and neutron fluence . The influence o f
the neutron fluence to the modulus of elasticity was not shown as the figure shows . The relationship of the
modulus of static elasticity and compressive strength were largely distributed around the AIJ curve of th e
relations as shown in the Figure 12 .

Fig. 9 Compressive strength along
with concrete depth

Fig. 10 Compressive strength distribution
along with fast neutron fluence
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NEA/CSNI/R(2002)7/VOL 2

Fig. 11 Modulus of static elasticity
along with fast neutron fluence

Fig. 12 Relationship of
compressive strength and modulus
of static elasticity

Figures 13 and 14 shows the relationship between Poisson’s ratio and fast neutron fluence and th e
relationship between tensile strength and fast neutron fluence, respectively, obtained by the strength test o f
the concrete core samples . The influence of the fast neutron fluence to the Poisons ratio and the tensil e
strength were not shown in the irradiation range of the test samples .

Fig. 13 Poisson’s ratio along with neutron
fluene

Fig. 14 Tensile strength alon g
with fast neutron fluence

Test on the chemical properties of the irradiated concrete

In order to investigate what influences were affected to the concrete microstructure components by th e
irradiation, the tests of the chemical properties of the concrete core samples were performed . Tests item s
were chemical element analysis, X-ray diffraction analysis, scanning electron microscope observation ,
porosity measurement, water of crystallization measurement, and differential heat analysis .

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NEA/CSNI/R(2002)7/VOL 2
Chemical element analysis : The analyses were performed regarding the nine principal elements of Si, Ti,
Fe, Al, Mn, Ca, Mg, Na, and K by the method of the Inductively Coupled Plasma Spectrometry . The
oxidation products of the nine principal elements were matched almost the same between the concrete
samples of reactor side and outer side as expected.
X-ray difraction analysis : The analyses were performed using the mortar specimens crushed into the
diameter less than 45 that were placed uniformly on the glass plate and the diffraction angles were
measured. The spacing of the crystal faces were obtained input the refraction angles into the Bragg’ s
condition and the crystals of the specimen were identified comparing with the standard samples . The
influences of the radiation dose effect to the crystallization in the concrete microstructures were not show n
from the diffraction pattern .
Scanning electron microscope : The observations were performed on two scanning field for one morta r
specimen using the mortar made by roughly crushed the concrete specimen after vacuum drying in scale s
up to 500, 1000, and 3000 times larger by the scanning electron microscope . From the scanning electro n
microscope observation at the pore where hydration crystal growth was observed the specimen of reacto r
side showed the large growth of the needle crystal than outer side .
Porosity measurement: The micro-pore size of the mortar extracted from the crushed concrete specime n
were measured in the range of 60 to 99,000 + using porosity gauge by the penetration method o f
pressurized mercury into the mortar in the pressure range from 0.9 to 2000 kg/cm2 . The micro-pore
diameter distribution of the reactor side concrete was distributed to the small size region than the outer sid e
concrete. The peak of the micro-pore distribution of the both side concretes were shown at the diamete r
130-250 .
Bound water measurements and diferential thermal analysis: They were performed using the microcrushed concrete specimen to the size of 45 . The quantities of the water of crystallization were obtaine d
from the differences after heated the micro-crashed specimen one hours each at the temperatures of 105 ,
400 , 650 and 950 by electro heater. The quantity of the free water measured from the loss of weight at
the heating temperature 105 in the reactor side concrete showed 12% larger than the outer side concrete .
The differential thermal analysis results were shown not much difference between the reactor side an d
outer side concrete.
Comparison tests on the mock-up concrete
The test concrete were exposed the thermal condition by heat of hydration during the massive concrete
placing and the heat of irradiation during the reactor operation . In order to clarify the influences of th e
irradiation effect and the environmental thermal effect, reference tests were performed on the physical ,
mechanical and chemical properties with the same test items using mock up specimens simulated th e
environmental thermal conditions that are the heat of hydration and the radiation exposure heat as t o
become the same conditions of the actual concrete samples . The test specimens were made to represent
four cases of environmental conditions as shown in the Figure 15 . The test cases represent the
environmental conditions of the actual concrete that are massive concrete cured in an air as case 1, massiv e
concrete cured in an air and radiation exposure as case 2, normal concrete cured in a water as case 3, and
normal concrete cured in an air as case 4 .

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NEA/CSNI/R(2002)7/VOL 2

Fig. 15 Curing conditions of the mock-up
test cases

Fig. 16 Compressive strength of th e
mock-up concrete tests

The test results showed that the effect of the steel lining brought good cure conditions for the concrete in
the both cases 1 and 2 to increase the compressive strength along with longer the curing term as shown i n
the Figure 16 . While the compressive strength of the mock up concrete without steel lining showed the
increasing rate of the long term strength became slower, especially, it were decreased significantly in th e
case 2 by the influences of long term heating .
The scanning electron microscope and the X-ray diffraction analysis test results showed not muc h
difference between the cases 1 and 2, but the porosity test results showed the pore size distribution of the
both cases 1 and 2 were shifted to the larger size in the case of without steel lining while in the case of wit h
lining it is not changed. The heat of hardening of massive concrete as represented the curing condition i n
the early stage of the cases 1 and 2 were increased the strength generation at the age of one month as larg e
as the same age strength of the case 3 . Then the heat of second stage in the case 2 to represent the condition
of the heat by radiation in the operating stage was affected to the concrete without lining to dry th e
concrete slowly and evaporate the water content, coarse the hardening body organization, and increased th e
total volume of the micro pores, that seemed to made the strength decrease at the age of 3 month than a t
the age of one month . The specific gravity of the concrete with steel lining did not change according with
the curing term but the concrete without shield lining decreased according with the curing term increase d
in the both cases 1 and 2 .

CONCLUSION
The evaluation of the radiation exposure dose effect to the concrete structures in nuclear power plants i s
important factor for the plant life management strategy considering aging of the concrete properties durin g
plant operations. The study using the actual concrete samples from the JPDR biological shielding concret e
was performed when the plant was decommissioning . The major results of the study is summarized as
follows,
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The maximum fast neutron and gamma ray dose rates at the reactor side of the biological shieldin g
concrete are 1 .11 × 10E+18 n/cm2 and 4 .77 × 10E+18 Gy, respectively, at the level of the reactor cor e
which is almost the same level as the one after 40 years operation of current commercial nuclear powe r
plants .
The biological shielding concrete are categorized two featured environmental conditions influenced to the
material properties that are 1) reactor side concrete with steel lining, and, 2) outer side concrete without
steel lining . These conditions were generated relatively high temperatures of concrete by the heat o f
hydration of massive concrete in the early stage and the long-term heat by radiation in the reacto r
operation .
The generation of heat within the biological shielding concrete by the hydration, neutron and gamma-ray
exposures caused large influences to the concrete properties and it were appeared as the variations of th e
compressive strength, the modulus of static elasticity, and the pore size distributions . Poisson’s ratio wa s
not shown the influence by the fast neutron dose rate .
The influences of the radiation exposure to the microstructures of concretes were appeared mainly in th e
behavior of the water contents by the long term heating of radiation exposure and it caused the wate r
dissipation slowly from the outer side concrete without lining. While, the reactor side concrete showed
good durability for the water dissipation by the steel lining even though the decomposition of wate r
contents by neutron exposure as suggested from the high radioactive tritium generation in the reactor sid e
concrete.
The compressive strength distribution of the biological shielding concrete were matched the tendency t o
increase along with the fast neutron fluencies increasing within the test range to 10E+17 n/cm2 . The
negative effects of the radiation dose to decrease the compressive strength of the concrete were no t
appeared within the dose range of the test samples to 10E+17 n/cm2 .
Scanning electron microscope observed the large crystal growth of the ettringite in the reactor sid e
concrete. In generally, the hydration of the Portland cement is said that aluminate phase C 3 A and gypsum
CaSO4 -2H2O generate the ettringite C 3 A-3CaSO 4 -32H2O in the early stage and then it is changing to th e
monosulfeto C3 A-3CaSO 4-12H2O. Large growth of ettrigite is said to swell the concrete organization an d
to obstruct the strength generation . The large crystal growth of the ettringite in the reactor side concrete i s
contradict with the general tendency . It is suggesting that there are still unknown factors about th e
influences of the radiation exposure to the concrete properties .

ACKNOWLEDGEMENT
This report was summarized the past study reports performed by the research people in the Japan Atomi c
Research Institute and the Kajima Corporation when the JPDR was decommissioning . The participants in
the study were greatly appreciated .

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REFERENCE S
Verrall, S ., et al ., Design concepts to minimize the activation of the biological shield of light-wate r
reactors, 1985 .
Hilsdorf, H .K., et al ., The effect of nuclear radiation on the mechanical properties of concrete, ACI SP55 10.
Idei, Y., Sukegawa T ., et al ., A study on aging of biological shielding concrete of JPDR (I . General study
plan and evaluation of environmental factors), Proceeding of the Fall Meeting of the Atomic Energ y
Society of Japan, pp .31, Oct., 1989 (In Japanese)
Akutu, Y., Idei, Y ., et al., A study on aging of biological shielding concrete of JPDR (II . Strength test
results), Proceding of the Fall Meeting of the Atomic Energy Society of Japan, pp .31, Oct., 1989 (In
Japanese)
Idei, Y., Kamata, Y ., Akutu, Y ., Kakizaki, M., et al ., Material properties of the JPDR biological shieldin g
concrete, JAERI-M, 90-205, Japan Atomic Energy Research Institute, Nov ., 1990 (In Japanese)
Konno, T., Suzuki, K., et al ., Study of aging for biological shielding concrete in nuclear facilities (I . Test
result of the concrete strength), Proceeding of the AIJ, pp .177-178, Sept ., 1991 (In Japanese)
Kurioka, H ., Kakizaki, M., et al ., Study of aging for biological shielding concrete in nuclear facilities (II.
Test result of the chemical properties), Proceeding of the AIJ, pp .177-178, Sept ., 1991 (In Japanese)
Kakizaki, M ., Idei, Y., Sukegawa, T ., et al ., Study on environmental and mechanical properties o f
irradiated concrete, J . Struct. Constr. Eng., AIJ, No . 488, 1-10, Oct ., 1996 (In Japanese)

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DETECTION OF REINFORCEMENT CORROSION AND ITS USE FOR SERVICE LIFE
ASSESSMENT OF CONCRETE STRUCTURE S
by C. Andrade*, I . Martinez*, J. Muïïoz*, J. Rodriguez**, M . Ramirez* *
*Institute of Construction Science “Eduardo Torroja”, CSIC, Madrid, Spai n
** Geotecnia y Cimientos S .A. (Geocisa), Madrid, Spai n

1 INTRODUCTIO N
Corrosion of reinforcement is one of the main durability problems of concrete structures . The corrosion i s
induced by two main factors : the carbonation of the concrete cover and the penetration of chloride s
providing from marine atmosphere or from chemicals in contact with concrete . Carbonation generally aim s
into uniform corrosion of the steel bar while chlorides mainly induce localised corrosion . Both types of
corrosion are of electrochemical nature .
There is a third type of corrosion named stress corrosion cracking, SCC, whose mechanism is not entirel y
electrochemical, but the mechanical stress co-operates for its development . This last type will not be
considered in present paper.
Reinforcement corrosion is not a common problem in nuclear power plants due to the limited life of thes e
installations, except in case of cooling towers where frequent corrosion problems have been noticed . It can
be however a key aspect to be taken into account when dealing with extension of power plant service life .
It is as well a very relevant aspect in long term storage or repository installations where lives beyond 30 0
years are usually targeted.
In present paper, it is described first how to measure reinforcement corrosion in order to obtain th e
corrosion rate of the steel . Then, the effects of the evolution of corrosion are listed, in order to b e
considered as limit states and therefore indicators of repair criteria . Finally, a 3D model is presented on th e
possible release of chlorides being part of the low and medium radioactive wastes stored in drums . Thes e
chlorides may diffuse through the surrounding cement mortar and reach the reinforcement of the concret e
containers used to encapsulate the drums . The model is part of a general one that will include not only th e
ionic diffusion, but also the corrosion of reinforcements and its evolution . This type of models, although
theoretical and simple, will help to understand the long-term performance of concrete structures regardin g
the corrosion of reinforcements.
2 ON-SITE TECHNIQUES FOR CORROSION MEASUREMEN T
2.1 Corrosion Potential and resistivity maps .
Up to the present the main techniques used on-site for appraising corrosion of reinforcements are o f
electrochemical nature due to that is the basis of the corrosion process .
Because of its simplicity, the measurement of Ecorr (rest or corrosion potential) is the method mos t
frequently used in field determinations . From these measurements, potential maps are drawn which reveal
those zones that are most likely to undergo corrosion in the active state 1 . However, such measurement s
have only a qualitative character, which may make data difficult to be interpreted 2. This is due the potentia l
only informs on the risk of corrosion and not in its actual activity . In addition, the developing of macrocell s
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may as well mislead the deductions because corroding zones polarize the surrounding areas, which ma y
seem corroding as well, when they are cathodic areas of the macrocell . In spite of which potential mappin g
still has a function to accomplish as a qualitative indication of the general performance and a complemen t
of the other on-site techniques .
The same that said for the potential can be stated on Resistivity, p, measurements 3, which sometimes are
used jointly with Ecorr mapping . The p values indicate the degree of moisture content of the concrete,
which is related to the corrosion rate when the steel is actively corroding, but which may mislead th e
interpretation in passive conditions. On figure 1 is represented a risk map of a slab . The risk level has been
calculated by a combination of these two parameters : Ecorr and p .

Figure 1 : Corrosion risk map on a reinforcement slab calculated from the combination of Ecorr and
measurements .
2.2 Polarization Resistance
2.2.1 Laboratory measurements
The only electrochemical technique with quantitative ability regarding the corrosion rate is the so-calle d
Polarization Resistance, R p 4 . This technique has been extensively used in the laboratory. It is based on the
application of a small electrical perturbation to the metal by means of a counter and a reference electrode .
Providing the electrical signal is uniformly distributed throughout the reinforcement, the AE/ I rati o
defines R p . The corrosion current, Icorr, is inversely proportional to Rp, Icorr= B/Rp where B is a constant . Rp
can be measured by means of D .C. or A.C. techniques 5, both of which have specific features in order t o
obtain a reliable corrosion current value in agreement with gravimetric losses .
2.2.2 On site measurements
Direct estimation of True R p values from AE/ I measurements is usually unfeasible in large real concret e
structures. This is because the applied electric signal tends to vanish with distance from the counte r
electrode, CE rather than spread uniformly across the working electrode, WE . Therefore, the polarizatio n
by the electric signal is not uniform, and it reaches a certain distance that is named the critical length, Lcrit .
Hence, AE/ I measurements on large structures using a small counter electrode provides an apparen t
app) that differs from the true R
polarization resistance (R p
p value depending on the experimenta l
conditions6. Thus, if the metal is actively corroding, the current applied from a small CE located on th e
concrete surface is ’drained’ very efficiently by the metal and it tends to confine itself on a small surfac e
area. Conversely, if the metal is passive and Rp is high, the current applied tends to spread far away (e.g.,
around 50 cm) from the application point . Therefore, the apparent Rp approaches the true R p for actively
corroding reinforcement, but when the steel is passive, the large distance reached by the current needs a
quantitative treatment.

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Modulated confinement of the current (guard ring) method .

There are several ways of accounting for a True R p value, among which the most extended one is the use o f
a guard ring6, in order to confine the current in a particular rebar area, as Figure 2 depicts . The
measurement is made by applying a galvanostatic step, lasting 30-100 seconds, from the central counter .
Then, another counter current is applied from the external ring, and this external current is modulated b y
means of the two reference electrodes called “ring controllers” in order to equilibrate internal and externa l
currents, which enables a correct confinement, and therefore, calculation of Rp . By means of this electrical
delimitation to a small zone of the polarized area, any localised spot or pit can be first, localised, an d
second its measurement can be made by minimising the inherent error of Rp . Not all guarded technique s
are efficient. Only that using a “Modulated Confinement” controlled by two small sensors for the guar d
ring control placed between the central auxiliary electrode and the ring, shown in figure 2, is able t o
efficiently confine the current within a predetermined area . The use of guard rings without this contro l
leads into too high values of the Icorr for moderate and low values, and the error introduced in the case o f
very localised pits, is very high.

Figure 2 : Modulated confinement of the current (guard ring) metho d
3 EMBEDDED SENSORS
The introduction of small sensors in the interior of the concrete, usually when placing it on-site is bein g
one of the most promising developments in order to monitor the long-term behaviour of the structures . The
most usual, as in the case of non-permanent on-site techniques, is to embed reference electrodes o r
resistivity electrodes . They can inform of the presence of moisture and on the evolution of corrosio n
potential . Others events that can be monitored are the advance of the carbonation or chloride fronts, th e
oxygen availability, temperature, concrete deformations and the corrosion rate .
A particular example of the use of embedded sensors is the case of storage facilities of low and medium
radioactive wastes in El Cabril (Côrdoba) 7 . There, a pilot container has been instrumented from 1995 by
embedding 27 set of electrodes (Figure 3) . The parameters controlled are : temperature, concrete
deformation, corrosion potential, resistivity, oxygen availability and corrosion rate . The impact o f
temperature on several of the parameters is remarkable, and therefore, care has to be taken whe n
interpreting on-site results .

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Figure 3 : Preparation of the embedded sensors in El Cabri l
3 RANGES OF CORROSION RATE VALUES MEASURED ON-SITE
The experience on real structures7 has confirmed the ranges of values previously recorded in laboratory
experiments4.
a0 .2 gA/cm2

Negligibl e

0.2 gA/cm2 < Icorr < 0.5 E A/cm2

Low

0.5 gA/cm2 < Icorr < 1 gA/cm2

Moderate

>1 E.,A/cm2

High

Table 1 : Ranges of corrosion rate and risk levels .
In general, values of corrosion rates higher than 1µA/cm2 are seldom measured while values between 0.11µA/cm2 are the most frequent. When the steel is passive very low values (smaller than 0.05-0 .1µA/cm2)
are recorded.
A comparison of on-site Icorr values to electrical resistivity has allowed the authors to also rank th e
resistivity ones.
Practical measurements of on-site Icorr in El Cabril (Spain )
In order to control long term performance of concrete containers used for low and medium radioactiv e
waste storage, Enresa is developing surveys of a set of parameters in the real structures . Corrosion rate o f
reinforcement is one among the parameters measured in the concrete cells in El Cabril – Cordoba – Spain .
Figure 4 shows the results of Icorr measured during several years . The values indicate the passivity o f
reinforcement as expected.

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Ecorr, WALL A

-400

-300

200

dic-94 jun-95 dic-95 jun-96 dic-96 jun-97 dic-97 jun-98 dic-98 jun-99 dic-99 jun-00 dic-00 jun-01 dic-0 1
A1

- A2

+ A3

- A4

- A5

- A6

RESISTIVITY, WALL A

dic-94 jun-95 dic-95 jun-96 dic-96 jun-97 dic-97 jun-98 dic-98 jun-99 dic-99 jun-00

A1

A1

_A2

-A2

-

A3 - A4 - A5

+A3

119

-

A4

-

A5

_A6

_A6

dic-00 jun-01 dic-0 1




NEA/CSNI/R(2002)7/VOL 2
Figure 4 : Results of Ecorr, resistivity and Icorr measured during several years over an internal wall of th e
container .
4 TRANSFORMATION OF ICORR VALUES INTO CALCULATIONS OF LOSS IN BAR
CROSS SECTION
Corrosion leads into four main structural consequences : 1)reduction of bar cross section, 2) reduction of
steel ductility, 3) cracking of concrete cover and, 4) reduction of steel/concrete bond (composite effect) .
All these effects occurring in isolation, or simultaneously, will result in a loss in the load bearing capacit y
of the structure 8 .
The primary information obtained from corrosion measurements is that concerning the loss in cross sectio n
of the bar. This parameter informs about all the other effects of the corrosion process . The attack
penetration P x is defined as the loss in diameter as is shows in Figure 4 . It is obtained through the
expression :
Px = 0. 0115 · IPEP tp
(1)
2
Being tp the time in years after corrosion started and 0 .0115 a conversion factor of gA/cm into mm/year
(for the steel) . This expression implies the need to know when the corrosion has started in order to account
for tp .
When the corrosion is localised (right part of figure 10), the maximum pit depth is calculated b y
multiplying expression (1) by a factor named a which usually takes a value of 10 . Hence expression (1 )
above becomes ,
(2)
Ppit= 0.0115 •I ~PEP • t p•a = 0. 1 1 5 •I~PEP • tp

UNIFORM
CORROSIO N

Figure 5 : Residual steel section loss considered for the cases of uniform and localised corrosion .
5 MODEL OF SERVICE LIFE PREDICTION OF REINFORCEMENT CORROSION
In general, low and medium radioactivity waste is disposed in concrete containers . In the case of Spain, the
primer container is a cube of 2x2x2 m with a wall thickness of 15 cm . Two layers of nine drums containing
the radioactive waste mixed with cement (cement matrix) are placed inside this container . Once the deck o f
the container has been set, the space left is completely filled with low porosity mortar minimising th e
number of non-desirable air bulbs that could be formed .
In order to study the service life of these containers, several research programs are developed by Enresa
(Spanish Agency for Nuclear waste management) . That concerning the service life of cementitiou s
materials in this type of disposal sites is made with the collaboration of the IETcc in Spain .

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5.1 Physical model
Experimental data has shown that transport of ions, from a macroscale point of view, can be describe d
through an apparent diffusion coefficient Da as a fickian diffusion process . Therefore the flux can be
expressed through the Fick´s First Law :
q = -Dap C(x)
(3)
Where q denotes the flux vector of chloride, Da is the apparent diffusion coefficient and C the
concentration . Although two materials with different diffusion properties are modelled (mortar an d
concrete), each one is considered as a sub-domain in which D a is constant . Besides, the conservation of the
total amount of ions implies :
(4)
aC =
D div(q)
at
From (3) and (4) we obtain, for a constant diffusion coefficient the governing equation can be written as :
(5)
aC
a2C a 2C a 2 C
= Dap AC = D ap a 2 +
+
t
x
ay 2
az 2
5.2 Numerical model
The whole model has been meshed with linear hexahedral elements . In some scenarios some regions have
been meshed with a coarse mesh and others with a denser one depending on the gradient of the chlorid e
flux expected . The three main parts of the container (drums, mortar and container walls) have been meshe d
separately in order to assign different material properties . Thirteen scenarios have been modelled.

Figure 6 : Section of the mesh used for the container
5.3

Predictions

Only the concentration history of some critical points has been plotted . A scheme of their location is shown
in figure 7 . Since the bottom face of the container has no mortar protection, a point B located on the
surface that contains the bottom bars (BS) is selected . B is within all the points in BS the one that reache s
the maximum concentration values . The same has been done for the point L but related to the lateral
surface (LS) . Curves in figure 7 show that for a constant surface concentration the level of chlorides i s
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continuously increasing (scenario 1) . On the contrary, preserving the total amount of chlorides that initiall y
the active drums contain the curves show a maximum that in all the cases appears before the first 15 0
years. This maximum is approximately 30 % of the maximum attained with the first approach . However,
this ratio is for the values at 300 years only 10-15 % .

100

150

20 0

T[years]

Figure 7 : Points monitored in the analysis and concentration histor y
0 .18 0 .16 0 .1 4
0 .1 2
0.1
0 .0 8
0 .0 6
0 .04
0 .02
100

5

10
T[years ]

0
200

250

300

0

100

T[years]

20 0

Figure 8 : Bottom and lateral critical points concentration histor y
Both graphics in figure 8 show that the values of the point B are in general higher than those of the point L .
Moreover, the decrease of the concentration values is more gradual in the second case. In scenario 13, the
more critical one, the final concentration levels are quite similar for both points . However, the bars in the
bottom wall of the container are subjected to more aggressive conditions during the whole history .
The start of the corrosion can be considered when the chloride concentration values reach the 0 .4-0.7% of
the cement weight . That is to say between 0 .068 and 0 .122 % of the total weight of the concrete for a
content of 400 kg/m 3 and a density of 2350 kg/m3 . The real surface concentration is variable and difficult
to determine, but it can attain 6000 ppm. However, the initial surface concentration that reaches th e
threshold level calculated for a final value of concentration of 0 .20 (scenario 13) is :
CS=0.122%/0 .20=0.61%=6100 ppm

(6)

Results show that a constant surface concentration is a roughly approach that must be verified for th e
model under study. Furthermore, the concentration of chlorides decreases when more real boundar y
conditions are imposed. The deterioration process is therefore determined not only by the maximu m
concentration level attained, but also by the whole history of the concentration profiles . New experiments
must be undertaken to show the behaviour of the bars under those conditions .
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The coupled effects of corrosion in concrete and the mechanical behaviour of the reinforcement are no w
under study. Since the chemical environment is very dependent on the materials, the geometry and time ,
the mechanical deterioration must be modelled jointly with the transport phenomena .
6 FINAL COMMENT S
Corrosion of reinforcement can be approximately model and accurately measured on-site . The periodical
corrosion rate measurements on its monitoring through embedded sensors seems very necessary to asses s
present conditions of concrete structures and is a very useful tool in the case of cooling towers of power
plants . Techniques based in the measurement of Polarization Resistance have been implemented i n
portable corrosion rate meters to obtain corrosion rate values, and corrosion-data-loggers are now operativ e
in pilot containers to monitor corrosion related parameters .
The in-situ techniques should be complemented by models that although still too oversimplified, may hel p
to make predictions of the advance of aggressive fronts towards the reinforcements and to predict ver y
long-term performance .
7 ACKNOWLEDGEMENTS
The authors thank to Enresa the funding provided to develop several of the researches presented in th e
paper. They thank as well the firm Geocisa for the results of corrosion rate measured by them in El Cabril .
8 REFERENCE S
1.
2.
3.

4.

5.

6.

7.
8.

9.

ASTM C876-91 . "Standard Test Method for Half Cell Potentials of Uncoated Reinforcing Steel n
Concrete" .
Elsener, B and B6hni, H . Corrosion Rates of Steel in Concrete, N .S. Berke, V.Chaker and D .
Whiting (Eds .), ASTM STP 1065, pp . 143-156. (1990)
Millard, S.G. and Gowers, K .R., "Resistivity assessment of in-situ concrete : the influence of
conductive and resistive surface layers", Proc. Inst. Civil Engrs . Struct. & Bldgs, 94, paper 9876 ,
pp .389-396 . (1992)
Andrade, C. and G6nzalez, J.A., "Quantitative measurements of corrosion rate of reinforcing steels
embedded in concrete using polarization resistance measurements", Werkst. Korros ., 29, 51 5
(1978) .
Andrade, C., Castelo, V ., Alonso, C. and Gonzâlez, J.A., "The determination of the corrosion rate o f
steel embedded in concrete by the Rp on A .C. Impedance methods," ASTM-STP 906, pp . 43-64 .
(1986)
Feliû, S . , Gonzâlez, J.A., Feliû, S .Jr., and Andrade, C ., "Confinement of the electrical signal or insitu measurement of Polarization Resistance in Reinforced concrete," ACI Mater. J., 87, pp 457.
(1990)
Andrade C; Sagrera J.L; Gonzalez J .A ; Jiménez F; Bolano J .A; Zuloaga P . "Corrosion monitoring of
concrete structures by means of permanent embedded sensors" . Niza. Eurocorr’9 6
Rodriguez, J ., Ortega, L .M. Garcia, A .M., Johansson, L. Petterson, K. "On-site corrosion rate
measurements in concrete structures using a device developed under the Eureka Project EU-401-Int .
Conference on Concrete Across Borders, Odense, Denmark, vol .I, pp .215-226 .
Rodriguez, J ., Ortega, L .M., Casal, J., Diez, J .M., "Assessing Structural conditions of concret e
structures with corroded reinforcements", Conference on concrete Repair, Rehabilitation an d
protection, Dundee (U .K.), Edited by R.K. Dhir and M.R. Jones, Published by E&FN Spon in Jun e
1996, pp.65-77 .
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Improved Detection of Tendon Ducts and Defects in Concrete Structures Using Ultrasonic Imagin g
Authors:

W. Müller, V . Schmitz *)
M. Krause, M . Wiggenhauser ** )

*)
**)

Fraunhofer Institut für Zerstdrungsfreie Prüfverfahren (IZFP )
Bundesanstalt für Materiaailforschung und –prüfung (BAM )

Introduction
At the beginning of the 90s the general opinion was, that ultrasonic inspection methods using pulse-ech o
technique were not suitable for the inspection of concrete because of the inhomogeneity and the stron g
scattering behavior of the embedded aggregates . In the meantime the progress in the development of ne w
equipment and inspection strategies in connection with ultrasonic imaging techniques turns the pulse-ech o
technique into a powerful tool to solve problems related to concrete materials . These imaging techniques –
developed for the inspection of homogenous materials like steel or aluminum – could be adopted to the
very low frequencies needed for concrete inspections .
Since 1994 the BAM and the IZFP cooperate in the field of concrete inspection . The BAM perform s
measurements using laser-vibrometers, probe-arrays, and pitch-and-catch arrangements with two probes .
The ultrasonic echoes received are digitized and stored on a computer and evaluated at the IZFP using th e
Synthetic Aperture Focusing Technique (SAFT) for 3-dimensional imaging . The results are very
encouraging with respect to detection and positioning of reinforcement structures inside the concrete lik e
tendon ducts and reinforcement bars as well as defects like compaction defects and voids (honey-combin g
represented by styrodur balls) and to detect and even size notches and natural cracks oriented vertical to th e
surface . These inspections are applied to bridges, nonballasted tracks, and foundation slabs . They were
performed in the laboratory as well as in the field at prestressed concrete bridges . In addition the Federal
Highway Research Institute (Bundesanstalt für Stral3enwesen, BAST) organized two round robin trials, one
on test specimen containing artificial defects, the other on a motorway bridge, which had to be replaced .
Examples of reconstructed ultrasonic images are presented . Further basic research work on ultrasoni c
imaging of concrete structures is carried out within a research group of the German Research Counci l
(FOR 384 of Deutsche Forschungsgemeinschaft, DFG) .
1.

A round robin trial with a test block containing artificial defect s

On the concrete support of railway tracks, vertical cracks have been observed . Those cracks allow water t o
penetrate which causes a corrosion of the tendon ducts and thus reduces the time of life of the supportin g
structure. In the literature different methods based on time-of-flight evaluation of the ultrasonic pulses are
reported. Difficulties arise from particles or water which act as ultrasonic bridges between the crac k
borders ; in those cases the true penetration depth of the cracks can be much larger than the actual measure d
depth extension . Modern research like /1/, /2/ and /2//3/ is aimed to improve the reliability of crack dept h
measurements .
During the propagation in concrete material with additives of different aggregate sizes, the ultrasoni c
pulses are scattered and change propagation direction into all possible directions . In the case of a crac k
between an ultrasonic transmitter and the receiver probe the time-of-flight of received scattered amplitude s
moves towards larger values /4/, /5/ .

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Most of the time-of-flight methods are based on the surface wave and the diffraction of the longitudina l
waves at the crack tip . The principle is shown in fig . 1 . For each probe position of the receiver the time-offlight will be measured . If a surface crack lies between transmitter and receiver the signal will disappea r
and come up again with a step in the time-of-flight curve . The time difference is a function of the longe r
sound path around the tip of the crack and proportional to its depth extension .

Fig. 1 Principle of crack depth measurement with acoustic surface and longitudinal wave s
This principle has been used in an experiment at a 20 cm thick concrete test block which contained a 10 c m
deep crack simulated by a notch. The center frequency of the ultrasonic probe was 100 kHz . If the crack i s
filled with water the crack gets penetrable to the ultrasonic pulses and one would expect that this metho d
will fail . In fig 2 the time-of-flight methods are displayed for both case s

Air-filled crack ; calculated depth: 91 mm Water-filled crack ; calculated depth : 22 mm
Fig. 2 Problems arise if the crack is filled with water
It is obvious that particles or water in a surface connected crack cause wrong measurmentsmeasurements ,
because only the first arrival of the signal is used for evaluation . With regard to cracks in concrete
elements reinforcement bars and uncracked aggregates can act as ultrasonic bridges. Therefore it i s
important to develop a more reliable procedure .

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Ultrasonic Synthetic Aperture Focusing Technique (SAFT) has a great potential to image cracks i n
concrete material . Its algorithm superimposes data obtained by pulse echo at many positions which lead s
to the suppression of structural noise and to a more reliable positioning of indications .
Computer-based implementations of such a procedure like SAFT were developed by /6/ for the inspectio n
of pressure vessels, a combination with ultrasonic holography for defect sizing and classification by /7/ an d
a three dimensional version implemented by /8/ .
In conventional ultrasonic testing, a specimen is scanned with a narrow search beam to determine th e
position of an object. The situation in concrete testing is different . Given an average velocity of 4000 m/ s
in concrete and a transducer of 40 mm diameter a frequency of 200 kHz leads to a divergency beam of 15° ,
a 100 kHz transducer to 31° . Hence the ultrasonic beam is not small enough to find the lateral position o f
an object . The SAFT algorithm removes this disadvantage .
The movement of a relative small probe imitates a large transducer by sampling its area at many points .
This can be done either by an array of transducers which is electronically scanned or by one or tw o
transducer which are moved step by step . Such a general arrangement is applied to the problem of imaging
a surface connected crack – fig. 3 . In pulse echo the transmitter acts as a receiver too and one has to mov e
a single probe across the whole surface . If the receiver is separated from the transmitter, it is possible to
keep the transmitter probe fixed and to move only the receiver to different probe positions or to change th e
position of the transmitter simultaneously . In the following a scanning laser Doppler vibrometer is used a s
a scanning ultrasonic receiver /8/ . In fig . 4 several transmitter positions have been selected and at each
transmitter position a two dimensional aperture on the surface opposite to the transmitter position has bee n
scanned by a laser vibrometer. The data have been superimposed to achieve best quality in the image .
Using these data the reconstruction calculation by means of 3D-SAFT is performed and results in a thre e
dimensional image of reflected and scattered objects from the inside of the specimen

Fig. 3 Scanning arrangements for acoustic imaging with SAFT

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The SAFT algorithm focuses the received signals to any point of the reconstructed image by coheren t
superposition; that amplitude of each A-scan which may originate from a given voxel due to ist time o f
flight value is calculated and averaged into the voxel. Scattered signals are statistical, true reflections not
and therefore reflectors can be imaged with a higher signal-to-noise ratio . Scatterer or indications in
concrete are localized at their geometric positions because they are not projected into a B-scan image wit h
the nominal insonification angle of the probe ; the beam opening angle has been introduced into the
algorithm and takes care of all different angles within the divergent sound field .
In fig. 4 an example of a specimen containing a notch is presented . From a 150 mm deep notch vertical
slices of the reconstructed image are shown from x- and y-directions .

Fig. 4 Side View SAFT-images from a 150 mm deep empty notc h
What advantages can we take from this procedure ?
If the crack is partially filled by particles, . they act as scattering center and are imaged in addition to the
other reflecting surfaces of the crack . This case was simulated in a specimen where a notch containt a n
aggregate which acted as a bridge from one side to the other side of the crack. – The result of the SAF T
image for this specimen is presented in fig . 5 . It shows the B-scans from the 3D .SAFT reconstruction in
two directions . The notch tip and the back-wall could still correctly imaged if one compares with th e
method explained in fig . 1 and fig . 2 . The signal from the aggregate is imaged too. In this special case the
image of the aggregate is not correctly located due to another effect . This is a mode conversion and
knowing this effect one can locate the bridge into the correct depth . The result shows, that measurement o f
a crack depth is possible under site conditions using the imaging system described . This has been verified
in addition by experiments at real cracks .

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s

Fig. 5 Correct depth sizing despite a contact spot in the notch due to SAFT-applicatio n
2.

Practical application of ultrasonic testing on a motorway bridg e

In the following the application of the ultrasonic array method including 3D-SAFT reconstruction is
demonstrated for the localization of honey combing and compaction faults of a base slab with a thicknes s
of 300 mm. The specimen has been produced using concrete with a maximum aggregate size of 32 mm an d
contains twofold layers of reinforcement bars on the top and at the back side . The diameter are 25 mm and
the mesh wide 125 mm. The faults were integrated as grains with one constant diameter and styrodur balls .

Fig . 6 Investigation of a 150 long bridge with a base plate and reinforcement
The transducer template was moved over the surface with a step width of 20 mm . The reconstruction in fig.
7 shows a vertical slice from the 3D-SAFT reconstruction . The upper and lower reinforcement laye r
perpendicular to the moving direction of the transducer template and the back wall echo are clearly
detected. At x = 880 mm and z = 160 mm the reflection from honeycombing is revealed . This is confirmed
by the shading of the back wall. A second compaction fault is seen from the shading of the back wall at th e
position of y = 120 mm without direct echo . The localization of the defects matches with the construction
plane with an uncertainty of several cm only. In addition it was possible to image the second layer of th e
reinforcement bars . They could not be detected using the impulse radar technology .

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Fig. 7 3D-SAFT : B-Scan of base plate with reinforcemen t
and two artificial flaws 88 mm x 34 m m

In a further application ultrasonic measurements on a post-tensioned concrete bridge deck have bee n
performed . Fig. 8a depicts a 1 m long section through the duct ; duct and back wall are visible on the left
side . A vertical crack – at the position of x = 150 mm –is the reason that neither the back wall nor the duc t
could be imaged. In fig . 18 b and fig. 18 c other sections of the concrete bridge are shown where an
overlap occurs .

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b.)

Fig. 8 Application of ultrasonic inspection on concrete motor way bridges
The results of a detailed analysis can be summarized :
The acquisition of high frequency ultrasonic data with signal processing by an imaging scheme lik e
SAFT allows to present an image where the direction and the concrete coverness thickness can b e
interpreted .

It has been demonstrated that the array system together with 3D-SAFT reconstruction calculation ca n
be used for the examination of transversal prestressed ducts having a concrete cover of about 100 mm .
The system has already been successfully used on site .

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3.

Acknowledgment s

The 3D-SAFT development was supported by the BMWi (Bundesministerium für Wirtschaft) in the fram e
work of reactor safety program . Part of the work was developed in the frame of research initiative „NDE o f
Concrete Structures using Acoustic and Electromagnetic Echo Methods“ by research funding of th e
Deutsche Forschungsgemeinschaft (German Science Foundation), of the Bundesanstalt für Straf3enwese n
(Federal Highway Research Instiutute) and by Deutsche Bahn AG (German Railway Corporation) . These
supports are gratefully acknowledged.
4.

Literature

/1/ Krause, M., Mielentz, F . und Milman, B .: „Machbarkeitsstudie zur zerstdrungsfreien
Risscharakterisierung in Festen Fahrbahnen“ ; Abschlussbericht und Anhang, BAM, Juni 2000
/2/ Mielentz, F ., Milman, B . und Krause, M .: „Machbarkeitsstudie zur zerstdrungsfreie n
Risscharakterisierung in Festen Fahrbahnen, Phase 2“ ; Weiterentwicklung der Verfahren für real e
Risse in Festen Fahrbahnen, BAM, Sept . 200 1
/3/ Mielentz, F ., Milman, B ., Krause, M . und W . Müller, Zerstdrungsfreie Risscharakterisierung i n
Betonbauteilen mit Ultraschall,in : Fachtagung Bauwerksdiagnose - Praktische Anwendunge n
Zerstdrungsfreier Prüfungen, 25 .-26 . Oktober 2001 in Leipzig, DGZfP-Berichtsband auf CD ,
Posterbeitrag 24, Berlin (200 1
/4/ Kroggel, O .: “Ril3beschreibung mittels Ultraschallreflexion “ Structural Faults and Repair-97 ;
Concrete and Composites . Vol . 2 . Proceedings of Seventh Int . Conf. On Structural Faults and Repair,
Edinburgh, GB, 9 th . July, 1997, p . 415-42 1
/5/ Frederick, J.R., Seydel,J .A., and Fairchild, R.C.: “Improved Ultrasonic Nondestructive Testing o f
Pressure Vessels, NUREG-0007-1 report ; U .S. Nuclear Regulatory Commission, University o f
Michigan, Department of Mechanical Engineering, Ann Arbor, Michigan, 197 6
/6/ Müller, W., Schmitz, V ., and Sch~fer, G. : “ Defect Sizing and Classification using HOLOSAFT” ,
6th Int.Conf
Proc .,
. NDE in the Nuclear Industry, American Society for Metals, Metal Park, Ohio, 153 158, 1983
/7/ Müller, W .: “Betonprüfung” ; IZFP Report No . 940109-E, Institut für Zerstdrungsfreie Prüfverfahren,
Saarbrücken, Germany, 1994
/8/ Krause, M ., Müller, W. and H. Wiggenhauser, “Ultrasonic Inspection of Tendon Ducts in Concret e
Slabs using 3D-SAFT”, Acoustical Imaging Vol . 23 (1997) pp . 433-439

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Structural Integrity Evaluation of Kori-1 NPP Steel Containment for the Replacement of Stea m
Generato r
Yong-Pyo Suh, Korea Electric Power Research Institute, KORE A
Jong-Rim Lee, Korea Electric Power Research Institute, KORE A
Yeon-Seok Jeong, Korea Institute of Nuclear Safety, KORE A

ABSTRAC T
In order to replace the steam generator at Kori unit 1 NPP, in 1998, a comprehensive program for th e
structural safety evaluation of steel containment was performed. Based on the replacement process, th e
check list for inspection was made to do the replacement work efficiently and prevent any mistakes .
Overall replacement process was appropriately formulated based on regulatory requirements related t o
nuclear design and repair. As a result of inspection according to the check list, several problems were
found such as adhesive defect at interface between old and fresh concrete and appropriately corrected .
FEM analysis in order to determine the ultimate load of cylindrical steel shell with opening wa s
performed using ABAQUS . Stress concentration and second order deformation due to crane load was
investigated through FEM analysis considering inelastic large deformation . It was verified that the current
approximation analysis using combined elastic buckling criteria gives conservative results . The analytical
result has shown that the structure follows elastic load-deflection behavior under the given crane loa d
condition with safety factor of 10 .3. The result of this study will give useful information to the replacemen t
of the steam generator of a nuclear power plant .
1. INTRODUCTION
In the field of electric power generation, an importance of nuclear power has been increased becaus e
of its large portion of electric facilities in Korea . Recently, extension of lifetime in the field of maintenanc e
of nuclear power plants has been main concern from economical point of view . Among equipment in plant,
the steam generator is one of the most important component that affect the lifetime of nuclear power plant .
KEPCO carried out the replacement of steam generator at Kori-1 nuclear power plant in 1998 . In
order to upgrade the steam generator to be safer and more stable, KEPCO planned the Kori-1 SGR(Stea m
Generator Replacement) project[1,3] . The containment vessel consists of a 32m in diameter and 44.5m in
height cylindrical shell with thickness of 36 .5mm and a spherical cap with thickness of 19mm .
In order to accommodate the steam generator replacement, an about 7m×7m opening hole was made
temporarily as shown in Fig. 1 and filled back after replacement. A built-in polar crane was to be used fo r
lifting and transporting the steam generators for the replacement .
To ensure the structural safety of the containment vessel, a comprehensive program for the structura l
safety evaluation of the Kori-1 Nuclear Power Plant was started in 1998 and was closed successfully . In
the program, the main processes of inspection were defined and detailed check list was developed. There
are many concerns in the constructing of safe structure. Specially, the stability of steel containment vesse l
against crane loading was considered as an important factor to guarantee the safety of the whole structure .
The non-linear analysis of the steel containment vessel was performed in order to determine the ultimat e
load of steel vessel with opening .

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Fig. 1 Containment Vessel Cross Section for Kori-1 SGR Project
2. MAIN PROCESS AND CHECK LIST FOR THE INSPECTIO N
The inspection to containment vessel should be performed in accordance with nuclear regulator y
requirements, such as regulatory guide, and ACI, ASME, and ASTM code, and so on . In order to ensure
structural soundness and functionality of containment vessel, main process for the inspection in this projec t
was defined as the following.
1.
2.
3.
4.
5.
6.

The cutting process of concrete shield building and rigging process of concrete block .
The cutting process of steel containment vessel and rigging process of steel plate .
Damage of containment facilities such as polar crane bracket during rigging process of stea m
generator .
The welding process of steel containment vessel .
Defect identification of welding part and repair process .
The process of concrete mix and production .
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7.
8.

Pressure test of steel containment vessel .
Defect identification of opening concrete containment and repair process .

Detailed check list for the inspection was developed to do the replacement work efficiently an d
prevent any mistake and attached in appendix A .

(a) Axial Compression
only

(b) Axial Compressio n
with Internal Pressur e

(c) External Pressur e
onl y
Fig. 2 Basic Loading Components of Buckling Stresses for Cylindrical Containment Vessel

3. STABILITY OF CONTAINMENT VESSE L

The primary concern in the Kori-1 SGR project was to investigate the stress distribution around th e
opening hole resulting from the crane operation during the period of replacement of steam generators . The
stress resultants from the SAP90 finite element analysis were put into the “buckling criteria” prepare d
originally for the design of Kori-1 nuclear power plant in the previous work[4] .

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Fig. 2 shows basic loading components of buckling stresses for the cylindrical containment vesse l
under various loading ; (a) the axial compression only, (b) the axial compression with internal pressure, (c )
the external pressure only, (d) the torsion only .
Equations for the basic components of buckling stresses were derived by using the theory o f
mechanics and supplemented by experimental observations . In application of the basic buckling stres s
components to the cylindrical containment vessel subjected to the various types of loading, they wer e
combined by two or three. The analysis revealed that the hole at the steel vessel did not decrease the
capacity of the vessel .
However, when a circular cylindrical containment vessel is under the various loading, shell element s
around the opening hole may be subjected to more severe stress concentration than the other parts . Sinc e
the probable failure mode of the shell elements in such a condition has not been clearly defined, th e
buckling criteria may not be directly applicable to this case, in which an opening hole exists, as was to the
containment vessel of Kori-1 SGR, in which an opening hole is not existing . It should be noted that the
buckling criteria were originally derived for the containment vessel in service in which there may be som e
internal pressure existing due to the operation of steam generators . However, in Kori-1 SGR project a
temporary opening hole is existing in the containment vessel that means any internal pressure cannot b e
generated during the period of replacement. Accordingly, it should be recognized that the loading
condition over the containment vessel with and without an opening hole is very much different. The
element around the opening hole may be subjected to rather yielding than buckling .
4. ULTIMATE STRESS ANALYSIS OF STEEL CONTAINMENT VESSEL
Because the element around the opening hole of a containment vessel may be subjected to rathe r
yielding than buckling, nonlinear analysis considering plastic deformation was required . So, it is performed
ultimate stress analysis, considering geometric and material non-linearity, of cylindrical containment vesse l
using finite element program ABAQUS in order to estimate the stability of containment vessel with a
7m×7m opening subjected to crane loading at both edges of polar crane . To evaluate the effect of stiffener
around the opening, the cylindrical containment vessel with or without stiffener was analyzed .

Fig. 3 3-D Finite Element Mode l

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Fig . 3 shows the 3-D analytical model for investigating the ultimate structural behavior o f
containment vessel . The load applied to finite element model was increased and the stress an d
displacement was calculated. Fig. 4 shows Von Mises stress distribution around the cylindrica l
containment vessel subjected to ultimate loading without stiffener and Fig . 5 shows Von Mises stres s
distribution around the cylindrical containment vessel subjected to ultimate loading with stiffener .
Fig . 4 depicts the relation of crane loading vs . displacement at node 2431 (which is located the pola r
crane in Fig . 3) . X axis in Fig. 4 shows the displacement at node 2431 and Y axis is the magnificatio n
factor of crane loading .
In the Fig. 5 and Fig . 6, it is shown that an installing stiffener around opening is effective to prevent
deformation and stress concentration of the opening . In the Fig . 7, when the value of magnification facto r
of crane loading is within 1, cylindrical containment vessel behaves linear elastic . The maximum value o f
Von Mises stress is 2 .47×103 t/m2 at area around opening with stiffener and is 9% value of Von Mises' s
yield criteria 2 .67×104 t/m2

25

20

0 .0 4

Vertical Disp . of Node 2431 (unit : M )

Fig. 4 The Relation of Crane Loading vs . Displacement at Node 243 1

Fig. 5
Von Mises Stress Distribution
subjected to Ultimate Loading without

Fig. 6 Von Mises Stress Distribution
subjected to Ultimate Loading with
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Stiffener

Stiffener

5. CONCLUSION
In order to replace the steam generator at Kori unit 1 NPP, in 1998, a comprehensive program for th e
structural safety evaluation of steel containment was performed. Based on the replacement process, th e
check list for inspection was made to do the replacement work efficiently and prevent any mistakes .
Overall replacement process was appropriately formulated based on regulatory requirements related t o
nuclear design and repair. As a result of inspection according to the check list, several problems were
found such as adhesive defect at interface between old and fresh concrete and appropriately corrected .
It was recognized that the direct application of the buckling criteria[2] for the Kori-1 SGR seems to
be inappropriate . Therefore, in order to estimate the ultimate capacity of containment vessel with a n
opening, ultimate stress analysis was performed in consideration of geometric and material non-linearity o f
cylindrical containment vessel against crane loading at both edges of polar crane . As a result, it is found
that when the cylindrical containment vessel is in service loading state, the vessel showed linear elasti c
behavior and service load is about 9% of ultimate load capacity . Therefore, it is concluded tha t
strengthening design due to opening is not needed .
REFERENCE S

Bechtel-Hyundai, "Calculation Sheet for Kori-1 Steam Generator Replacement,"
Job No .23213 ,
1996 .
Lee, A .J.H., "Buckling of Shells under Various Types of Loading," Technical
Report, Gilbert
.
Associates, Inc., 1971
Westinghouse Electric International Company, "Containment Vessel Design Report for Kori
Nuclear Power Station - Unit #1," 1972 .
Lee, K.J., et al ., “Structural Integrity Evaluation of Kori-1 NPP Steel Containment for th e
Replacement of Steam Generator”, SMIRT 15, Seoul, Korea, 1999 .

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APPENDIX A
Check List for the Inspection in the Replacement of Kori-1 Steam Generato r
Object
Check Point s
installation of temporary opening o n 1 . corresponding work procedure and design drawings
the wall of steel containment vessel 2 . cutting process and condition of temporary openin g
after cutting
installation of temporary opening o n 1 . whether the cutting work of concrete block is
the wall of the concrete shield
performed under the appropriate operating mode or
building
not.
2. whether rigging of concrete block damages the stee l
containment vessel and concrete shield building o r
not.
recovery of temporary opening on 1 . whether the surface treatment and cleaning o f
the wall of the steel containment
welding part are performed appropriately or not.
vessel
2. welding procedure and qualification of welding
personnel and welding material
3. defect identification of welding part and
appropriateness of repair work
4. structural integrity evaluation of steel containment
(pressure test)
recovery of temporary opening on 1 . corresponding work procedure and design drawings
the wall of the concrete shield 2. cutting process and condition of temporary opening
building(rebar work)
after cutting
3. appropriateness of mechanical rebar splice method
4. appropriateness of rebar welding method
5. whether rebars are installed appropriately or no t
recovery of temporary opening on 1 . whether the physical and chemical characteristics o f
the wall of the concrete shiel d
cement and admixture are in compliance with th e
buildin g
corresponding technical guidelines or no t
(concrete mix and production)
2. whether the certification and confirmation tests o f
fine and coarse aggregate are performed
appropriately and the those results are acceptable o r
no t
3. certification of testing personnel and appropriateness
of testing equipment
4. appropriateness of administrating test
5. appropriateness of concrete mix desig n
6. certification of batch plant and inspection and
correction of measuring devices
7. whether the row material and temperature control o f
hot weather concrete are appropriate or not
recovery of temporary opening on 1 . appropriateness of chipping work, cleaning, wettin g
the wall of the concrete shield
in the contact surface between old and fres h
building(concrete work and curing)
concrete
2. appropriateness of form work design and installation
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Object

Check Point s
3. whether the mortar thickness at the contact surfac e
between old and fresh concrete is conformed to the
requirement of technical guidelines or no t
4. whether the concrete pouring work is performed i n
compliance with the temperature requirement of ho t
whether concrete or no t
5. whether the concrete quality assurance is performed
in compliance with technical guidelines and tha t
result is appropriate or not.
6. appropriateness of the pouring and vibration of fresh
concrete
7. occurrence of cold joint in hardened concrete
8. appropriateness of form removal based on th e
technical guideline s
9. appropriateness of curing method
10 . occurrence of defects such as honeycombing an d
voids etc .
11 . appropriateness of concrete repair material and
method
12 . appropriateness of disposition method for the no n
conformance things and those result s
structural integrity of containment 1 . whether rigging of steam generator damages the
internal structures
operating floor and steel containment and concret e
shield building or not
2. whether rigging of steam generator damages the
polar crane bracket or not

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New Methods for the Reconstruction of Safety Compartments of Nuclear Power Plant s

Dieter Busch, RWE Solutions AG, Essen, German y
Prof. Dr . H.-D . Kdpper, Zerna, Kdpper & Partner, Ingenieurgesellschaft für Bautechnik,
Bochum, Germany
Peter Holdt, Zerna, Kdpper & Partner, Ingenieurgesellschaft für Bautechnik ,
Bochum, Germany

Safety Compartments – Challenges and Construction Specialitie s
Safety compartments are essential structures for the operation of nuclear power plants . Due to their
importance these concrete structures have to be carefully observed . Certain special requirements must b e
fulfilled in the construction phase of the structures . When these requirements are not fulfilled the chance o f
damages occurring is great . Simultaneously these structures demonstrate a series of distinctiv e
constructional features :

Picture 1 : Total view of a safety compartment whilst reconstructio n

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A safety compartment is a symmetrical structure with a cylindrical base and crowned with a half-dome
(Pic . 1) . The wall thickness varies up to 2 meters in width . Furthermore the diameter of the reinforcemen t
is very large. Due to constructional changes, the concrete cover can vary and in some instances very small .
Simultaneously the high number of and large diameter of stirrups present near the concrete surface are als o
a cause of the damages .
Found mainly in older structures, the distribution of reinforcement for the reduction in surface cracking i s
missing, and therefore cracking does occur .
Damages to be restore d
From the inspections of the concrete surfaces, the following was discovered :
The carbonization depth was up to 15 mm and in some areas the concrete showed detoriation . (Pic . 2)
Due to the weather conditions, rain water containing CO 2 and the influence of wind, the concrete surfac e
became washed out . Additional, found mainly on the top of the safety compartments, microorganism wer e
observed. These organisms have started to produce acids, which added to the wash out of the surface.

Picture 2 : Concrete surface of a safety compartment with damages and detoriations
Of greater importance are the surface cracks present . These cracks have reached deep into the concrete,
some of which have reached the reinforcement . These cracks have occurred either due to temperatur e

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changes that were not taken into account during the construction design, or even from shrinkage crackin g
occurring shortly after the completion of construction .
Demands for upgrade

At none of the inspected safety compartments were reconstruction measurements necessary . All the works
carried were to minimize future repairs and to keep the state of art . There was at no time any doubt for the
safety of the concrete structure or the power plant . Furthermore, to reassure the public, nuclear powe r
plants should present a perfect optical appearance .
The main objective of the works at the surface of the safety compartments was to stop any furthe r
carbonization . Simultaneously the further development of cracking had to be stopped and existing crack s
had to be closed.

Concepts for restoration
The development of the restoration concept was based on the data and experiences gained from the repai r
of other concrete surfaces such as cooling towers, stacks, foundation walls, bridges, etc . Their evaluation
as well as the first tests on the safety compartments showed that stiff repair systems even with flexibl e
coatings would not achieve a satisfactory solution . Therefore the aspect of crack bridging and weather
protection was addressed in two steps, simultaneously looking for an architectural satisfactory solution .
This concept with its optimal results was only possible due to the new developments of repair and coatin g
materials achieved in the recent years . Table 1 shows the different measures taken in the last twelve year s
on German concrete safety departments .

Table 1 : Development of repair systems for concrete surfaces in the last twelve year s
The following concept resulted from these experiences :
-

-

Preparation of the concrete by detecting and removing all loose areas by hand and with the use of a
hammer
Blasting of the complete concrete structur e
Preparation of all corroded reinforcement and application of a protective coating .
Injection of cracks when possible, i .e. the wide cracks
Reprofiling of the damaged and deteriorated areas with a PCC Mortar
Closing of pores with a stiff mortar
Application of a mineral based crack-bridging coatin g
Final application with an elastic, pigmented top coat based of pure acrylic resi n

An area of concern was at the edges of the hollow areas, most of which had occurred due to architectura l
reasons during the construction. As shown in Picture 3 – at the edge there is the danger that the coverag e
may become too thin while on the inside cracks could occur due to too much material being place d

143

Picture 3: Cross – section of a fluting, showing minimal and excess material thicknes s
In former times with the materials available, it was not possible to solve these problems and so hollow s
were completely filled with a PCC mortar before applying a crack-bridging mortar . At the “minimum
restored areas” new materials were used. These new materials were able to take care of the problems . None
of the different techniques used showed a lack in the optical appearance .

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For all the materials that were developed and tested severely, it was of great importance that the material s
0
remain flexible even at temperatures of minus 20 C, as well as keeping at least 50% of its flexibility eve n
after 20 years . These demanding requirements were tested on existing safety compartments . The results
showed crack movements only due to temperature changes .

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Scaffolding s
An important device for undertaking the repair work is the use scaffolding . In the past a scaffolding wa s
constructed around the whole compartment . This was not only expensive and ineffective, but also did not
comply to the existing safety standards within a German nuclear power plant . These scaffoldings had to b e
guarded to ensure that only restricted personnel were allowed to step on the scaffolding . Another problem
of this scaffolding construction was the many anchor points into the concrete surface . These anchor point s
had to worked on after the scaffold was removed .
Therefore a movable scaffolding/platform was developed, which is fixed to one huge anchor bolt on th e
very top of the dome . This scaffolding/platform is circulating on wheels around the dome . This system
worked well with the old protection systems coating the concrete . But when the flexible protective system s
were used, the forces of the wheels started to damage the applied mortar, as shown in picture 5 .

Picture 5: Damages on the repair system caused by high load on scaffolding wheel s
Therefore a new light weight cradle-type scaffolding was designed, as shown in picture 6 . This lighter
scaffolding was used without any problems on the last safety compartments . The light weight scaffolding
could move in all directions more easily, and like the former model was also fixed to an anchor bolt on th e
top.

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Picture 6 : New designed light weight scaffolding/platfor m

Concrete protection with crack – bridging system s
The repair of the concrete surface is started, as is the norm with all other concrete structures, by blastin g
and cleaning the concrete and the all open reinforcement. The reinforcement has then a protective coatin g
applied, and then the concrete is re-profiled . What is unique about the reported system here is the use of a
crack–bridging system for the first time. Due to the costs and time it was agreed to use a fine mortar to
close all the pores and small caverns in the concrete surface before applying the flexible system .
Through experience it was observed that leaving out the scratch mortar would result in damaging th e
bridging layers . This is due to the lack of thickness or small holes, as shown in picture 7 .

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Picture 7: Surface of the crack – overbridging material with pores and hollow s
The optimal application method for crack–bridging, is to wet spray apply it onto the surface . By optimizing
every application step it is possible to achieve the required thickness of the layer and to achieve a surfac e
which can accepted the final coating .
An important area of concern during the application is to pay attention to the change of climate, especiall y
the change in humidity. If these conditions are not regarded carefully, “Calcium-hydroxide / Calcium carbonate” may develop, which will influences the color and physical properties .

Picture 8 : Perfect surface of the crack overbridging surface protection system
148

NEA/CSNI/R(2002)7/VOL 2

Quality Contro l
Most important, beside the use of proper materials and correct techniques is the installation of a qualit y
control system from the very beginning . Before the start up of works a quality handbook was created by
the contractor, the owner of the plant and the supervising engineer . This quality handbook contains all the
quality checks and the production documentation. Its contents also include the various times and
specifications that the contractor has to control during the repair .
It was also agreed upon how often the supervising engineer would make his quality control checks . And
that before the next stage of repair commenced the supervising engineer had given his approval . The
reason for this is that the preparation priming coat can not be checked after the coating later .

One of the most important measurements is the determination of the film thickness, for example with the
use of a penetrometer, but other methods without destroying the coating are better . The main goal for the
repeated controls, for which the costs were calculated in advance and paid separately, was to reach a
constant and optimal thickness of the crack–bridging and the final layer of coating . If these layers are too
thick possible problems may occur in the future .

Picture 9 : Partial View of a restored concrete surface of a safety compartmen t
Future Aspects
There are no problems with the applied coatings at present . The materials and techniques used on the
safety compartments can be used of course on other concrete structures such as cooling towers and stacks .
One problem in Germany has still not been solved . As shown in picture 10, the pigeons are attracted to
these domes, and since their droppings are quite aggressive . Therefore the coating in the pole area has to b e
checked carefully in the future .

149

NEA/CSNI/R(2002)7/VOL 2

Picture 10 : Completely restored surface, biological attack caused by birds residue s

150

NEA/CSNI/R(2002)7/VOL 2
E. LIST OF PARTICIPANTS

CANAD A

Mr. POPOVIC
Manager of Civil Engineerin g
AECL
2251 Speakman Drive
MISSISSAUGA, Ont L5K 1B 2

Tel: +1 (905) 823 9060 X213 5
Fax: +1 (905) 855 947 0
E-mail: [email protected]

Mr . Claude SEN I
Mattec Engineering Ltd.
William Carson Crescent 217-21 8
North York
Ontario M2P 2G6

Tel: +1 (416) 224 575 1
Fax: +1 (416)224 575 1
E-mail: [email protected]

BELGIU M
Mr. Luc DE MARNEFFE
Principal Engineer
TRACTEBEL
7, Avenue Arian e
B-1200 BRUSSEL S

Mr. Roland LASUDRY
TRACTEBEL
7, Avenue Ariane
B-1200 BRUXELLES

CZECH REPUBLIC
Mr. Jan MALY
Energoprojekt Praha a.s .
Vyskocilova 3
P .O. Box 15 8
140 21 Prague 4

Tel: +32 2 773 81 4 8
Fax : +32 2 773 89 7 0
Eml : luc .demarneffe@tractebel .be

Tel:
Fax:
Eml : roland .lasudry@tractebel .be

Tel : +420 2 41006 EXT 42 0
Fax : +420 2 41006 40 9
Eml : [email protected]

Mr. Ladislav PECINKA
Senior Research Worker
Division of Integrity and technical Engineering
NRI Rez
Vltavska 2
25068 REZ

Tel : +420 2 2094 11 12
Fax : +420 2 2094 051 9
Eml : [email protected]

Mr. Jan STEPAN
Energoprojekt Praha a.s .
Vyskocilova 3
P.O. Box 15 8
140 21 Prague 4

Tel : +420 2 4100642 1
Fax : +420 2 4100640 9
Eml : stepan@egp .cz

15 1

NEA/CSNI/R(2002)7/VOL 2

Mr. Petr STEPANEK
Associate Professor
Brno University of Technology
Faculty of Civil Engineerin g
Department of Concrete and Masonry Structure s
Udolni 53 - 602 00 Brn o

FINLAND
Dr. Pentti E. VARPASUO
Fortum Engineering Ltd .
POB 10, 00048 Fortum,
Rajatorpantie 8, Vantaa
FIN-0101 9
Erkki VESIKARI
Lic . Sc . (Tech.), Senior Research Scientist
VTT Technical Research Centre of Finlan d
Kemistintie 3, Espo o
P .O. Box 1805, FIN-02044 VTT, Finland

FRANCE
Mr. TOURET
EDF-Septen
Basic Design Dept .
Engineering and Construction Division
12-14 Dutriévoz Avenue
69628 VILLEURBANNE CEDEX

Tel : +420 5 4114 620 5
Fax : +420 5 4321 210 6
Eml : stepanek.p@fce .vutbr .cz

Tel: +358 10 45 3222 3
Fax : +358 10 43 32022
Eml : [email protected]

Tel : +358 9 456 6922
Fax : +358 9 456 700 3
E-Mail : [email protected]

Tel: +33 4 72 82 71 9 4
Fax: +33 4 72 82 77 0 7
E-mail: [email protected]

Mr. Jean Mathieu RAMBACH
DES/SAMS CE FAR
CEA/IPSN
60-68 Avenue du General-Lecler c
B .P. 6
92265 Fontenay aux Rose s

Tel: +33 (0)1 4654 8028
Fax : +33 (0)1 4746 101 4
Eml : mathieu.rambach@ipsn .fr

Mr. Olivier STRICH
IPSN
Département d'Evaluation de Sûreté
B .P. 6
92265 Fontenay aux rose s

Tel: +33 (0)1 46 54 93 2 8
Fax : +33 (0)1 47 46 10 1 4
Eml : olivier.strich@ipsn .fr

GERMANY
Mr. Dieter BUSCH
Junior Assistant Manager
RWE Energie AG Bereich Bau
Kruppstrasse 5
45117 ESSEN

Tel: +49 + 49 201 12 24 47 6
Fax: +49 + 49 201 12 22 48 6
Eml : dieter.busch@rweplus .com

152



NEA/CSNI/R(2002)7/VOL 2
Mr. Peter HOLDT
Project Managaer
Zerna, Kdpper und Partner
Industriestrasse 27
44892 Bochum

Tel: +49 234 9204 18 5
Fax: +49 234 9204 15 0
E-mail: hol@zkp .de

Mr. Christoph NIKLASCH
Research Assistant
University of Karlsruhe
Institut für Massivbau
P.O. Box 698 0
D-76128 Karlsruhe

Tel : +49 721 608 227 5
Fax : +49 721 608 2265
Eml : christoph .niklasch@ifmb .uni-karlsruhe .de

Dr. Rüdiger MEISWINKEL
E.ON Kernkraft GmbH Zentrale
Nuclear Systems & Components Division
Tresckowstrasse 5
D-30457 Hannover

Tel : +49 0511 439 290 6
Fax : +49 0511 439 4144
E-mail:[email protected] m

Dr. Volker SCHMITZ
Head of Department Quantitative NDE
Fraunhofer IzfP
Universitaet, Geb 3 7
D-66123 Saarbruecke n

Tel : +49 681 9 302 387 0
Fax : +49 681 9 302 593 0
Eml : schmitz@izfp .fhg.de

Dr. Friedhelm STANGENBERG
Stangenberg und Partners, Consultants
Ingenieur-GmbH
Viktoriastrasse 4 7
D-44787 BOCHUM

Tel : +49 + 49 (0) 234 961301 2
Fax : +49 + 49 (0) 234 9613048
Eml : [email protected]

Dr. Herbert WIGGENHAUSER
Tel : +49 (0) 30 8104 1440
Fax
Bundesanstalt für Materialforschung
: +49 (0) 30 8104 1447
und -prüfung
Eml : herbert.wiggenhauser@bam .de
Division IV .4 Non-Destructive Damag e
Assessment and Environmental Measurement Method s
Unter den Eichen 8 7
D-12205 BERLIN
Herr Rüdiger DANISCH
FRAMATOME ANP GmbH
NDA2
P.O. Box 322 0
91050 Erlangen

Tel: +49 (0) 9131 189 342 6
Fax : +49 (0) 9231 189 759 9
E-Mail : ruediger.danisch@framatome-ANP .de

Herr Andreas KOCHAN
MC Bauchemie Bottrop GmbH
Am Kruppwald 2-8
46238 Bottrop

153

NEA/CSNI/R(2002)7/VOL 2

HUNGARY
Mrs . Katalin GYARMARTH Y
Engineer
Nuclear Power Plant Co .
PAKS

Tel : +36 75 50 88 6 3
Fax : +36 75 50 65 3 5
Eml : gyarmathy@npp .hu

Prof. Dr. Peter LENKEI
Professor of Structural Engineering
PÉCS University, College of Engineering
H-7624 PECS
Boszorkany U.2

Tel : +36 72 224 268 ext 723 7
Fax : +36 72 214 26 8
Eml : lenkeip@witch .pmmf.hu

Mr. Csaba NYARADI
Systems Technologist
Nuclear Power Plant Paks
H-7031 PAKS P .O.B. 7 1

Tel: +36 7550 705 4
Fax : +36 7550 7334
Eml : nyaradi@npp .hu

Mr. Oliver KAKASY
Resident Inspector
Hungarian Atomic Energy Authority
Nuclear Safety Directorat e
H-1539 Budapest 11 4
P.O.B. 67 6

Tel : +36 (75) 508 93 9
Fax: +36 (75) 311 47 1
E-mail: [email protected] .hu

ITALY
Mr. Alberto TAGLIONI
ENEA
Via Anguillarese 301
I-00060 ROMA
Dr. Lamberto D’ANDREA
SOGIN S.p.A .
Via Torino 6
00184 Rome

JAPAN
Mr. Takaaki KONNO
Secretariat of Nuclear Safety Commission
Cabinet Office
Technical Counsellor
3-1-1 Kasumigaseki, Chiyoda-k u
Tokyo 100-8970

Tel : +39 06 30483 3628
Fax: +39 06 3048 630 8
Eml : [email protected]

Tel: +39 (0) 6 83 04 03 5 0
Fax: +39 (0) 6 83 04 04 7 4
E-mail: [email protected]

Tel : +81 3 3581 9842
Fax : +81 3 3581 983 6
Eml : tkonno@op .cao .go.jp

154

NEA/CSNI/R(2002)7/VOL 2

KOREA (REPUBLIC OF )
Dr. Yun Suk CHUN G
Principal Research Engineer
Korea Institute of Nuclear Safety
19 Guseong,Yuseong
Taejon 305-33 8

Tel : +82 42 868 053 3
Fax : +82 42 861 994 5
Eml : k063cys@kins .re.kr

Mr. Yun-Suk CHUNG
Research Project Manager
Korea Institute of Nuclear Safety
19 Guseong, Yusung,
Taejon 305-33 8

Tel: +82 42 868 053 3
Fax : +82 42 861 994 5
Eml : k063cys@kins .re.kr

Dr. Jeong-Moon SEO
Project Manager
Korea Atomic Energy Research Institut e
P.O. Box 105
Yusong
Taejeon 305-60 0

Tel : +82 42 868 839 1
Fax : +82 42 868 837 4
Eml : jmseo@kaeri .re.kr

Mr. Yong-Pyo SU H
Senior Member of Technical Staff
Korea Electric Power Research Institute
103-16 Munji-Dong
Yusong-Gu
Taejon 305-380

Tel: +82 42 865 579 1
Fax: +82 42 865 5504
E-mail: ypsuh@kepri .re.kr

SLOVAKIA
Mr. Juraj NOZDROVICKY
Project Manager
VUEZ, a.s.
Sv . Michala 4,
P.O. Box 15 3
934 80 LEVIC E
Mr. Milan PRANDORFY
Project Manager
VUEZ, a.s.
Sv. Michala 4
P .O. Box 15 3
934 80 Levice

Tel : +421 36631 366 5
Fax : +421 36631 366 3
Eml : mechanika@pobox .sk

Tel : +421 3663 13665
Fax : +421 3663 13663
Eml : [email protected]

155



NEA/CSNI/R(2002)7/VOL 2
SLOVENIA
Mr . Lojze BEVC
Head of Structural Department
Slovenian National Building and
Civil Engineering Institut e
Dimiceva 12,
St-1000 LJUBLJANA
Mr. Bozo KOGOVSEK
Project Manager
IBE Consulting Engineers
Hajdriho va ul 4
1000 Ljubljana

Tel : +386 1 28 04 48 7
Fax : +386 1 28 04 484
Eml : lojze .bevc@zag .zi

Tel: +386 1 477 62 03
Fax: +386 1 251 05 27
E-mail : bozo [email protected] i

SPAIN
Mrs . Dora LLANOS
Head of Civil Structure Section
NUCLENOR, S .A .
Calle Hern àn Cortés 2 6
39003 Santander

Tel: +34 942 245 100
Fax : +34 942 245 12 3
Eml : dora .llanos@nuclenor .es

Mr. Jesus GARCIA ROCASOLANO
TECNATOM
Avenida Montes de Oca No . 1
28709 San Sebastian de los Reye s
MADRID

Tel: +34 91 659 8726
Fax : +34 91 659 86 7 7
E-mail: jroca@tecnatom .es

Mr. Juan SABATER
Civil Engineer
C.N. ASC Ô / C.N. VANDELLÔ S II
ct . n.340 km. 1123
Apartado de Correos 4 8
43890 L'Hopitalet de l'Infant

Tel : +34 97781870 0
Fax : +3497781872 0
Eml : jsabater@anacnv .co m

Mrs . MARTINEZ SIERRA
Instituto Eduardo Torroja
Spanish Research Council
Serrabi Galvache Street
ES 28033 Madrid, Spain

Tel : +34 91 3020440
Fax : +34 91 3020700
E-Mail : [email protected]

SWEDE N
Dr. Behnaz AGHILI
Swedish Nuclear Power Inspectorate
Klarabergsviadukten 90
Stockhol m
SE-106 58

Tel: +46 8 698 869 2
Fax : +46 8 661 908 6
Eml : behnaz .aghili@ski .se

156

NEA/CSNI/R(2002)7/VOL 2

Mr. Patrick ANDERSON
Division of Structural Engineerin g
Lund University
P .O. Baox 1 8
SE-22100 LUND

Tel : +46 46 2984 3
Fax : +46 46 2421 2
Eml : [email protected]

Mr. Gabriel BARSLIVO
Swedish Nucler Power Inspectorate
Department of Structural Integrit y
(SKI)
S 10658 Stockhol m

Tel : +46 8 698 866 0
Fax : +46 8 661 908 6
Eml : gabriel .barslivo@ski .se

Mr. Jonas BERGFORS
Project Manager
Oskarshamn Nuclear Power Plant
SE-572 83 Oskarshamn

Tel: +46 491 78 79 4 3
Fax : +46 491 78 60 3 8
Eml : jonas .bergfors@okg .sydkraft.se

Mr . Jan GUSTAVSSON
Manager Y2K Projec t
Ringhals Nuclear Power Plant
Vattanfall AB
Ringhal s
S-430 22 VAROBACKA

Tel : +46 340 66 79 5 0
Fax : +46 340 66 83 8 9
Eml : jan [email protected]

Mr. Thomas ROT H
Department of Structural Engineering
Royal Institute of Technology (KTH )
SE 10044 Stockholm, Sweden

Tel : +46 (0)8 790 813 6
Fax : +46 (0)8 21 69 4 9
Eml : [email protected] .se

Mr. Thomas VIBERG
Project Manager
Oskarshamn Nuclear Power Plant
SE-572 83 OSKARSHAMN

Tel : +46 0491 786 20 5
Fax : +46 0491 786 038
Eml : thomas.viberg@okg .sydkraft.se

SWITZERLAND
Mr. Jean-Baptiste DOMAGE
Monitoring Development Manage r
VSL Schweiz AG
Industriestrasse 14
CH-4553 Subingen
UNITED KINGDOM
Mr. Robin BALDWIN
Mott MacDonald Limite d
Materials Technology Unit
Bristol Office - 0117 906 9529
Prince House, 43-51 Prince Stree t
Bristol BS1 4PS

Tel : +41 32 613 30 7 2
Fax : +41 32 613 30 7 5
Eml : jbdomage@vsl-schweiz .ch

Tel : +44 (0)181 774 200 0
Fax : +44 (0)181 681 5706
Eml : [email protected]

157



NEA/CSNI/R(2002)7/VOL 2

Dr. Tony MCNULTY
NII, HSE
Nuclear Safety directorate
St Peter’s House
Balliol Road
Bootle, Merseyside L20 3JZ

Tel: +44 151 951 362 4
Fax: +44 151 951 4163
Eml : tony .mcnulty@hse .gsi .gov.uk

Dr. Leslie M . SMIT H
Senior Civil Engineer
British Energy Generation (UK) Lt d
3 Redwood Crescent, Peel Park
East Kilbride G74 5PR
GLASGOW

Tel : +44 (13552) 6238 5
Fax : +44 (13552) 6245 9
Eml : les.smith@british-energy .com

UNITED STATES OF AMERICA
Dr. Dan NAUS
Oak Ridge National Laboratory
PV Tech Sect, Eng Tech Div
P.O. Box 2009, Bldg .9204- 1
OAK RIDGE
TN 37831-805 6
Dr. James F . COSTELLO
Office of Research
Division of Engineering Technology
Mail Stop T10-L 1
US Nuclear Regulatory Commissio n
Washington, DC 20555

Tel: +1 865 574 065 7
Fax: +1 865 574 203 2
E-mail: nausdj@ornl .gov

Tel: +1 (301)415-600 4
Fax: +1 (301)415-5074
E-mail: jfc2@nrc .gov

International Organisation s
International Atomic Energy Agency, Vienna
Mr . Paolo CONTRI
International Atomic
Energy Agency
IAEA/NS/NSNI/ES S
Wagramerstrasse 5
P.O. Box 100 A-1400 VIENNA

Tel : +43 1 26000 26426
Fax : +43 1 26007
Eml : [email protected]

158

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