Waste Management

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Dr. P. M. Satya Sai, Professor, HBNI, SO-H, Deputy General Manager, WMPC, NRB, BARC, Kalpakkam



Every industry, plant or factory requires certain inputs such as raw materials and energy to convert them into useful products and by products. In the process materials not required by consumers are also produced. The materials not required are rejected as wastes. It is quite likely that materials rejected, as waste may become by-products or important raw materials for other industries in due course. Nuclear industry is no exception. Nuclear industry for power generation depends on harnessing of fission where fissile isotopes split to produce: (i) (ii) (iii) Enormous amount of heat which is converted into electricity, Fission products i.e. lighter nuclei from fissioning of fissile materials. Activation products i.e. nuclei of fuel and surrounding materials after absorbing neutrons become radioactive.

The first of these streams is the desired product and the second and third is the price we pay. Though the nuclear wastes are not alone being dangerous to life, they have acquired special visibility in last few decades. This lecture covers the origin, nature, treatment and disposal of radioactive waste into environment.



The nuclear fuel cycle encompasses the major physical and chemical activities or steps, which are directly related to power production. The steps include:



Ores of both Uranium and Thorium are found in our country. Uranium content of ore generally range from 0.1% to 0.3%. Thorium content of monazite ore is about 8-13%. The Uranium ore is concentrated at mills where it is crushed and screened. Wet grinding is carried out in ball mills. The resultant product is reacted with H2SO4 and other oxidizing agents. The tailings are let off as waste. The solution is solvent extracted and Uranium is precipitated as Ammonium or Magnesium di-uranate. Liquid effluents from mills are waste solutions from leaching, grinding, extraction and washing cycles of the mills. The solutions have pH of 1.5 to 2 containing unreacted portion of acid used and some silica.

The air borne releases include dust, containing Uranium and its daughters released from ore piles, tailing retention system, crushing and grinding system and from general ventilation system. Radon gets released from leach tank vents, ore piles and tailing systems and ventilation.



Mag. Di-uranate cake – the yellow cake- frommills is processed to yield uranium pellets. It is dissolved in HNO3 and then precipitated as ammonium diuranate at pH 8-9 by liquor ammonia. This is then filtered, dried and sintered to produce fuel pellets. The major pollutants are liquid containing ammonia, nitrate, fluoride, detergents and uranium and gases containing NOx, and uranium dust. At Nuclear Fuel Complex at Hyderabad where besides above enriched uranium is also processed NH4F is major effluent.



This is an important step in fuel preparation for light water reactors which require enriched fuel. In our country the mainstay is PHWR which does not require enrichment. So we skip this step for time being. However a pilot plant is being operated by BARC on centrifuge technology



Nuclear reactor in India can be classified into three main categories based on coolant used. They are:


Light water reactors:

They are light water moderated and cooled reactors using enriched Uranium as fuel. TAPS is boiling water reactor (BWR) where light water converts to steam in reactor core and drives turbine to generate power. New reactors under construction in Kudankulam are pressurized water reactors (PWR) where water in core gets pressurized and no phase change takes place. The hot pressurized water produces steam from light water in boiler.


Heavy water reactors:

They are heavy water moderated and cooled reactors using natural Uranium as fuel. This is mainstay of our nuclear energy programme. This is similar to PWRs and online refueling is possible in our design.


Liquid metal fast breeder reactors:

They are generally Sodium metal cooled reactors using enriched U or U-Pu mix fuel. No moderator is required. Hot molten Sodium transfers heat from fuel to steam generators. Since neutron flux is high, it is used as a breeder, where Th 232 and U238 are converted into U233 and Pu239 respectively, which are fertile materials. In order to exploit vast reserves of Thorium< Fast Breeder Technology is

an important step in Indian nuclear energy programme. A test reactor, FBTR is operational and a prototype reactor PFBR is under construction at Kalpakkam.



Unlike fossil fuel plants that discharge ash with no fuel content, spent fuel discharged from nuclear power stations contain appreciable quantities of unburnt fissile elements and neutron activated fertile elements, which can be made use of. Fuel elements are removed from reactor core before they are fully consumed primarily because of build up of fission product, which capture neutrons essential for sustaining chain reaction. Spent fuel is reprocessed for recovery of Plutonium, which can be used as a fuel in Fast Breeder reactors. In fuel reprocessing plant fuel elements are chopped and dissolved in HNO3. The hulls are rejected as solid waste. Using TBP extraction U & Pu are separated and recovered and raffinate is stored as high active waste. Other types of waste are personal showers and washes, chemical waste from laboratories, IX regenerants, rod handling pond water and evaporator condensates etc.



Radioactive wastes are generated at every step of fuel cycle. However, nuclear reactor and reprocessing plant generate bulk of waste requiring special treatment before they are discharged into environment. Other steps generate small quantity of wastes containing toxic chemicals and natural Uranium and its daughters.


Radioactive nuclides are generated as follows:-


Fission of Uranium & Plutonium:

Though entire periodic table with exception TRU are produced in binary/ternary fissions, the following are major elements thus produced: 1. 2. 3. 4. Noble gases Kr & Xe Alkali metals Cs & Rb (Cs-137 – 30y) Alkaline earth Sr & Ba (Sr-90 – 28y) Halogen Br & I ( I-131 – 8 d)


Neutron Activation:
Neutron absorption in the coolant & structural material:

Due to heavy neutron density elements found in coolant and structural material absorb neutrons and get activated. Coolant activation products are generally of short half life except Tritium & Carbon 14. They are: 1. A41 2. F18

3. 4. 5. 6.

N13 & N16 019 H3 C14

Activation products formed in structural material have considerable half lives. They normally remain fixed but do enter into coolant from corrosion and erosion from many parts. They are Zr. Mn, Ni, Fe, C, Cr, Co, Cu. In reactor system, radioactivity is normally contained in the system and many barriers are provided which prevent release of activity into enrivonment. These barriers are: 1. Fuel cladding 2. Reactor system 3. The reactor & auxiliary buildings Release of radioactivity occurs when these barriers are penetrated. The penetration can occur due to presence of structural defects, leakages from pumps or other components or intentional release as a consequence of the particular plant design. Major portion of fission products is retained in the ceramic matrix of UO 2 pellets. More volatile elements like hydrogen s and noble gases can diffuse into gap between pellets & cladding. Daughters such as Sr, Ba, Ce etc., also accumulate. Diffusion is accelerated due to high temperature. Nuclides movement also takes place from cracks & fissure in cladding due to thermal stress. However, pin hole defect are more common compared to large failure of cladding, as it is impossible to remove the defect completely. From fuel cladding defects & diffusion, radioactive materials are transported into PHT and other auxiliary system. The radioactivity is generally retained in the system but small quantity leaks out from the gaps and other machinery. Part of liquid is decontaminated in built-in IX system which generates exhausted resins as waste.



This is chemical operation where spent fuel is decladded, dissolved in HNO 3 and U-Pu separated and purified by solvent extraction. Fission product solution generated as first cycle raffinate constitutes a well defined high level waste. Other liquid wastes are generated from equipment washings, process drains evaporator condensate, Ion-exchange regenerants etc. Hulls from decladding operation and other rejects constitute solid waste.



Various classification systems are used to categorise radioactive waste. Classification can be based on the specific activity, the dose rate or the radiotoxicity. Other classification principles might be the origin of waste, its physico-chemical nature, or type of radiation and half life of the nuclides. Each

of these classification systems has its advantages and disadvantages depending on the purpose and basis applied. Primarily, a general classification of radioactive wastes can be made on the basis of their physical state i.e. a. Liquid waste b. Solid waste c. Gaseous waste Another distinction can be made between a. Combustible, and b. Non-combustible waste In addition to this general classification, there is more detailed classification on the basis of the radiological properties of waste.



The IAEA has suggested a classification system based upon specific activity. Specific activity per unit volume can be measured in liquids relatively simply without serious problems. Four categories of low and intermediate level liquid waste according to this classification is shown below: Category I II III IV V Activity (mCi/l) < 10 –6 10 –6 TO 10 –3 10 –3 TO 10 –1 10 –1 TO 10 4 >10 4 Remarks May be discharged after monitoring Requires treatment Requires treatment Shielding is required during treatment Self boiling liquid requires heavy shielding



Untreated radioactive solid waste is mainly classified according to the radiation dose on its surface instead of activity per volume, because of the heterogeneous nature of solid waste. The classification includes three dose rate categories. Category I <0.2mR/h No shielding required Category II Category III Category IV >0.2 - <200 mR-hr >200 mR/h  contaminated Some shielding Shielding required

The first category includes solid waste, which can be transported and treated without any shielding. Intermediate level waste will always have to be handled and processed under radiation protection conditions.

Additionally solid waste is also categorized into combustible, compactable and non-combustible non-compactable categories.



The gaseous waste is characterized by the specific activity per unit volume as proposed by the IAEA. 1. 2. 3. Category I Category II Category III <10-10 Ci/m 3 >10-10 - <10-4 Ci/m 3 >10-6 Ci/m 3


A three pronged strategy is adopted for management of waste: i. ii. iii. Dilution and dispersion Concentration and containment Delay and decay



The most nuclear facilities and installations have several drainage systems, all arranged that the minimum amount of effluent arising on site requires treatment and can be segregated and that only a small proportion of the active liquids need to be diverted to higher level liquid treatment plant. These wastes are collected by pipelines. For Cat III and above, pipe in pipe construction is used dwith provision for leak detection. Small quantities of wastes can be collected in shielded tankers with appropriate safety provisions. The following system is often used:


Potentially active waste (Category I):

This waste arises from personal washes & showers. It can also include liquids from secondary cooling systems of nuclear facilities. It contains virtually no activity and can be discharged well within the authorized limits after necessary dilution.


Low level liquid wastes (Categories 2 and 3):

These wastes arise at laboratories and other areas where radioactive work is carried out, or as cleaning water or decontaminated waste streams of the higher category. In laboratories they are normally discharged into a piping system inside the building or drained into tanks outside. When the tank is full, the liquid can be mixed, sampled and analysed and, if satisfactory, discharged to the main low level treatment plant. If the activity level is too high the liquor can be diverted to the intermediate level treatment plant either by tanker or a special drain lines.

Water from cooling ponds can go by normal drainage directly or to the low level treatment plant or if considered too high, be diverted to the intermediate level plant. The bulk of the active arisings in nearly every establishment should be in this category.


Intermediate level liquid wastes (Category 4):

These streams are segregated in a variety of ways. They can be collected at source in small containers and transported to the intermediate level waste treatment plant, especially in research establishments or even have their own separate drainage system. Intermediate level liquids usually have their own special treatment facilities.


High level liquid waste (Category-5)

These wastes are generated during reprocessing of spent nuclear fuel and have very high radio activity and will be boiling. After providing cooling period of 5 to 10 years, these wastes are sent to waste immobilization plant.


Treatment of Low & Intermediate Level liquid wastes :

The objective of liquid waste treatment is to generate two different streams, one with very low volume and most of the activity and another with almost same volume as the original waste, but with very low activity, compared to the original waste. The concentrate of low volume is solidified and disposed suitably. The other stream, treated waste, is diluted and disposed into a water body. Decontamination factor in the liquid waste treatment is defined as the ratio of specific activities of the original waste and treated waste. This is the most important factor in the treatment of liquid waste. DF = Specific activity of original waste/Specific activity of treated waste The low and intermediate level liquid effluents from sumps, laboratories and decontamination are treated suitably. Four technical processes are available for decontamination of contaminated effluents. i. Solid phase separation ii. Chemical precipitation iii. Ion-exchange iv. Evaporation Some other novel processes like reverse osmosis, electrodialysis, ultra filtration etc., are under development but upto date they have not reached a technical state of plant level utilization. The high level liquid waste are treated by evaporation and calcinations. The calcined mass is fixed in glass or synrock.


Solid phase separation

Solid phase separation is carried out to remove suspended and settled solid matter from effluents. There are several types of separation equipment available, all based on those which have been regularly used in the conventional water and effluent treatment plants in the industry. The most popular types are: Filters Sand filters Pre-cast filters Back flushable filters Cartridge filters Bulk centrifuges High speed separators


Centrifuges Hydroclones

Generally, the need for separation equipment is to remove particles which could interfere with subsequent liquid waste treatment processes, e.g. ion exchange or with the re-use of the water. Decontamination factor for solid phase separation may range from 1 to 100. Nomally, values lie between 2 and 10 because of the presence of dissolved radionuclides in the waste streams.


Chemical precipitation

Chemical precipitation methods based on the coagulation flocculation separation principle are mostly used for the treatment of liquid effluent from research establishments and at reprocessing plants. Most radionuclides can be precipitated, co-precipitated and absorbed by insoluble compounds, e.g. hydroxides, carbonates, phosphates and ferrocyanides and are thus suspended particles from the solution by physical entrainment. However, the separation is never complete for several reasons, and the decontamination factors achieved can be relatively low. For this reason, chemical treatment is usually used only for low and intermediate level liquid waste treatment. It can be used in combination with other more efficient methods. Laboratory tests are necessary to establish the correct conditions for operation and it is very important that this work be carried out using samples of the real radioactive waste to be treated. Chemical treatments have always to be connected with physical methods such as sedimentation, filtration and centrifugation.


Ion exchange

Ion exchange methods have extensive application in nuclear fuel cycle operations and other activities involving radioactive materials. Examples of these include the clean up of primary and secondary coolant circuits in water

reactors, treatment of fuel storage pond water at nuclear power plants and reprocessing plants. Liquid radioactive wastes usually have to satisfy the following criteria to be suitable for ion-exchange treatment: The concentration of suspended solids in the waste should be low. The waste should have low (usually less than 1 gpl total salt content) Radionuclides should be present in suitable ionic form. Filter pre-coated with powdered resin can be used to remove colloids. Decontamination factors can range between perhaps 10 to 10 7 for very sophisticated systems; however, values of 10 2 to 103 are relatively common. Generally, the lower DFs are based on gross activity and the higher figures are nearly always for specific radionuclides for which the system has been designed. Ion-exchange procedures have been applied to the treatment of liquid wastes for many years. The technique is well established and its value has been widely demonstrated.



The best decontamination effect, compared with the other techniques, is achieved by evaporation. Depending on the composition of the liquid effluents and the types of evaporators, decontamination factors between 10 4 and 104 are obtained. Evaporation is a proven methods for the treatment of liquid radioactive waste providing both good decontamination and good concentration. Water is removed in the vapour phase of the process leaving non-volatile components such as salts and most radionuclides. Evaporation is probably the best technique for wastes having relatively high salt content with a wide heterogenous chemical composition. Although, it can be considered a fairly simple operation which has been successfully applied in the conventional chemical industry for many years its application in the treatment of radioactive waste can give rise to some problems such as corrosion, scaling or foaming. Such problems can be reduced by appropriate provision, e.g. (i) Adjusting the pH value to reduce corrosion (ii) Removing the organics to reduce foaming or add anti-foaming agents (iii) Cleaning the evaporator system by nitric acid to eliminate scaling and subsequent (iv) Passivation of construction material There are numerous types of evaporators suitable for the treatment of radioactive effluents. According to the operational experience the types of evaporators that are used at nuclear facilities could roughly be divided into three main categories.

The i. ii. iii.

categories of evaporators: Pot or kettle types Natural circulation Forced circulation

All decontaminated effluents i.e. filtrates, regenerants and condensates can be used again in the plant or can be discharged to the environment after monitoring. For further treatment all radioactive concentrates are stored temporarily in concentrate tanks.


Immobilisation of concentrates

The liquid concentrates from effluent treatment must still be transformed into solid products for final disposal. Immobilisation processes involve the conversion of concentrates to chemically and physically stable forms that reduce the potential for migration or dispersion of radionuclides from the wastes by natural process during storage, transport and disposal. From the different effluent treatment processes the following waste concentrates have been identified. i. ii. iii. iv. Sludges Spent ion exchange resins Evaporator bottoms Calcined mass

Various techniques are developed for transforming aqueous concentrates in a solid form or immobilization in a matrix. Solidification of concentrates, filter sludges and ion exchange resins in a matrix is necessary for final storage, because in the safety assessment of the storage site, flooding of water and accordingly potential contamination of the ground water due to leaching of the waste, cannot be completely excluded provided simple drums are used. The three matrix material are known for incorporation of concentrates arising from nuclear power plants and other nuclear facilities. i. ii. iii. iv. cement bitumen plastics glass or synrock

Cementation process

A relatively simple technique for immobilization of radioactive concentrates is mixing with cement. The main reasons for using cement are: i. ii. iii. iv. the relative simplicity of handling extensive experience in civil engineering operations the availability of raw material the relatively low cost

v. the high density(shielding effect) and the mechanical strength of cement products vi. the compatibility of water with the matrix material To improve specific properties of the waste form, compatible additives can be used. The cementation processes for nuclear waste immobilization, with and without additives has been commonly used on a industrial scale for several years all over the world. The various cementation processes can be classified according to how this mixing is achieved. (i) In-drum mixing process: There is drum mixing cementation process based on feeding cement concentrate and any additives separately into a container which is also the final shipping container. In the container the components are mixed until homogeneous mixture is obtained; after mixing the cement composite is allowed to set. Depending upon the activity level of the waste concentrates to the processed, the complexity of cementation systems will vary. Drums which are pre-filled with cement are lightly flanged onto the cell from below by means of double lid system. The liquid concentrates are transferred into a dosage vessel within the box by means of a vacuum pump and from here into the drums. Concentrates and cement are mixed with a planetary stirrer. (ii) In-line mixing process: The in-line mixing cementation process based on metering cement and waste separately into one end of the mixer. From the mixer the cement product is released directly into the storage container. It has been found in cementation that the stability of the cement/salt product quickly decreases with increase in concentration of salt in the final product.

Bituminization Processes

The bituminisation process for nuclear waste immobilization has been commonly used on an industrial scale for several years in different countries. Basically, the process consists of mixing solutions, sludges or solids with bitumen at elevated temperatures. The water contained in the waste is evaporated and the residual particles are uniformly coated with a thin layer of bitumen. The bitumen mix is released in suitable containers, where it cools down and solidifies. The various bituminisation processes can be classified into batch processes and continuous processes. Extruders can be used to embed all kinds of concentrates in bitumen, such as evaporator concentrates, precipitates and filter sludges, powder and ion exchange resins. The high capital costs for this system are regarded additionally as disadvantage of bituminisation, especially for processes involving only small amounts of concentrates. In the future, especially in connection with large units

of NPS, or sites with two or three units, bituminisation or radioactive waste will meet the growing interest of waste producers.

Plasticising processes

Incorporation of radioactive waste into plastics is a relatively new immobilization process when compared to incorporation in cement or bitumen. Depending on the polymer used, the immobilization can take place either at ambient temperatures or with hot liquids upto 60 deg.c. All the polymer processes are essentially batch processes where a catalyst is generally mixed with the wastes and polymer either in a pre-mixer vessel or in the container itself. The main plastic materials used for this purpose are: Polyethylene Urea Formaldehyde Polyester Polystyrene Several methods have been under development all over the world for sometime already in which waste concentrates, specifically those coming from nuclear power plants are embedded in plastics. This technique offers the advantage of working without any kneading or mixing step. The resins are first transferred with water into container designed for disposal such as a drum in a concrete container and then the water is removed by a vacuum. In a separate vessel, the necessary quantities of monomers and reaction gents are premixed and transferred to the resins by vacuum. After an induction period, polymerization is started and completed after some days, depending from the initiator.

5.3 Treatment & Fixation of High Level waste
High Level Waste is evaporated to suitable concentration prior to fixation in glass. High level calcined waste is mixed with glass frit and whole mass is melted in induction furnace or joule melter. In induction furnace heat is provided by indirect heating through induction coils while in joule melting hest is provided directly by special metal electrode. Both processes are being used.

Material for immobilization (Fixation) of HLW
Sine it is minatory to achieve maximum safety during the storage of HLW, the waste material must be rendered both immobile and insoluble. Thus the desirable properties of the waste form material are as follows: 1. Good capacity of accept all the elements in the waste. 2. Composition range flexible enough to accommodate variations in the waste. 3. Low melting point to facilities production. 4. High thermal conductivity to dissipate the heat produced by radioactive decay. 5. Good resistance to leaching by waste. 6. Good mechanical integrity elevated temperatures.

7. Good resistance to radiation damage. Glass meets these criteria and is available in plenty. Glass forming technology is well established and is familiar to mankind for centuries. Glasses formed centuries ago have been seen to be maintaining



Nearly 60% of solid radioactive wastes produced in nuclear facilities like nuclear power plants, reprocessing plants or nuclear research centers can be classified as combustible waste and 40% as noncombustible or bulky waste. Experience has shown, that more than 90% of solid waste, produced in nuclear facilities have dose rate below 200 mR/hr that means low level waste and only 10% have higher. Solid waste is composed of, for instance, paper, plastics, used protective clothing laboratory wastes and filters as well as wastes from maintenance and repair work. The objective of solid waste treatment is to minimise the volume of the disposal space required. Thus, the most important factor in solid waste treatment is Volume Reduction Factor (VRF), which is defined as the ratio of the volumes of waste before and after treatment. VRF = Volume of waste before treatment/Volume of waste after treatment



Pretreatment of solid waste is mainly at an appropriate preparation of the waste facilities the subsequent waste treatment steps. Emphasis must be placed on an as informative as possible description of the waste, stating origin, nature, activity, isotopes, radiation level and all further information relevant to the safe evacuation of the subsequent treatment steps. Sorting and segregation is done at the place of origin, i.e. into combustible and non-combustible and they should be combined with the packaging into appropriate containers usually plastic bags, or other boxes. The filled plastic bags are collected normally in 200 l. drums bearing adequate indication i.e. colour codes for transport to the corresponding waste treatment facilities. -bearing waste generally produced from reprocessing and research labs are classified based on the content and are segregated from other   wastes. The steel drums used for transportation are covered with coat of paint easy to decontaminate and resistant against acid alkali attack. It is highly recommended to store the waste in segregated categories from the beginning.



One of the essential aims in the treatment of solid waste is to reduce as much as possible the waste volumes to be stores or disposed of, and to concentrate and immobilize as much as possible the radioactivity contained in

the waste. The two well -known techniques to do this are compaction and incineration.



Volume reduction of low level solid wastes by compaction aims essentially at an increased in the overall density of the wastes material. This mechanical volume reduction method is widely used in waste treatment. Commercially available presses and compacting devices or compactors are frequently used in radioactive waste treatment after appropriate adaptation to the specific task. Volume-reduction factors obtained depend largely on the waste material and the pressured applied, but in general are between 3 and 10. Based on economics and practice, compaction can be divided into two main categories i.e. low-pressure and high-pressure units. Force applied can vary between 4.5 and 1500 MT. Pressures vary normally between 2 and 800 kg/cm 2. Both hydraulic and pneumatic presses are in use. A typical application of the low pressure compaction technique is the simple compression of bags of trash into a 200 liter drum. Implicit in the design of such a device is the requirement that adequate containment and off-gas treatment is provided to meet safety criteria. High pressure compaction is used to reduce finally the compatible waste to an optimum density. The unsorted waste is collected in 200 l. drums and automatically transported into the press and compacted with a total force of 1500 t. The compacted waste looking like a pancake is transferred to another drum and covered with concrete. Compaction of solid wastes is a widely used treatment technique in waste management. It is characterized by comparatively low capital and operating costs.



Incineration of combustible waste results in a high volume reduction to a factor of 50 to 80, depending on the content of non-combustible waste. Incinerators are in operation in different Nuclear Research Centres of the world. Numerous types of methods are used for combustion of radioactive waste. Incinerators are classified by different characteristics according to the amount of air used, such as controlled or excess air, according to special characteristics of the combustion chambers. A typical incinerator consists of : i. ii. iii. iv. material – loading mechanism combustion chamber after burner off-gas treatment system

v. as removal equipment


Size reduction and decontamination of bulk solid waste

More complicated and still more expensive techniques must be applied for the management of large and bulky equipments. Glove boxes, master slave manipulators and other process components which are very often contaminated with Plutonium are dismantled and decontaminated or scrapped respectively in special treatment facilities.


Treatment of intermediate and high level solid waste

In addition to the low level solid waste, nuclear facilities also produce activated bulky plant components with higher radiation dose rates. Hot cells or shielded boxes are needed to handle this category of waste, in order to prepare it for disposal.


Treatment of  bearing waste

 bearing waste, due to high toxicity and tendency to become air borne require special treatment. Acid digestion processes for cellulosic and plastic waste such as H2SO4 + HNO3 and H2O2 + H2SO4 digestion in presence and absence of catalyst like Iron, Selenium etc. is being developed. It is low temperature and wet process. Recovery of TRUs is also possible. Incineration process for such waste is also under development.



Radionuclides released from different parts of the nuclear facilities may be in the inform of inert gases, vapours and particulate suspensions. Accordingly, retention of radionuclides is achieved as a result of certain physical, chemical or physio-chemical processes. Physico-chemical processes are a combination of the former two, chemisorption e.g. collection on impregnated charcoal, followed by interaction with an impregnant. For retention of gaseous radionuclides and particulars, the following methods are used: i. particulate filter systems ii. iodine sorption system iii. noble gas delay system High Efficiency Particulate Air(HEPA) filters are normally used for effective particulate removal. They are usually made in standard sizes for defined flow rates of effluent air or gas and consist of pleated glass fiber paper in a wooden or metal case which is sealed. Individual HEPA filters should have an efficiency of atleast 99.15%.

The removal of iodine species from the gaseous effluents of NPPs is nearly always performed with the help of impregnated activated charcoal. Impregnation is required in order to trap the organic iodine compounds from humid gas. Noble gases cannot be retained by physical or chemical means, except with complex and expensive equipment. However, since the noble gases with the exception of Krypton-85 have radioactive half-lives of no more than a few days, delaying their release will allow radioactive decay processes also greatly reduce the quantities finally release to the environment. Two delay techniques are used: storage of the gases and passage through a delay bed: Delay beds consist of a number of vessels filled with charcoal, which selectively retards the passage of krypton and xenon relative to the carrier gas and allows radioactive decay to take effect.



In a FRP, the airborne radionuclides of primary concern are Tritium (H-3), Carbon(C-14), Krypton-85(Kr-85), Iodine-129(I-129) and aerosols containing fission products. In the normal chop-leach process, up to 50% of the Tritium inventory would be retained in the Zercaloy hulls and the remained would end up in the high level liquid waste (HLLW). If HLLW were concentrated or solidified, the Tritium would be released as water vapour to the stack or as low level aqueous wastes to the ground. Carbon-14 is assumed to be evolved during fuel dissolution as 14CO2. Its recovery is complicated by a small mass of C-14 being diluted with a large mass of 12CO2and air and requiring the recovery of both 14CO2 and 12CO2. Removal of CO2 may be impractical if larger air streams must be treated. The technologies for removal are caustic scrubbing, molecular sieve adsorption, Fluorocarbon scrubbing and adsorbent bed fixation. Most of Kr-85 remains in the spent fuel until it is sheared and dissolved. Since Krypton is chemically inert, recovery methods based on physical separation from the off-gas have been developed. The technologies are cryogenic distillation and fluorocarbon absorption which use the selective solubility of Krypton in liquid Nitrogen or fluorocarbons to separate it from air streams. The recovery processes also collect xenon, a chemically stable fission product that is present about 10 times the Krypton concentration. In the cryogenic system, xenon would be separated in a second column by bath distillation in the fluorocarbon process or cold adsorbent beds. Iodine is chemically reactive and may be found as an element or in many of several chemical compounds in the process of off-gas. Many compounds of iodine are not stable, and conversion from one chemical to another may occur depending on the environment. Most recovery methods convert the iodine to a soluble or insoluble iodide or iodate either on adsorbent or in solution. The well-

known recovery systems are caustic scrubbing, the Mercurex and Iodox process, silver reactors and silver loaded adsorbents.



The basic objectives to be followed in disposal of radioactive wastes to the environment after their treatment are: a) b) c) Man is not endangered; As far as practicable, man is not inconvenienced; and No damage is caused to natural resourced and no restrictions are imposed on their development

The first of these requirements must be regarded as absolute. The other two are a matter of compromise between cost and operational convenience on one hand and the desirability of keeping all exposures to radiation as low as practicable. While at a glance this requirement may look to be too restrictive, in practice, on careful evaluation of environmental capacity to absorb, disperse or attenuate radionuclides, the disposal can be planned in such a way that by the time the radionuclides come in the environment or diluted so much that their hazardous effect on man is eliminated. The disposal philosophy based on ICRP recommendation is that at no time discharge of effluent will not produce public exposure more than 1mSv/yr(100mR/yr) at plant boundary. In this context it may be of interest to note that background radiation due to radioactive material present in earth in our country varies from 0.3mSv/yr to 5.75 mSv/yr. In word of Lord Marshall ex-chairman UKAEA “In my own country UK I would like to point out that the average Englishmans’ garden occupies 1/10 of acre. By digging down one meter, we can extract 6 kg of Thorium, 2 kg of Uranium and 7000 kg of Potassium all of them radioactive. In a sense all of them is radioactive waste, non man made but residue left over when GOD created earth.



Contamination of the sea by radioactive wastes is incompatible with the principle of freedom of the high seas. The danger to the life and well being of mankind from contamination of the sea by radioactive wastes, the threat it constitutes to international shipping and harmful effect it has on fisheries have been repeatedly emphasized. Therefore, in planning discharge of radioactive wastes into the marine environment, elaborate pre-operational survey is conducted to find the critical radionuclide, the critical path through which it can reach man and the allowable body burden. Based on all the above parameter, the maximum permissible concentration limits (MPC) are derived. The important aspect to be noted with respect to disposal of radioactive waste into the water environment is that the MPC limits recommended for radionuclides are very stringent as compared to their inactive isotopes. The MPC

limits should be applied to the effluents in the discharge line and no allowance is made for the dilution of the effluent by the water bodies.



Stack dispersion or the use of atmospheric dilution has been employed for both gas and particulate dispersion by industries for decades. Use of such tall stacks for dispersion from reactors or chemical reprocessing operation is also followed by nuclear industry but generally after suitably treating the gaseous wastes by means of cleaning devices at the source itself. The stack height selection is rather complex without going into details, it is just relevant to mention that the height to be selected is chosen to produce the stipulated ground level concentration.



Traditionally, ground has been used a receptacle for the waste materials. When ground disposal is being considered, it is necessary to have information on: a) Geographical features including physical features of the area, vegetation present, climate particulars precipitation and evaporation rate, location of adjacent water bodies and population density; Geological aspects including type of soil, permeability of formation, and chemical and mineral compositions of formation, and Hydrological conditions including depth of water table, direction and rate of flow of the ground water and relative rate of movement of radioactive substances.

b) c)


Ground disposal facilities

Ground disposal facilities are located in a specially chosen area and is fully fenced all around. Some of the typical ground disposal facilitieis are described below:Earth trenches:- These are excavations made in the disposal site with suitable slope. The dimensions are fixed according to sub-surface conditions and operational convenience. Only potentially active wastes are disposed off in these trenches. Reinforced concrete trenches:- These trenches are of reinforced cement concrete consutruction and are usually 2m deep 1m wide and 50m long. Additional water proofing to the outside concrete walls has been found necessary. Higher active solids and slidges are disposed of in these RC trenches. When full, these RC trench compartments are concreted on top giving enough shielding to bring down the dose on the surface. Tile-holes:- Tile-holes are generally of 1m dia meter spun concrete pipes with steel lining driven to a depth of 5m below ground. All the external surfaces of

the tile-holes are provided with additional water proofing. These are used for the disposal of very high active wastes.


Movement of radionuclides in ground

The radionuclides move along the same pathways as the ground water. These are retained in the ground due to exchange processes. The chemistry of the waste solution is altered continuously as it moves through the formation. The absorbed radionuclides may be again desorbed with the influx of fresh waste solution. Therefore, it can be seen that the movement of radionuclides in space and time is a complex phenomenon.



Even though vigorous steps are taken to evaluate the suitability of a disposal site before its selection, surveillance of the biosphere and study of the impact of radioactive waste on the environment are essential. General methods adopoted to monitor the environment include drilling of a series of borewells around a disposal site and collection of water samples for periodical analysis. This would help in locating the moving fronts of pollutants and taking remedial measures to prevent their entering sources of water. In addition to the water samples, soil samples are also collected in and around a disposal site and analysed.



Though we do not know all that we should know about the environment, we have a wide knowledge of terrestrial geology than either oceanography or outer space. Also it is evident that escape of radioactivity from a solid environment is limited to diffusion in the solid, which is much smaller than in a fluid medium. The state of knowledge existing today indicates that land is generally preferred as a safe sink for radioactive wastes.



The high level wastes have twin properties of high decay heat and high activity. Therefore, the immobilized waste is kept for 20 to 25 years in air or water cooled vaults to remove the decay heat. Then they are transferred to deep repositories as they are going to remain potent for about 10000 years. 8.6.1


Emplacement of wastes in the appropriate facility at a depth of at least several hundreds meters without the intention of retrieval forms the basis of geological disposal concept. A facility for radioactive waste disposal located underground( usually several hundred meter below the surface) in a suitable for host rock to be provide long term isolation of radio nuclides from the accessible environment(biosphere)


A geological disposal system can be defined us combination of conditioned and packaged solid wastes (like Vitrified waste canisters)and other engineered barriers with in an excavated or drilled repository located at a depth of hundreds of meters in a stable geological environment . The geological formation, in which the waste is emplaced, refer to us ‘host rock’ generally constitutes the most important isolation barriers. Host rock types considered for geological disposal Host rock types Crystalline rocks Granite Gneiss Argilaceous formations Strongly consolidated clays: Clay stone, mudstone consolidated clays: shale, marl Plastic clay Rock salt Bedded salt Salt domes Volcanic tuffs Above water table USA Belgium USA Germany France, Switzerland Hungary countries Canada, Sweden China, Finland, India,

   Waste form i. e. vitrified waste Over pack i. e. inner mild steel and outer stainless steel cover i.e. container Engineered barrier system (EBS)- It comprises of buffer and back fill zone, support system i.e. the immediate zone between over pack and host rock Natural geological formation (host rock)


1. Forms stable Continental Region 2. Large lateral &depth ward persistence 3. Known to have remained unaltered over millions of years 4. Narrow compositional range 5. Lack of mineral deposits

6. Doesn’t support forest cover &groundwater movements 7. Allow underground construction with ease 8. Good thermal conductivity &rock mechanical strength 9. Known to occur through out the country And the site should have 1. Tectonic stability 2. No major groundwater bodies 3. No surface water bodies 4. Minimum forest cover &population & rainfall. 5. Away from know mineral deposit or Archaeological monuments 6. Favourable socio political environment Studies on the development of a Deep Geological repository are at different stages of development in different countries and caN be implemented when there is sufficient inventory.



There are various misapprehensions regarding the generation and management of radioactive wastes. The quantity of radioactive waste generated per unit energy generated is much smaller compared to thermal power generation. High Level Waste is only a small fraction of the radioactive waste generated. Stringent limits are stipulated for the disposal of radioactive waste into the environment and there is well established organisation to monitor these discharges. Instruments are available to measure the radioactivity in these discharges, which is very low level concentration. The radioactive wastes that have been produced so far have been managed in a more satisfactory manner than any other industry, much to the credit of the nuclear industry. Research and development work carried out has led to robust and highly reliable technologies for treatment and disposal of radioactive waste. There are sound and reassuring reasons to believe that the nuclear industry is capable of safety and economically managing its waste products.

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