Webserver Date: 23-March-2017

Nuclear India

Published by the
Department of Atomic Energy
Government of India


VOL. 33/NO. 9-10/March-April 2000


Demonstration Plant for Radiation Processing of Spice



The Demonstration Plant for Radiation Processing of Spices set up by the Board of Radiation and Isotope Technology (BRIT) at Vashi, Navi Mumbai has started commercial operation from January 1, 2000. BRIT is a constituent unit of the Department of Atomic Energy established to promote the applications of radiation and isotope technology.


The Plant will process, to start with about 3,000 tones of spices a year and will go up to 12,000 tones per year in the next couple of years.


Technical Features:

  1. Designed for round the clock fail-safe automatic operation.
  2. Flexibility in processing a range of whole and ground spices.
  3. Tote (product) box size: 59cm X 45cm X 110cm (l x w x h).
  4. ~30 tones of spices can be processed per day at maximum cobalt-60 loading of one million curies.

Fuel and Materials Development for Prototype Fast Breeder Reactor


Baldev Raj
T. P. S. Gill
Head, Materials Joining Section, Metallurgy and Materials Group
Indira Gandhi Centre for Atomic Research



The successful commissioning and sustained operation of the Fast Breeder Test Reactor (FBTR) at Kalpakkam is considered a unique event in the history of the Indian nuclear power programme. Mastering of a host of new and complex technologies has been the hallmark of this achievement. Experience in designing, constructing, commissioning and operating FBTR has enabled attainment of in-depth knowledge in various fields and provided a solid foundation to face the challenge of designing and building the 500 MWe Prototype Fast Breeder Reactor (PFBR). The FBTR has not only provided experience in the area of high technology, but also serves as a test bed for developing fuels, fuel reprocessing methods and structural materials.


The R&D laboratories at IGCAR are fully equipped to develop and characterize fuel and construction materials for PFBR and also efficiently reprocess the spent fuel. Such a powerful combination of state-of-the-art facilities and highly trained and motivated scientific manpower has given us immense confidence in taking up the construction of PFBR.


Fuel Chemistry


The development of fuel for the PFBR is a multidisciplinary task. While the Nuclear Fuel Complex (NFC), Hyderabad will have the responsibility of fabricating the stainless steel components for the fuel subassemblies, the development as well as fabrication of the fuel per se will be carried out jointly by the Bhabha Atomic Research Centre (BARC) and IGCAR. Development of the flow sheet for fuel fabrication has already been initiated at BARC. The development of materials for the subassemblies is being undertaken at IGCAR and NFC.


The fuel chosen for PFBR is uranium-plutonium mixed oxide with plutonium content of 21% and 27% in two different regions of the reactor. Even though this fuel has been studied extensively, specific studies are needed to master some aspects of the fuel cycle.


The fast reactor fuels operate in high temperature regimes. As a result, the fuel behavior is much influenced not only by the composition, but also other parameters such as linear heat rating (which decides the temperature gradients) and burn-up. Therefore, it becomes very important to understand and predict the fuel behavior during irradiation in order to realize safe operation to high burn-up. Fortunately, there is a wealth of data in the literature on the mixed oxide fuel, chosen for PFBR. However, to generate more confidence in the modeling of fuel behavior, experimental programmes have been initiated at IGCAR and BARC to measure some of the properties such as thermal conductivity, heat capacityand oxygen potentials on the as-fabricated fuel material.


The fast reactor fuels are normally reprocessed by the conventional PUREX process. The vast experience gained in our country on this process with regard to the thermal reactor fuels and the experience that will be gained with the high plutonium content fuel of FBTR will provide the necessary input to the design of the process flow sheet for PFBR. Programmes are also being pursued at IGCAR to identify specific chemical aspects of aqueous reprocessing which would provide necessary data for optimizing the process for the PFBR fuel.


While reprocessing through the PUREX route will be adopted for the PFBR, it has been recognized that non-aqueous and pyroelectrochemical processes are potential candidates for application to fast reactor fuels, since they have several advantages which would greatly simplify waste management. IGCAR has gained valuable experience in the development of pyrochemical process. Based on this experience, studies are being undertaken to develop pyrochemical processes for oxide fuels as well.


The fuel pellets of the first core of PFBR will be prepared by the conventional powder-pellet route. However, for the future cores, it is important to keep in mind the alternative method of fuel preparation through the sol-gel process, which has a number of advantages over the conventional powder route. This route is amenable for fuel pin production through the pellet route as well as the vibropac route. A joint programme between IGCAR and BARC has been initiated to fabricate test fuel pins of mixed oxide through sol-gel/pelletisation and sol-gel/vibropac routes. The fuel pins produced will be irradiated in the FBTR. These steps will provide valuable experience in the fabrication of the fuel through these routes. This experience will provide the options for the future fast reactors, as well as the future cores of PFBR.


For ensuring the operation of PFBR with a few failed fuel subassemblies, it would be necessary to provide a cover gas purification system so that the cover gas let out into the atmosphere is free from the fission gases. For this purpose, a cover gas purification system based on ambient temperature charcoal absorption system is being designed.


Chemical and Physical Properties of Fuel Materials


Thermal conductivity of the fuel is an important property that determines the maximum specific power that can be drawn from it. This property is being measured for PFBR oxide fuel employing laser flash technique. A mass spectrometric method has been standardized for the measurement of the burn-up of fast reactor fuels. This method which has been recently used for the measurement of the burn-up of FBTR fuel can be employed for the determination of the burn-up of PFBR fuels as well. During the life of the fuel in the core, volatile fission products accumulate in the fuel-clad gap and some of the fission products can chemically interact with constituents of stainless steel. Tellurium being one such fission product, thermochemical data on tellurides of iron, nickel, chromium, manganese and molybdenum are being generated by measuring their vapor pressures by using Knudsen cell mass spectrometry. In order to study the vapourisation behavior of fuel/fission product materials, a high temperature mass spectrometer facility is being set up in a glove box.


Calculations on safety aspects of fuels require reliable equation of state, which can be derived from their vapor pressures at very high temperatures. A Laser Induced Vapourisation Mass Spectrometry (LIV-MS) system has been developed in-house and used for measuring vapor pressures of materials such as uranium oxide, thorium oxide, uranium carbide at 3000-7000K as well as for identifying the high temperature vapor species. By employing matrix isolation technique, their vibrational spectra have been obtained so that partition functions can be deduced.


Sodium Chemistry


Chemical interaction of constituents of structural materials with non-metallic impurities present in liquid sodium, particularly O and C can influence the corrosion damage and activity transport even though the actual corrosion rates may be low. However, in view of long life of a fast reactor, a complete understanding of the chemical aspects of corrosion is helpful. Phase diagrams of constituents of stainless steel with oxygen and sodium, such as those of Na-Mo-O, Na-Fe-O and Na-Ni-O systems have been established. Thermochemical data of ternary compounds that form during the oxygen assisted corrosion in sodium have been measured and the threshold oxygen levels in sodium for their formation have been established. In the event of a clad breach in PFBR, chemical interaction of irradiated oxide fuel with coolant can result in the formation of voluminous products and their release into sodium. This chemical interaction of the fuel and cladding is also being studied in detail.


Development of Sensors for Process Monitoring


Manufacturing defects growing during the course of operation of the steam generator can lead to steam leak into sodium and sodium-water reaction which calls for detection of the leak at its inception itself. To meet this demand, very sensitive electrochemical meters capable of measuring ppb levels of hydrogen (formed as a result of sodium and water reaction) in sodium have been developed and put into service in FBTR. An on-line meter for measuring hydrogen level in argon cover gas has also been developed for detecting steam leaks during low power operation or during start up of the reactor. Solid electrolyte based oxygen meters have also been developed for continuously monitoring oxygen in sodium. Sodium in the coolant loops is circulated by centrifugal pumps with their shafts cooled by hydrocarbon based oils. An oil leak into liquid sodium can result in carburization of steels leading to deterioration in their mechanical properties. To detect this leak, an electrochemical meter that can measure carbon potential in sodium has also been developed and it is being used successfully in FBTR service. Currently, miniaturization of these sensors is in progress for future use in PFBR. Carbon activities of D-9 alloy and low alloy steels to be used in PFBR are being measured as a function of carbon concentration using the electrochemical carbon meter and other techniques.


A view of the laboratory scale set up (Inside a glove box for injection casting of metallic fuel materials)


Studies on Radioactivity Transport and Trap Development


With a view to forecast the radioactivity transport in sodium coolant circuits in PFBR, model calculations on release behavior of Mn-54 from clad material were made. Various materials have been tested as scavengers for removing Cs-137 released into sodium during a fuel clad failure. Reticulated vitreous carbon has been shown to have the potential for this application. A trap using this scavenger for use in PFBR is being designed.


Sodium Removal and Decontamination


Techniques based on treatment with steam or alcohol is being studied for removing sodium from sodium-wetted components. In view of the possible explosive reactions of sodium alcoholates under certain conditions, the various possible reactions in sodium-alcohol systems are under investigation. This will help to optimize safe sodium removal for cleaning before repair, maintenance or disposal.


Development of a Low Temperature Method for Sodium Production


Production of sodium metal is highly energy intensive. Research has been undertaken at IGCAR on an alternative low temperature process. The experimental cell has worked well. Plans to upscale the process are being worked out.


Analytical Chemistry


Analytical methods have been developed and validated for the determination of trace levels of elemental impurities in sodium, uranium, steels, water and organic solvents by using Inductively Coupled Plasma-Mass Spectrometry and Atomic Absorption Spectrometry. Methods have also been established for the direct and rapid analysis of stainless steels, ferritic steels and Ni-base alloys for the major, minor and trace constituents using the Direct Reading Spark Emission Spectrometer.


Post Irradiation Examination for Fuel Assemblies


Being a prototype fast breeder reactor, PFBR needs detailed post-irradiation examination (PIE) of irradiated fuel to validate its performance and to optimize fuel fabrication parameters. This would also facilitate standardizing effective reactor operation procedures such as fuel handling schedule and power ramps. A comprehensive state of the art PIE facility has been set up at IGCAR, which is designed to cater to the needs of a variety of plutonium rich fuels. Rich experience exists at IGCAR in the area of design, construction and operation of PIE facilities for plutonium rich fuel and structural materials for FBRs. The specifications for the mixed oxide fuel for PFBR has been already finalized and trial fabrication of fuel pellets is undertaken at BARC. PIE of a few fuel pins fabricated as per PFBR specifications and irradiated in FBTR will be carried out in this hot-cell facility to gain further confidence and fine tune the specifications. This facility will be used for detailed PIE of PFBR fuel along with a dedicated hot cell facility, which will be set up at PFBR complex, to cater to the examination needs of the indigenously developed fuel and structural material for PFBR.


Fuel Reprocessing


Fast reactors employ natural uranium and plutonium mixture as fuel, generating power and at the same time converting the fertile species U-238 into fissile Pu-239. To sustain the fast reactor program, the Pu-239 needs to be recycled as quickly as possible. This calls for reprocessing the irradiated fuel immediately on discharge from the reactor. Hence the reprocessing is carried out in a high radiation field. The demanding requirements in equipment and process calls for a complex technology which is being developed at IGCAR.


The well-established PUREX process used for the thermal reactor irradiated fuels is modified to meet the exacting needs of the fast reactor fuel reprocessing. Spent fuel transportation, fuel chopping, dissolution and the solvent extraction flow sheet are the major steps to accomplish this.


Sixteen Stage Centrifugal Extractor Bank


Shipping casks have to be designed for alpha-tight transfer requirements. For handling high decay heats, they have to be provided with water cooling or forced air cooling, especially for large quantity fuel transports.


For some typical advanced fuels like the ones used in FBTR, because of the pyrophoric nature, special chopping devices have been designed. The dissolution of the carbide fuels in nitric acid generates soluble organic acids. These organics interfere in subsequent solvent extraction. A novel dissolution method, Electro Oxidative Dissolution Technique (EODT), has been developed which generates Ag2+ at anode and decomposes the organics. Also the Pu4+ is oxidized to Pu6+ which accelerates the plutonium dissolution. This technique has been validated for the dissolution of irradiated FBTR fuel (25 GWD/t burn up). Cost effective Pt electroplated anodes have also been developed in collaboration with Central Electrochemical Research Institute, Karaikudi and fabricated for this application. The performance of an alternative electrode developed in house is also satisfactory. A titanium electrolytic dissolver is fabricated and installed for dissolving FBTR fuel.


The insoluble residues not only lead to loss of plutonium, but also create problems in solvent extraction by accumulating at the interfaces. To remove these fines which are of the order of 0.5 mm, high-speed centrifuges that can be remotely operated and maintained are used.


Flow sheet formulation is very important for the success of reprocessing. Solvent extraction modeling has been successfully done to analyze the extraction flow sheet. The SIMPSEX computer code developed, not only helps in optimizing the operating conditions, but also aids in fixing the margins in operating parameters so that maloperating conditions can be predicted. This is of importance in the high plutonium bearing solutions where second organic phase is formed, because of the limited solubility of the extracted plutonium complex in the extractant phase.


The high radiation levels in the solution demand the design and development of short residence time contactors such as centrifugal extractors to minimize the solvent damage. The residence time in this type of contactor is about 10 seconds as against a few minutes in pulse columns or mixer settlers used in thermal reactor fuel reprocessing. Such contactors have been developed for a wide range of flow rates with individual stage driven by electrical motor. Also mini mixer settlers (20 ml hold up per stage) which are ideally suited for flow sheet simulation studies in glove boxes have been developed and experiments were conducted with U-Pu solution. Electrolytic methods are also developed to partition the purified plutonium and uranium.


The plant and equipment design for the fast reactor fuel reprocessing should meet the requirement of a-tight cell operation. The maintenance philosophy should be mostly based on remote technology and less on contact type.


The pilot plant, Lead Mini Cell (LMC) which is in the commissioning phase is to reprocess the first few cores of FBTR fuel assembly. All the components of LMC are totally developed and built in-house. This facility will serve as the test bed to understand the process problems, gain operational experience and provide solutions to the prototype plant, Fast Reactor Fuel Reprocessing Plant (FRFRP) which is under construction.


A view of Lead Mini Cell


PFBR Materials


At IGCAR, materials-related research has been playing a crucial and significant role towards design and development of PFBR. Supported by comprehensive research and testing facilities, expertise in various disciplines is used to cater to the needs of PFBR. Major inputs have gone into:

  1. Selection of materials and finalization of various specifications
  2. Indigenization of base materials and welding consumables
  3. Mechanical, chemical and physical evaluation of various materials in air, inert and sodium environments
  4. Fabricability and weldability of materials
  5. Surface coating and hardfacing technologies
  6. In-service inspection and non-destructive evaluation
  7. Damage assessment and life prediction approach

Indigenization of Materials


Based on extensive interactions between designers, research scientists and engineers, the following materials have been selected and specifications have been finalized for 500MWe PFBR:

  1. Structural: Types are 316LN and 304LN austenitic stainless steels
  2. Core material: Alloy D9 (a Ti-stabilized stainless steel)
  3. Steam Generator Material: Modified 9Cr-1Mo ferritic steel
  4. Roof Slab Material: A48P2 carbon steel
  5. Welding Consumables: E316-15 16-8-2 and Mod. 9Cr-1Mo

A major step taken towards total self-reliance in materials is to make the Indian industry as partners in our quest for indigenization. The specifications of these alloys are quite stringent with respect to chemistry, inclusion content and properties. Materials meeting our specifications have not been available in India. Suitable Indian collaborators now have been identified and process of producing these materials is underway. To make these materials indigenously, certain facilities for heat treatment, quenching, pickling and passivation need to be augmented or added. None of the Indian industry has plates in their scope to the dimensions required for PFBR. The Steel Authority of India Ltd. has agreed to be collaborators in this challenging project and would augment their facilities to produce type 316LN, Mod. 9Cr-1Mo and A48P2 steels indigenously.


Indigenization of Alloy D9 has been carried out in close co-operation with MIDHANI. After standardizing the melting and casting procedures, the optimization of forging temperature and solution annealing treatment schedule was arrived at after extensive trials on laboratory heats. Based on these studies, industrial production of Alloy D9 with three Ti/C ratios (4, 6 and 8) was successfully carried out at MIDHANI. The indigenous developed alloy satisfies the PFBR design requirements. On the basis of pre-irradiation property evaluation these heats have been fully characterized and Ti/C ratio of 6 and cold work level of 20% has been found to meet the requirements. The selected heat has been produced by MIDHANI for production trials for the manufacture of clad and hexagonal wrapper tubes at Nuclear Fuel Complex (NFC).


The welding consumables E316-15 and Mod.9Cr-1Mo have been developed indigenously. IGCAR interacted with the industry at the level of core wire selection and flux design to optimize the weld metal composition. The challenge of controlled addition of various minor elements, especially nitrogen, vanadium and niobium, through flux has been met successfully.


Technology Development


Successful construction of PFBR requires absorbing, mastering and developing new innovative technologies. On this front several breakthroughs have been achieved only a few important ones find mention here.


The bimetallic joint between austenitic stainless steel and Cr-Mo steel is used in steam generators of almost all power plants and is plagued with the problem of premature failure. To circumvent the problem, an improved joint-the trimetallic joint (TMJ)-has been developed. A piece of Alloy 800 is inserted between austenitic stainless steel and Cr-Mo ferritic steel. Extensive accelerated testing has shown a four-fold increase in the life of TMJ over that of bimetallic joint.


The PFBR steam generator tubes will be supported by corrugated Inconel 718 strips. Aluminizing of Inconel 718 strips has been chosen for this application because of the excellent performance of aluminide coatings in reducing impact fretting wear of the tubes due to flow-induced vibrations. A new process for aluminizing has been developed based on direct spraying of aluminum metal, followed by diffusion treatment in vacuum. This process enabled aluminizing to be carried out at a much lower cost than the conventional process of pack aluminizing. The weldability of aluminized strips has also been found satisfactory. A large number of strips have been aluminized for the steam generator module being fabricated under the technology development programme.


In an industrial forming operation, the microstructure and thereby the properties of the final product can be controlled by varying mainly two parameters: the temperature and the deformation speed. Processing maps based on dynamic materials model have been developed for all the PFBR materials. Based on these maps, the processing parameters can be identified to process the materials with homogeneous microstructures. The processing window identified based on the maps has been validated with industrial scale forming operations. The processing maps may be adopted as useful tools in industrial forming operations and will be helpful in providing guidelines to industry engaged in indigenous production of PFBR materials.


Materials Property Evaluation


Mechanical properties of the core and structural materials undergo significant changes during service because of exposure to high temperature, sodium environment and neutron flux. Creep is an important consideration for design of safe and reliable structural components for long term operation. The presence of liquid sodium with good heat transfer properties and austenitic stainless steels with poor thermal conductivity and high coefficient of thermal expansion lead to situation where the stresses associated with start-ups, trips and other operational changes would be quite large and induce low cycle fatigue damage. This along with the presence of steady loads leads to complex creep-fatigue interaction. Extensive high temperature studies carried out on base and weld materials have unambiguously established the influence of minor elements, grain size, microstructural instability and thermomechanical history on the useful life of the material. The input from such detailed studies has proved invaluable in enhancing our confidence in the selection of materials for PFBR.


AISI type 316LN stainless steel has superior mechanical and corrosion properties over type 316 used in FBTR. The addition of nitrogen and lowering of carbon content have increased the resistance to sensitization of type 316LN many fold over that of type 316. This data gives confidence to the fabricators about the robustness of the PFBR material either during welding or during post weld heat treatment.


Effects of long term exposure of AISI type 316 SS to flowing liquid sodium on corrosion and mechanical properties have indicated that the change in mechanical properties is mainly due to thermal effects rather than due to sodium. These studies have also established that the indigenous sodium (having higher carbon content) can be used for PFBR without any adverse effect on stainless steel properties. The wall thinning and degraded layer in AISI 316 SS due to exposure to sodium for 20 and 30 year of service have been predicted.


Magnetic Barkhausen Noise Studies for Microstructural Characterization


Non-destructive Testing and Evaluation


In the areas of NDT&E, testing techniques and procedures for various PFBR components have been established. In addition, development of special instrumentation, sensors and software for application to various inspection tasks envisaged in PFBR have been carried out.


As part of the technology development for PFBR steam generator, microfocal X-radiography technique using a rod anode has been standardized for the inspection of critical tube-to-tube sheet joints. Root concavity is one of the defects that are difficult to qualify by microfocal radiography. Appropriate hardware along with silicon consumables has been developed to obtain replicas of welds from ID (internal diameter) of the tubes. The weld ripples can be clearly discerned and root concavity could be established from studies on the replicas. Eddy current testing of the steam generator (SG) tubes during the manufacturing stage has been established using the DC saturation method with a sensitivity of 10% of the wall thickness. The methodology was successfully applied to first batch of indigenously manufactured tubes at NFC.


For the in-service inspection of PFBR steam generator, a partial saturation eddy current technique has been developed. Ultrasonic and eddy current testing for the quality assurance of cladding tubes has been standardized. Magnetic Barkhausen Noise (MBN) technique has been developed for characterization of microstructure in Cr-Mo steel and would be extremely useful in future for damage assessment and residual life analysis studies.


A robotic system has been designed and developed indigenously for pre-service and in-service inspection of SG of PFBR. The robot can be deployed through the manhole of the SG and a CCTV camera monitors the movements of the robot. The robotic inspection system can automatically position various NDE probes such as eddy current probes, ultrasonic transducersand miniature cameras are required for ISI of PFBR SG tubes. Using this system, the overall time of inspection for all the tubes can be substantial lower than manual inspection


A mobile scanner (MOBSCAN) has been developed indigenously for NDE of large metal plates of PFBR. The system is also capable of venturing into inaccessible areas, carrying with it suitable cameras for inspectionand radiation surveillance etc. A six degrees of freedom robot is being developed for standardizing remote repair welding procedures for PFBR components.




The IGCAR was set up in 1971 with the objective of carrying out R & D in the frontier areas of advanced technologies for the construction and commissioning of Fast Breeder Reactors. A holistic approach was adopted and several specialized branches of science and engineering have been nurtured to meet this objective. Over the years, various programmes have dealt successfully with indigenous development of these reactors and associated technologies.


The mixed oxide fuel and its cladding, structural and steam generator materials and welding consumables, various on-line sensors for monitoring sodium purity, etc. have been developed indigenously. Novel electrolytic dissolution method and Pt coated anodes have been developed for fuel reprocessing. Several non-destructive evaluation methods have been established to ensure the reliability of components. The newer approaches of life assessment and enhancement have also been given due consideration. In short, various technologies needed for successful construction of PFBR are available to take up this challenging task. The future directions of the work have been carefully decided and are being implemented to construct, commission and operate PFBR with high availability. Expertise and support in the area of metallurgy, chemistry and reprocessing would not be found lacking for the success of PFBR.


A Perspective on Fast Breeder Reactor Safety


S. M. Lee
Director, SHINE Group
Indira Gandhi Centre Atomic Research



Risk vs Benefit


Every purposeful human activity has an associated benefit with an associated risk. The most important point to realize in any discussion on risks associated with an electric power producing system is that it should be relative to:

  1. The risk associated with any alternate electric power producing system
  2. The risk associated with not producing the power at all

These risks are fairly well quantified and relative comparisons are available (Fig 1). The lower risks for nuclear option in particular for FBRs are noteworthy. The Fig.2 shows correlation between population health and commercial energy use. India is at about the middle of the curve with still scope for several fold increase in energy consumption and corresponding benefits. The needed multifold increase of electric power production (from the present 90 GWe to over 400 GWe ) will not be possible in a sustained manner in India from the available fossil and hydro resources. The contribution from non-conventional energy resources will be also limited. These considerations make development of the nuclear resource potential essential. However, the decision on producing electricity by a particular mode varies from country to country depending on the available primary energy resource and the state of development of the country.


Essential Concern


The essential concern of nuclear safety is to avoid release of radioactivity which can lead to adverse health effects on the population. There are three types of radioactive releases to be addressed:

  1. Routine radioactive releases during normal operation of a nuclear power plant: These releases are kept by design well within internationally accepted safe limits. It is not commonly known that a coal fired station (on account of the inevitable uranium and thorium content in the coal) in fact releases routinely more radioactivity than a nuclear power plant.
  2. Radioactive releases during emergency and accident conditions: The design of the nuclear reactor is such that the consequences of radioactive releases during accidental situations is low enough to be acceptable compared to other means of electricity production. In fact due to adverse public perception of nuclear power the risks associated with nuclear power production has been brought much lower than other means of power production.
  3. Possible radioactive releases from storage of long lived radioactive wastes: In fact this risk is the lowest of all as the technology of passively storing long lived radioactive wastes till it decays to the level of radioactivity in the original uranium ore is already available in several countries including India.


Fig 1: Mortality Risks due to Electricity Production


Reactor Safety


The safety goals followed in the design, construction and operation of nuclear power plants are:

  1. Safety of the general public
  2. Safety of the plant workers
  3. Minimum environmental impact

Meeting the safety goals is ensured by parallel approaches at different levels providing a "defence in depth". The first level is the design of an inherently safe plant with an adequate control system which can operate with a high degree of reliability. Inherent safety features vary with the reactor type. In addition, to negative feedback reactivity coefficients FBRs are characterized by low pressure coolant with large margin between operating temperature (5500C) and sodium coolant boiling point (9000C), provision of intermediate sodium loop for isolation, double walled primary system and multiple radial coolant entry to subassemblies. Further, sodium has excellent heat transfer properties for emergency cooling, while pool type FBRs have large bulk of sodium to absorb abnormal heat generation without boiling with no possibility of coolant loss by pipe rupture. Primary sodium pumps are provided with flywheel action to continue action even after power supply loss and FBRs can be designed with the ability to ride out a loss of grid incident with all pumps coasting down and failure of all control rods to act with no fuel melting or coolant boiling. In addition, special emphasis is placed on the quality of materials and workmanship of components and sub-systems. The careful design and construction is followed by equally careful operations with regular surveillance checks so as to reduce the probability of accidents to a very low level.



At a second level, safety is provided by means of a comprehensive plant protection system (PPS) consisting of monitoring systems, shutdown systems and decay heat removal systems. The PPS can handle a wide range of conceivable abnormal incidents and malfunctions, shutdown the reactor and safely remove decay heat. The PPS includes a variety of instruments and sensors to monitor the state of the plant and to take protective as well as sympathetic control actions to prevent any abnormality leading to an accident. Features of the design are redundancy, diversity and fail safe operation. This means that for safety no dependence is placed on any single system or item or equipment. Ample backup is provided so that sequential failure of several systems even does not lead to unsafe conditions. Electric power is provided by alternate supplies and backed up by diesel generator and batteries. The shutdown system consists of independently acting control rods divided into two groups. The two groups of rods are based on different designs. The reactor shuts off if any one group rods act. The shutdown is effected by rapid fall of the control rods under gravity (< 1 sec) and the system is of fail safe nature (i.e. any failure leads to the reactor shutdown).


Similarly, the decay heat removal system has a normal mode of operation backed up by a failure mode of operation. In addition an independent emergency decay heat removal system is provided.


Fig 3: Multiple Barriers for Radioactivity Containment


The third level of safety is provided by engineered safety features such as reactor containment vessel and building, secondary shutdown systemsand emergency core cooling systems which limit the consequence of certain highly unlikely accidents, which are assumed to occur in spite of the first and second level safety measures. The fission product radioactivity is in fact contained by multiple barrier system. The solid fuel matrix provides the first barrier by retaining the solid fission products. The fuel is itself contained within a high strength, high integrity stainless steel clad which contains the gaseous fission products too and acts as the second barrier. The cladded fuel elements are arranged in fuel subassemblies which reside in the reactor vessel which is part of the closed primary sodium system and forms the third barrier. This third barrier is designed to withstand severe reactor accidents and is itself contained in another vessel called Safety vessel. The final barrier is provided by housing the entire reactor system in a "Containment Building" which is designed, constructed and tested to withstand the pressure generated in accidents and prevent radioactive release.


Finally, a careful choice of site is madeand exclusion areas defined such that routine as well as accident radioactive releases do not affect the general population. Exclusion areas are defined by a radius of 1.5 km around the power plant where entry of general public is prohibited. Further, the region from 1.5 km to 5 km radius is treated as a controlled zone.


100 kJ Condenser Bank


Training & Qualification


To avoid accidents and mishaps great deal of importance is given to the rigorous training and qualification of all the operating personnel to a much higher degree than in conventional power plants. Following requirements are to be met:

  1. Basic academic qualification commensurate to the given responsibility. Senior operating staff holds an Engineering Degree.
  2. Minimum experience on the specific system operated varying from 4 to 12 years according to the level of responsibility before being given independent charge.

For initial qualification personnel have to complete checklists, written examinations, walk-through test and viva-voce. Periodic re-qualification is also necessary.


FBR Operating Experience


Modern FBRs show extreme stability of operation with no need of automatic power regulation. On account of the concept of isolated primary sodium circuit and negligible tritium production, the radioactive discharges in normal operation and radiation exposure of operating staff are much lower than in conventional water cooled reactors.


At present there are FBRs operating in France, Russia, Japan and India. Other countries which have operated test and prototype FBRs are USA, UKand Germany. Countries like China and South Korea now have fast reactor programmes.


Benefits of Nuclear Power


The comparison of the risks of nuclear power generation shows that it is lower or comparable to other forms of power production. In addition, there are some real benefits of nuclear power production. Fossil fuel substitution results in great environmental benefits. A 1000 MWe coal fired station consumes 3 million tones of coal per year producing 7 million tones of carbon dioxide, 120 thousand tones of sulphur dioxide, 20 thousand tones of nitrogen oxide and three quarters of a million tones of ash. These emissions produce much environmental damage including global warming through the green house effect. Similarly, for hydel projects, large environmental effects and loss of land occurs. For the Kaiga nuclear power plant, a land requirement of 0.06 hectare/MWe was required, whereas for the hydel projects in Karnataka forest area cleared is 18.6 hectares/MW or a factor 300 times higher.


Global levels of the main green house gases - carbon dioxide, the chlorofluorocarbons, methane and nitrous oxide are increasing as a result of rising energy demand and industrial expansion. Carbon dioxide from fossil fuel combustion accounts for about 40 percent of the global warming. Hence control of green house gas emissions is essential. In addition to increase of energy production and utilization efficiency it is essential to substitute fuels like coal by nuclear power.


FBR Safety Research


Extensive research has been done to establish the safety principles of fast reactor design and operation. The safety experiments have been held to standardize computer codes for modeling various kinds of accidental situations. In fact, much of the research has led to increase in the understanding so as to be able to improve the stringent, conservative designs used when the information on materials and components behavior was limited.


Air Scrubber for Sodium Fires


Radiation Exposure Limits


The limits are as follows:


General Public: 100 millirem/year
Radiation Workers: 2000 millirem/year


Rem is a unit of radiation exposure.

It must be noted that irrespective of nuclear reactor operation the whole of humanity exists in a sea of natural background radiation. This background varies widely as indicated below.


Place Background Millirem Year
India Average 243
Kerala, Tamilnadu (Coastal Areas) 1152
Central City, Colorado,USA 4400
Guarapari, Brazil 34,000


The actual operations of nuclear power plants in fact give much lower exposure than permitted for the public, less than 1% of the permitted doseand for nuclear workers about 10% of the permitted dose.


FBR Safety Research at IGCAR


  1. 100 kJ Condenser Bank: A 100 kJ, fast condenser discharge facility has been set up at IGCAR to carry out studies related to response of scaled down reactor components under simulated transient loading condition for their safety assessment. Facilities also exist to carry out experiments simulating the slug impact of primary coolant on the roof slab for assessment of its loading under accidental condition by pressurized gas bubble release.
  2. Slow Heating of UO2 Pellets: Simulated electrical heating of UO2 pellets is being carried out to assess the pellet integrity under slow power transients of about 100 W/s and to measure the slumping rate at higher power ramps. Under such situation, the pellet may generate aerosol that need to be characterized as well as filtered suitably.
  3. Air-cleaning Concept for Sodium Fire Containment: To ensure that the levels of sodium aerosols in the exhaust gases released from the plant ventilation system into the environment is within the permissible levels (2 mg/cu.m of sodium hydroxide), a suitable filtering method needs to be employed. The conventional air filters used are not suitable for sodium. Hence, a packed bed type wet-bed scrubber has been developed. Based on experiments their aerosol trapping efficiency has been assessed to be better than 95%. Such a device has the advantage of high mass loading capability by virtue of which they are compact in size.
  4. Sodium-Concrete Interaction Studies: If the burning sodium comes in contact with concrete floor or wall, sodium concrete interactions can occur, liberating gases like hydrogen that are explosive beyond certain levels. Concrete can also get damaged in the process. Based on experimental trials in argon atmosphere and in air, resistance of concrete to thermal and chemical attack of sodium is assessed. These studies pave the way for development of suitable sodium resistance concrete or a liner material on concrete surface, to mitigate the consequences of such interactions.

Sodium Concrete Interaction Studies




Any concept of large scale energy conversion has some element of associated risk which must be balanced against the corresponding benefits from the power generation. In an overall sense, there is a little difference between the risks from nuclear power generation by FBRs or conventional water reactors. Experience and safety evaluation studies have shown that in terms of the probability of a major accident nuclear power is safer than many other forms of large scale energy conversion. Under routine operating conditions, nuclear power generation is much less harmful to the environment and public health than fossil fired power generation.


Interview with Professor Roger Clarke, Chairman, (ICRP)


Professor Roger Clarke, Chairman, International Commission on Radiological Protection (ICRP), on his visit to India sometimes back was interviewed by Dr. K. S. Parthasarathy, Secretary, Atomic Energy Regulatory Board. Present here are the excerpts of the interview published earlier in the AERB News letter (Vol. 11 No. 4)


Q: I remember that it is not your first visit to India.


RC: I came to India a few years ago. But I am attending the Medical Physics Conference for the first time.


KSP: How did you get into the field of radiological protection?


RC: I was a reactor physicist working for the Central Electricity Generating Board (CEGB). One day my head of Division came to me. He wanted an answer for the question, with how many failed fuel pins you can operate an Advanced Gas Cooled Reactor (AGCR). It was in the Sixties. AGCR fuel was more expensive. Primarily the emphasis was on the basis of dose to the public. I collected relevant data. In the process I calculated the inventory of the fission products. I got introduced to health physics. I did some environmental modeling, got interested in radiobiology. Later, I joined the National Radiological Protection Board (NRPB) to establish dose assessment capability after gathering inputs on the environmental concentration of radionuclides. There was a group working on the movement of radionuclides in the biosphere. This was 20 years ago. This group was primarily interested in the study of radiation doses due to releases of radionuclides.


Q: Do you agree that the scholarly discussion on the Linear-Non Threshold (L-NT) hypothesis has contributed to the notion that there is no safe level of radiation. Has it not sensitized the large public to greater and unreasonable levels?


RC: I agree. When experts disagree, the credibility of specialists suffers. If experts do not agree how people can decide which side of the argument is believable. I cannot deny that the arguments on L-NT have created some difficulties. The situation could be bad because there is an increasing possibility that decisions in science may be made by judges and juries in court rooms and not by professional associations or by Royal Societies! The judiciary system may not be able to convince itself about the increased possibilities of radiation effects.


Q: Don’t you think that it is futile to try to get for a deterministic answer to a purely probabilistic question?


RC: Yes. But I do not understand why some people wanted to establish that there is a threshold dose below which there will not be any radiation effect. One of the major difficulties is in tackling the problem of old contaminated sites. Small radiation doses due to residual radioactivities left behind at certain sites can cause very tiny amount of radiation dose. But when these doses are integrated over several thousand years, one may end up with getting significant amount of doses. ... But I believe that we must have started dialogue on acceptable risks.


Q: I am sure we must exclude voluntary risks such as risk due to smoking while we consider acceptable risks.


RC: Yes. I agree. Only involuntary risks are to be considered. In general, I am worried that the philosophy of protection has become somewhat complex. .... We must develop simpler concepts. ICRP must start consultations with other groups and collect ideas for reviewing and consolidating the system of protection. We have started to do that already. It explains why some of our recent documents are better than earlier ones. Consultation with others will help to improve the documents.


Q: Everyone was keen on the on-going L-NT controversy. While the ICRP and the NRPB supported the argument that there is no threshold for the effects of ionizing radiation, the US Health Physics Society was unconvinced.


RC: Certainly different professional groups looked at this issue very differently. American Health Physics Society has its own stated view. ... The arguments put forth by the Health Physics Society are outdated with respect to the recent findings on the Japanese survivors of the atomic weapons. They did not then have the occasion to see the data. The recent data indicated that there could be significant risk at doses as low as 50 mSv, of course with much uncertainty. I do not still understand why they are looking for a threshold. There are many unknown cellular phenomena to be understood. Genomic instability for instance.


Q: Biological effect of radiation has been studied for the past 100 years. The stochastic effects such as cancer would not have even been thought about, but for the long and expensive epidemiological studies. Is it not unfair to spend too much of resources, in fact, vast sums of money to carry out studies about an agent which is now known to be much less hazardous than hundreds of toxic chemicals about which practically nothing is known.


RC: I may say that physicists should take the blame for it. The study of nuclear physics progressed rapidly. Some of the best brains entered the profession. The study of physics was intellectually satisfying and scientifically stimulating. Unfortunately, the same was not true for chemicals. Of late, biologists are also starting to use more and more mathematical formulations. Probably natural sciences are getting ready to make quantitative estimates.


The biggest injustice done is to attach a speaker to a Geiger counter. You must remember that nobody attaches a speaker to a gas chromatograph. In case higher values of hazardous chemicals are detected, the speaker howling is more dramatic and will create more concern than the flickering of a needle. The sound will definitely arrest the attention and create a problem.


Q: Releases from a nuclear facility can be controlled by appropriate methods. It may cause increases in the collective doses to workers. In some instances the collective doses to workers may be far more than those to public. Which option will be acceptable in your opinion?


RC: Both occupational and public exposures are being reduced by optimization procedures. Storage of waste on-site may actually cause a potential for accidental exposure of the public as well as exposing workers. As long as individual doses – workers and public – are acceptably low, the situation is optimized.


Q: Based on the impression that ICRP may revise the dose limits downward, AERB had its first comprehensive review of occupational exposures in 1989, a year before ICRP 60 was issued. We have implemented the recommendations in a phased way. The AERB recommendations are similar to those of ICRP except that the maximum dose limit in any year recommended by AERB is 30 mSv instead of 50 mSv recommended by ICRP.


Our experience is that among the various groups using ionizing radiation, industrial radiography is the most important. In India this field is probably one of the most regulated. The Regulatory Board issues authorizations only if certified radiographers, a site-in-charge and appropriate radiation measuring instruments and protective accessories are available at every site.


RC: Industrial radiography has certain peculiarities. In this field, the workers are likely to be exposed to high radiation doses. The field has more potential for accidental exposures.


Q: Since reducing dose in medical X-ray practice is easier and less expensive, is it not more appropriate to allocate resources more prudently to achieve substantial reduction in collective doses in diagnostic radiology? ICRP should come out with such clear statements on avoiding needless exposure.


RC: We have gone a long way. The International Basic Safety Standards for Protection against Radiation and Safety of Radiation Sources has given certain guidance values. The United Kingdom also accepted certain guidelines. With these precautions, the collective dose can be reduced by about 40%. But this was more than offset by the predominant use of CT scan units.


Q: Though conceptually it is clear that exposure at the dose limit is just tolerable, exposing everyone to the dose limit all the time is not acceptable. Was it not more appropriate for ICRP to recommend a range of values rather than a single value?


RC: The current dose limit of 20 mSv per year average for 5 years offers this operational flexibility. A single number is administratively convenient. It is obvious that the body does not know whether the exposure occurred in one calendar year or another. Biology does not identify this.


Q: The recommendations of ICRP are scientifically the best available. But you will agree with me that these recommendations have enormous social impact. Is it justifiable for over a dozen specialists in purely scientific disciplines to take such decisions which have enormous social impact? Don’t you think that the representation in ICRP should be broadened to include social scientists and economists?


RC: There are various components to this question. ICRP recommendations reflect the best scientific information. We do not say what is acceptable to society or not. There is one recent development. ICRP is currently engaged in more and more consultations with specialists by providing drafts of their recommendations to other specialists and concerned people for review. It would certainly reveal whether there is any inconsistency in the concept and approach put forward by ICRP. It will help to find out whether there is any fallacy in our approach. I believe the recent ICRP documents bear testimony to this.