Founder's Day 2007 - Mr. S.V. KUMAR's Speech


I consider it a great honour to stand before you today to pay homage to Dr. Homi Bhabha, the architect of the Indian Nuclear Energy programme. He had the vision to create blue print for this programme and laid solid foundation on which successive generations of the DAE family have built and brought it to its present stage. It is very satisfactory to note that the world today recognizes India as a country with advanced nuclear technology.

Today, I wish to share with you some of my thoughts on the development of spent fuel reprocessing technology in India and the role it plays in harnessing nuclear energy for the progress of this country. I would also like to share with you some of my personal experiences during this journey.

Energy security is vital for the economic development of our country. The proven resources of uranium in our country are insufficient to support a long-term and meaningful contribution to India’s energy demand by way of nuclear energy. Closing the nuclear fuel cycle by reprocessing the spent fuel and recycle of uranium and plutonium back into appropriate reactor systems helps in exploiting the full potential of nuclear power and maximizes the resource utilization. The Indian nuclear resources have been estimated to be around 61000 tonnes of uranium and around 225000 tonnes of thorium. 

The strategy adopted by the Indian Nuclear Power Programme is to use the country’s modest Uranium and vast Thorium resources resources. In line with this strategy a three stage programme is envisaged. The first stage is based on setting up the Pressurised Heavy Reactors (PHWRs) using indigenously available natural uranium producing electricity and Plutonium. This will be followed by the second stage by plutonium fuelled Fast Breeder Reactors (FBRs) producing electricity and additional quantity of Plutonium and also Uranium-233 from Thorium. The third stage of reactors will be based on Thorium-Uranium233 cycle.

The nuclear fuel, as you all know, after its useful life in the reactor contains unused Uranium, Plutonium and fission products. There are two ways of handling this spent fuel, either to dispose it off as high active waste called the “Open Nuclear Fuel Cycle” or Direct Disposal option, Or to reutilize the unused uranium and plutonium for future generation of energy – called the “Closed Nuclear Fuel Cycle”. It makes economic sense to adopt the closed nuclear fuel cycle and we have chosen to adopt this option. Reprocessing of spent fuel, therefore, is a vital link in closing the fuel cycle. Closing the nuclear fuel cycle by reprocessing the spent fuel and recycle of uranium and plutonium back into the reactor system helps in exploiting the full potential of nuclear power and maximizes the resource utilization.

From waste management considerations, environmental benefits using the closed cycle option are in terms of reduction in the long term radiological risk as a result of uranium and plutonium extraction from the spent fuel.

In the ‘Open Cycle’ or Direct disposal option the plutonium inventory would not be eliminated and would keep increasing with increased nuclear power generation.

The comparison of hazards over extended time scale clearly indicates the reprocessing and plutonium recycle to be a safer option.

Reprocessing followed by vitrification of high level waste generated thereby as widely accepted strategy. This not only reduces the overall toxicity of the waste but also the final volume to be disposed off.

Recycling is more proliferation resistant than the “once through” option because recycling can consume all plutonium while ‘once through’ leaves behind huge stocks of spent fuel which contain recoverable plutonium. The spent fuel from the “once through” option buried for disposal, represents virtually a plutonium mine which would need to be protected indefinitely. 

The decision to build a plutonium plant was taken in 1958 under the leadership of Dr. H.N Sethna and Mr. N. Srinivasan. I joined the group in 1959 after completing the Training School. Three of us – Shri G.R. Balasubramanian, Shri M.K.T. Nair and myself, were sent to the Radio-chemistry laboratory in south site. We were to carry out the process related studies. Shri Nair took up the Ion Exchange studies, Shri Balasubramanian on metal production studies with Late Shri P.R. Roy and I took up the Solvent Extraction studies. We were in a strange situation – three engineers working among a group of chemists! But this gave us an opportunity to associate and appreciate the importance of radio-chemistry and handling radioactive substances. In fact, once Late Dr. M.V. Ramaniah asked me to give a talk in the monthly colloquium about how the determination of distribution coefficient of uranium and plutonium in TBP would help in designing solvent extraction column. I had to describe the equilibrium curve, operating line and the Mcable Thiele diagram! Our stay in Radiochemistry laboratory also helped me to understand the importance of interdisciplinary interaction which is one of the best features of this great Institution. This concept was very useful in subsequent years of my career. 

Dr. Bhabha used to bring visitors to the Radiochemistry laboratory. There were two laboratories called the Alpha lab and the Beta lab. He would leave the visitors with Dr. H.D. Sharma in one of the labs and drift to the other lab. On one such occasion, I was working in the glove box and a soft voice from behind asked me what I was doing. I turned round and saw Dr. Bhabha! It was such a surprise and I explained the experiment I was doing. He was very informal and eager to know things. Sometimes Mr. E.C. Allardice who was the Controller used to accompany Dr. Bhabha. He was a tall man and was very particular about cleanliness. He had the habit of running his fingers on top of the glove boxes to see any dirt! Once, our health physicist told him it may be radioactive dust! He stopped the practice after that!

After about 9 months we moved to the plant site as the construction work had started. The reprocessing programme was launched with the design, construction and commissioning of the demonstration plant at Trombay. Preliminary design work was started in January 1959 and completed by January 1961. During this period experiments with pulsed perforated columns were carried out to confirm design data. This was followed by finalization of the process and equipment design, and fabrication and installation of equipment and piping in the process cells and associated systems. The plant was commissioned in 1964 to reprocess the spent fuel from CIRUS. The PUREX process employed in this plant comprised a decontamination cycle, a partition cycle and two separate parallel cycles for the purification of uranium and plutonium. The reductant used in the partitioning stage was ferrous sulphamate solution in nitric acid medium. The final purification of plutonium nitrate solution was by ion exchange. The Trombay plant had a nominal capacity of 30 tonnes HM/year. Being first of its kind, the design philosophy was completely based on direct maintenance concept. The Trombay plant was preceded by a very limited amount of laboratory experiments. The experience in operating the plant and in assessing future requirements in reprocessing served as the basis for R&D programme in the field.. The successful operation of the plant also helped in providing plutonium for pursuing various programmes of nuclear research and development. 

We were all curious to know how a plutonium plant would look like. When the process cells were ready I was entrusted with the erection of the plutonium purification cell (Cell-5). We had only the engineering flow sheet and the piping drawings were like diffraction grafing! So we erected the cell using only the engineering flow sheet. Dr. Sethna used to come to the site very frequently. He was a thorough engineer. Initially, we used to tidy up the cells during his visit. He told us not to bother and waste time. He told things would be scattered only when you are working! The Trombay Plutonium Plant began operation in mid-1964. It was a memorable day when the first irradiated CIRUS rod was charged into the dissolver! With this India became one among the five countries in the world (others being US, UK, France and the Soviet Union) and demonstrated capabilities in the advanced technology of nuclear fuel reprocessing.

I must mention here that the contribution of Shri N. Srinivasan was outstanding. He brought together a group of young scientists and engineers and succeeded building the first demonstration plant at Trombay.

Some of the major problems encountered during the operation of Plutonium plant were failure of weld joints due to high corrosive nature and high temperature in certain equipment in the cells. This was attributed at the time to the material of construction namely SS 347. There were also some component failures like the control valves and diaphragm pumps. 

It was decided in 1968 to build a plant to reprocess the spent fuel from power reactors. In the early 1970s construction was started for the Power Reactor Fuel Reprocessing Plant (PREFRE). Many innovative features based on our experience of operation of the Plutonium Plant were incorporated in this plant. Innovation in the face of difficulties has been the hallmark of human development. Simultaneously, intense R&D activities on various areas like pilot plant engineering studies, process chemistry studies using laboratory mini mixer settlers etc. were initiated. The Trombay plant gave sufficient impetus to continue R&D in the domain of reprocessing. In particular, these included solvent degradation studies, development of equipment and systems for higher plant throughput and bringing about improvement in performance, representative sampling and analysis, on-line instrumentation and data acquisition system for process control and operation safety. The results of these efforts were integrated into the design of the second reprocessing plant constructed at Tarapur for the treatment of spent zircalloy clad oxide fuel from Tarapur and Rajasthan Atomic Power Stations. PREFRE, Tarapur was commissioned in 1975. 

This plant uses a chop-leach technique for the head end and uranous nitrate stabilised by hydrazine as the reductant for partitioning. The flowsheet was subsequently modified to have co-decontamination cum partitioning cycle. The ion exchange purification of plutonium was replaced with a 20% TBP solvent extraction/stripping cycle to cater to the needs of higher Pu throughput. Several innovations such as pneumatic pulsers in place of mechanical pulsing, air lift as a metering device for radioactive process solutions, thermosyphon evaporators, solvent wash systems for aqueous streams emanating from solvent extraction etc were introduced in the plant. The experience gained from the earlier plant gave enough insight into the material of construction used for critical equipment fabrication and qualification. Austenitic stainless steel variety 304 L was the choice available at that point of time. Except for the head-end, which had provision for remote maintenance of in-cell equipment, the concept used for the rest of the plant was that of direct maintenance. The plant had a nominal capacity of 100 tonnes HM/year. 

This plant also provided experience in the design of appropriate packages and safe in-land transportation of spent fuels which is a vital input for locating the reprocessing facilities. Since 1975 there have been several shipments of spent fuel involving several kilometers with no property damage or personal injury, no breach of containment, and very low dose rate to the personnel involved.

Mr. M.K.T. Nair moved to Tarapur as the Plant Superintendent. We – Mr. A.N. Prasad, Mr. M.K. Rao and myself – used to go to Tarapur every week to provide support and review the operations. This we did for almost two years!

As the plant was to process the safeguarded fuel from the power reactors TAPS & RAPS, there was need to gear up for this. Safeguards requirement for the reactors is simpler since the fissile material is confined in the fuel bundles and item counting and surveillance would be adequate. Whereas in the reprocessing plant, the fissile material is in an open condition as a solution and accounting procedures are challenging. There was need to accurately calibrate the Accountability Tanks, apply statistical methods to determine the hold up, determine the sampling and analytical errors and maintain the Material Unaccounted For (MUF) within the stipulated limits. Several campaigns were carried out under the IAEA Safeguards and its stood the test of International Inspection.

The PREFRE Plant demonstrated our capability to design and operate bigger plants to reprocess the oxide fuels from power reactors with higher burn ups and subject the plant to satisfactory international inspection.

The Trombay plant needed to be refurbished to process the DHRUVA spent fuel. This necessitated decommissioning of the Plant. Valuable experience was gained in the decontamination and decommissioning of a plant that had handled very high activity. 
Six out of the eight process cells were decontaminated, decommissioned and reconditioned for installation of new process equipment for the new plant. During decommissioning and reconditioning, radiation levels inside the process cells were reduced to the maximum extent possible and the surfaces were made free of contamination to carry out the erection of new equipment and associated piping without undue exposure and personnel contamination. 

About 50 process equipment and associated piping ( ~ 40 kms length) were decontaminated using the various chemicals such as nitric acid of varying strength, Sodium hydroxide and sodium carbonate solution, caustic solution + tartaric acid mixture, 1M nitric acid + 0.05M Sodium fluoride (to limited extent), 0.2 M oxalic acid + 0.2 M hydrogen peroxide etc. Radiation level of ~ 100 R/hr on few process equipment were decontaminated to levels below 250 mR/hr during the internal decontamination. 
In most of the cells, the cell floors were indicating radiation level ranging from 2 to 7 R/hr. In cell-3, hot spots upto 2500 R/hr were also present. External decontamination using water, EDTA, nitric acid, trisodium phosphate, caustic soda- tartaric mixture etc was carried out to reduce the floor radiation levels. Hot spots on the cell wall made of hematite concrete were removed by chipping the concrete. 

For the entire decommissioning operation carried out in six cells, ~ 8150 personnel entries were made in the process cells spending ~ 4200 man-hours. The total personnel exposure was 867 man-rem. The success of the decommissioning operations could be gauged from the insignificantly low background levels of radiation fields (<5mR/hr) ultimately achieved, the absence of transferable contamination on cell surfaces, and the fact that personnel exposures were well with in regulatory limits. This resulted in salvaging of most of the cells and permitted almost unrestricted access into them for carrying out fresh installation work. The feed back information and experience obtained during the execution of the above mentioned jobs once again emphasised the importance of making provisions for decommissioning to be incorporated at the design stage of reprocessing plants. 

The decommissioned cells along with the piping embedments and the auxiliary areas were used with suitable design modifications for the installation of new equipment and associated piping for the augumented plant with enhanced capacity ofr processing of spent fuel from Cirus and Dhruva.

This was indeed a great achievement and is an example of proper planning and coordinated execution. Very few countries in the world have carried out such a difficult task. The expanded Trombay plant was put back into service in 1983 and had improved features like pneumatic pulsing of solvent extraction columns, use air lift for interface control etc.

Introduction of metering airlifts in addition to metering pumps, steam ejectors with flooded suction, column interface controls using either airlifts or diaphragm control valves, thermosyphon evaporators in waste cycle, use of uranous nitrate solution instead of ferrous sulfamate , diluent wash columns for trapping of organic from aqueous streams, change in layout of reagent heads of metering pumps were the major design/process/layout changes incorporated in the new plant. 

After the successful operation of the power reactor reprocessing facility at Tarapur and the experience gained during the decommissioning operation of Trombay plant, the need arose to augment the reprocessing capacity to treat the spent fuel from the increased nuclear power generation. To cater to the needs of reprocessing zircaloy clad natural uranium oxide spent fuel from Madras atomic power station, a new plant was designed near the power station location, Kalpakkam, with 100 tonnes HM /year capacity. The execution of this plant was carried out with the involvement of industry in the fabrication of equipment, their installation and piping work. 

This plant has many innovative features like computer aided design of piping layout, indigenously designed and manufactured chopper unit, more sophisticated instrumentation etc. The plant was commissioned in 1996.

To sustain the second stage of our nuclear power programme large quantities of Plutonium will be required for the proposed PFBRs. This calls for reliable reprocessing plants of higher capacities.

233U required for the third stage breeder reactors will obtained by irradiation of thorium in PHWRs and FBRs. An Advanced Heavy Water Reactor (AHWR) is being developed at BARC, Trombay to expedite the transition to Thorium based systems. The third stage calls for reprocessing capability of thorium based fuels. In view of the importance of utilization of thorium in our programme, small laboratory trials were carried out initially from separation of 233U from irradiated Thoria rods from Cirus reactor. 

Most of the experience in THOREX domain has come from the recovery of low amounts of 233U bred in irradiated ThO2 from CIRUS research reactor. Two small engineering pilit setups were operated at Kalpakkam and Trombay. Subsequently, an engineering demonstration facility (Uranium Thorium Separation Facility, UTSF ) was operated in this domain. The old evaporation plant of Phoneix plant was decommissioned and three out of four process cells were reconditioned and utilized for erection of equipments required for irradiated Thoria rod processing. This plant was commissioned in 2003 and was operated for ~ 6 months to process irradiated Thoria rods from research reactors. Cal-mix type mixer settlers were used for extraction process. THOREX process using 3% TBP was used for selective extraction of 233U. Final purification of 233U was carried out using cation exchange column. 

Further studies are required to deal with the reprocessing of irradiated ThO2 bundles used in the initial flux flattening of PHWRs. A facility for processing of Thoria bundles used in PHWRs, Power Reactor Thoria Reprocessing Facility (PRTRF) is being constructed at Trombay. This facility will provide further experience in processing of Thorium based fuels from power reactors. 

AHWR, a hybrid reactor, is an innovative reactor design meeting the dual objective of interim utilisation of the PHWR produced Pu and introducing the available Th resources to enhance the fissile inventory. This reactor uses Th-Pu fuel pins along with Th-233U fuel pins.

The spent fuel cluster before reprocessing would undergo disassembly for segregation of (Th-Pu)O2 pins, (Th-U233)O2 pins, structural materials and burnable absorbers. The (Th-U233)O2 pins will require a two stream reprocessing process i.e. separation of thorium and uranium whereas the (Th-Pu)O2 pins will require a three stream reprocessing i.e. separation of thorium, uranium and plutonium. At the reprocessing facility, the pins containing (Th-Pu)O2 MOX and pins containing (Th-U233)O2 MOX will need to be processed in separate set of cycles. The rejects generated in fuel fabrication facility (~20% of the pellets produced) with high γ-dose will have to be brought back to fuel reprocessing plant for reprocessing again. 

For partitioning of Pu from U, generally large amount natural uranous salt is added from a specially designed electrolyser to generate natural uranous salt. But for AHWR fuel reprocessing, isotopic dilution of U233 is not desirable by addition of such large amount of natural U+4 salt. This calls for development efforts for process technology like use of Hydroxyl Amino Nitrate, reduction with H2 in presence of a catalyst, electrochemical in-situ reduction along with the corresponding process equipment for in-situ reduction of Pu+4 to Pu+3 needed for partitioning. 

The reprocessing of AHWR spent fuel gets further complicated due to extreme radiological hazards and the need for a special off-gas filtration system. Further, the higher energy (n, 2n) reactions encountered by Th-232U during the irradiation in Th-U233 fuel also lead to the formation of long lived Pa-231 (t ˝ = 3.27 x 104 yr. Even though production of minor actinide in Th-U233 fuel is orders (~102 – 106 time) times less compared to those in U-U235 fuel, production of minor actinides in AHWR is expected to be higher than that in PHWR due to the use of 239Pu in AHWR fuel which leads to higher generation of Am and Cm isotopes. 

The extreme chemical inertness of ThO2 fuel calls for catalyst assisted dissolution process. After reprocessing, 233U is always associated with 232U, whose daughter products are hard gamma emitters. The radioactivity of 232U associated with 233U starts increasing after separation. This poses radiation exposure problems during its transportation, handling and refabrication. Hence, it is targeted to minimize delay between separation of 233U and its refabrication into fuel. The 233U based fuel needs to be fabricated in shielded facilities due to radioactivity associated with 232U. This also requires considerable enhancement of automation and remotization technologies used in fuel fabrication. 

Reprocessing of spent Fast Reactor fuels would throw up new challenges essentially due to higher radioactivity and higher inventory of fissile plutonium. Therefore, attention needs to be given to have adequate criticality control. This will require a conceptual change in the extraction equipment such as fast contactors, radiation damage to the solvent used in the process also requires special attention and there would be need to have appropriate solvent purification systems. The non aqueous and pyro-metallurgical process offer attractive alternatives. Compact Reprocessing Facility for Advanced Fuels (CORAL) was commissioned and operated to demonstrate successfully the reprocessing of FBTR fuel assembly. Simultaneous with the commissioning of CORAL, a host of R&D activities have been initiated towards development of processes as well as equipment for fast reactor fuel reprocessing. CORAL will continue to be used as a test bed for equipment and technologies developed for reprocessing of fast reactor spent fuel. 

Based on the experience gained form CORAL, a Demonstration Fuel Reprocessing Plant (DFRP) is being designed and constructed at Kalpakkam. For reprocessing of fuel from fast breeders and re-fabricating into MOX fuel assemblies, an integrated Fast Reactor Fuel Cycle Facility (FRFCF) is being set up at Kalpakkam to close the fuel cycle. Some more R&D efforts are needed in this direction and I am sure the young scientists and engineers entrusted with this task would come up with appropriate solutions.

Any description on reprocessing activity would not be complete without a discussion on the waste management strategies. The safe management of radioactive waste including its disposal has been given utmost importance right from the inception of India’s nuclear power programme. Waste management facilities have been set up and are operational at various sites all over the country. Some of these facilities are co-located with the nuclear power plants. There are also seven near surface disposal facilities operating in the country. An interim storage facility for storage of vitrified high level waste overpacks spanning over a period of 30 to 40 years is also operational. Valuable experience gained during the operation of these facilities are being utilized for the design and operation of new waste management facilities. 

In the present context, for management of low and intermediate level waste, attention is focused towards technology development that can lead to design of compact equipment, effective decontamination and minimization of secondary waste. In this direction, technologies finding a vital role include synthesis and use of specific sorbents, ultra filtration, development of advanced oxidation technologies for destruction of spent organic resins etc. 

For vitrification of HLW, present plants in India are based on induction heated metallic melter and joule heated ceramic melter technologies. Globally emerging vitrification technology based on cold crucible induction melting is under study to address various requirements such as high temperature availability, high waste loading, high specific capacity, compatibility with new matrices etc. In order to keep pace with expanded atomic energy programme, thrust areas identified for future vitrification plants are (a) throughput enhancement (b) actinide separation. 

Studies on evaluation of vitrified waste product under simulated conditions have been conducted in specially designed hot cells at Tarapur and this programme is pursued on long-term basis. 

Presently each reprocessing plant has a dedicated waste tank farm for interim storage of radioactive waste generated during reprocessing. With the development of vitrification technology and successful operation of vitrification plants at Trombay and Tarapur, it has been decided to setup integrated reprocessing and waste management facilities in future. In this direction, an integrated plant with a capacity of 300 TPY is being planned under XIth plan at Tarapur. A similar plant is also being planned at BARC campus, Vizag. 

The reprocessing technology is further to be developed to cater to the needs of the different reactor systems and also to be augmented to meet the additional requirements. Two spent fuel storage facilities each of 800 Tonnes at Tarapur and Kalpakkam been constructed and commissioned. These storage facilities, besides reliving the burden of storage at reactors will also ensure continuous availability of spent fuel for reprocessing plants. To cater to the requirements of Plutonium for fast breeder programme, two more reprocessing plants, one each at Tarapur(100 TPY) and Kalpakkam (150 TPY are currently under advanced stages of construction. 

In conclusion, I may say sufficient expertise and experience has been obtained in reprocessing of spent fuels from power reactors to facilitate construction and operation of larger plants. Reprocessing of spent fuel from fast reactors and thorium reactors will be meeting with new challenges and require more R&D efforts.

I thank you all for giving me a patient hearing. I once again thank Dr. Kakodkar and Dr. Banerjee for having given me this opportunity to speak on this auspicious occasion.

Thank you.