A perspective on fast reactor fuel cycle in India

Progress in Nuclear Energy; Vol. Available online at www.sciencedirect.com

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A P E R S P E C T I V E ON FAST R E A C T O R F U E L C Y C L E IN INDIA BALDEV RAj,1 H.S. KAMATH, 2 R. NATARAJAN 1 and P.R. VASUDEVA RAO 1 1Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, India 2 Bhabha Atomic Research Centre, Trombay, Mumbai 400 085, India ABSTRACT The growing energy needs of India can be fulfilled only by judicious mix of all the fuel resources. It is possible to achieve energy security and sustainability through the introduction of fast reactors in an expeditious manner and closing the fuel cycle. This approach is inevitable in view of the limited uranium resources in India~ The Fast Breeder Test Reactor (FBTR) built by India uses mixed carbide as fuel and the 500 MW(e) Fast Breeder Reactor Project (PFBR), to be operational in 2010, will use mixed oxide as fuel. It has also been decided that fast reactors beyond 2020, with enhanced safety features and having better economy, will use metallic fuel. Having successfully operated FBTR with carbide fuels, we need to develop the fuel cycles for both the mixed oxide fuel in the near future and the metallic fuel expeditiously. The progress achieved so far and the plans for implementation are discussed in this paper. © 2005 Elsevier Ltd. All rights reserved KEYWORDS Fast Reactor; Fuel Cycle; Fabrication; Reprocessing 1. INTRODUCTION

The total energy demand of India in 2050 is envisaged to be about 1000 GWe in terms of installed capacity. The contribution of nuclear energy has to be increased at the fastest possible pace to be able to meet about a quarter of the national electricity demand in 2050. This can be realized by India only through the effective use of the limited resources of uranium by employing FBRs, as otherwise, the available uranium can support a power generation of only 10 GWe through the P H W R route. Development of fast

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reactors and their utilisation for breeding 233U from the thorium based blankets and later using the 233U thus produced in thermal or fast breeder reactors form the second and third stages of the Indian nuclear power development programme. Closing of the fuel cycle is an important part of the strategy to ensure the growth of nuclear energy at the desired pace. The Indian fast reactor programme started with the 40 MWth FBTR, commissioned in 1985 at Kalpakkam. It is the only reactor in the world which uses the uranium-plutonium mixed carbide as driver fuel. The choice of the mixed carbide fuel for FBTR was necessitated by the technological problems anticipated in the use of high Pu content MOX fuel, and the non-availability of enriched uranium. For the first core (Mark I), a Pu/(U+Pu) ratio of 0.7 was required in the fuel. For the Mark II expanded core, the Pu/(U+Pu) ratio is 0.55. The fuel cycle of FBTR is being successfully closed, thanks to the multidisciplinary, inter-institutional research programmes which have been pursued in a focused manner, in the areas of reactor physics, chemistry, fuel fabrication, characterisation, materials science and engineering, and reprocessing For the PFBR, a uranium-plutonium mixed oxide (with PuO2 content of 21% in inner zone and 28% in outer zone) has been chosen as the driver fuel. It has also been decided that only PFBR and the next four FBRs will have the mixed oxide as the fuel and the future FBRs will use metallic alloy of U-Pu-Zr as the fuel. The decision is based on the potential of metallic alloy fuel for high breeding and high burn-up, and better safety features of fast reactors based on metallic alloy fuels (Cahalan et al., 1988). This paper describes the progress achieved so far and our future plans for the development of fast reactor technology. 2. FUEL FABRICATION 2.1 FBTR fuel fabrication The fabrication of the uranium-plutonium mixed carbide fuel of FBTR has to be carried out in inert atmosphere glove boxes because of its highly reactive and pyrophoric nature. Since the carbon/metal (C/M) ratio decreases with burn-up, to ensure that the metal phase does not form during irradiation, the fuel is fabricated as a monocarbide with 5 to 15 vol.% of sesquicarbide. The specifications of the fuel were arrived at based on thermochemical modeling, and the limited experimental data on the thermochemical properties of the fuel, available at that time. The fabrication flow sheet, incorporating the inspection steps for quality control is given in Fig. 1. The Mark I mixed carbide fuel has performed extremely well, reaching a burn-up of 125 GWd/t, without any fuel pin failure. This success story validates the selection of the fuel as well as the technical excellence of the various institutions of the Department of Atomic Energy.

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FABRICATION FLOW-SHEET FOR MIXED

PLUTONIUM-URANIUM CARBIDE FUEL PINS lNSPECTED STAINLESS STEEL HARDWARE: CLAD-TUBES, END-PLUGS, PLENUM*TUBES, DISCS, SPRINGS

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ELIVERY CHECK 1 O2/PuO2: O/M, I, IC,] MIXING, GRINDING, BLENDING

I CL ",NGANDDR ,NG

I'°'' I METALLOGRAPHY I pOF~ET~JP EL%D ~ 4 ~

~ ' FIRST EN D'PiUG WELDING [

I BUFFINGCL,ANINGDHYINO I

HITE: I, SA

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TABLETLNG, CARBOTHERMICREDUCTION I CRUSItING AND MILLING

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MLXING, PRE-COMPACTION, GRANULATION, FINAL COMPACTION

i) U, Pu, O, M,, C, I ]

it) MC, M=C~,MO~ ii) SA, A.D

DEWAXING & SLNTERING !) U, Pu, O, M, C, I i) M e , z~c~, MO=. ii) SA, AD

lii) METALLOGRAPHY OF END-

L "°T"

PELLET INSPECTION i) PI, DM, D, LM ii) MT, AR, MC, 1V/tC.j, MC~ iii) U, Pu, C, O, N, I, TG FORMING FUEL-PELLET STACKS OF ACCEPTED PELLETS

I LEGEND: O/M-OXYGEN TO METAL RATIO SA S U R F A C E A R E A A D - A P P A R E N T DENSITY I-IMPURITIES I C - ISOTOP C C O M P O S TION iDM - DIMENSIONS 'LM - LINEAR MASS D - DENSITY P I - PELLET INTEGRITY AR - ALPHAAUTORADIOGRAPHY MT - M E T A L L O G R A P H Y f I G - TOTAL G A S i

LOADING ACCEPTED FUEL PELLETS,

METALLOGRAPHY OF SET-UP AND PROCESS WELDS

i IPJ SECOND END-PLUG WELDING t V B~F~NTGA~ILI~AA~/N~c)F FUEL

DECONTAMINATION CHECK, He-LEAK TEST, RADIOGRAPHY, METROLOGY, X-GAMMA AUTO RADIOGRAPHY

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WIRE-WRAPPING, PINCHING AND SPOT WELDING WIRE TO TOP END-PLUG

M ETROLOGY, VISUAL, GAMMA-RADIOMETRY, WEIGHING [NUMBERING FUEL-PIN, CLEANING, DRYING I

DOCUMENTATION AND DESPATCH

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Fig. 1. Flow sheet for FBTR fuel fabrication 2.2 PFBR fuel fabrication The fuel proposed to be used in PFBR, as mentioned earlier, will contain 21 and 28 mole percent PuO2 in solid solution with depleted uranium dioxide The fuel pin consists of annular MOX fuel pellets, of nominal diameter, 5.6 mm, encapsulated in a D9-SS clad tube Extensive trials were carried out with

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Table 1 PFBR Fuel Pellet Specifications No. 1

2

3

Characteristics of Fuel Physical Characteristics a) Pellet Diameter b) Length of the pellet c) Linear mass d) Pellet inner-diameter e) Fuel stack length Chemical Characteristics Pu content U content Maximum permissible impurities O/M Dissolution test (Residue) Metallurgical Characteristics Grain Size Micro-homogeneity (PuO2 particle size)

Specifications 5.58 ± 0.05 mm 8+2mm 2.184-0.07 g/cm 1.904-0.2 mm 1000.0+2.5 mm 18.7% / 24.4% 69.9% / 63.7% 5000 ppm 1.994-0.01 < 1.0 w/o 5 to 50 gm < 100 gm

natural UO2 and PuO2 to finalize the process flow sheet and standardize the process parameters. The pellets were fabricated from the feed powder by cold compaction and sintering. The specifications of PFBR fuel pellet are given in Table 1. To validate the PFBR fuel pin design, a test fuel sub-assembly containing 37 fuel pins with annular MOX fuel pellets is being irradiated in FBTR. For the first time in India, rotary compaction press, designed and built in India, was inducted for fabrication of nuclear fuel pellets for the above fuel pins. The test fuel has the composition (U0.71 Pu0.29)O2, with enhanced uranium fissile fraction achieved by addition of E33u in order to achieve the desired linear rating. 3. POST-IRRADIATION EXAMINATION OF THE FUEL

A programme for carrying out detailed post-irradiation examination (PIE) of the irradiated FBTR fuel was drawn up and equipment have been set up in hot cells for the visual examination as well as various measurements such as dimensional measurements of the subassembly, eddy current testing, x-radiography and metallography. The micrographs of the middle of the fuel column obtained during the examination of the FBTR fuel after 25, 50 and 100 GWd/t bum-up are shown in Fig. 2. The circumferential cracks in the cross-section of the fuel after 100 GWd/t bum-up are indicative of the closure of the fuel clad gap. The fuel was qualified for higher levels of burn-up, of the order of 150 GWd/t. based on the measurement of residual ductility as well as swelling of clad. Conventional mass spectrometry technique (Balasubramanian et al., 1999; 2003) as well as a new technique based on HPLC (Sivaraman et al., 2002) were used for the determination of the burn-up of the fuel.

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Fig. 2. Micrographs of the middle of the fuel column after various burn-ups 4. FUEL REPROCESSING The success of Fast Breeder Reactor technology of the Indian nuclear program depends on the reprocessing of irradiated fuels with a high burn up of over 100 GWd/t after short cooling (typically about 6-12 months). Modified PUREX process based on chop-leach process will be employed for the reprocessing of the fast reactor fuels. The technology of reprocessing fast reactor fuel is being established in four phases. The first phase of the program, namely the developmental phase, is to understand and provide solutions for the challenges in fast reactor fuel reprocessing. The second phase, namely the pilot plant phase, is the construction and operation of a pilot plant called, Lead Mini Cell, (LMC), to demonstrate the process flow sheet, optimize the process parameters and to validate various equipments under hot environment. The third phase namely, the demonstration phase, is the construction and operation of a Demonstration Fast breeder Reprocessing Plant (DFRP) to gain experience in the reprocessing of fast reactor fuel with high availability factors and plant throughput. In-service inspection techniques for critical equipments like dissolver and underground waste storage tanks, on line monitoring of raffinate and product streams etc would also be demonstrated in this plant. In the fourth phase, construction and operation of the commercial plant, namely, Fast Reactor Fuel Reprocessing Plant (FFRP), to reprocess the spent fuel from PFBR will be carried out. In the first phase of development, development of electrolytic dissolution techniques for the fuel and Ti based materials and alloys to withstand the corrosive environment of the dissolver, their fabrication and joining techniques with stainless steels (Kamatchi Mudali et al., 1993) have been carried out. The solvent extraction flow sheet has to be designed for low plutonium losses with reasonable number of extraction stages. Due consideration to process upsets under plant conditions has to be given. The extraction process has been modeled by a simulation code SIMPSEX. Codes have also been developed for studying the extraction behavior of U, Pu as well as fission products such as ofTc, Ru, Zr and Np (Shekar Kumar et al., 1996). The developments of equipment for chopping, centrifuges for clarification, centrifugal extractors (Koganti et al., 1986; 1994) and electrolytic partitioning equipment (Nair et al., 1990) have been completed.

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Remote handling equipment such as manipulators, in-cell cranes, sampling systems and in-cell analytical gadgets have been developed. The LMC facility comprises of a compact lead shielded hot cell with 250 mm or 200 mm thick lead shielding (depending upon the [3, 7 radioactivity in different zones) housing an or-tight stainless steel containment box. The cell is provided with radiation-shielding windows and gadgets for remote operation and maintenance. About 35 process vessels and 30 equipments are installed in this compact hot cell. To carryout the processing in this compact cell, about 2 km of intricate stainless steel piping involving 3000 bends and 2000 X-radiography joints were carried out within the limited area of the facility. The process flow sheet being followed for reprocessing irradiated carbide fuels is shown in Fig. 3. The recovery and the decontamination factors achieved during reprocessing 25 GWd/t fuel were found to be good. With the success of the first two phases, the third phase of the programme, namely, design, construction and operation of DFRP has been undertaken with the main objective of reprocessing FBTR spent fuel for closing the fuel cycle and demonstration of recovery and decontamination factors. The other objectives include waste volume reduction, acid recovery and reuse, demonstration of laser dismantling of fuel subassemblies, demonstration of multi-pin chopping of PFBR fuel pins etc. The plant is to be commissioned by March 2007. FFRP will be operational in 2012 and will be used for reprocessing the fuel discharged from PFBR.

Irradiated Carbide Fuel Pins [

I

I

~ Electro-oxidativedissolution I Diss°lver s°lutiOn [ 30% TBP, HeavyNormal Paraffin Organic phase ~ ~ Aqueousphase [ U, Pu [ [ FissionProducts ] Electrolytic Partitioning Aqueous phase Organic Phase r

7

I 0 r ~ Stripping U in Aqueous solution ]

1

I Reconversionthrough oxalate route ] [ Reconversionthrough ADU route J

Fig. 3 Process flow sheet for reprocessing the carbide fuels of FBTR

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5. R&D FOR IMPROVING THE FBR FUEL CYCLE 5.1 Short term R&D related to fuel fabrication With a view to building up experience on high plutonium containing mixed oxide fuels for limiting burn-up case study and achieving comprehensive experience on oxide fuels, it has been proposed to irradiate mixed oxide fuel pins with 45% plutonium oxide in FBTR. The process flow sheet for the fabrication of these fuel pellets has also been standardized. The thermal diffusivity and heat capacity of these mixed oxides have been measured using the laser flash and drop calorimetric techniques. Studies on the fuel- sodium compatibility are also being carried out. The PFBR fuel fabrication plant will be part of the PFBR Fuel Cycle Facility to be set up at Kalpakkam comprising the PFBR Reprocessing plant and waste management plants also. This co-location concept obviates the need for transportation of fuel through public domain. The integration of fuel reprocessing and fuel fabrication, through the use of sol-gel process, is an area, which is receiving focused efforts. Sol-gel technology based on internal gelation has been developed for urania, thoria and U,Pu mixed oxide with PuO2 content up to 15% (Vaidya et al., 1981). The development of sol-gel process for mixed oxide fuel with high Pu content (as required for FBTR) is now being undertaken. A laboratory scale facility is being set up for the demonstration of remotisation of the fabrication of test fuel pins with mixed oxide microspheres produced by sol-gel route. It has been proposed to irradiate test fuel pins containing mixed oxides with 45% PuO2 in FBTR to generate data on the performance of these fuels in fast rectors. 5.2. Short term R&D related to development of advanced clad materials Development of high burn-up fuels capable of long residence time in reactors is primarily limited by clad and wrapper, due to dimensional changes caused by swelling and creep. Hence development of advanced clad and wrapper materials for achieving high burn-up is being carried out by a combination of alloy design and thermo-mechanical processing routes. Alloy D9 has been chosen as the clad for PFBR A detailed characterization of D9 was carried out to establish the optimum amount of minor alloying elements needed for better creep, fatigue and creep-fatigue interaction resistance. The kinetics of precipitation and solute distribution of minor elements in the clad and wrapper are the key concerns and are thus areas of research at the center. Positron lifetime techniques have indicated that Ti/C = 6 with a cold work of 20% is the optimum for achieving maximum TiC precipitation within the grain. The fine scale precipitation characteristics of titanium carbides in alloy D9 have also been investigated by lattice imaging techniques using a high resolution transmission electron microscope (HRTEM) (Fig. 4). To obtain insights into void swelling behaviour of D9 alloy, carefully designed ion beam irradiation studies were carried out using a 1.7 MV Tandetron accelerator. Analysis of the S-Parameter obtained from positron studies of the irradiated specimens suggested the peak swelling temperature to be 873K Detailed investigations to develop modified grades of D9 capable of withstanding up to 200 dpa for future FBR's are in progress.

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Fig. 4. High resolution TEM image showing fine coherent TiC precipitates seen as Moire fringes 5.3 Short term R&D related to reprocessing Reduction of waste volume is proposed to be achieved by employing techniques such as microwave denitration and hull compaction. For accurate metering of active solutions, constant volume feeders are being developed. R&D work has been initiated for the on-line monitoring of Pu in various process streams as well as waste materials using on-line alpha monitors and neutron collars. Pu in the hull in LMC will be estimated by a method using 144Ce. Neutron interrogation techniques are under development for deployment in FFRP. Inspection of LMC dissolver is proposed to be carried out with immersion ultrasonic technique for wall thinning measurement and laser triangulation for inner surface profiling. Visual inspection techniques with suitable robots will be tried in existing waste tank farm. Suitable gadgets are under development for inspection of evaporators of FFRP. It is known that tri-butyl phosphate (TBP) has a few disadvantages such as tendency to form third phase at high Pu loadings, significant solubility in the aqueous phase and formation of deleterious products by degradation at high radiation doses. R & D has been planned on possible alternate extractants such as higher homologues of TBP and long chain monamides to avoid these problems. The extractant for PFRP would be finalized based on the results of development of these extractants. Corrosion of materials in the high molar nitric acid, high temperature and high radiation environment is an important problem which is of interest to the reprocessing industry (Baldev Raj et al., 2001). Work is in progress to improve the performance of the AISI type 304L austenitic stainless steel presently used as the material of construction for the FBR reprocessing and waste management plants, and for developing corrosion resistant alloys like Ti-5%Ta-l.8%Nb for dissolver and evaporator. Special (Baldev Raj et al., 2000) coating technology for Ti components has been developed to improve their corrosion resistance. A unique nitric acid corrosion loop has been set up to study the corrosion of materials in nitric acid, in liquid as well as vapour phase.

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The management of high level liquid waste generated in reprocessing is one of the crucial aspects of nuclear fuel cycle which ultimately influence the acceptance of nuclear energy. Development

and

demonstration of the flow sheets based on CMPO or substituted diamides as the extractants, for the recovery of minor actinides from high level waste have also been taken up for incorporation in the reprocessing scheme for PFBR fuel for the minor actinide recovery. 5.4. Long term R&D issues related to FBR fuel cycle

5.4.1. Development of advanced clad and wrapper materials. In the light of international alloy development efforts, and in-reactor experiences, D9I is recommended for future cores of PFBR, for achieving higher burn-ups. Though the basic composition will remain same as D9, stringent control of minor alloying elements like phosphorous (0.025 - 0.04wt%), silicon (0.7 -0.8%), carbon (0.04 - 0.05%) and boron (0.004

0.006%) have been recommended, that have a strong influence on the void swelling resistance of

the alloy. Ti/C ratio of about 4 to 5 will be ensured, with the titanium concentration not exceeding 0.25wt%. Along with D9I, advanced ferritic steels also are under development to realize 200 GWd/t target burn-up of FBR fuel. This would result in significant economy in fuel cycle cost of FBRs and make them competitive and more environment friendly. Modified chrome-moly steels are promising candidate materials for wrapper applications in future FBRs. However, the increase in ductile to brittle transition temperature (DBTT) due to irradiation, is a cause of concern for ferritic steels. Detailed alloy design, microstructural mapping and modeling studies to improve the void swelling resistance and DBTT behaviour is under way for both D9I and modified 9Cr-1Mo steel, with efforts for indigenous development of these alloys. Extensive evaluation of creep, fatigue, creep-fatigue and toughness parameters of both wrought and weld metals of these advanced alloys have been undertaken for aiding the alloy development process. 5.4.2 Development of pyrochemical reprocessing technology. Molten salt electrorefining process, a pyrochemical method, developed by Argonne National laboratory, USA, is ideally suited for reprocessing the metallic fuels and hence a well focused work programme for the development of this technology is being pursued. This will culminate in an industrial scale plant based the molten salt electrorefining process in about 20 years time. The unit operations of the molten salt electrorefining process have been studied using a laboratory scale facility which is being operated for the last 12 years (Prabhakara Reddy et al. 1997).

For generating engineering scale experience, a pilot plant scale facility is being set up.

Development of remote fabrication technology is also being pursued to enable refabrication of the fuel pins using the fuel materials from pyrochemical reprocessing which has low decontamination factors. 5.4.3. Waste management The emphasis of R&D on fast reactor fuel reprocessing has been to develop flow sheets which result in waste minimization. It is necessary to recover minor actinides and long lived fission products from the waste produced in fast reactor fuel cycle. In view of higher amounts of noble metal fission yields from the fast fission of Pu, it is also necessary to evolve methods for recovering noble metal fission products such as palladium and rhodium. Development of matrices other than glass have to be

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developed for the immobilisation of the high level waste constituents for long term storage and disposal. Pyrochemical reprocessing of metallic fuels will generate different kinds of waste streams, namely, salt waste and metallic waste, demanding a different approach. Development of ceramic waste matrices for the immobilisation of waste from aqueous as well as pyrochemical processing routes is also being pursued. 6. CONCLUSION In India, technologies have been developed to master both the oxide fuel and carbide fuel based fast reactor fuel cycles. With the experience on these fuel cycles, we are poised to develop the technology for the metal fuel cycle in about twenty years which will help us to meet the high growth rate envisaged for the next fifty years. NOMENCLATURE Acronyms CMPO- Octylphenyl (N,N) disiobutylcarbamoyl methylene

IGCAR- Indira Gandhi Centre for Atomic

phosphene oxide

Research

DBTT

LMC- Lead Mini Cell

-

Ductile to Brittle Transition Temperature

DFRP- Demonstration Fast Reactor Fuel Reprocessing Plant

MOX-

FBR-Fast breeder Reactor

MSM-Master slave manipulators

FBTR- Fast Breeder Test Reactor

PHWR- Pressurised Heavy Water Reactor

FFRP- Fast reactor Fuel Reprocessing Plant

PIE-Post Irradiation Examination

HPLC

PFBR- 500 MW(e) FBR Project

High Performance Liquid Chromatography

HRTEM- High Resolution Transmission Electron Microscope

Mixed oxide

FFRP- Prototype Fast reactor Fuel Reprocessing Plant

IAEA-International Atomic Energy Agency

TBP- Tributyl phosphate

ACKNOWLEDGEMENT The authors gratefully acknowledge contributions of a large number of colleagues, participants from industries, a number of academicians and students, who have worked with great confidence and passion with the mature scientists and engineers of the Department of Atomic Energy India, to achieve remarkable success. The authors are grateful to Dr. K. Nagarajan for useful discussions and his help in preparing the manuscript.

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REFERENCES Balasubramanian R., Nalini S., Sai Baba M., Darwin Albert Raj D., and Viswanathan R.(1999), Mass spectrometric determination of burn-up of a spent fuel pin from the mixed-carbide fuelled fast breeder test reactor, Nuclear and Radiochemistry Symposium 1999, p. 389. Mumbai, 19-22 Jan. Balasubramanian R., Swaminathan K, Viswanathan R. and Venkatasubramani C.R. (2003), Mass spectrometric determination of burn-up of FBTR fuel, p.349, Nuclear and Radiochemistry Symposium 2003, p.349, Mumbai, 10-13, Feb. Baldev Raj and Kamachi Mudali U. (2001), Materials challenges in the back end of the fuel cycle, 12th Annual Conference of Indian Nuclear Society, p.93,. Indore, Oct. 10-12 Baldev Raj, Kamachi Mudali U., Jayakumar T., Kasiviswanathan K.V. and Natarajan R. (2000), Meeting the challenges related to material issues in chemical industries, Sadhana, 25,519. Cahalan J.E, Kramer J.M., Marchaterre J.F, Mueller C.J., Pedersen D.R. , Sevy R.H. , Wade D.C. and Wei T.Y.C. (1988), Integral fast reactor safety features,, International Topical Meeting on Safety of Next Generation Power Reactors, p. 103, Seattle, Washington, May. Kamachi Mudali U., Dayal R.K and Gnanamoorthy J.B. (1993), Corrosion studies on materials of construction for spent nuclear fuel reprocessing, J. Nucl. Mater, 73-82, 203. Koganti S.B., Sreedharan V. and Balasubramanian G R. (1986), Development of Low Capacity Centrifugal Extractors for Small Scale Reprocessing Facilities, International Solvent Extraction Conference (ISEC- 86), Vol. 1, p.I-413, Munchen, Germany, 11-16, Sep. Koganti S.B., Periasamy K, Rajagopal C.V., Sreedharan V and Balasubramanian. G R. (1994), Experience with Centrifugal Extractors, 4th International Confeence on Nuclear Fuel Reprocessing and Waste Management (RECOD '94), Vol. 1, London, United Kingdom, 24-28, Apr. Nair M.K.T., Singh R.K., Venugopal A.K., Bajpai DD., Singh R.R. Thomal M. and Gurba P.B. (1990), Application of electrochemical processes in Fuel Reprocessing, Bhabha Atomic Research Centre Report BARC/I-994. Shekhar Kumar and Koganti S.B. (1996), Dynamic simulation of high plutonium flow sheets proposed for reprocessing FBR fuels, International Solvent Extraction Conference(ISEC- 96), p. 1013, Melbourne, Australia, 21-25, Mar. Sivaraman N., Subramaniam S., Srinivasan T.G. and.Vasudeva Rao P.R. (2002), Burn-up measurements on nuclear reactor fuels using High Performance Liquid Chromatography, J. Radioanal.Nucl Chemistry 35, 253 Vaidya V.N., Kamat R.V., Joshi J.K, Iyer V.S., Pillai K.T. and Sood, D.D.(1981), Preparation of (U,Pu)O2 microspheres by sol-gel method,, Nuclear and Radiochemistry Symposium (NUCAR 1981), p. 549, Varanasi, India, 3-7 Nov.