Mixed Uranium-plutonium Research Articles

Advanced x-ray spectrometric techniques for characterization of nuclear materials: An overview of recent laboratory activities

Advancements in x-ray spectrometric techniques at different stages have made this technique suitable for characterization of nuclear materials with respect to trace/major element determinations and compositional uniformity studies. The two important features of total reflection x-ray fluorescence spectrometry: 1) requirement of very small amount of sample in ng level 2) multielement analytical capability, in addition to other features, make this technique very much suitable to nuclear materials characterization as most of the nuclear materials are radioactive and the radioactive waste generated and radiation hazards to the operator are minimum when such low amount of sample is used. Similarly advanced features of energy dispersive x-ray fluorescence e.g. better geometry for high flux, reduction in background due to application of radiation filters have made the measurements of samples sealed inside thin alkathene/PVC covers possible with good sensitivity. This approach avoids putting the instrument inside a glove box for measuring radioactive samples and makes the operation/maintenance of the instrument and analysis of the samples possible in easy and fast manner. This approach has been used for major element determinations in mixed uranium–plutonium samples. Similarly μ-XRF with brilliant and micro-focused excitation sources can be used for compositional uniformity study of reactor fuel pellets. A μ-XRF study using synchrotron light source has been made to assess the compositional uniformity of mixed uranium–thorium oxide pellets produced by different processes. This approach is simple as it does not involve any sample preparation and is non-destructive. A brief summary of such activities carried out in our laboratory in past as well as ongoing and planned for the future have been discussed in the present manuscript.

Melting point determination of uranium nitride and uranium plutonium nitride: A laser heating study

Understanding of the behaviour of nuclear material in extreme conditions is essential for the analyses of the operation limits of nuclear fuels, and prediction of possible nuclear reaction accidents. In this context, the high temperature behaviour of uranium nitride and mixed uranium–plutonium nitrides has been studied in the present work by laser heating under controlled atmosphere coupled with fast multi-wavelength pyrometry. Such an approach has allowed performing a thermal arrest analysis and establishing the solid–liquid phase boundaries in the investigated compositions, whereby non-congruent vaporisation was avoided by setting a suitable nitrogen overpressure. In addition, the normal spectral emissivities of the current samples were determined by radiance spectroscopy. Besides revealing a slightly more metallic optical behaviour in plutonium-containing compositions, this latter characterisation led to the determination of the real melting/solidification temperatures of the investigated nitrides. It is confirmed that UN melts congruently at (3120±30) K in a nitrogen pressure of 0.25MPa (2.5bar). The melting/solidification temperatures decrease in plutonium containing samples, reaching (3045±25) K for x(PuN)=0.2, a composition of interest for potential applications of this material as a nuclear fuel. Besides their fundamental importance, the current results are useful for a deeper understanding of the nitride fuel behaviour under accidental conditions, whereby uncontrolled thermal excursions might occur in the nuclear reactor core.

Controlling the oxygen potential to improve the densification and the solid solution formation of uranium–plutonium mixed oxides

Diffusion mechanisms occurring during the sintering of oxide ceramics are affected by the oxygen content of the atmosphere, as it imposes the nature and the concentration of structural defects in the material. Thus, the oxygen partial pressure, p(O2), of the sintering gas has to be precisely controlled, otherwise a large dispersion in various parameters, critical for the manufacturing of ceramics such as nuclear oxides fuels, is likely to occur.In the present work, the densification behaviour and the solid solution formation of a mixed uranium–plutonium oxide (MOX) were investigated. The initial mixture, composed of 70% UO2+30% PuO2, was studied at p(O2) ranging from 10−15 to 10−4 atm up to 1873K both with dilatometry and in situ high temperature X-ray diffraction.This study has shown that the initial oxides UO2+x and PuO2−x first densify during heating and then the solid solution formation starts at about 200K higher. The densification and the formation of the solid solution both occur at a lower temperature when p(O2) increases.Based on this result, it is possible to better define the sintering atmosphere, eventually leading to optimized parameters such as density, oxygen stoichiometry and cations homogenization of nuclear ceramics and of a wide range of industrial ceramic materials.

Sustainability issues of plutonium recycling in light water reactors: Code evaluations up to 2050

Plutonium recycling in light water reactors is a viable and mature technology capable of improving several indicators of great importance in the evaluation of the sustainability of nuclear energy development in the near- and long-term such as the shortage of natural uranium resources, the increase of spent fuel inventories and the proliferation risks of high-level radioactive waste. This paper, after a brief review of the status of plutonium recycling, presents the results of a scenario analysis carried out in the hypothesis that the share of nuclear fleet loaded with mixed uranium–plutonium oxide fuel (MOX) is kept constant up to the middle of the century. Beside mentioned indicators, the paper discusses the needs for complex and costly fuel cycle infrastructures required for the reprocessing of spent nuclear fuel and the fabrication of MOX. Assuming that the deployment of fast reactors occurs beyond the middle of the century, the article focuses on the comparison of an open fuel cycle strategy with a closed fuel cycle strategy where plutonium is recycled.Presented calculations, according to moderate and high projections of nuclear energy development, confirm that plutonium recycling, although deployed to a limited extent, could be beneficial in reducing the stockpiles of nuclear spent fuel and in reducing the risks of proliferation due to the amount of fissile plutonium in the system. The improvement found in the consumption of natural uranium resources was limited promoting to this purpose the deployment of next-generation fast reactors. If a moderate development of nuclear energy is confirmed, the current capacity for reprocessing and MOX fuel fabrication could be sufficient to cope with the foreseen demand, on the contrary, significant investments could be necessary in case of steep increase of installed nuclear energy. Calculations were performed by means of the DESAE code (Dynamic Energy System – Atomic Energy), a tool developed within the IAEA INPRO project.

Characterization of polymeric films subjected to lithium ion beam irradiation

Two different polymeric materials that are candidate materials for use as binders for mixed uranium–plutonium oxide nuclear fuel pellets were subjected to Li ion beam irradiation, in order to simulate intense alpha irradiation. The materials (a polyethylene glycol 8000 and a microcrystalline wax) were then analyzed using a combination of mass spectrometry (MS) approaches and X-ray photoelectron spectroscopy (XPS). Samples of the irradiated PEG materials were dissolved in H2O and then analyzed using electrospray ionization-MS, which showed the formation of a series of small oligomers in addition to intact large PEG oligomers. The small oligomers were likely formed by radiation-induced homolytic scissions of the C–O and C–C bonds, which furnish radical intermediates that react by radical recombination with H and OH. Surface analysis using SIMS revealed a heterogeneous surface that contained not only PEG-derived polymers, but also hydrocarbon-based entities that are likely surface contaminants. XPS of the irradiated PEG samples indicated the emergence of different carbon species, with peak shifts suggesting the presence of sp2 carbon atoms. Analysis of the paraffinic film using XPS showed the emergence of oxygen on the surface of the sample, and also a broadening and shifting of the C1s peak, demonstrating a change in the chemistry on the surface. The paraffinic film did not dissolve in either H2O or a H2O–methanol solution, and hence the bulk of the material could not be analyzed using electrospray. However a series of oligomers was leached from the bulk material that produced ion series in the ESI-MS analyses that were identified octylphenyl ethoxylate oligomers. Upon Li ion bombardment, these shifted to a lower average molecular weight, but more importantly showed the emergence of three new ion series that are being formed as a result of radiation damage. Surface analysis of the paraffinic polymers using SIMS produced spectra that were wholly dominated by hydrocarbon ion series, and no difference was observed between unirradiated and irradiated samples. The studies demonstrate that for the PEG-based polymers, direct evidence for radiolytic scission can be observed using ESI-MS, and suggests that both radiolytic pathways and efficiencies as a function of dose should be measurable by calibrating instrument response to the small oligomeric degradation products.

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported.For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6Li, 7Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra.The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such instances can often be related to nuclear data for specific non-fissile elements, such as C, Fe, or Gd. Indications are that the intermediate and mixed spectrum cases are less well described.The results for the shielding benchmarks are generally good, with very similar results for the three libraries in the majority of cases. Nevertheless there are, in certain cases, strong deviations between calculated and benchmark values, such as for Co and Mg. Also, the results show discrepancies at certain energies or angles for e.g. C, N, O, Mo, and W.The functionality of MCNP6 to calculate the effective delayed neutron fraction yields very good results for all three libraries.

A physical model for evaluating uranium nitride specific heat

Nitride fuel is one of perspective materials for the nuclear industry. But unlike the oxide and carbide uranium and mixed uranium–plutonium fuel, the nitride fuel is less studied. The present article is devoted to the development of a model for calculating UN specific heat on the basis of phonon spectrum data within the solid state theory.

GLOBAL NUCLEAR ENERGY ARCHITECTURE: A KEY TO ENERGY SECURITY

Click to increase image sizeClick to decrease image size Notes 1. A.P. Aleksandrov, and N.N. Ponomarev-Stepnoi, Nuclear Energy and Technical Progress: 20 Years of Nuclear Energy (Moscow: Atomizdat, 1974); N.N. Ponomarev-Stepnoi, “Nuclear-hydrogen Energy,” Atomnaya Energiya 96, No. 6 (2004), pp. 411–426. 2. Statement by Rosatom Director-General Sergey Kiriyenko at a plenary meeting of the 55th session of the IAEA General Conference, 2011. 3. A.Y. Gagarinsky, V.V. Ignatyev, N.N. Ponomarev-Stepnoi, S.A. Subbotin, and V.F. Tsibulsky, “Role of Nuclear Energy in the Structure of Global Energy Production in the 21st Century,” Atomnaya Energiya 99, No. 5 (2005), pp. 323–335; N.N. Ponomarev-Stepnoi, “Role of Nuclear and Nuclear-Hydrogen Energy in the Structure of Global Energy Production in the 21st Century,” Vesti v elektroenergetike No. 1 (2005), pp. 7–12; N.N. Ponomarev-Stepnoi, “Nuclear Energy and its Future,” Scientific Session of the General Assembly of the Russian Academy of Sciences “Russian Energy Sector: Problems and Outlook,” 2005, December 20–21; “Nuclear Energy: Expert Assessments of Development in 1949–2008,” Kurchatov Institute, Moscow, 2008. 4. N.N. Ponomarev-Stepnoi, and V.F. Tsibulskiy, “Assessing the Effectiveness of Using Mixed Uranium-Plutonium Fuel in VVER-type Reactors,” Atomnaya Energiya 103, No. 5 (November 2007), pp. 275–277. 5. V.N. Ponomarev-Stepnoi, and V.F. Tsibulskiy, “Choosing the Fast-neutron Output Range for the Russian Nuclear Energy Sector,” Atomnaya Energiya 103, No. 2 (August 2007), pp. 83–89. 6. V.N. Bobolovich, N.N. Ponomarev-Stepnoi, S.M. Feinberg, and others, “Fast Breeder Reactors with Helium Coolant,” Report at the 11th Symposium of Comecon countries on fast reactors, Obninsk, 1973. 7. N.N. Ponomarev-Stepnoy, “Feasibility of Putin's Initiative: Ways to Ensure Sustainable Energy Supply,” Yaderny Kontrol No. 2 (2001), p. 30. 8. V. Sukhoruchkin, N. Ponomarev-Stepnoi, A. Rumjantsev, V. Shmelev, J. Toevs, and R. Wagner Jr., “A New Look at Metrics for Proliferation Resistance,” Report presented to the INMM 41th Annual Meeting, 2000, July 16–20, New Orleans, Louisiana, USA. 9. V.M. Murogov, N.N. Ponomarev-Stepnoi, V.V. Artisyuk, A.G. Kalandarashvili, Y.A. Korovin, and A.I. Voropayev, “International Initiative of IAEA: From Innovative Nuclear Technologies to the Global Role of Nuclear Education,” Atom for Peace: An International Journal 1, No. 2/3 (2006). 10. A.Y. Gagarinsky, op. cit. 11. V.B. Ivanov, V.S. Kagramanyan, A.K. Polushkin, N.N. Ponomarev-Stepnoi, A.V. Ubeev, and A.N. Chebeskov, “Nuclear Renaissance: Russian in Global Context,” Security Index No. 3 (85, 2008), pp. 73–82.

Une curieuse carcinose péritonéale

Nitrides have been proposed to be a suitable material for fast neutronic systems from beginning of the development of nuclear fuel. Starting with the production of uranium nitride and sesquinitrides up to mixed plutonium uranium nitrides, today's developments are inert nitride matrix materials to burn plutonium or to transmute long-lived actinides in accelerator-driven sub-critical systems (ADS) or fast reactors (FR). Several authors proposed zirconium nitride as possible inert matrix material for this reason. Mixed zirconium nitrides can be fabricated by carbothermal nitridation of the oxides in a narrow temperature window. Obtaining high quality material with low carbon and oxygen content is still the major challenge. Producing mixed nitride fuels by special shaping methods, as for example direct coagulation casting or freeze drying, in comparison to conventional powder compaction enables to use this material in special shapes to optimize burn up.The history of nitride fuels at PSI will be shown up to today's CONFIRM project, dealing with plutonium zirconium nitride fuels.

Safety properties of bn-1200 with mixed oxide and nitride fuel: a comparison

Along with mixed oxide fuel, the possibility of using in BN-1200 dense nitride fuel, making it possible to attain higher technical-economic performance, is also studied. However, safety analysis will determine the choice of fuel type. In this connection, it is important to perform a comparative analysis of the inherent safety properties for variants of the BN-1200 core with mixed uranium-plutonium and nitride fuel for the most serious unanticipated loss-of-power accident with failure of all emergency protection organs of the reactor simultaneously. A two-dimensional version of the COREMELT computer code was used in the calculations. The computational analysis showed that the inherent safety of BN-1200 is much greater with nitride than with mixed uranium-plutonium fuel.

The Chemistry of (U,Pu)O2 Dissolution in Nitric Acid

The focus of this paper is the chemistry of mixed uranium plutonium oxide (MOx,) in nitric acid. An overview of dissolution chemistry is discussed by comparing the differences in the dissolution characteristics of uranium and plutonium oxides. An overview of batch dissolution experiments, studying the dissolution chemistry of high surface area MOx powders and low surface area MOx pellets with reference to the effects of nitrous acid, nitric acid and temperature are described. The results are discussed in terms of the autocatalytic mechanism and mass transfer limited dissolution.

On the melting behaviour of uranium/plutonium mixed dioxides with high-Pu content: A laser heating study

The melting behaviour of mixed uranium–plutonium dioxides (MOX) has been investigated by laser heating under controlled atmosphere in the PuO 2-rich composition range (with amount-of-substance fraction x(PuO 2) ⩾75%). The observed solidus/liquidus points are in agreement with the newly measured melting point of pure plutonium dioxide (3017 K). They suggest the existence of a minimum freezing temperature at a composition x(PuO 2) between 50% and 80%, in contrast with earlier research carried out with traditional furnace heating methods. The current results have been obtained under optimised experimental conditions aimed at maintaining the integrity and original composition of the samples throughout the laser heating cycles. With this goal in mind, experiments have been carried out under controlled gas pressure (pressurised air or argon) and short time duration (<0.25 s). A critical discussion of the present results highlights the fact that the formation of the gas phase has to be taken into account in the study of the high-Pu MOX behaviour at high temperature. The experimental results obtained thus correspond to slightly hypo-stoichiometric (U, Pu)O 2− x compositions in equilibrium with the gas phase.

FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress–strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium–uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal–hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

Accuracy of criticality calculations made using current nuclear data libraries for MOX fuelled fast critical assemblies

Criticality calculations have been made for a set of ten mixed plutonium–uranium oxide (MOX) fuelled fast critical assemblies using the current nuclear data libraries, JEFF-3.1, JEFF-3.1.1, JENDL-3.3 and ENDF/B-VII.0. The results obtained using the different libraries are compared and conclusions drawn concerning the accuracy of criticality calculations made for MOX fuelled fast reactors.

Recommendations for calculating the rate of irradiation-induced creep of oxide fuel when analyzing fuel-element serviceability

Recommendations for calculating the characteristics of irradiation-induced creep of mixed uraniumplutonium oxide fuel and uranium dioxide for programs analyzing fuel-element serviceability are developed on the basis a physical model of deformation. The deformation processes in the model include the mechanism of modular interaction of radiation defects with dislocations and radiation diffusion. It is shown on the basis of an analysis of the dynamics of radiation point defects that the rate of irradiation-induced creep is determined by the sum of thermal and athermal components. Thermal creep makes a large contribution at temperatures above one-half the melting temperature. The model takes account of the effect of temperature, stress, fission density, and fuel burnup and density on the rate of irradiation-induced creep. The use of physical models to obtain computational relations made it possible to improve by approximately a factor of 10 the agreement between calculations and experimental data as compared with empirical relations previously used to describe the characteristics of creep.

Development and benchmarking of a mechanical model for GFR plate-type fuel

The reference fuel design currently being considered within the Generation-IV Gas-cooled Fast Reactor (GFR) project is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium–uranium carbide, while the matrix material is silicon carbide. The present paper describes the mechanical part of a thermal–mechanical model being developed for studying the transient behavior of this highly heterogeneous fuel type. Benchmarking has been carried out against detailed finite-elements modeling (FEM). The resultant thermal–mechanical model can provide reliable fuel and cladding (matrix) stress/strain conditions to evaluate temperatures and neutronic feedbacks. As such, it has been integrated into PSI’s coupled code system “FAST”, which aims at the comprehensive safety analysis of advanced fast reactor systems. The detailed FEM analysis of the GFR fuel has been useful not only for benchmarking the new model, but also for obtaining an in-depth understanding of fuel stress/strain characteristics, which cannot be reproduced with simplified models. Thereby, the range of applicability of the new model has clearly been defined. In particular, the 3D FEM analysis has revealed a concentration of stresses at the pellet corners during pellet/matrix contact, which could lead to fuel element failure. This effect is found to be mitigated considerably, if the fuel pellets are shaped in a manner which enhances the contact area.

Compatibility of fuel-element cladding materials with gallium-containing uranium and uranium–plutonium oxide fuel

Powders of uranium oxide powder and mixed fuel containing 5 and 20 mass % plutonium and 0.4 and 5 mass % gallium were prepared by coprecipitation from nitrate solutions. Pelleted samples for testing were made by cold pressing and sintering. The compatibility of uraniwn oxide fuel and mixed uranium-plutonium fuel, containing 0.4 and 5 mass % gallium, with the zirconium alloy E- 110 at 400 and 500°C and ChS-68 corrosion-resistant steel at 650 and 750°C over periods of 1000, 2000, and 3000 h was investigated. Metallographic and x-ray spectral microprobe analyses of diffusion samples established that there was no interaction and penetration of gallium into the zirconium alloy and steel. In addition, the diffusion coefficient of metallic gallium in zirconium alloy and the distribution of the elements on interaction of gallium with ChS-68 steel were evaluated.

A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse ‘A VVER-1000 LEU and MOX Assembly Computational Benchmark’ and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group ‘JEFF31GX’ cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor ( k ∞ ). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium–Gadolinium (UGD) pin, fission rate distributions in UGD, UO 2 and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.

Direct carbothermic reduction of actinide oxalates: Example of Nd(III) oxalate–carbon mixtures conversion

Mixed uranium–plutonium monocarbide was obtained by carbothermic reduction of mixture of UO 2, PuO 2 and C at around 1500 °C under vacuum or argon for 4 h followed by sintering at around 1650 °C under mixture of Ar + 8%H 2 for 10 h. Moreover, PuO 2 was synthesized by oxalic precipitation followed by calcination at around 700 °C (PUREX process). This work deals with a process simplification, i.e. fabrication of carbide by calcination of actinide oxalate, precipitated in presence of carbon, at 1600 °C for 13 h under argon. In this preliminary study, neodymium was used as a surrogate of actinides. No carbon influence was noticed during calcination of neodymium oxalate into oxide. Carbon content was the same before and after calcination making possible conversion to carbide directly from mixture of neodymium oxalate and C.

Experience on mixed carbide fuels with high ‘Pu’ content for Indian fast breeder reactor – An overview

Hyperstoichiometric Plutonium rich (70%) mixed Uranium Plutonium carbide fuel was proposed as the driver fuel for Indian Fast Breeder Test Reactor (FBTR) after some deliberations. The issues of mixed carbide fuel and of plutonium rich (70%) carbide fuel, in particular, are: presence of ‘O’ impurity, clad carburization, lower solidus temperature, lower thermal conductivity and pyrophorocity. The paper highlights the extensive research and development carried out to address the above issues. The modification incorporated in the different process steps of fabrication flow sheet is also presented. The performance of the fuel has been assessed by periodic Post Irradiation Examination (PIE) of the fuel sub-assemblies for extension of life. Till date the fuel has seen a burn up of 155 Gwd/t without any pin failure.