Papers by Keyword: Fast Reactor

Abstract: The feasibility of employing the indigenously developed Ferroboron alloy as an alternate neutron shield material in combination with 9Cr-based ferritic steel clad in future Indian Fast Breeder Reactors (FBR), has been investigated from a metallurgical perspective. In this regard, extensive studies have been undertaken to estimate quantitatively the nature of interaction between Ferroboron and P91-ferritic steel at high temperatures. It is found that in the temperature range 550 to 600°C, 9Cr-based ferritic steel is fully compatible with Ferroboron. However, at higher temperatures, Feroboron interacts with ferritic steel; but the maximum estimated loss of clad thickness is restricted to about 250 μm for 60 years of service.

Abstract: SOCRAT-BN code is developed for the analysis of design and beyond design basis accidents at sodium cooled fast reactors. To simulate the behavior of the coolant in the reactor core heat transfer and friction in rod bundle geometry are required to consider. The article describes the validation of the code SOCRAT-BN on the experiment with fuel rod imitators in the triangular geometry with wire-wound taking into account experiment and some code model uncertainties.

Abstract: Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. An application of the hafnium hydride has been investigated as a neutron absorber in the Fast Breeder Reactors (FBRs). Fast neutrons are efficiently moderated by hydrogen in Hf hydrides and are absorbed by Hf. Since three isotopes of Hf have large cross sections, increase in the life of control rod is considered by Hf hydride. Results of design study of the core with Hf hydride control rods shows that the long lived hafnium hydride control rod is feasible in the large sodium-cooled FBR. Results of irradiation test conducted in BOR-60 has demonstrated the integrity of the capsules during irradiation. Na bonded capsule has an advantage in confinement effect of hydrogen compared with He bonded one. An application of hydride technique to transmutation target of MA was also discussed. MA hydride target is able to enhance the transmutation rate in FBR.

Abstract: Fuel cycle cost of sodium cooled fast reactors (SFRs) is strongly dependent on the in-reactor performance of core structural materials, i.e., cladding and wrapper tube materials of the fuel subassembly, which are subjected to intense neutron irradiation during service, leading to unique materials problems like void swelling, irradiation creep and helium embrittlement. In order to increase the burnup of the fuel and thereby reduce the fuel cycle cost, it is necessary to employ materials which have high resistance to void swelling as well as better high temperature mechanical properties. The Indian fast reactor program began with the commissioning of the 40 MW_t Fast Breeder Test Reactor (FBTR). The core structural material of FBTR is 20% cold worked 316 austenitic stainless steel (SS). For the 5000 Me_t Prototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction at Kalpakkam, 20% cold-worked alloy D9 (14Cr-15Ni-Ti SS) has been selected as the cladding and wrapper tube material for the initial core. The target burnup of the fuel is 100 GWd/t. Advanced austenitic stainless steel and oxide dispersion strengthened steels are being developed for achieving fuel burnup higher than 100 GWd/t. An advanced alloy D9 containing controlled amounts of titanium, silicon and phosphorous has been developed. This alloy named as IFAC-1 (Indian Fast Reactor advanced Clad-1) SS is aimed at thermal creep properties comparable to that of alloy D9, and superior void swelling resistance upto a target burn-up of about 150 GWd/t. The nominal chemical composition of IFAC-1 SS is 14Cr-15Ni-.25Ti-.75Si-.04P. The chemical composition has been optimized after extensive evaluation of the tensile, creep and microstructural stability of fifteen laboratory heats with different amounts of titanium, silicon and phosphorous. Void swelling behavior was studied using ion irradiation. IFAC-1 SS contains higher levels of low melting eutectic phase forming elements such as phosphorous, and so is susceptible to solidification cracking. Extensive pulsed TIG welding trials have been carried out on IFAC-1 SS/316LN SS weld joints with varied weld parameters to find out the feasibility of obtaining solidification crack-free welds and the optimum welding parameters have been established. This paper gives an overview of the development of this advanced core structural material for SFRs.

Abstract: Fuels for future fast reactors will not only produce energy, but they must also actively contribute to the minimisation of long lived wastes produced by these, and other reactor systems. The fuels must incorporate minor actinides (MA = Np, Am, Cm) for neutron transmutation into short lived isotopes. Within Europe oxide fuels are favoured. Transmutation can be considered in homogeneous or heterogeneous reactor recycle modes (i.e. in fuels or targets, respectively). Fabrication of such fuels can be made by advanced liquid processing methods, enabling property determination and screening irradiation experiments. This paper will describe these fabrication processes, and discuss properties and fuel irradiation experiments made to date. Both fertile and inert matrix fuel types are considered.

Abstract: Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. Fast neutrons are efficiently moderated by hydrogen in metal hydrides. Metal hydrides have been studied for their potential application as nuclear materials in fast reactors (FRs). Two types of the utilizations of metal hydride in FRs are discussed in this paper. One is the utilization for transmutation target of long-lived nuclear wastes. Hydride fuel containing 237Np, 241Am and 243Am has been studied as a candidate transmutation target to reduce the radioactivity of long-lived nuclides included in reprocessed nuclear wastes. An application of the hafnium hydride has been investigated as neutron absorber in FRs. The core design has been performed to examine its characteristics and to evaluate the cost reduction effect. Demonstration of fabrication of hydride pin has been done with hydride pellets and stainless steel cladding. Coating technique of inner cladding surface has been also developed to reduce the permeation of hydrogen through stainless steel cladding. Physical and chemical properties of the pellet have been measured for designing the hydride pin. The integrity of the pellets at high temperature has been tested and their compatibility with sodium has also been tested. Irradiation test of hydrides has been performed in the fast experimental reactor, JOYO, at Japan Atomic Energy Association (JAEA).