Moriyasu Tokiwai

Development of trail, a simulation code for the molten salt electrorefining of spent nuclear fuel

Abstract A simulation code for the molten salt electrorefining of spent metallic nuclear fuel from the Integral Fast Reactor has been developed. This code (named trail ) employs diffusion layer theory in the vicinity of the electrodes. Model parameters such as the diffusion layer thickness were determined from polarization data measured with uranium at different concentrations in the molten salt electrolyte and liquid cadmium anode of an electrorefining cell. Calculations were made to verify the code with experimental data for various operational modes. Good agreement with the data was obtained. It was also found that this code can provide useful information to aid in understanding the electrotransport process within the electrorefiner.

Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y2O3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

Continuous TEM observation of cavities in nickel under helium irradiation

Abstract Helium irradiation experiments at 550°C on pure nickel specimens and annealing experiments at 650°C on the specimens were performed in a TEM. The results of this study show that the shape of the cavities is spheroidal when their growth rate is small, while it changes to polyhedral when their growth rate becomes larger. The polyhedral cavities are hexahedra bounded by (100) planes. It is also found that the polyhedral cavities spheroidize during the annealing. It is inferred that there exists preferential directions for vacancy inflow to an overpressurized spherical cavity, and this could assist the formation of the polyhedral cavities in nickel.

Hardness of 12Cr8Mo ferritic steels irradiated by Ni ions

Abstract 12Cr Mo and 12Cr 8Mo 0.1Y2O3 ferritic steels were irradiated with 4-MeV Ni+ ions up to 300 dpa at 525°C. Microstructural evolution was examined by transmission electron microscopy (TEM) and mechanical properties were evaluated with a depth-sensing ultra-low load indentation hardness tester at room temperature with a maximum load of 1 gf. Effects of aging at 650°C for 115 h and heat treatment at 525°C for 50 h were also investigated. TEM observations reveal that these steels exhibit no void swelling in the present irradiation condition. Aging and heat treatment induces precipitation of Laves phase and ion-irradiation enhances precipitation. The induction and enhancement of precipitation strengthened the specimens.

Effects of Cyclic Tensile Loading on Stress Corrosion Cracking Susceptibility for Sensitized Type 304 Stainless Steel in 290 C High Purity Water

Abstract The effects of load waveform on intergranular stress corrosion cracking (IGSCC) susceptibility have been examined for sensitized Type 304 stainless steels in a 290 C high purity water loop. Concerning the strain rate in the trapezoidal stress waveform, it was found that IGSCC susceptibility was higher for smaller values of the strain rate. It was also shown that IGSCC susceptibility became higher when the holding time at the upper stress was prolonged, and when the upper stress was high. The occurrence of IGSCC for sensitized Type 304 stainless steel became easy due to the application of cyclic tensile stress in 290 C high purity water.

The amount of chlorine contamination for prevention of stress corrosion cracking in sensitized type 304 stainless steel

Abstract Stress corrosion cracking (S.C.C.) of sensitized Type 304 stainless steel was investigated under conditions of chloride contamination with varying tensile stress level and relative humidity as an environmental condition. The results obtained were summarized and the allowable chlorine concentration of the contaminant for the prevention of S.C.C. was determined. The effect of relative humidity on S.C.C. was also discussed.

The improvement of irradiation-enhanced copper embrittlement in FeCu alloys

Abstract The effect of alloying elements on neutron irradiated FeCu alloys has been investigated in order to obtain the fundamental information on the irradiation-enhanced copper embrittlement for power reactor vessel steels. The mechanism of copper-induced irradiation embrittlement in the copper-containing iron alloys was proved to be due to both the interaction of copper atoms with irradiation-produced complex defects within grains, and the preferred grain boundary segregation of copper atoms existing near grain boundaries. The former effect causes the increase of yield strength, and the latter results in the ductility loss and grain boundary crackings. The addition of titanium up to 0.4 wt% to the Fe-0.1 wt% Cu alloy was found to be extremely effective in the improvement of both the irradiation-induced ductility loss and strength. Aluminum and silicon were not as effective as titanium.

Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel

The tolerance capability against ATWS for the FBR core with metallic fuel can be improved by employing a fuel with high thermal conductivity (HTC fuel) instead of the conventional metallic fuel, UPu-Zr. To investigate the self-controllability for the HTC-fueled core with U-Pu-Al alloy fuel, having one order of magnitude higher thermal conductivity than that of the U-Pu-Zr, the core employing the U-Pu-Al fuel was evaluated against ULOF and UTOP. Based on the systematic calculation, it was found that the larger temperature margin between the steady state and ULOF/UTOP conditions caused the excellent tolerance capability against ULOF and UTOP for the HTC-fueled core compared with that for the Zralloy-fueled core. Also, the conditions of the core reactivity coefficients required for neither fuel melting nor coolant boiling were investigated by using a “self-controllability map” consisting of effective fuel and coolant reactivities. As a result, the self-controllable region was found to be expanded especially for UTOP in the case of the HTC-fueled core.

Development of Metallic Uranium Recovery Technology from Uranium Oxide by Li Reduction and Electrorefining

Abstract The purpose of the study is to develop technology for pre-treatment of oxide fuel reprocessing through pyroprocess. In the pre-treatment process, it is necessary to reduce actinide oxide to metallic form. This paper outlines some experimental results of uranium oxide reduction and recovery of refined metallic uranium in electrorefining. Both uranium oxide granules and pellets were used for the experiments. Uranium oxide granules was completely reduced by lithium in several hours at 650°C. Reduced uranium pellets by about 70% provided a simulation of partial reduction for the process flow design. Almost all adherent residues of Li and Li2O were successfully washed out with fresh LiCl salt. During electrorefining, metallic uranium deposited on the iron cathode as expected. The recovery efficiencies of metallic uranium from reduced uranium oxide granules and from pellets were about 90% and 50%, respectively. The mass balance data provided the technical bases of Li reduction and refining process flow for design.

Development of Eutectic Free Cladding Materials for Metallic Fuel

Abstract Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10mass%Zr alloy were conducted at a temperature of 1027K for 2250hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr rich layer was observed in all of these materials. Zr rich layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN.