Masatoshi Hayashi

Release of fission products from irradiated aluminide fuel at high temperatures

Irradiated uranium-aluminide fuel plates of 40% /sup 235/U enrichment were heated for the determination of the amounts of fission products released at temperatures up to and higher than the melting point of the fuel cladding material. The release of fission products from the fuel plate at temperatures below 500/sup 0/C was negligible. Three stages of fission product release were observed. The first rapid release was observed at about 561/sup 0/C along with blistering of the plates. The next release, which occurred at 585/sup 0/C, might have been caused by melting of the Type 6061 aluminum alloy. The last release of fission product gases occurred at 650/sup 0/C, which probably corresponds to the eutectic temperature of the uranium-aluminum alloy. The released material was mostly xenon, and small amounts of iodine and cesium were observed.

Effect of reducing fuel enrichment on the void reactivity - part II: analytical study

The results of analyses on the void reactivity measurements performed in the Kyoto University Critical Assembly using medium-enriched uranium fuel as well as highly enriched uranium fuel are provided. In consideration of the heterogeneity of a complex core, four-group constants were generated by SRAC, a standard thermal reactor code system for reactor design and analysis at the Japan Atomic Energy Research Institute. The eigenvalue and perturbation calculations were subsequently performed by the 2D-FEM-KUR code which is a two-dimensional diffusion code based on the finite element method. The calculated eigenvalue k /SUB eff/ agreed with the measured value to within 0.5% in the calculated-to-experiment ratio. The void reactivity calculated by perturbation theory approximately reproduced the experimental data including the spatial dependence. The discrepancy between the calculated and measured void reactivity was < 0.05 x 10/sup -3/ ..delta..k/k per voided flow channel.

Effect of Reducing Fuel Enrichment on the Void Reactivity — Part I: Experimental Study

Reactivity of void in the channels between the fuel plates is measured, and the impact of core conversion from using highly enriched uranium (HEU) to medium-enriched uranium (MEU) in the light-water-moderated cylindrical core with heavy water reflector is investigated on this quantity at the Kyoto University Critical Assembly. The void was generated in the flow channels by producing nitrogen gas bubbles through a small needle-like nozzle and the reactivity effect was measured. The void fraction was measured in an outof-pile experiment. The results indicate that the void effect on reactivity is slightly larger (more negative) in the MEU core than in the HEU core. It is also shown that the interference effect of reactivity by bubbling in two adjacent channels simultaneously is within the experimental error.

Analysis of Critical Experiments Using Medium-Enriched-Uranium Fuel in Kyoto University Critical Assembly (KUCA)

The critical experiments using medium-enriched-uranium (MEU) fuel in the Kyoto University Critical Assembly (KUCA), a light-water-moderated and heavy-water-reflected cylindrical core, were started in May 1981, as a part of the international Reduced Enrichment for Research and Test Reactors (RERTR) program. The following KUCA critical experiments were analyzed: (1) the criticality measurements for high-enriched-uranium (HEU) and MEU cores and (2) the reactivity effect measurements of boron burnable-poison (BP) for MEU cores. Five-group constants were generated using the EPRI-CELL code, and two-dimensional diffusion calculations were performed using a conventional finite-difference code DIF3D(2D), and a finite-element code 2D-FEM-KUR. Some of the results from the two diffusion codes were compared with each other. Advantage was taken of the finite-element method for the application of the 2D-FEM-KUR code to a detailed analysis of the BP effect measurements. Differences between the results of calculations and...

Measurement and analysis of capture reaction rate of 237Np in various thermal neutron fields by critical assembly and heavy water thermal neutron facility of Kyoto University

Abstract Capture reaction rate ratios of 237Np relative to 197Au were measured in 11 thermal neutron fields provided by the Kyoto University Critical Assembly and the Kyoto University Reactor Heavy Water Neutron Irradiation Facility. In the measurement, both samples of 237Np and 197Au were irradiated at the same time, and their gamma activities were measured. The typical experimental error was 3.5%. The analysis was performed by three steps: full-core calculation, self-shielding correction of the sample, and perturbation correction of the sample. Three full-core calculations by a continuous-energy Monte Carlo code (MVP), a transport code (TWOTRAN), and a diffusion code (CITATION) were made with the JENDL-3.2 library. The self-shielding factors were derived by an analytical formula, and the perturbation factors were calculated by another MVP calculation. The reaction rates were derived by multiplying the neutron spectrum, the two correction factors, and the capture cross sections of 237Np and 197Au. As a result, the three full-core calculations provided almost the same neutron spectra at the sample position and gave almost the same calculated-to-experimental values (C/Es) for the capture reaction rate ratios of 237Np relative to 197Au. Based on the capture cross section of 237Np taken from the JENDL-3.2 library, the C/Es were between 0.97 and 1.04, and the average C/E among the 11 cores was 1.01. On the other hand, the C/Es using the ENDF/B-VI and the JEF-2.2 were 1.02 to 1.06 for harder spectrum cores, whereas the C/Es for the softer spectrum cores were 1.08 to 1.16. It is concluded that the JENDL-3.2 library has good accuracy for the capture cross section of 237Np but the ENDF/B-VI and the JEF-2.2 libraries overestimate that of 237Np >10% in the thermal neutron energy region.