| Project/Area Number |
02452294
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| Research Category |
Grant-in-Aid for General Scientific Research (B)
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| Allocation Type | Single-year Grants |
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| Research Field |
Nuclear engineering
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| Research Institution | KYOTO UNIVERSITY |
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Principal Investigator |
SHIROYA Seiji Kyoto University, Research Reactor Institute(KURRI), Associate Professor, 原子炉実験所, 助教授 (80027474)
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| Co-Investigator(Kenkyū-buntansha) |
KANDA Keiji KURRI, Associate Professor, 原子炉実験所, 助教授 (10027419)
UNESAKI Hironobu KURRI, Research Associate, 原子炉実験所, 助手 (40213467)
ICHIHARA Chihiro KURRI, Research Associate, 原子炉実験所, 助手 (90027475)
HAYASHI Masatoshi KURRI, Research Associate, 原子炉実験所, 助手 (20027444)
KOBAYASHI Keiji KURRI, Research Associate, 原子炉実験所, 助手 (30027445)
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| Project Period (FY) |
1990 – 1992
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| Project Status |
Completed (Fiscal Year 1992)
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| Budget Amount *help |
¥6,200,000 (Direct Cost: ¥6,200,000)
Fiscal Year 1992: ¥200,000 (Direct Cost: ¥200,000)
Fiscal Year 1991: ¥1,100,000 (Direct Cost: ¥1,100,000)
Fiscal Year 1990: ¥4,900,000 (Direct Cost: ¥4,900,000)
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| Keywords | Nuclear Data / Neutronics Calculation / Critical Experiment / Tight-Pitch Lattice Core / Moderator-to-Fuel Volume Ratio / Self-Shielding Effect / Effective Cross Section / KUCA / 核計算コード / 中性子スペクトル / ENDF / B-IV / JENDL-3 / 減速材対燃料体積化 / 燃料板バンチング / ウラン核デ-タ / 核計算コ-ド / 評価 / 平均ウラン235濃縮度 / 燃料分布 / 非等方散乱 / 輸送効果 |
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| Research Abstract |
A study on the assessment of the nuclear data and neutronics calculations was performed through the critical experiments on the tight-pitch lattice cores loaded with uranium fuel and moderated by polyethylene.
In this series of critical experiments using the Kyoto University Critical Assembly (KUCA), the moderator-to fuel volume ratio (V_m/V_f ratio) and the average enrichment of ^<235>U in the core were systematically varied to obtain the benchmark data for the assessment of the nuclear data and neutronics calculations. Monte Carlo calculations were executed to assess the validity of the nuclear data and the computational methods used in the neutronics calculation.
Through the present study, it was found that the use of the current version of evaluated nuclear data file in Japan, JENDL-3, tends to underestimate the effective multiplication factor in comparison with the measured one and the cal-culated result by ENDF/B-IV which is a previous version of evaluated nuclear data file in the United States, whereas the use of a previous version in Japan, JENDL-2, tends to overestimate the effective multiplication factor. For the generation process of effective cross sections, it was found that one should properly take into account the interference effect in resonance self-shielding of uranium contained in two adjacent regions. It was also found that the trans-port effect can not be neglected in the diffusion calculation and the result com-parable to the accurate transport calculation can be obtained on the basis of the diffusion calculation by correcting the transport,energy group and mesh effects.
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