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Luu, V. N.; 谷口 良徳; 宇田川 豊; 勝山 仁哉
Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)For near-term application, coated-Zr alloy claddings show potential for enhancing safety by providing better oxidation resistance and minimizing hydrogen absorption under design-basis accidents (DBA). This benefit could extend the burnup and operational cycles of fuel rods. In assessing safety, reactivity-initiated accidents (RIA) are considered as one of the DBA conditions. The current safety criteria for high-temperature oxidation failure, one of the failure modes linked to RIA, are defined by peak fuel enthalpy values that range from 205 to 270 cal/g. This wide variability presents challenges when attempting to generalize criteria for modified-Zr alloy claddings with superior oxidation resistance. Therefore, it may be more relevant to apply failure criteria based on embrittlement mechanisms, such as oxygen concentration in the 
-Zr phase. This study aimed to assess the failure based on both peak fuel enthalpy and cladding embrittlement by analyzing previous NSRR experiments conducted with conventional materials using the RANNS fuel performance code. The findings suggest that the failure criteria associated with cladding embrittlement can provide a rational evaluation of failure behavior compared to the existing criterion based on peak fuel enthalpy. The local failure criterion leading to the formation of through-wall cracks during quenching is consistent with Chung's proposal (NUREG/CR-1344): -Zr thickness of 
0.9 wt% oxygen is less than 0.1 mm, and this corresponds to approximately 35% BJ-ECR.
谷口 良徳; 三原 武; 垣内 一雄; 宇田川 豊
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
被引用回数：1 パーセンタイル：18.87(Nuclear Science & Technology)A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5
cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.
谷口 良徳; 宇田川 豊; 天谷 政樹
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
被引用回数：1 パーセンタイル：8.04(Nuclear Science & Technology)The fuel-failure-limit data obtained in the simulated reactivity-initiated-accident experiment SPERT-CDC 859 (SPERT859) has entailed a lot of discussions if it represents fuel-failure behavior of typical commercial LWRs for its specific pre-irradiation condition and fuel state. The fuel-rod conditions before and during SPERT859 were thus assessed by the fuel-performance codes FEMAXI-8 and RANNS with focusing on cladding corrosion and its effect on the failure limit of the test rod. The analysis showed that the fuel cladding was probably excessively corroded even when the influential calculation conditions such as fuel swelling and creep models were determined so that the lowest limit of the cladding oxide layer thickness was captured. Such assumption of excessive cladding corrosion during pre-irradiation well explains not only the test-rod state before pulse irradiation but also the fuel-failure limit observed. Such understanding undermines anew the representativeness of the test data as a direct basis of safety evaluation for LWR fuels.
and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8宇田川 豊; 三原 武; 谷口 良徳; 垣内 一雄; 天谷 政樹
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
被引用回数：3 パーセンタイル：25.23(Nuclear Science & Technology)This paper reports a computer-code analysis on the base-irradiation behavior of the chromia-and-alumina-doped BWR rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction (PCMI) at the lowest fuel-enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced more intense PCMI loading due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests and thus had been prone to experience enhanced radial-hydride formation. The significant difference in the cladding hoop-stress more than 50 MPa discriminates the OS-1 rod from other BWR rods and supports the interpretation that enhanced radial-hydrides formation differentiated the PCMI-failure behavior observed in the test OS-1 from the previous BWR-fuel tests.
宇田川 豊
no journal, ,
This presentation reports a computer-code analysis on the base-irradiation behavior of the chromia-and-alumina-doped BWR rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction (PCMI) at the lowest fuel-enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced more intense PCMI loading due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests and thus had been prone to experience enhanced radial-hydride formation. The significant difference in the PCMI-related parameters between the OS-1 rod and other BWR rods supports the interpretation that enhanced radial-hydrides formation differentiated the PCMI-failure behavior observed in the test OS-1 from the previous BWR-fuel tests.
谷口 良徳
no journal, ,
軽水炉燃料の燃料破損基準の妥当性確認において重要となるCN-1実験で観察された燃料破損モード(高温破裂)の原因及び高温破裂限界を明らかにするため、高燃焼度MOX燃料を用いたRIA模擬実験(CN-2実験)及びCN-1実験の照射後試験を実施した。これらの実験結果から、反応度事故(RIA)時の核分裂生成物(FP)ガス放出が同破損の主な駆動力となっていることが分かった。この結果を踏まえ、RIA条件下での動的なFPガス放出挙動を調べるため、高燃焼度MOX燃料を対象に、JAEAが開発したNSRRを用いて動的なFPガス放出挙動を調べる技術であるFission Gas Dynamics(FGD)実験技術を用いたRIA模擬実験(FGD-3/CN-3)を実施した。同実験時のオンラインデータ(FGDチャンバー内圧)と、既往の高燃焼度燃料UOを対象としたFGD-2実験の同データの比較から、MOX燃料のFPガス放出挙動はUO燃料のそれと同様であり、FPガス放出はエネルギー印加直後から開始し、急速に完了段階に達したことが分かった。また、燃料挙動解析コードを用いてプレナムへの軸方向ガス移行挙動を解析した結果、FPガス移行量はFGD-3実験の方がFGD-2実験に比べて大きいことが分かった。
