Development of phenomenological degradation models for Cr-Coated Zr alloy cladding under high-temperature oxidation conditions

谷口 良徳; Luu, V. N.; 田崎 雄大; 宇田川 豊; 勝山 仁哉

Annals of Nuclear Energy, 231, p.112177_1 - 112177_16, 2026/06

https://doi.org/10.1016/j.anucene.2026.112177

Advanced technology fuels (ATF) with improved oxidation resistance are under development to enhance the safety of light water reactors. Cr-coated Zr alloy cladding, a promising near-term ATF, exhibits excellent oxidation resistance below the Cr-Zr eutectic temperature. However, its gradual loss of protective effect over time, even without mechanical damage, indicates the need to understand its degradation mechanisms. This article presents a phenomenological model describing degradation due to high-temperature oxidation, focusing on Zr ingress into the Cr coating and the formation of oxygen pathways that accelerate oxygen uptake into the Zr matrix. The model was validated against experimental data at 1200C and 1300C, reproducing key trends such as oxide growth, weight gain, and oxygen concentration profiles. Applying the same parameters to a different PVD-coated cladding test gave reasonable agreement at 1200C, while discrepancies at 1300C suggest Cr-Zr eutectic reactions from local temperature variations, highlighting the model's sensitivity near the eutectic point.

Measurement of transient fission gas release from high-burnup MOX fuel under a simulated reactivity-initiated accident condition using fission gas dynamics testing technique

谷口 良徳; 浦野 建太; 三原 武; 宇田川 豊; 垣内 一雄; 勝山 仁哉

Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10

To investigate the fission gas release behavior of MOX fuel under reactivity-initiated accident (RIA) conditions, a RIA-simulated test on a high-burnup MOX fuel irradiated up to about 64.5 GWd/t (Test FGD-3) was conducted at the Nuclear Safety Research Reactor (NSRR) in JAEA by using recently developed Fission Gas Dynamics (FGD) testing technique. The concept of the FGD tests is to evaluate fission gas release during RIA-simulated test by measuring the pressure transient inside a rigid chamber containing the test fuel rod. We utilize Linear Variable Differential Transformer (LVDT)-type pressure sensor which less affected by gamma and/or neutron field in the NSRR core than conventional strain gauge-type pressure sensor. The maximum fuel enthalpy during Test FGD-3 was evaluated as 276 J/g, which is almost the same value as that of a previous FGD test on a high-burnup UO fuel (about 61 GWd/t) (Test FGD-2). The measured pressure increased from 0.1 MPa to eventually stabilized at about 0.75 MPa: this increase of pressure roughly corresponds to a transient FGR of about 28%, which is higher than that obtained in Test FGD-2 (about 18%). Sensitivity analyses of effective gas permeability for axial gas communication inside the FGD-3 test fuel rod using fuel performance code RANNS showed that apparent gas permeability of the FGD-3 fuel was much higher than that of the FGD-2 fuel. These results suggest that transient fission gas release from high-burnup MOX fuel exceeds that from UO fuel with similar burnup levels, and a significant portion released shortly after energy injection.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

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

https://doi.org/10.1016/j.anucene.2023.110144

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.

Radiochemical analysis of the drain water sampled at the exhaust stack shared by Units 1 and 2 of the Fukushima Daiichi Nuclear Power Station

島田 亜佐子; 谷口 良徳; 垣内 一雄; 大平 早希; 飯田 芳久; 杉山 智之; 天谷 政樹; 丸山 結

Scientific Reports (Internet), 12(1), p.2086_1 - 2086_11, 2022/02

https://doi.org/10.1038/s41598-022-05924-2

2011年3月12日に福島第一原子力発電所の1号機のベントが行われ、1・2号機共用スタックから放射性ガスが放出された。本研究ではこのベントにより放出された放射性核種の情報を有していると考えられる、1・2号機共用スタック基部のドレンピットから採取したドレン水の放射化学分析を実施した。揮発性のIやCs, Csだけでなく、Co, Sr, Sb, 1号機由来安定Moが検出された。1号機由来安定Moの量はCsの量よりもはるかに少ないことから、事故時の炉内状況ではCsMoOの生成は抑制されたと考えられる。また、2020年10月時点では、約90%のIがI、約10%がIOで存在した。Csより多いIが観測されたことから、事故時にIはCsIというよりも分子状のヨウ素として放出されたことが示唆された。2011年3月11日に減衰補正したCs/Cs放射能比は0.86で、2号機や3号機由来と考えられる放射能比より低いことが示された。

Simulation of the effect of radially oriented hydride precipitates on failure limit of high-burnup BWR fuel cladding under PCMI loading

谷口 良徳; 三原 武; 宇田川 豊

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

Scattering of hydride precipitates in a fuel cladding tube was simply modeled by mapping of multiple cracks in finite element system based on the image-processed hydride morphologies observed in post-test cladding samples and the mechanical interactions of these cracks were simulated by damage mechanics calculation. This is a part of ongoing efforts to analyze the effect of the radially oriented hydride precipitates in the cladding tube on the fuel-failure limit observed in Test OS-1: a reactivity-initiated accident (RIA)-simulated test on the BWR fuel with additives irradiated to 64 GWd/tU, which resulted in a fuel failure with the lowest failure limit among the tests ever performed at the NSRR for high burnup BWR rods. The LS-1 test fuel rod, with similar burnup to the OS-1 rod, was selected as another RIA-simulated test rod to be compared with. Sensitivity was examined for damage model parameters, which dominate strain level at which a finite element becomes softened and finally loses its load-carrying capacity, and two sets of plasticity model parameters calibrated for irradiated and unirradiated materials. In the calculation, large stress concentration occurred in the regions between the tips of two adjacent cracks, and one pair of such cracks, typically one of the longest radial cracks existing in the outer periphery of the cladding, then linked to form a longer crack. The simulated macroscopic circumferential strain at failure of the OS-1 cladding model was lower than that of the LS-1 cladding model by about 40% or more. Limited sensitivity of the damage and plasticity model parameters, observed for the investigated range, suggests that the reduction of failure strain primarily reflects the difference in crack distributions between the two simulated rods. The results support the interpretation that the radially oriented hydrides contributed to the low PCMI-failure limit observed in Test OS-1.

The Effect of base irradiation on failure behaviors of UO 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

AA2019-0372.pdf:0.81MB

https://doi.org/10.1016/j.anucene.2019.107268

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.

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 宇田川 豊; 三原 武; 天谷 政樹; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

A pulse-irradiation test CN-1 on a high-burnup MOX fuel with M5 cladding was conducted at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Although the transient signals obtained during the pulse-irradiation test did not show any signs of the occurrence of PCMI failure, the failure of the test fuel rod was confirmed from the visual inspection carried out after test CN-1. Analyses using fuel performance codes FEMAXI-8 and RANNS were also performed in order to investigate the fuel behavior during normal operation and pulse-irradiation regarding the test fuel rod of CN-1, and the results were consistent with this observation result. These experimental and calculation results suggested that the failure of test fuel rod of CN-1 was not caused by hydride-assisted PCMI but high-temperature rupture following the increase in rod internal pressure. The occurrence of this failure mode might be related to the ductility remained in the M5 cladding owing to its low content of the hydrogen absorbed during normal operation.

燃料安全研究国際会議(Fuel Safety Research Meeting)2018

谷口 良徳; 垣内 一雄; 天谷 政樹

核燃料, (54-1), p.16 - 19, 2019/03

http://www.aesj.or.jp/~fuel/Pdf/kaihou/kaihou_2019_54-1r2.pdf

日本原子力研究開発機構(JAEA)は、国内外の専門家との間で軽水炉燃料の安全性に係る情報交換や議論を目的とした国際会議「燃料安全研究国際会議(Fuel Safety Research Meeting: FSRM)」を開催している。本報は、2018年10月30-31日に茨城県水戸市で開催した、FSRM2018の概要について述べたものである。

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA has conducted a research program called ALPS-II program for advanced fuels of LWRs. In this program, the tests simulating a RIA and a LOCA have been performed on the high burnup advanced fuels irradiated in European commercial reactors. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by the pulse irradiation tests at the NSRR in JAEA. The information about pellet fragmentation etc. during the pulse irradiations was also obtained from post-test examinations on the test rods after the pulse irradiation tests. As for the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the RFEF in JAEA. The fracture limits under LOCA and post-LOCA conditions etc. of the high-burnup advanced fuel cladding have been investigated, and it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined.

Analyses of SPERT-CDC test 859 by FEMAXI-7 and RANNS codes

谷口 良徳; 宇田川 豊; 天谷 政樹

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.229 - 238, 2016/09

In the current Japanese regulation concerning fuel safety, the criterion of fuel failure due to pellet-cladding mechanical interaction (PCMI) in a burnup range of 25-40 GWd/t is determined substantially based on the result of SPERT-CDC test 859 (SPERT859). In this study, the oxide thickness of the cladding formed on the cladding outer surface of SPERT859 test rod and its fuel enthalpy at failure due to PCMI under this corrosion condition were analyzed by using fuel performance codes FEMAXI-7 and RANNS. These results of FEMAXI-7 and RANNS showed that the cladding of the test rod had excessive corrosion and suggested that the fuel enthalpy at failure of SPERT859 was affected by the excessive corrosion on the cladding of the test rod and was likely lower than that of the typical fuel for light water reactors.