垣内 一雄; 天谷 政樹; 宇田川 豊
Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06
In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, an irradiation growth testing in the Halden reactor of Norway was conducted on various fuel cladding materials including the improved Zr alloy. In this paper, the effect of hydrogen, which was absorbed in the cladding tube due to corrosion, on the irradiation growth behavior was evaluated. Comparison between the specimens with or without pre-charged hydrogen revealed that the effect of hydrogen absorption, accelerating irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Analysis based on this understanding derived growth acceleration effect (0.060.01)%/100 ppm, whose denominator is defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter.
田崎 雄大; 宇田川 豊; 天谷 政樹
Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03
This study develops a new fission gas release (FGR) model for mixed oxide (MOX) fuels with a fundamentally heterogeneous microstructure. The model adopted in FEMAXI-8 was applied to irradiation Instrumented Fuel Assembly (IFA)-626 and 702 tests in which two types of MOX fuels had different heterogeneity in their microstructure, while the other spec were similar. Upon analyzing these fuels, the original FGR model predicted lower FGR from the fuel with a remarkably heterogeneous microstructure than the other MOX fuel. This estimation contradicts the experimental observation. However, the new FGR model improved the consistency because of the early release of fission gas from Pu agglomerate region, and showed issues for aiming further improvement. Therefore, the above results confirmed a certain validity of the developed model for studying heterogeneity effect.
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
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。
三原 武; 宇田川 豊; 杉山 智之; 天谷 政樹
Journal of Nuclear Science and Technology, 58(8), p.872 - 885, 2021/08
To evaluate the effects of the hydride morphology and initial temperature of fuel cladding on the pellet-cladding mechanical interaction failure under reactivity-initiated accident (RIA) conditions, RIA-simulated experiments were performed on high-burnup fuels with stress-relieved annealed (SR) and recrystallized (RX) M-MDA cladding at room and high ( 280C) temperatures. The results demonstrated that the failure-limit trend of RX-cladded fuels being lower than that of SR-cladded fuels for a similar hydrogen content holds up to at least about 700 wtppm. The observation of the fracture surfaces of failed RX cladding suggests a contribution of radially-oriented hydrides to the crack formation and/or penetration, which coincides with the aforementioned failure-limit trend. The temperature effect, namely the failure-limit rise at a high temperature, is evident irrespective of the hydride morphology, while the degree of the temperature effect decreases as the hydrogen content increases.
垣内 一雄; 宇田川 豊; 天谷 政樹
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
In order to investigate fission gas release behavior of high-burnup mixed-oxide (MOX) fuel pellet for LWR under reactivity-initiated accident (RIA), the tests called BZ-3 and BZ-4 were conducted at the Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Agency (JAEA). Electron probe microanalysis and rod-puncture tests were performed on the fuel pellets before and after pulse irradiation tests, and from the comparison between the puncture test results and the results evaluated from EPMA, it was suggested that fission gas release from not only the Pu-spot but also the Pu-spot-excluded region.
永瀬 文久; 成川 隆文; 天谷 政樹
JAEA-Review 2020-076, 129 Pages, 2021/03
与能本 泰介; 中島 宏*; 曽野 浩樹; 岸本 克己; 井澤 一彦; 木名瀬 政美; 長 明彦; 小川 和彦; 堀口 洋徳; 猪井 宏幸; et al.
JAEA-Review 2020-056, 51 Pages, 2021/03
小澤 正明*; 天谷 政樹
日本原子力学会和文論文誌, 19(4), p.185 - 200, 2020/12
Negyesi, M.; 天谷 政樹
Oxidation of Metals, 94(3-4), p.283 - 299, 2020/10
Oxidation tests of Zry-4 fuel cladding in steam at 1273 K were carried out in this study. The effect of specimen surface roughness on the oxidation behavior was investigated. Steam was applied either at room temperature or at experimental temperature. Weight gain kinetics was evaluated by post-test weight measurement. Metallographic analysis was conducted using optical microscopy. Hydrogen pick-up was measured by gas extraction technique. The effect of specimen surface roughness on the oxidation kinetics as well as on the hydrogen absorption has not clearly been evidenced. The breakaway oxidation was suppressed significantly when the steam was applied at RT. The oxide breakaway was related to grain size of the base metal. Higher hydrogen absorption before the kinetic transition in the condition when steam was applied at 1273 K suggested enhanced oxide porosity.
岡田 裕史; 天谷 政樹
Annals of Nuclear Energy, 145, p.107539_1 - 107539_8, 2020/09
In order to evaluate the fracture resistance of fuel rods against a seismic loading which might be applied following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated after the Fukushima-Daiichi Nuclear Power Plant accident. In consideration of previous studies and results, the effect of the amount of oxidation on the maximum bending stress of pre-hydrided cladding tube with a small amount of ballooning was investigated in this study. According to the obtained results, it was suggested that the decrease in the maximum bending stress of the cladding tube experienced LOCA conditions is mainly determined by the hydrogen concentration in the cladding tube after simulated LOCA test, irrespective of pre-hydriding. It was also suggested that the decreasing trend of the maximum bending stress with increasing the hydrogen concentration would be expressed by a form of exponential function, in which the maximum bending stress at a hydrogen concentration of 1500 ppm was estimated to be about a half of that at 0 ppm.
Li, F.; 三原 武; 宇田川 豊; 天谷 政樹
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
Fuel cladding may be subjected to biaxial tensile stress in axial and hoop directions during pellet-cladding mechanical interaction (PCMI) of a reactivity-initiated accident (RIA). Incipient crack in the hydride rim assisted by the scattered hydrides in the metal phase may lead to failure of the cladding at small hoop strain level during PCMI. To get insight of such phenomenon, biaxial-EDC tests under axial to hoop strain ratios ranging from 0 to 1 were performed with pre-cracked (outer surface) and uniformly hydrided Zircaloy-4 cladding tube samples with final heat-treatment status of cold worked (CW), stress relieved (SR) and Recrystallized (RX). Results showed dependencies of failure hoop strain on pre-crack depth, strain ratio, hydrogen content and final heat-treatment status on fabrication, but no apparent dependencies were observed on the distribution pattern of hydrides (with similar hydrogen contents and hydrides predominantly precipitated in hoop direction) and the heat-treatment process for hydrogen charging. J integral at failure seems to be available to unify the effect of pre-crack depth.
成川 隆文; 天谷 政樹
Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07
To evaluate the fracture resistance of high-burnup advanced fuel cladding tubes during the long-term core cooling period following loss-of-coolant accidents (LOCAs), laboratory-scale four-point-bend tests were performed using the following advanced fuel cladding tubes with burnups of 73 - 84 GWd/t: low-tin ZIRLO, M5, and Zircaloy-2 (LK3). Three four-point-bend tests were performed on the high-burnup advanced fuel cladding tube specimens subjected to the integral thermal shock tests which simulated LOCA conditions (ballooning and rupture, oxidation in high-temperature steam, and quench). During the four-point-bend tests, all the specimens that were oxidized at 1474 K to 9.9% - 21.5% equivalent cladding reacted exhibited brittle fractures. The maximum bending moments were comparable to those of the conventional Zircaloy cladding tube specimens. Furthermore, the effects of oxidation and hydriding on the maximum bending moment were comparable between the high-burnup advanced fuel cladding tube specimens and the unirradiated Zircaloy-4 cladding tube specimens. Therefore, it can be concluded that the post-LOCA fracture resistance of fuel cladding tubes is not significantly reduced by extending the burnup to 84 GWd/t and using the advanced fuel cladding tubes, though it may slightly decrease with increasing initial hydrogen concentration in a relatively lower ECR range ( 15%), as observed for the unirradiated Zircaloy-4 cladding tubes.
Li, F.; 三原 武; 宇田川 豊; 天谷 政樹
Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06
To better understand the failure limit of fuel cladding during the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA), pre-cracked and hydrided cladding samples with base metal final heat-treatment status of cold worked (CW) and recrystallized (RX) were tested under biaxial stress conditions (axial to hoop strain ratios of 0 and 0.5). Displacement-controlled biaxial-expansion-due-to-compression (biaxial-EDC) tests were performed to obtain the hoop strain at failure (failure strain) of the samples. The conversion of the failure strains to J-integral at failure by finite-element analysis involving data of stress-relieved (SR) cladding specimens from our previous study revealed that the failure limit in the dimension of J-integral at failure unifies the effects of pre-crack depth. About 30 to 50 percent reduction in the J-integral at failure was observed as the strain ratio increased from 0 to 0.5 irrespective of the annealing type, pre-crack depth, and hydrogen content. the rate of fractional decreases of J-integral at failure with increase of hydrogen content are in the order of CWSRRX, which are essentially independent of strain ratio for the CW and SR samples. The results were incorporated into the failure prediction model of the JAEA's fuel performance code in the form of a correction factor that considers the biaxial loading effect.
谷口 良徳; 宇田川 豊; 天谷 政樹
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
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.
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
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.
垣内 一雄; 天谷 政樹
日本原子力学会和文論文誌, 19(1), p.24 - 33, 2020/03
岡田 裕史; 天谷 政樹
Annals of Nuclear Energy, 136, p.107028_1 - 107028_9, 2020/02
In order to evaluate the fracture resistance of fuel rods against a seismic loading following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated since the Fukushima-Daiichi Nuclear Power Plant accident. In this study, four-point-bending-tests were performed using Zircaloy-4 cladding tubes which experienced a simulated LOCA sequence in order to investigate the effects of oxidation and secondary hydriding occurring during a LOCA on the bending strength of fuel cladding. According to the obtained results, it was suggested that the maximum bending stress would be affected by the oxygen concentration in the prior-beta layer as well as the thickness of prior-beta layer. It was considered that the decrease in maximum bending stress by secondary hydriding is probably expressed by multiplying a factor of 0.37 by the maximum bending stress which solely takes account of the effect of oxidation.
成川 隆文; 天谷 政樹
Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01
To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73 - 85 GWd/t: M-MDA, low-tin ZIRLO, M5, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10% - 30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520 - 530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.
宇田川 豊; 杉山 智之; 天谷 政樹
Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12
Negyesi, M.; 天谷 政樹
Oxidation of Metals, 92(5-6), p.439 - 455, 2019/12
The paper deals with the effect of air fraction in steam on the embrittlement of Zry-4 fuel cladding exposed under steam-air atmospheres (air fractions of 10-100%) in the temperature range of 1273-1573 K. Ring compression tests were carried out in order to evaluate the embrittlement of fuel cladding. Furthermore, the microhardness of prior -phase was measured and fractured surfaces were observed under SEM. The degree of the embrittlement was discussed against the results of metallographic and hydrogen analyses. The microstructure and the hydrogen pick-up were substantially affected by nitride formation. Accelerated oxidation kinetics enhanced shrinking of the prior -region. The enhanced hydrogen absorption resulted in the increased microhardness of prior -phase. The degree of fuel cladding embrittlement, expressed by the plastic strain at failure and the maximum load, correlated well with the microhardness and the thickness of prior -phase.