垣内 一雄; 宇田川 豊; 山内 紹裕*
JAEA-Research 2022-001, 21 Pages, 2022/06
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.
Li, F.; 三原 武; 宇田川 豊
Journal of Nuclear Science and Technology, 10 Pages, 2022/04
The mechanical properties of fuel cladding near the elastic limit are essential in considering its failure limit during a pellet-cladding mechanical interaction phase under reactivity-initiated accident (RIA) conditions. The mechanical properties of a Zircaloy-4 cladding tube, such as orthotropic elasticity and anisotropic constants for Hill's plasticity law, were evaluated based on the biaxial stress test data, focusing on the equivalent plastic strain up to ~2.5%. Samples with various fabrication conditions, such as cold-worked, recrystallized, and stress relieved after cold-work with Q-factors of 2, 3, and 4 were investigated. The cold worked samples and recrystallized samples showed high yield stress and Young's modulus, respectively. The evaluated mechanical properties of the stress relieved samples revealed a limited impact of Q-factors on mechanical behavior, including their anisotropic feature. The derived mechanical properties were applied to evaluate the fracture mechanics parameter, J-integral, based on failure limit data from biaxial-expansion-due-to-compression tests on precracked tubes. This evaluation produced systematically lower J-integral values of the stress relieved tube than previously evaluated based on the failure limit data from in-pile RIA-simulated tests.
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.
三原 武; 垣内 一雄; 谷口 良徳; 宇田川 豊
Journal of Nuclear Science and Technology, 14 Pages, 2022/00
Fuels with additives are expected to provide enhanced fuel performance in fission gas retention owing to their large grain size, which elongates fission gas migration path. To investigate behavior of the fuels during a reactivity-initiated accident (RIA), RIA-simulated experiments OS-1 and LS-4 were performed on ADOPT (chromia- and alumina-doped UO) fuel of 64 GWd/t and chromia-doped UO fuel of 48 GWd/t, respectively. The OS-1 rod failed at a fuel enthalpy increase of 160 J/g due to pellet-cladding mechanical interaction failure, which was the lowest failure limit among the test results ever obtained at the NSRR on high-burnup fuels from 40 to 65 GWd/tU. Comparison of the hydride morphologies in the cladding metallic layer between the rods subjected to the past NSRR tests suggests the contribution of radially oriented hydrides during base irradiation to the low failure limit. The LS-4 rod survived for a peak fuel enthalpy increase of 549 J/g, which resulted in cladding deformation of 2.4% in the residual hoop strain and FGR of 1.4%-6.1%. Whereas the low fission gas release exhibits the effect of additives, the cladding deformation is within the range explained by the deformation mechanism essentially identical to those recognized for high-burnup undoped fuels.
古本 健一郎; 宇田川 豊
Journal of Nuclear Science and Technology, 12 Pages, 2022/00
In order to contribute to better modeling and evaluation of fuel fragmentation, relocation, and dispersal expected under loss of coolant accident (LOCA) conditions, LOCA-simulated cladding burst experiments were performed on as-received nonirradiated 17 17 type Zircaloy-4 cladding specimens that were internally pressurized. The experiments were designed to terminate at burst occurrence to focus on ballooning and rupture opening formation and to investigate the effects of various factors. The postburst cladding hoop strain decreased with the increase in azimuthal temperature distribution (ATD) of the cladding, as found previously. The rupture opening size increased with the increase in ATD and the increase in energy of the pressurized gas stored inside the pressure boundary of the test sample system. Comparison with the existing database, which included tests on irradiated rods containing fuel pellets, suggested that formation of the rupture opening was influenced by the characteristic behavior of high burnup fuels, such as limited gas migration in the cladding tube due to fuel-cladding bonding and interaction of the ejected fuel fragments with the cladding tube.
三原 武; 垣内 一雄; 谷口 良徳; 宇田川 豊
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Test OS-1, the reactivity-initiated-accident (RIA) -simulated test on 64 GWd/tU high burnup fuel with ADOPT (chromia-and-alumina-doped UO) pellets resulted in a failure at the lowest fuel enthalpy increase among the tests ever performed at the NSRR on high burnup fuels from 40 to 65 GWd/tU. Roles of both fuel pellets and cladding behaviours in this remarkable observation are being investigated. A comparative RIA-simulated test OS-2 was thus performed on undoped fuel that had been base-irradiated in the identical fuel assembly with the OS-1 rod. The transient records acquired during Test OS-2 indicated that the rod survived without fuel failure. Radially projected hydride lengths in the cladding metallic layer were evaluated from the metallograph images observed in the vicinity of the OS-2 test rod and compared with other failure test cases. The comparison suggested that the hydride morphologies affected the low failure limit of the OS-1 rod and also explains the survival of the OS-2 rod, to some extent. Nevertheless, as the OS-2 rod survived 100 J/g higher peak fuel enthalpy than the OS-1 failure limit, further experimental and analytical studies are desired to pursue other possible causes: additional loading specific to ADOPT pellets, difference in the pellet/cladding bonding condition, and so on.
成川 隆文; 宇田川 豊
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
To clarify the mechanism and temperature threshold for fuel fragmentation during loss-of-coolant accidents (LOCAs), out-of-pile heating tests on bare fuel pellet pieces taken from a high-burnup PWR UO fuel rod (segment average burnup: 81 GWd/tU) were performed. The fuel pellet pieces taken from various regions in the radial direction of the fuel pellet were inductively heated with no cladding restraint in vacuum up to 1473 K at a rate of 5 K/s. During the heating tests, the fission gases released from the fuel pellet pieces were continuously analyzed in-situ using a quadrupole mass spectrometer. Following the heating tests, microstructural observation of the fuel pellet fragments was carried out. Based on the relationship between the extent of fuel fragmentation and the terminal temperature, and the time history of fission gas release, temperature thresholds for minor fuel fragmentation and slightly more fuel fragmentation were estimated to be 973 - 1073 K and 1173 - 1273 K, respectively. The extent of fuel fragmentation and the amount of fission gas release became more pronounced with increasing temperature. Further, the microstructural observations after the heating tests revealed that most of the fuel fragments smaller than approximately 500 - 750 m have microstructures consisting of many micropores and subgrains, which are characteristic of the dark zone or high-burnup structure. On the basis of these results, the mechanism of fuel fragmentation during LOCAs was discussed.
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.
宇田川 豊; 田崎 雄大
JAEA-Data/Code 2021-007, 56 Pages, 2021/07
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.
日本原子力学会誌ATOMO, 62(10), p.555 - 559, 2020/10
宇田川 豊; 更田 豊志*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
This article aims at providing a general outline of fuel behavior during a reactivity-initiated accident (RIA) postulated in light water reactors (LWRs) and at showing experimental data providing technical basis for the current RIA-related regulatory criteria in Japan.
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.
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.
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
三原 武; 宇田川 豊; 天谷 政樹; 谷口 良徳; 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.544 - 550, 2019/09
In order to assess effects of additives for fuel pellet on the fuel behavior during a reactivity-initiated accident (RIA), fuels with additives irradiated in commercial light water reactors (LWRs) in Europe up to high burnup were subjected to pulse-irradiation experiments in Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Two tests were performed: test LS-4 with chromia-doped UO and Zry-2 cladding with liner and test OS-1 with ADOPT (chromia-and-alumina-doped UO) pellet and Zry-2 cladding with liner. The test fuel rod of LS-4 did not fail. The test fuel rod of OS-1 was considered to be failed by hydride-assisted pellet-cladding mechanical interaction (PCMI). The fuel failure limit in OS-1 was the lowest among the test results ever obtained at the NSRR in similar burnup range. The morphology of the hydrides precipitated in the fuel cladding of OS-1 was investigated by metallography and compared with previous results obtained in JAEA in connection focusing fuel failure limit. It was suggested that the observed lower limit of fuel failure was related to the amount and length of the hydride precipitated along the radial direction of cladding.
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.