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論文

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

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:0 パーセンタイル:100(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.

論文

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ 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:0 パーセンタイル:100(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.

論文

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

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

反応度事故時のペレット・被覆管相互作用により生じる軽水炉燃料の破損に関して、我が国の規制基準改訂の検討に資するため、原子炉安全性研究炉NSRRを用いて得られた近年の研究成果を総括する。これに基づき、現行基準の妥当性及び現行基準に代わりうる新たな判断基準としての燃料破損しきい値とその考え方について議論する。

論文

Behavior of LWR fuels with additives under reactivity-initiated accident conditions

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$$_{2}$$ and Zry-2 cladding with liner and test OS-1 with ADOPT$$^{rm TM}$$ (chromia-and-alumina-doped UO$$_{2}$$) 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.

論文

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$$^{TM}$$ 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$$^{TM}$$ cladding owing to its low content of the hydrogen absorbed during normal operation.

論文

Fracture behavior of recrystallized and stress-relieved Zircaloy-4 cladding under biaxial stress conditions

三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(8), p.724 - 730, 2019/08

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to failure of high-burnup fuel rods. Zircaloy cladding tubes are subjected to biaxial stress states induced by PCMI loading. This type of stress state, specific to PCMI, presumably makes the tubes more susceptible to failure. To clarify the influence of the anisotropic mechanical properties of Zircaloy cladding tubes on their fracture behavior under biaxial stress conditions, biaxial tensile tests were performed, and the measured stresses and strains were converted to reduced parameters such as equivalent strain, equivalent stress, and stress triaxiality by using the anisotropic constants of the Hill yield function derived in our previous study. The minimum fracture strains for recrystallized (RX) and stress-relieved (SR) specimens were located where the stress ratio of axial to circumferential is 0.75 in the measured range. The equivalent plastic fracture strains tended to decrease monotonously with increasing stress triaxiality, which is a typical trend observed in ductile fracture, in the range of 0.65-0.78 for both specimens. In the case of SR specimens, however, the analysis with stress triaxiality did not reduce the fracture strains well to a single trend curve, showing that the anisotropic constants used in the present work or Hill yield function itself is not enough to describe the whole anisotropy involved in the fracture process of SR material.

論文

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

 被引用回数:2 パーセンタイル:23.33(Nuclear Science & Technology)

FEMAXI-8は、軽水炉燃料の通常運転時及び過渡条件下の挙動解析を目的として原子力機構が開発・整備を進めてきた解析コードである。主に実験データ解析や燃料設計等研究/開発ツールとして利用されてきたFEMAXI-7に対し、ペレットクラックや核分裂生成物ガス挙動の新規モデル開発、既存モデルの改良及び拡充、プログラムのデータ/処理構造見直し等の改良を行い、性能向上を図った。本論文では最近のモデル改良を経たFEMAXI-8を対象に、168ケースの照射試験ケースで得られた実測データを用いた総合的な予測性能検証を実施し、燃料中心温度やFPガス放出率について妥当な予測を与えることを示した。また別途実施したベンチマーク解析により、数値計算の安定性や計算速度についても前バージョンからの大幅な改善を確認した。

論文

The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(5), p.432 - 439, 2019/05

 被引用回数:3 パーセンタイル:12.61(Nuclear Science & Technology)

Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided samples was found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm.

報告書

燃料挙動解析コードFEMAXI-8の開発; 軽水炉燃料挙動モデルの改良と総合性能の検証

宇田川 豊; 山内 紹裕*; 北野 剛司*; 天谷 政樹

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8は、軽水炉燃料の通常運転時及び過渡条件下の挙動解析を目的として原子力機構が開発・整備を進めてきたFEMAXI-7(2012年公開)の次期リリースに向けた最新バージョンである。FEMAXI-7は主に実験データ解析や燃料設計等研究/開発ツールとして利用されてきたが、燃料挙動に係る現象解明やモデル開発等の燃料研究分野における適用拡大並びに燃料の安全評価等への活用を念頭に、原子力機構ではその性能向上及び実証を進めた。具体的には新規モデル開発、既存モデルの改良及び拡充、プログラムのデータ/処理構造見直し、旧言語規格からの移植、バグフィックス、照射試験データベース構築等のインフラ整備、体系的な検証解析を通じた問題の発見と修正等を行うとともに、各種照射試験で取得された144ケースの実測データを対象とした総合的な性能評価を実施した。燃料中心温度について概ね相対誤差10%の範囲で実測値を再現する等、解析結果は実測データと妥当な一致を示した。

論文

Behaviors of high-burnup LWR fuels with improved materials 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 Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Fuels for light water reactors (LWRs) which consist of improved cladding materials and pellets have been developed by utilities and fuel vendors to acquire better fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate adequacy of the present regulatory criteria in Japan and safety margins regarding the fuel with improved materials, Japan Atomic Energy Agency (JAEA) has conducted ALPS-II program sponsored by Nuclear Regulation Authority (NRA), Japan. In this program, the tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on the high burnup advanced fuels irradiated in commercial PWR or BWR in Europe. This paper presents recent results obtained in this program with respect to RIA, and main results of LOCA experiments, which have been obtained in the ALPS-II program, are summarized.

論文

Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 55(2), p.151 - 159, 2018/02

 被引用回数:4 パーセンタイル:23.21(Nuclear Science & Technology)

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while be the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.

論文

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.

論文

Biaxial-EDC test attempts with pre-cracked zircaloy-4 cladding tubes

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07

The failure behavior of cladding tube was investigated by using the improved EDC test apparatus. Cold-worked, stress-relieved and recrystallized Zircaloy-4 tubes with a pre-crack were used as test specimens: this pre-crack simulated the crack which is considered to form in the hydride rim of high-burnup fuel cladding at the beginning of PCMI failure. In the EDC test, a tensile stress in axial direction was applied and displacement-controlled loading was performed to keep the strain ratio of axial/hoop as a constant. The data of cladding deformation had been achieved in the range of strain ratio of 0, 0.25, 0.5 and 0.75 and pre-crack depth of 41-87 micrometers. Failures in hoop direction were observed in all the tested samples, and a general trend that higher strain ratio and deeper crack depth lead to lower failure limit in hoop direction could be seen. Different crack propagation mode was observed between recrystallized and stress relieved and cold worked samples.

論文

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

篠崎 崇*; 宇田川 豊; 三原 武; 杉山 智之; 天谷 政樹

Journal of Nuclear Science and Technology, 53(9), p.1426 - 1434, 2016/09

 被引用回数:8 パーセンタイル:19.39(Nuclear Science & Technology)

In order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency (JAEA). The specimens with an outer surface pre-crack were prepared by using RAG (Rolling After Grooving) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain ($$varepsilon$$$$_{tz}$$) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain ($$varepsilon$$$$_{ttheta}$$) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state.

論文

Behavior of high-burnup advanced LWR fuels under accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.53 - 62, 2016/09

軽水炉用改良型燃料について、現行の安全基準の妥当性及び安全余裕を評価するため、また今後の規制のためのデータベースを提供するため、原子力機構ではALPS-IIと呼ばれる原子力規制庁からの委託事業を開始した。この事業は、商用PWR及びBWRで照射された高燃焼度改良型燃料を対象として、主として反応度投入事故及び冷却材喪失事故を模擬した試験から構成されている。最近、高燃焼度改良型燃料のRIA時破損限界がNSRRにて調べられ、パルス照射試験後の燃料を対象とした照射後試験が行われている。LCOA模擬試験に関しては、インテグラル熱衝撃試験及び高温酸化試験が燃料試験施設で行われ、高燃焼度改良型燃料被覆管の破断限界、高温酸化速度等が調べられた。本論文では、この事業で取得された最近のRIA及びLOCA模擬試験結果について主に述べる。

論文

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.

論文

Crack formation in cladding under LOCA quench conditions

Wu, H.; 宇田川 豊; 成川 隆文; 天谷 政樹

Nuclear Engineering and Design, 303, p.25 - 30, 2016/07

 被引用回数:2 パーセンタイル:67.65(Nuclear Science & Technology)

Loss-of-Coolant-Accident (LOCA) is a design basis accident that is considered in the safety analyses for LWR. This paper discusses crack formation in one-side oxidized Zircaloy-4 cladding with LOCA one-side oxidation quench experimental data. The experimental data suggest that the order of cracks formed in cladding during LOCA quench conditions should be, first in the alpha-Zr(O) layer, and then in the oxide, finally in the prior-beta layer when the fracture of cladding occurs. Both the experimental data and RANNS computation suggest that the formation of crack in the oxide could be related to the heat capacity inside the cladding and off-center pellets during quench.

論文

Validation of updated RANNS with effect of oxygen-dissolved metallic zircaloy-4 under LOCA quench condition

Wu, H.; 宇田川 豊; 成川 隆文; 天谷 政樹

Nuclear Engineering and Design, 300, p.249 - 255, 2016/04

 被引用回数:2 パーセンタイル:67.65(Nuclear Science & Technology)

Loss-of-Coolant-Accident (LOCA) is a classical design basis accident considered in LWR safety analyses, and LOCA simulation technique can be used to gain a better understanding of local cladding behaviors. This paper first summarizes equations regarding the oxygen-dissolved metallic Zircaloy-4 layer (ODMZ). These equations have been added to the updated RANNS code, which is validated using LOCA quench experimental data. The update RANNS code is then used to examine the influence of ODMZ and the oxide layer on its axial load under LOCA quench conditions. The results suggest that the contribution of both the ODMZ and the oxide layer to the axial load increase with oxidation time, and the latter increases more in a fixed length of oxidation time. This study shows the importance and necessity of considering the effect of the ODMZ when computing the axial load on cladding in LOCA quench conditions.

論文

Recent research activities using NSRR on safety related issues

宇田川 豊; 杉山 智之*; 天谷 政樹

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04

JAEA launched ALPS-II program in 2010 in order to obtain regulatory data for advanced fuels. Five new reactivity-initiated accident (RIA) simulated tests on the advanced fuels have been performed. The first two fuels tested, VA-5 and VA-6, were 17$$times$$17-PWR-type with stress-relieved and recrystallized M-MDA cladding tube, and irradiated to ~80 GWd/tU. The cladding failed due to the pellet-cladding mechanical interaction. Fission gas dynamics tests to promote a better understanding of the behavior of fission gas during an RIA are planned. A recent qualification test on a prototype pressure sensor demonstrated its ability to obtain history data of transient fission gas release. JAEA also launched a new experiment program using NSRR to investigate fuel degradation behaviors in the temperature region beyond-DBA LOCAs.

論文

Behavior of high burnup advanced fuels for LWR during design-basis accidents

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 杉山 智之

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09

高燃焼度領域での燃料性能を向上させるとともに既設の原子炉の安全性を向上させるため、高耐食性被覆管や核分裂生成ガス放出を抑えたペレットで構成された改良型燃料が事業者や燃料メーカによって開発されてきた。このような改良型燃料の現行の規制基準や安全裕度の妥当性を評価するため、またこれらに係る将来の規制のためのデータベースを提供するため、原子力機構はALPS-IIと呼ばれる新しい研究プロクラムを開始した。このプログラムは、欧州から輸送された高燃焼度改良型燃料を対象とした反応度事故(RIA)模擬試験及び冷却材喪失事故(LOCA)模擬試験から主に構成されている。本論文では、このプログラムの概要及び現在までに得られているRIA及びLOCA模擬試験結果について述べる。

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