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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

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

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

 パーセンタイル:100(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

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dost$'a$l, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; 天谷 政樹; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

The benchmark on PCMI was initiated by OECD/NEA Expert Group on Reactor Fuel Performance (EGRFP) in June 2015 and is currently in the latter stages of compiling results and preparing the final report. The aim of the benchmark is to improve understanding and modelling of PCMI amongst NEA member organisations. This is being achieved by comparing PCMI predictions of different fuel performance codes for a number of cases. Two of these cases are hypothetical cases aiming to facilitate understanding of the effects of code-to-code differences in fuel performance models. The two remaining cases are actual irradiations, where code predictions are compared with measured data. During analysis of participants' results of the hypothetical cases, the assumptions for number of radial pellet cracks and the pellet-clad friction coefficient (which can be zero, finite or infinite) were identified to be important factors in explaining differences between predictions once pellet-cladding contact occurs. However, these parameters varied in the models and codes used originally by the participants. This fact led to the extension of the benchmark by inclusion of two additional cases, where the number of radial pellet cracks and three different values of the friction coefficient were prescribed in the case definition. Seven calculations from six organisations contributed results were compared and analysed in this paper.

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:1 パーセンタイル:38.14(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

The Fracture behaviors of non-irradiated zircaloy-4 fuel cladding with a pinhole under simulated LOCA conditions

小宮山 大輔; 天谷 政樹

Journal of Nuclear Science and Technology, 54(12), p.1338 - 1344, 2017/12

 パーセンタイル:100(Nuclear Science & Technology)

In a case where a pinhole leak occurs in a fuel rod incidentally, it is possible that coolant enters the fuel rod through the pinhole. Since knowledge about the behavior of the fuel rod with a pinhole under LOCA condition is limited, semi-integral quench tests consisting of heat-up, isothermal oxidation and quenching processes were performed with non-irradiated fuel cladding tubes with a pinhole in order to investigate the difference in the fracture behaviors between normal and leaker fuels under LOCA condition. Isothermal oxidation temperature and time ranged from 1100 $$^{circ}$$C to 1225 $$^{circ}$$C and 0 seconds to 4200 seconds. Ballooning and rupture during the heat-up process did not occur in the case of test rods with a pinhole. While the existence of the pinhole affected neither the inner-surface-oxidation behavior nor the fracture boundary of the cladding tube, the shape and size of the opening including the pinhole and the rupture opening might affect the rate and amount of steam entering the test rod. Initially injected water affected the inner-surface-oxidation behavior. While the fracture boundary of the test rods with initially injected water tended to be higher than that of the test rods without injected water at short oxidation time, vice versa at longer oxidation time. This tendency may be related to the amount of the steam which remained in or entered the test rod during the test.

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

EBSD解析によるステンレス鋼の疲労損傷検出; EBSDパターンクオリティの適用性

黒田 雅利*; 釜谷 昌幸*; 山田 輝明*; 秋田 貢一

日本機械学会論文集(インターネット), 83(852), p.17-00072_1 - 17-00072_7, 2017/07

これまでに、電子後方散乱回折(EBSD)測定により得られる結晶方位データを解析することで得られる平均局所方位差(Averaged Local Misorientation, Mave)(Kamaya, 2009)とオーステナイト系ステンレス鋼の低サイクル疲労の損傷量との間に相関があることを報告している。ここでは、実用上の観点から、Maveと、市販ソフトウエアで得られるイメージクオリティ(Image Quality: IQ)とを比較することで、IQの疲労損傷評価に対する適用性について検討し、各パラメータの有用性と特徴を明らかにした。また、X線回折(X-Ray Diffraction: XRD)測定も実施し、得られたXRDデータとIQ値分布とを比較することで、疲労損傷の蓄積に伴うIQ値分布の変化はすべり変形によりもたらされたことを示した。

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

The Effect of oxidation-and-quenching process during a LOCA on the behavior of the oxidation and embrittlement of Zircaloy-4 cladding under reheating transients

三輪 英紀; 天谷 政樹

Journal of Nuclear Science and Technology, 53(12), p.2090 - 2097, 2016/12

 被引用回数:3 パーセンタイル:42.29(Nuclear Science & Technology)

冷却材喪失事故時の高温酸化や急冷却が、長期炉心冷却機能喪失時の酸化・脆化挙動に与える影響を確認するために、急冷後再昇温を模擬した温度推移条件で両面酸化試験、リング圧縮試験およびクエンチ試験を実施した。未照射ジルカロイ-4被覆管の試料を水蒸気流中で1173Kから1473Kの温度で酸化し、その後に室温の冷却水にて急冷した。再昇温試験はこの試料を用いて水蒸気流中で1173Kから1473Kの温度で実施した。特定の急冷後再昇温の条件では酸化膜成長や重量増加の低減が見られた。しかしながら、急冷後再昇温を含む温度推移が被覆管脆化に与える影響は顕著でなかった。急冷後再昇温を含む温度推移条件下での被覆管の脆化挙動はBaker-Justの式を用いて算出したECRで評価できることが分かった。

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:6 パーセンタイル:17.95(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.

報告書

Zircaloy-4の高温酸化挙動に及ぼす固体ホウ酸の影響

小宮山 大輔; 天谷 政樹

JAEA-Research 2016-013, 20 Pages, 2016/08

JAEA-Research-2016-013.pdf:6.05MB

PWRの冷却材喪失事故(LOCA)において、流路の閉塞等により燃料棒の冷却が十分に行われない場合、燃料被覆管表面に冷却材中のホウ酸が析出する可能性が考えられる。通常運転温度域では、実機での実績からホウ酸水はZircaloy-4の酸化挙動に影響を及ぼさないと考えられるが、LOCAを想定した高温域におけるホウ酸とZircaloy-4との反応に係る知見は十分に得られていない。本研究では、固体ホウ酸を載せたZircaloy-4の板材を900$$^{circ}$$Cまでの温度及び複数の雰囲気で酸化させることにより、固体ホウ酸の高温時挙動、ホウ酸とZircaloy-4との反応の有無、及びホウ酸がZircaloy-4の酸化挙動に及ぼす影響を調べた。実験結果から、高温酸化雰囲気においてZircaloy-4表面に固体ホウ酸の脱水により生成する無水ホウ酸が存在すると、この無水ホウ酸がZircaloy-4と雰囲気との接触を断つことでZircaloy-4の酸化を抑制することが示唆された。また、酸化膜付きZircaloy-4の表面に固体ホウ酸が付着し高温まで加熱された場合は、形成している酸化膜の空隙に無水ホウ酸が浸透することでその後の酸化を抑制することがうかがえた。

論文

Benchmark analyses of probabilistic fracture mechanics for cast stainless steel pipe

北条 公伸*; 林 翔太郎*; 西 亘*; 釜谷 昌幸*; 勝山 仁哉; 眞崎 浩一*; 永井 政貴*; 岡本 年樹*; 高田 泰和*; 吉村 忍*

Mechanical Engineering Journal (Internet), 3(4), p.16-00083_1 - 16-00083_16, 2016/08

鋳造ステンレス鋼に対する非破壊検査が計画されているが、鋳造ステンレス鋼のような二相ステンレス鋼では、超音波の低い透過性などの理由から、許容欠陥寸法が定められていない。鋳造ステンレス鋼の許容欠陥寸法を合理的に決定するためには、確率論的破壊力学(PFM)は有用である。本研究では、鋳造ステンレス鋼配管を対象に、PFM解析コードの適用性や信頼性に係るベンチマーク問題を提案した。破損モードとしては、疲労亀裂進展、塑性崩壊、及び延性亀裂進展を考慮し、それらの相互作用を考慮した条件でPFM解析を行った。6機関が参加して実施されたベンチマーク解析による破損確率の比較を行った。その結果、各機関で様々なPFM解析コードで得られた破損確率はよく一致し、鋳造ステンレス鋼配管に対するPFMの適用性が確認された。

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:1 パーセンタイル:76.09(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:57.25(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.

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