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論文

LOCA時燃料破断限界評価の信頼性向上を目指して; 不確かさ定量化手法の開発と高燃焼度化の影響評価

成川 隆文

日本原子力学会誌ATOMO$$Sigma$$, 63(11), p.780 - 785, 2021/11

冷却材喪失事故時の軽水炉燃料被覆管の破断限界評価の信頼性向上を目指した原子力機構の取り組みとして、ベイズ統計手法による不確かさの定量化手法の開発、並びに燃焼の進展及び被覆管材質の変更の影響評価に関する研究を紹介する。

論文

Steam oxidation of silicon carbide at high temperatures for the application as accident tolerant fuel cladding, an overview

Pham, V. H.; 倉田 正輝; Steinbrueck, M.*

Thermo (Internet), 1(2), p.151 - 167, 2021/09

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000$$^{circ}$$C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600$$^{circ}$$C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

報告書

軽水型動力炉の非常用炉心冷却系の性能評価指針の技術的根拠と高燃焼度燃料への適用性

永瀬 文久; 成川 隆文; 天谷 政樹

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

軽水炉においては、冷却系配管破断等による冷却材喪失事故(LOCA)時にも炉心の冷却可能な形状を維持し放射性核分裂生成物の周辺への放出を抑制するために、非常用炉心冷却系(ECCS)が設置されている。ECCSの設計上の機能及び性能を評価し、評価結果が十分な安全余裕を有することを確認するために、「軽水型動力炉の非常用炉心冷却系の性能評価指針」が定められている。同指針に規定されている基準は1975年に定められた後、1981年に当時の最新知見を参考に見直しが行われている。その後、軽水炉においては燃料の高燃焼度化及びそれに必要な被覆管材料の改良や設計変更が進められたが、それに対応した指針の見直しは行われていない。一方、高燃焼度燃料のLOCA時挙動や高燃焼度燃料への現行指針の適用性に関する多くの技術的な知見が取得されてきている。本報告においては、我が国における指針の制定経緯及び技術的根拠を確認しつつ、国内外におけるLOCA時燃料挙動に係る最新の技術的知見を取りまとめる。また、同指針を高燃焼度燃料に適用することの妥当性に関する見解を述べる。

論文

Transient response of LWR fuels (RIA)

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.

論文

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

 被引用回数:2 パーセンタイル:62.87(Nuclear Science & Technology)

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$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, 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.

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:0.01(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:0.01(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

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01

 被引用回数:1 パーセンタイル:23.13(Nuclear Science & Technology)

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$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, 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.

論文

Behavior of high-burnup advanced LWR fuel cladding tubes under LOCA conditions

成川 隆文; 天谷 政樹

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09

To evaluate behavior of high-burnup advanced light-water-reactor fuel cladding tubes under loss-of-coolant accident conditions, laboratory-scale isothermal oxidation tests and integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73-85 GWd/t: M-MDA$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5textregistered, and Zircaloy-2 (LK3). The isothermal oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube and was slower than that given by the Baker-Just oxidation rate equation. Therefore, the oxidation kinetics is considered to be not significantly accelerated by extending the burnup and changing the alloy composition. During the integral thermal shock tests, the high-burnup advanced fuel cladding tube specimens did not fracture under the oxidation condition equivalent to or lower than the fracture limit of the unirradiated Zircaloy-4 cladding tube. Therefore, the fracture limit of fuel cladding tubes is considered to be not significantly reduced by extending the burnup and changing the alloy composition, though it may slightly decrease with increasing initial hydrogen concentration.

論文

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

天谷 政樹; 垣内 一雄; 三原 武

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

New fuel cladding alloys of which composition was changed from conventional ones have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rod and/or fuel assembly during normal operation, the irradiation growth behavior of the improved Zr-based alloys for light-water reactor fuel cladding was investigated. The coupon specimens were prepared from fuel cladding tubes with various kinds of improved Zr-based alloys. The specimens were loaded into test rigs and had been irradiated in the Halden reactor in Norway under several coolant temperature conditions up to a fast-neutron fluence of $$sim$$7.8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$ 1 MeV). Irradiation conditions such as specimen temperatures had been continuously monitored during the irradiation. During and after the irradiation, the amount of irradiation growth of each specimen was evaluated as a part of the interim and final inspections. The effect of the difference in alloy composition on the amount of irradiation growth seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication and the irradiation temperature were the same.

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

Overview of accident-tolerant fuel R&D program in Japan

山下 真一郎; 井岡 郁夫; 根本 義之; 川西 智弘; 倉田 正輝; 加治 芳行; 深堀 智生; 野澤 貴史*; 佐藤 大樹*; 村上 望*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09

福島第一原子力発電所事故を教訓に、冷却材喪失等の過酷条件においても損傷しにくく、高い信頼性を有する新型燃料の開発への関心が高まり、世界中の多くの国々において事故耐性を高めた新型燃料の研究開発が進められている。本プロジェクトは、経済産業省資源エネルギー庁からの委託を受けて2015年10月から2019年3月までの3年半の間実施され、新型燃料部材を既存軽水炉に装荷可能な形で設計・製造するために必要となる技術基盤を整備することを目的に、国内の軽水炉燃料設計,安全性評価,材料開発を実施してきた人材,解析ツール,ノウハウ、及び経験を最大限活用して進められてきた。本論文では、プロジェクトの総括として、各要素技術について3年半の研究開発の成果をまとめ、日本の事故耐性燃料開発の現状と課題を整理した。

論文

Benchmark of fuel performance codes for FeCrAl cladding behavior analysis

Pastore, G.*; Gamble, K. A.*; Cherubini, M.*; Giovedi, C.*; Marino, A.*; 山路 哲史*; 加治 芳行; Van Uffelen, P.*; Veshchunov, M.*

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1038 - 1047, 2019/09

耐酸化性FeCrAl鋼が軽水炉の事故耐性燃料用被覆管として提案されている。IAEAの研究プロジェクトの一環として、FeCrAl被覆管挙動に関する燃料ふるまいモデリングのベンチマークを実施した。この中で、FeCrAl被覆管材料の燃料棒問題に対して種々の燃料ふるまいコードを用いて計算を行い結果を相互に比較した。

論文

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

 被引用回数:6 パーセンタイル:81.33(Nuclear Science & Technology)

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective effect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.

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:3 パーセンタイル:57.07(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

Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

湯村 尚典; 天谷 政樹

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

 被引用回数:3 パーセンタイル:45.99(Nuclear Science & Technology)

To investigate the relationship between the fracture resistance of a cladding tube and the amount of deformation of the cladding tube due to ballooning and rupture during a loss-of-coolant accident (LOCA), four-point-bending tests were performed using non-irradiated Zircaloy-4 cladding tubes which experienced a LOCA-simulated sequence (ballooning, rupture, high temperature oxidation and quench). According to the obtained results, it was found that the maximum bending stress of the cladding tube after the LOCA-simulated sequence, which was defined as the fracture resistance, correlated to the average thickness of prior-$$beta$$ layer in the cladding tube. Based on the average thickness of prior-$$beta$$ layer, the fracture resistance of the cladding tube with ballooning and rupture was expressed as functions of isothermal oxidation time and temperature and the maximum circumferential strain on the cladding tube.

論文

Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions with several temperatures

高畠 容子; 安倍 弘; 佐野 雄一; 竹内 正行; 小泉 健治; 坂本 寛*; 山下 真一郎

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10

事故耐性軽水炉燃料の燃料被覆管として開発されているFeCrAl-ODS鋼の硝酸腐食評価を、使用済燃料再処理工程に対して燃料被覆管腐食生成物が与える影響を評価するために実施した。3mol/L硝酸における腐食試験を、60$$^{circ}$$C, 80$$^{circ}$$C,沸騰条件において実施し、浸漬試験の試験片に対してはXPS分析を行った。沸騰条件にて最も腐食が進展し、腐食速度は0.22mm/yであった。酸化被膜内のFe割合は減少しており、CrとAlの割合は増加していた。腐食試験の結果、FeCrAl-ODS鋼は高い硝酸腐食耐性を持つため、再処理工程中の溶解工程において許容可能であることを確かめた。

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

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

上羽 智之; 根本 潤一*; 石谷 行生*; 伊藤 昌弘*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

 被引用回数:3 パーセンタイル:45.99(Nuclear Science & Technology)

高速炉MOX燃料ピンの照射挙動を計算するコードと燃料ピン内のCsの挙動に特化して計算するコードを連成することにより、Cs挙動が燃料ピンの熱・機械的挙動に及ぼす影響を解析できるようにした。連成した計算コードを高燃焼度MOX燃料ピンの照射挙動解析に適用し、Csの燃料ピン内軸方向分布やCs化合物による燃料ペレットと被覆管の機械的相互作用などを評価した。

論文

Oxidation kinetics of Zry-4 fuel cladding in mixed steam-air atmospheres at temperatures of 1273 - 1473 K

Negyesi, M.; 天谷 政樹

Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10

 被引用回数:5 パーセンタイル:58.67(Nuclear Science & Technology)

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam_air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0 up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation applicable for thermomechanical analysis codes of nuclear power reactor under severe accidents. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transient and post-transient regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transient regime.

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