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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

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

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

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

The reduction of epistemic uncertainty for safety-related events that rarely occur or require high experimental costs is a key concern for researchers worldwide. In this study, we develop a new framework to effectively reduce parameter uncertainty, which is one of the epistemic uncertainties, by using the Bayesian optimal experimental design. In the experimental design, we used a decision theory that minimizes the Bayes generalization loss. For this purpose, we used the functional variance, which is a component of widely applicable information criterion, as a decision criterion for selecting informative data points. Then, we conducted a case study to apply the proposed framework to reduce the parameter uncertainty in the fracture boundary of a non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimen under loss-of-coolant accident (LOCA) conditions. The results of our case study proved that the proposed framework greatly reduced the Bayes generalization loss with minimal sample size compared with the case in which experimental data were randomly obtained. Thus, the proposed framework is useful for effectively reducing the parameter uncertainty of safety-related events that rarely occur or require high experimental costs.

論文

平成29年度核燃料部会賞(学会講演賞)を受賞して,1

成川 隆文

核燃料, (53-2), P. 5, 2018/08

日本原子力学会2017年秋の大会における発表「非照射ジルカロイ-4被覆管のLOCA時破断限界の不確かさ評価」が評価され、同学会の平成29年度核燃料部会賞(学会講演賞)を受賞した。今回の受賞に関する所感を同部会報に寄稿する。

論文

Uncertainty quantification of fracture boundary of pre-hydrided Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Nuclear Engineering and Design, 331, p.147 - 152, 2018/05

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

To quantify the fracture boundary uncertainty for non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimens under loss-of-coolant accident conditions at a light-water reactor, data from integral thermal shock tests obtained by an earlier study are analyzed statistically and the fracture boundary is estimated in terms of probability, as follows. First, a method is proposed to obtain the specimens' fracture probability curve as a function of equivalent cladding reacted (ECR) and initial hydrogen concentration using Bayesian inference with a generalized linear model. A log-probit model is used, modified to reflect the effect of the initial hydrogen concentration on the fracture boundary and the ECR evaluation uncertainty, and scaled to improve convergence. Second, using the modified log-probit model, it is shown that the boundary representing a 5% fracture probability with 95% confidence for the pre-hydrided cladding tube sample is higher than 15% ECR, for initial hydrogen concentrations of up to 800 wppm.

論文

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Journal of Nuclear Materials, 499, p.528 - 538, 2018/02

 被引用回数:2 パーセンタイル:16.17(Materials Science, Multidisciplinary)

For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best model to estimate the fracture probability. It was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

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

The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

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

In order to investigate the effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 (Zry-4) cladding tube, laboratory-scale experiments on non-irradiated Zry-4 cladding tube specimens were performed under transient-heating conditions which simulate loss-of-coolant-accident (LOCA) conditions by using an external heating method, and the data obtained were compared to those from a previous study where an internal heating method was used. The maximum circumferential strains ($$varepsilon$$s) of the cladding tube specimens were firstly divided by the engineering hoop stress ($$sigma$$). The divided maximum circumferential strains, ${it k}$s, of the previous study, which used the internal heating method, were then corrected based on the azimuthal temperature difference (ATD) in the cladding tube specimen. The ${it k}$s for the external heating method which was used in this study agreed fairly well with the corrected ${it k}$s obtained in the previous study which employed the internal heating method in the burst temperature range below $$sim$$1200 K. Also, the area of rupture opening tended to increase with increasing of the value which is defined as $$varepsilon$$ multiplied by $$sigma$$. From the results obtained in this study, it was suggested that $$varepsilon$$ and the size of rupture opening of a cladding tube under LOCA-simulated conditions can be estimated mainly by using $$sigma$$, $$varepsilon$$ and ATD in the cladding tube specimen, irrespective of heating methods.

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

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

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

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

In order to investigate the effect of oxidation and crystal phase condition on the ballooning and rupture behaviors of cladding tube under simulated loss-of-coolant-accident (LOCA) conditions, laboratory-scale experiments were performed in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) cladding specimens were heated to burst in steam and argon gas conditions. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. The dependence of the normalized value on burst temperature and the relationship between the normalized value and the length, width and area of rupture opening were evaluated. The correlation between the normalized value and the burst temperature suggested that the fraction of the $$beta$$ phase in Zry-4 cladding specimens affected the strain in the specimens and the oxidation of specimens suppressed the amount of ballooning of the specimens. The relationship between the normalized value and the length, width and area of rupture opening indicated that the length, width and area of rupture opening depended on the crystal phase condition in Zry-4 cladding specimens irrespective of atmosphere in the case of the heating rate of $$sim$$3 K/s.

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。

論文

Ballooning and rupture behavior of Zircaloy-4 cladding under transient-heating conditions

成川 隆文; 天谷 政樹

Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 7 Pages, 2014/09

Phenomena of fuel fragmentation, relocation and dispersal have been observed in several experiments on very-high-burnup fuels under simulated loss-of-coolant-accident (LOCA) conditions using a test reactor. In order to improve the prediction of the phenomena, ballooning and rupture behaviors of cladding under simulated LOCA conditions were investigated by performing laboratory-scale experiments in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) claddings were heated to burst. The maximum circumferential strains of the ballooned claddings were strongly dependent on burst temperature and the trends seemed to depend on the heating rate in the experiment. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. A correlation between the normalized values and the burst temperatures suggests that the fraction of $$beta$$ phase in Zry-4 affects the extent of the strain of cladding ballooning.

口頭

リスク情報を活用した意思決定における低頻度高影響度事象の評価手法; 事故時の燃料挙動に関する研究課題検討への利用

成川 隆文; 杉山 智之; 天谷 政樹

no journal, , 

原子力安全規制にかかわるリスク情報活用においては、低頻度高影響度事象 に対処し、原子炉施設のリスクを抑制するために、深層防護思想とPRA(確率論的リスク評価)に代表されるリスク評価手法をいかに融合させるかが課題となっている。そこで本研究では、特にリスク情報の活用が進む安全確保活動の変更におけるリスク情報を活用した意思決定において、PRAを深層防護の有効性を測る尺度として用いて、リスクの観点から深層防護の有効性を評価する指標を提案するとともに、この指標の有用性に基づく知見から事故時の燃料挙動に関する研究課題を検討した。その結果、提案指標により深層防護の各防護層においてリスク寄与因子を評価することが可能となり、本指標が低頻度高影響度事象を評価するうえで有効であることが確認された。また、本指標の有用性に基づく知見から、設計基準事故を超えた後、燃料が実際に破損、溶融に至るまでの限界値を知ることで、リスク評価を精緻化することが重要であるとの結論を得た。

口頭

Investigation of the occurrence of a load spike in LOCA quench experiment

Wu, H.; 宇田川 豊; 成川 隆文; 天谷 政樹

no journal, , 

The cause of a load spike which is sometimes observed at the time of quench in one-sided oxidation Loss-of-coolant accident (LOCA) quench experiments was investigated. The experimental results implied that it could be related to the pressure fluctuation occurred in the experimental system during the experiment, and two solutions were proposed and verified.

口頭

LOCA模擬実験時の加熱方法がZircaloy-4被覆管の膨れ及び破裂挙動に及ぼす影響

成川 隆文; 天谷 政樹

no journal, , 

未照射ジルカロイ-4被覆管を対象とし、冷却材喪失事故(LOCA)を模擬する試験時の加熱方法が同試験時の被覆管の膨れ及び破裂挙動に及ぼす影響を調べた。被覆管の最大周方向歪みを破裂直前の内圧に基づく公称応力で規格化した値で整理したところ、その絶対値は加熱方法の違いよりも加熱時の試験燃料棒の周方向温度分布の影響を受けること、また、破裂開口部の面積は加熱方法の違いの影響を受ける可能性のあることが示唆された。

口頭

社会と共存する魅力的な軽水炉が有するべき特性

山本 章夫*; 成川 隆文; 堺 紀夫*

no journal, , 

平成26年12月に設立された「社会と共存する魅力的な軽水炉の展望」調査専門委員会は、東京電力福島第一原子力発電所事故、並びに2015年に策定されたエネルギー基本計画及び軽水炉安全技術・人材ロードマップ等を踏まえ、エネルギー政策の基本視点である3E+S(エネルギー安定供給、経済性、環境適合性+安全性)と調和する軽水炉の在り方(軽水炉が備えるべき社会的受容性及び技術的特性)に関する調査・研究を実施し、軽水炉の設計を基本に立ち返って考え、さらに若年世代あるいは他学術・技術分野の研究者・技術者や学生が挑戦に値する魅力的な展望を描くことを目的としている。本企画セッションでは、本委員会の概要に加え、これまでに検討してきた社会的受容性とそれから展開される基本要件、及び技術的論点と検討経緯について報告する。

口頭

Status and plan of LOCA study at JAEA

成川 隆文

no journal, , 

JAEA has conducted studies on fuel behaviors under loss-of-coolant-accident (LOCA) conditions with both unirradiated and high-burnup advanced fuel cladding tubes. As a result, various kinds of information have been obtained on behaviors of these cladding tubes under LOCA conditions: oxidation, ballooning and rupture, thermal shock resistance (fracture/non-fracture conditions), post-LOCA mechanical strength, etc. In addition, new LOCA tests are planned at JAEA for the purpose of investigating effects of phenomena of fuel fragmentation, relocation and dispersal (FFRD) on fuel behaviors and coolability of reactor core during LOCA. It is expected that these results including those obtained by the future study provide necessary information for future regulation on high-burnup fuels with advanced cladding alloys.

口頭

社会と共存する魅力的な軽水炉が有するべき特性,3; 社会的受容性とそれから展開される基本要件

成川 隆文

no journal, , 

日本原子力学会に平成26年12月に設立された「社会と共存する魅力的な軽水炉の展望」調査専門委員会は、東京電力福島第一原子力発電所事故、並びに2015年に策定されたエネルギー基本計画及び軽水炉安全技術・人材ロードマップ等を踏まえ、エネルギー政策の基本視点であるS+3E(安全性+エネルギー安定供給、経済性、環境適合性)と調和する軽水炉の在り方(軽水炉が備えるべき社会的受容性及び技術的特性)に関する調査・研究を実施し、軽水炉の設計を基本に立ち返って考え、さらに若年世代あるいは他学術・技術分野の研究者・技術者や学生が挑戦に値する魅力的な展望を描くことを目的としている。本シンポジウムでは、これまでに検討してきた社会的受容性とそれから展開される基本要件について報告する。

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