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fuels with additives under RIA conditions三原 武
no journal, ,
JAEA have obtained regulatory data for advanced or further high burnup fuels in order to provide data and knowledge on fuel behavior needed for safety regulation. Two RIA tests on test rods with doped UO pellets (LS-4 and OS-1) which had been irradiated in commercial BWRs were successfully performed at room temperature condition. The doped UO pellets were designed to have a larger grain size compared to standard UO fuel with adding additives. The results of LS-4 and OS-1 tests showed no failure and failure, respectively. The fuel enthalpy increase at failure of OS-1 was lower than those obtained in the failure cases of the fuel rods with RX cladding in a cladding hydrogen content of 200-300 ppm. The result of transient fission gas release during LS-4 test suggested an effect of the large-grained fuel pellet. According to the post-test appearance and hydride morphology in the cladding of the reference rod of OS-1, one of the reasons why the OS-1 showed the low fuel enthalpy increase at failure may be the decrease in the failure limit of the cladding tube due to the hydrides precipitated and radially oriented in the tube wall.
塩津 弘之; 伊藤 裕人*; 杉山 智之; 丸山 結
no journal, ,
水蒸気雰囲気での大規模FP放出・移行挙動実験であるVERDON-2実験について、化学平衡を仮定するFP移行挙動評価コードVICTORIA2.0を用いた解析を実施し、平衡論に基づいたCsおよびIに係る化学挙動予測機能の評価を行った。解析の結果、平衡論により高温領域でのCs化学挙動を定性的に評価できることを明らかにした。一方で、Iや低温領域でのCs化学挙動では実験結果との定性的な一致が見られず、手法に課題があることが示された。
Li, F.
no journal, ,
The effect of hydride morphology on the failure behavior of cladding tube was investigated using biaxial-EDC testing method. Cold-worked, stress-relieved and recrystallized Zircaloy-4 tubes with pre-crack were used as test specimens. The specimens were pre-hydrided. A decrease in the failure strain (hoop strain at failure) with increasing hydrogen content was observed at stress-relieved samples. The failure strain of the sample was more sensitive to hydrogen content under higher strain ratio. Micro-crack formation in hydrides is considered to be the microscopic failure mechanism of the decrease in failure strain. A decrease in the failure strain with increasing hydrogen content was observed similarly at cold-worked samples. No significant decrease in failure strain with increasing hydrogen content (up to about 280 ppm) was observed at recrystallized samples. This is probably because the enhanced ductility of recrystallized samples suppressed the effect of micro-cracks formed in hydrides on the decrease in failure strain.
天谷 政樹
no journal, ,
原子力機構における燃料安全研究の目的は、軽水炉燃料に関する現在の規制基準の妥当性評価、新しい燃料材料からなる改良型燃料に関する規制のための技術的知見の取得整備、及び規制に活用可能な燃料挙動解析技術の開発、等である。本発表では、原子力機構における最近の反応度事故模擬試験、冷却材喪失事故模擬試験、及び燃料挙動解析コード改良の状況について報告する。
山下 真一郎; 井岡 郁夫; 根本 義之; 川西 智弘; 加治 芳行; 深堀 智生; 野澤 貴史*; 渡部 清一*; 村上 望*; 佐藤 寿樹*; et al.
no journal, ,
In order to increase accident tolerance of light water reactors (LWRs), fuel rod, channel box and control rod with new materials and concepts have been considered and developed in Japan. Since 2015, Japan Atomic Energy Agency has conducted and coordinated the Japanese R&D program of accident tolerant fuel (ATF) for establishing technical basis of ATF under a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). ATF candidate materials considered in this METI program are silicon carbide (SiC) composite and FeCrAl steel strengthened by dispersion of fine oxide particles (FeCrAl-ODS). SiC composite is a highly attractive material because of its lower hydrogen generation rate and lower reaction heat in comparison with conventional Zircaloy. Therefore, practical uses for a fuel cladding of pressurized water reactor (PWR) and for the fuel cladding, channel box of boiling water reactor (BWR) are expected. On the other hand, FeCrAl-ODS steel is a promising material and is considered to apply to the fuel cladding of BWR. Until now, we have been accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the candidate materials, evaluated fuel behavior simulating operational and accidental conditions by the developed code. In this paper, we will report the updates of out-of-pile data and evaluation results.
逢坂 正彦
no journal, ,
原子力機構において実施している軽水炉シビアアクシデント時の核分裂生成物化学に関する基礎研究の成果を報告する。
岡田 裕史
no journal, ,
It is important to evaluate the fracture resistance of fuel rod against a seismic loading following a LOCA since the accident at the Fukushima Daiichi Nuclear Power Plant. In consideration of this, the effects of oxidation and secondary hydriding on the fracture resistance were investigated by four point bending test using the cladding tube with drilled holes. Based on the obtained results, it was suggested that the oxygen concentration in prior-beta layer affects the maximum bending stress. The maximum bending stress related to the secondary hydriding was estimated as about a half compared with that at the rupture opening position.
山内 紹裕*; 天谷 政樹
no journal, ,
DSCを用いて、冷間加工、応力除去焼きなまし、再結晶焼きなましZry-4の水素固溶限を測定した。加熱中の水素固溶に対応する固溶限は冷間加工材が最大であり、次いで応力除去焼きなまし材、再結晶焼きなまし材の順であることがわかった。冷却中の水素化物析出に対応する固溶限は各試料間で同等であった。これは、DSC測定中に試料が加熱されたことで、被覆管製造時の格子欠陥が回復したことが原因と考えられる。本試験の結果より、被覆管製造工程における熱処理条件が水素化物の固溶及び析出挙動に影響を及ぼすことが示唆された。
Negyesi, M.
no journal, ,
This study deals with high-temperature steam oxidation behavior of Zry-4 fuel cladding. The experimental conditions correspond to the loss of coolant accident (LOCA) of LWRs. Oxidation tests were carried out at 1273 K employing two experimental techniques. The effect of the heat-up conditions on the subsequent isothermal oxidation rate were investigated. Furthermore, the effect of specimen surface roughness on the oxidation behavior was treated. The oxidation rate during the isothermal exposure was substantially suppressed at higher degrees of oxidation upon the heat-up. Consequently, the well-known "breakaway" effect, which results in accelerated oxidation rate, was found to be significantly delayed. The effect of the surface roughness on the oxidation behavior appeared to be rather minor under the test conditions of this study. The effect of the degree of oxidation upon the heat-up showed to be crucial and needs to be further addressed when evaluating the "breakaway" effect.
宇田川 豊
no journal, ,
原子力機構安全研究センターが取り組んでいる安全研究の内、反応度事故時の燃料挙動に関する研究の状況と最近の成果について概説する。特に2018年実施した高燃焼度低スズZIRLO被覆燃料を対象とした高温条件下のRIA模擬実験、最近実施された高燃焼度燃料実験の照射後試験結果を中心に紹介し、応力除去焼鈍タイプの被覆管については破損限界に関する従来の理解とこれに基づく予測の妥当性が改めて確認されたこと、核分裂生成物ガスの過渡放出挙動に及ぼす破損時燃料エンタルピや拘束力の影響が見られること等、最新の実験結果に基づく考察を示す。この他、原子力機構における機械特性試験を用いた分離効果実験技術の発展、燃料挙動解析コードの開発、検証の進捗についても紹介する。
成川 隆文
no journal, ,
JAEA has carried out studies on fuel behaviors under loss-of-coolant-accident (LOCA) conditions with both unirradiated and high-burnup advanced fuel cladding tube materials. As a result, various kinds of information have been obtained: oxidation, ballooning, rupture, thermal shock resistance (fracture/non-fracture conditions), post-LOCA mechanical strength, etc. In addition, new LOCA tests are planned at JAEA to investigate the effects of the phenomena of fuel fragmentation, relocation and dispersal (FFRD) on the fuel behaviors and the coolability of the reactor core during a LOCA. These results, including those obtained from the future study, are expected to provide the necessary information for future regulation on high-burnup fuels with advanced cladding alloys.
谷口 良徳
no journal, ,
欧州で高燃焼度まで照射されたM5被覆MOX燃料について、反応度事故(RIA)時における安全評価に必要なデータ及び知見を取得するため、NSRRを用いたRIA模擬実験(CN-1実験)を実施した。実験後の燃料棒の外観、過去報告されているVVER燃料のRIA時破損形態の特徴等から、CN-1実験に供した試験燃料棒は燃料ペレット-被覆管相互作用(PCMI)破損よりも高温時の破損形態である膨れ破損であることが示唆された。PCMI破損を示さなかった原因の一つとして、通常運転中のM5被覆管の腐食量が少ないことが考えられた。