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1
Identification of penetration path and deposition distribution of radionuclides in houses by experiments and numerical model
廣内 淳; 高原 省五; 飯島 正史; 渡邊 正敏; 宗像 雅広
Radiation Physics and Chemistry, 140, p.127 - 131, 2017/11
The dose assessment for people living in preparation zones for the lifting of the evacuation order is needed with the return of the residents. However, it is difficult to assess exactly indoor external dose rate because the indoor distribution and infiltration pathways of radionuclides are unclear. This paper describes indoor and outdoor dose rates measured in eight houses in the difficult-to-return zone in Fukushima prefecture to examine the distribution of radionuclides in a house and the main infiltration pathway of radionuclides. In addition, it describes also dose rates calculated with a Monte Carlo photon transport code to understand thoroughly the measurements. These measurements and calculations provide that radionuclides can infiltrate mainly through ventilations, windows, and doors, and then deposit near the gaps, while those infiltrate hardly through sockets and air conditioning outlets.
2
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
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.
3
JASMINE Version 3による溶融燃料-冷却材相互作用SERENA2実験解析
堀田 亮年*; 森田 彰伸*; 梶本 光廣*; 丸山 結
日本原子力学会和文論文誌, 16(3), p.139 - 152, 2017/09
Among twelve FCI cases conducted in the OECD/NEA/CSNI/SERENA2 test series using two facilities, six steam explosion cases, five from TROI and one from KROTOS, were analyzed by JASMINE V.3. Major model parameters were categorized into "focused zone", a core part of interest, and "peripheral zone", the initial and boundary conditions given intentionally for each test case. For the former, base values established through past validation studies of JASMINE V.3 were applied. The code was modified to implement the measured distribution of entrained droplet size acquired in TROI-VISU. For the latter, melt release histories were given as a combination of time tables of jet diameter and release velocity that were estimated based on image data and transit timing data of the melt leading edge. The base values were shown to predict impulse responses of SERENA2 systematically with a reasonable error band. A statistical analysis based on the LHS method was performed. Uncertainty ranges were given based on measurement errors and past validation studies in the JASMINE development. Underlying mechanisms causing apparent differences in the mechanical energy conversion ratio between two facilities were studied from the view point of breakup length and trigger timing.
4
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.
5
ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test
竹田 武司; 大津 巌
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
An experiment using PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with LSTF on a cold leg small-break loss-of-coolant accident with an accident management measure in a PWR. The rate of steam generator secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.
6
Uncertainty evaluation of seismic response of a nuclear facility using simulated input ground motions
崔 炳賢; 西田 明美; 村松 健*; 高田 毅士*
Proceedings of 12th International Conference on Structural Safety & Reliability (ICOSSAR 2017) (USB Flash Drive), p.2206 - 2213, 2017/08
本稿では、モデル化手法の違いが原子力施設の地震応答解析結果のばらつきに与える影響を明らかにするため、多様な模擬入力地震動を用いた地震応答解析を実施し、応答のばらつきの統計的分析を行った。特に、建屋せん断壁の最大加速度応答に着目し、モデル化手法による応答結果への影響、応答のばらつき要因について分析を行い、得られた知見について報告する。
7
原子力緊急時支援・研修センターの活動; 平成27年度
原子力緊急時支援・研修センター
JAEA-Review 2017-011, 54 Pages, 2017/07
日本原子力研究開発機構(JAEA)は、災害対策基本法及び武力攻撃事態対処法に基づき、「指定公共機関」として、国及び地方公共団体その他の機関に対し、災害対策又は武力攻撃事態等への対処において、JAEAの防災業務計画及び国民保護業務計画に則り、技術支援をする責務を有している。原子力緊急時支援・研修センター(NEAT)は、緊急時には、全国を視野に入れた専門家の派遣、防災資機材の提供、防護対策のための技術的助言等の支援活動を行う。また、平常時には、我が国の防災対応体制強化・充実のために、自らの訓練・研修のほか、国、地方公共団体の原子力防災関係者のための実践的な訓練・研修、原子力防災に関する調査研究及び国際協力を実施する。平成27年度、NEATでは、日本原子力研究開発機構の新たな第3期中期計画に基づき、以下の業務を推進した。(1)NEATの基盤整備及び運営体制の維持、(2)機構内専門家の研修及び支援活動訓練の企画実施並びに国、地方公共団体の原子力防災関係者の人材育成及び研修・訓練、(3)原子力防災に係る調査・研究の実施及び情報発信、(4)国が実施する緊急時の航空機モニタリングへの支援についての必要な準備の実施、(5)国際機関と連携を図ったアジア諸国への原子力防災に係る技術的貢献
8
Verification of probabilistic fracture mechanics analysis code PASCAL through benchmark analyses with FAVOR
Li, Y.; 宇野 隼平*; 勝山 仁哉; Dickson, T.*; Kirk, M.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by the Japan Atomic Energy Agency to evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events base on Japanese data and Japanese methods published for or prescribed in the Japanese regulations and standards. To verify this code, benchmark analyses were carried out with FAVOR code which was developed in United States and has been utilized in nuclear regulation. Through the benchmark analyses, the applicability of PASCAL in failure frequency evaluation of Japanese RPVs was confirmed with great confidence. The outline of PASCAL, the benchmark analysis conditions and results are provided in this paper.
9
Probabilistic fracture mechanics analysis models for Japanese reactor pressure vessels
Lu, K.; 勝山 仁哉; 宇野 隼平; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessments of reactor pressure vessels (RPVs) by considering the inherent probabilistic distributions of various influence factors. For practical applications, several evaluation models are improved, and have been implemented into the current PASCAL code. In this paper, the improvements of PASCAL are introduced firstly, such as the evaluation method for underclad cracks, treatments of the complicated welding residual stress distribution, and evaluation models for the warm pre-stressing effect. In addition, the effects of these improvements on failure probability or failure frequency of RPVs are investigated by performing PFM analyses for domestic RPVs using PASCAL. From the analysis results, the effects of the improved evaluation models are discussed.
10
Guideline on probabilistic fracture mechanics analysis for Japanese reactor pressure vessels
勝山 仁哉; 小坂部 和也*; 宇野 隼平; Li, Y.; 吉村 忍*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
確率論的破壊力学(PFM)に基づく構造健全性評価手法は、経年劣化に関連する様々な因子の確率分布を考慮して原子炉圧力容器(RPV)の破損頻度を評価できる合理的な手法である。我々は、中性子照射脆化や加圧熱衝撃事象(PTS)を考慮してRPVの破損頻度を評価するPFM解析コードPASCALを開発してきた。また我々は、国内におけるPFMの適用性向上を図るため、破壊力学に関する知識を有する解析者がそれを参照することでPFM解析を行い亀裂貫通頻度を評価できるよう、標準的解析要領を整備した。本要領は、本文、解説及び付属書で構成されており、PFM解析に関する技術的根拠や最新知見が取りまとめられたものになっている。本論では、本要領の概要について述べるとともに、本要領とPTS評価に関する国内データベースに基づき得られた国内モデルRPVに対する破損頻度の評価結果について述べる。
11
A Study for evaluating local damage to reinforced concrete panels subjected to oblique impact of deformable projectile
西田 明美; 太田 良巳*; 坪田 張二; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
剛飛翔体の衝突に伴う構造物の局部破壊については、その破壊様式に応じて 多くの評価式が提案されている。既往の評価式は、構造物に対して垂直に衝突する実験から導かれた実験式が主であり、斜め衝突に関する研究はほとんど行われていないのが現状である。本研究では、実験結果およびシミュレーション結果に基づき斜め衝突に対する評価式を提案することを目的とする。本論文では、既往の衝撃実験結果のシミュレーション解析により妥当性が確認されたシミュレーション手法を用いて、柔飛翔体の斜め衝突を受ける鉄筋コンクリート版の局部損傷シミュレーションを実施し、衝突角度の違いによる局部損傷の低減効果について得られた知見を報告する。
12
Closed-form stress intensity factor solutions for deep surface cracks in cylinders subjected to global bending
東 喜三郎*; Li, Y.; 長谷川 邦夫; Shim, D. J.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
Materials made of alloy 82/182/600 used in pressurized water reactors are known to be susceptible to primary water stress corrosion cracking. The depth, ${it a}$, of flaws due to primary water stress corrosion cracking can be larger than the half of crack length ${it c}$, which is referred to as cracks with large aspect ratios. The stress intensity factor solution for cracks plays an important role to predict crack propagation and failure. However, Section XI of the ASME Boiler and Pressure Vessel Code does not provide the solutions for cracks with large aspect ratios. This paper presents the stress intensity factor solutions for circumferential surface cracks with large aspect ratios in cylinders under global bending loads. Finite element solutions were used to fit closed-form equations with influence coefficients ${it G}$gb. The closed-form solutions for coefficient ${it G}$gb were developed at the deepest points and the surface points of the cracks with aspect ratios ranged from 1.0 to 8.0.
13
Study on the relationship between interaction factors and stress intensity factor for elliptical flaws
東 喜三郎*; Li, Y.; 長谷川 邦夫
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 6 Pages, 2017/07
The interaction of multiple flaws in close proximity to one another may increase the stress intensity factor of the flaw in structures and components. This interaction effect is not distributed uniformly along the crack front. For instance, the strongest interaction is generally observed at the point closest to a neighboring flaw. For this reason, the closest point shows a higher value of the stress intensity factor than all other points in some cases, even if the original value at the point of the single flaw is relatively low. To clarify the condition when the closest point shows the maximum stress intensity factor, we investigated the interaction of two equal elliptical flaws in an infinite model subjected to remote tension loading. The stress intensity factor of the elliptical flaws was obtained be performing finite element analysis of a linear elastic solid. The results indicated that the interaction factors along the crack front can be expressed by a simple empirical formula. Finally, we show the relationship between geometrical features of the flaw and the stress intensity factor at the closest point.
14
Closed-form stress intensity factor solutions for deep surface cracks in plates
東 喜三郎*; Li, Y.; 長谷川 邦夫; Xu, S.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
Materials made of alloy 82/182/600 used in light-water reactors are known to be susceptible to stress corrosion cracking. It is known that the depth ${it a}$ of some cracks due to primary water stress corrosion cracking is larger than the half of crack length ${it c}$. The stress intensity factor solution for cracks plays an important role to predict crack propagation and failure. However, Section XI of the ASME Boiler and Pressure Vessel Code does not provide the solutions for cracks with large aspect ratios. In this study, closed-form stress intensity factor influence coefficients for deep surface cracks in plates are discussed. The crack tip stress distribution was represented by a fourth degree polynomial equation. Influence coefficient tables obtained by using finite element analysis in previous studies were used for curve fitting. The closed-form solutions for the coefficient were developed at the surface points, the deepest points, and the maximum points of the cracks with aspect ratios ranged from 1.0 to 8.0.
15
Verification methodology and results of probabilistic fracture mechanics code PASCAL
眞崎 浩一; 宮本 裕平*; 小坂部 和也*; 宇野 隼平*; 勝山 仁哉; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
国内の原子炉圧力容器を対象とした加圧熱衝撃事象時の破損頻度評価を行うため、確率論的破壊力学(PFM)解析コードPASCALが整備されている。一般的に、PFM解析コードは試験との比較等を通じた機能確認を行うことができないことから、その信頼性確認は困難である。本論文では、PFM解析コードの信頼性確認に係る方法を示すとともに、解析コードに含まれた確率変数、アルゴリズムや解析機能に関する検証を実施し、解析コードの有効性を明らかにした。
16
Fracture toughness evaluation of neutron-irradiated reactor pressure vessel steel using miniature-C(T) specimens
Ha, Y.; 飛田 徹; 高見澤 悠; 西山 裕孝
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07
圧力容器鋼の破壊靱性評価へのミニチュアC(T)試験片の適用性を調べるため、中性子照射された圧力容器鋼材のシャルピー試験片からミニチュアC(T)試験片を加工するとともに破壊靱性試験に供し、参照温度$$T_{o}$$を評価した。その結果、ミニチュアC(T)試験片で得られる$$T_{o}$$は疲労予亀裂入りシャルピー型破壊靭性試験片から得られる値とよく一致すること、ミニチュアC(T)試験片から得られる1T-C(T)相当の破壊靱性値のばらつきは疲労予亀裂入りシャルピー型破壊靭性試験片等から得られるものと大差が無いこと、参照温度$$T_{o}$$とシャルピー吸収エネルギー41Jレベルの延性脆性遷移温度の関係は、米国データのばらつきの範囲内にあることが明らかになった。
17
Improvement of probabilistic fracture mechanics analysis code PASCAL-SP with regard to primary water stress corrosion cracking
真野 晃宏; 山口 義仁; 勝山 仁哉; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
亀裂を有する原子力構造物の健全性評価には、決定論的な破壊力学に基づく手法が用いられている。一方で、影響因子の不確実性の考慮及び構造物の破損確率の定量評価が可能であるという理由から、確率論的破壊力学(PFM)に基づく手法の実用性が期待されている。原子力機構ではこれまでに、沸騰水型原子炉水質環境中における粒界型応力腐食割れや疲労等の経年劣化事象を考慮した原子力配管系の破損確率の評価を目的として、PFM解析コードPASCAL-SPの開発を進めてきた。近年国内では加圧水型原子炉一次系水質環境中応力腐食割れ(PWSCC)に起因する亀裂がニッケル合金溶接部において確認されていることから、その構造健全性評価が重要となっている。本論文は、PWSCCを考慮した一次系配管の評価を目的としたPASCAL-SPの改良について示すものである。PWSCCに関連する確率論的評価モデルとして、亀裂の発生、進展及び非破壊検査による亀裂の検出等のモデルを整備した。また、応力拡大係数の計算精度の向上を図った。評価事例としてPWSCCに起因する周方向及び軸方向亀裂を有するニッケル合金溶接部を対象とした破損確率の評価を示した。加えて、非破壊検査が破損確率に及ぼす影響を評価した。評価結果を踏まえて、改良したPASCAL-SPがPWSCCを考慮した一次系配管の破損確率評価に有用であると結論付けた。
18
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.
19
High temperature oxidation of Zry-4 in oxygen-nitrogen atmospheres
Negyesi, M.; 天谷 政樹
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
Zry-4 fuel cladding tubes were exposed in mixtures of oxygen and nitrogen at temperatures of 800-1380 $$^{circ}$$C. The influence of various flow rates of oxygen and nitrogen as well as specimen height on the weight gain was examined. The overall weight gain was substantially affected by both the applied flow rates and the height of specimens. The oxidation kinetics in air was assessed based on the results of weight gain measurements. A transition in the kinetics was observed at 800 and 1000 $$^{circ}$$C. The kinetics in the post-transient regimes was rather accelerated than linear. The equation proposed in this study for air condition was in good agreement with the Leistikow-Berg correlation and the Baker-Just correlation. Prior $$beta$$-phase shrinked when the oxide scale along with the $$alpha$$-Zr(O) layer progressed. Eventually, both the specimen plastic strain and maximum load decreased due to the shrinkage and increasing embrittlement of the prior $$beta$$-phase.
20
A Study for evaluating local damage to RC panels subjected to oblique impact, 1; A Study for evaluating local damage caused by oblique impact of rigid projectiles
太田 良巳*; 西田 明美; 坪田 張二; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
剛飛翔体の衝突に伴う構造物の局部破壊については、その破壊様式に応じて多くの評価式が提案されている。既往の評価式は、対象構造物に対して垂直に衝突する実験から導かれた実験式が主であり、斜め衝突に関する研究はこれまでほとんど行われていないのが現状である。そこで本研究では、実験結果およびシミュレーション結果に基づき斜め衝突による局部損傷の評価式を提案することを目的とする。本稿では、既往の斜め衝突実験結果をもとに既往の垂直衝突実験式である修正NDRC式を修正し、剛飛翔体の斜め衝突による局部損傷の新たな評価式を提案する。