The Effect of base irradiation on failure behaviors of UO 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

https://doi.org/10.1016/j.anucene.2019.107268

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.

Closed-form stress intensity factor solutions for surface cracks with large aspect ratios in plates

東 喜三郎*; Li, Y.; Xu, S.*

Journal of Pressure Vessel Technology, 142(2), p.021207_1 - 021207_10, 2020/04

https://doi.org/10.1115/1.4045029

Alloy 82/182/600, which is used in light-water reactors, is known to be susceptible to stress-corrosion cracking. The depth of some of these cracks may exceed the value of half-length on the surface. Although the stress intensity factor (SIF) for cracks plays an important role in predicting crack propagation and failure, Section XI of the ASME Boiler and Pressure Vessel Code does not provide SIF solutions for such deep cracks. In this study, closed-form SIF solutions for deep surface cracks in plates are discussed using an influence coefficient approach. The stress distribution at the crack location is represented by a fourth-degree-polynomial equation. Tables for influence coefficients obtained by finite element analysis in the previous studies are used for curve fitting. The closed-form solutions for the influence coefficients were developed at the surface point, the deepest point, and the maximum point of a crack with an aspect ratio a/c ranging from 1.0 to 8.0, where a is the crack depth and c is one-half of the crack length. The maximum point of a crack refers to the location on the crack front where the SIF reaches a maximum value.

OSCAARコードパッケージの使用マニュアル

安全研究センター リスク評価研究ディビジョン 放射線安全・防災研究グループ

JAEA-Testing 2020-001, 65 Pages, 2020/03

JAEA-Testing-2020-001.pdf:4.42MB

日本原子力研究開発機構安全研究センターは、原子炉事故の確率論的リスク評価(PRA: Probabilistic Risk Assessment)研究の一環として、レベル3PRAコードOSCAARの開発を進めている。OSCAARは、レベル2PRAで得られたソースタームを基に、様々な気象条件に対し、環境に放出された後に拡散・沈着した放射性物質から公衆が受ける被ばく線量、防護措置による被ばく低減効果を考慮した上で、それに起因する健康影響等を確率論的に評価する計算コードである。このOSCAARを基に、Windows上にて、OSCAARの解析実行に加え、入力データファイルの作成、出力データファイルの後処理まで効率良く実施できるOSCAARコードパッケージを整備した。本報告は、OSCAARコードパッケージの使用方法を示したマニュアルである。

Experimental and analytical investigation of formation and cooling phenomena in high temperature debris bed

堀田 亮年*; 秋葉 美幸*; 森田 彰伸*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Skld, P.*; et al.

Journal of Nuclear Science and Technology, 57(3-4), p.353 - 369, 2020/03

https://doi.org/10.1080/00223131.2019.1691078

Key phenomena in the cooling states of debris beds under wet cavity conditions were classified into several groups based on the complicated geometry, nonhomogeneous porosity and volumetric heat of debris beds. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. The modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three additional tests, DEFOR-A, PULiMS and REMCOD were employed to validate implemented models. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. It also tries to report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the PULiMS tests.

Effects of oxidation and secondary hydriding during simulated Loss-Of-Coolant-Accident tests on the bending strength of Zircaloy-4 fuel cladding tube

岡田 裕史; 天谷 政樹

Annals of Nuclear Energy, 136, p.107028_1 - 107028_9, 2020/02

https://doi.org/10.1016/j.anucene.2019.107028

In order to evaluate the fracture resistance of fuel rods against a seismic loading following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated since the Fukushima-Daiichi Nuclear Power Plant accident. In this study, four-point-bending-tests were performed using Zircaloy-4 cladding tubes which experienced a simulated LOCA sequence in order to investigate the effects of oxidation and secondary hydriding occurring during a LOCA on the bending strength of fuel cladding. According to the obtained results, it was suggested that the maximum bending stress would be affected by the oxygen concentration in the prior-beta layer as well as the thickness of prior-beta layer. It was considered that the decrease in maximum bending stress by secondary hydriding is probably expressed by multiplying a factor of 0.37 by the maximum bending stress which solely takes account of the effect of oxidation.

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

崔 炳賢; 西田 明美; 村松 健*; 高田 毅士*

日本地震工学会論文集(インターネット), 20(2), p.2_1 - 2_16, 2020/02

https://doi.org/10.5610/jaee.20.2_1

本研究では、原子力施設の確率論的地震リスク評価の信頼性向上に資するため、原子炉建屋の地震応答解析結果におけるモデル化手法の違いによる影響を評価し、フラジリティ評価における認識論的不確実さを定量化することを目的としている。入力地震動として、偶然的不確実さを考慮するため、ハザード適合地震波を用いた。現実的な応答を得るため、原子炉建屋の3次元詳細モデルによる地震応答解析を実施し、建屋の壁および床の最大加速度について、中央値と対数標準偏差により統計的評価を行った。本研究により、建屋の上層部や床の開口部周辺において面外変形などの3次元効果が現れることが分かった。

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

https://doi.org/10.1080/00223131.2019.1659873

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, low-tin ZIRLO, M5, 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.

Rapid clogging of high-efficiency particulate air filters during in-cell solvent fires at reprocessing facilities

大野 卓也; 田代 信介; 天野 祐希; 吉田 涼一朗; 阿部 仁

Nuclear Technology, 206(1), p.40 - 47, 2020/01

https://doi.org/10.1080/00295450.2019.1620057

再処理施設におけるセル内溶媒火災事故は、近年の日本の規制基準において注目されている事象である。本研究では、再処理施設の換気系を模した試験系でリン酸トリブチル/ドデカン混合溶媒を燃焼させ、発生したエアロゾルをHEPAフィルタに負荷させた。その結果、ドデカンが混合溶媒中から焼失した後にフィルタ差圧が急激に上昇する現象が確認された。我々は、この現象を急速目詰まりと呼び、ドデカン焼失に起因する現象であると考えている。急速目詰まりとドデカン焼失の関係を把握することは、再処理施設における規制基準の構築にとって有用であると思われる。また、エアロゾルの分析結果からは、急速目詰まりが生じている間に生成されたエアロゾルは、粒径が大きく、未燃溶媒成分に富んでいることが明らかになった。これらの結果より、急速目詰まりの原因は、未燃溶媒の放出又は未燃溶媒蒸気と煤煙の相互作用によるものであると考えられる。総じて、この研究が示唆するところによると、火災事故時においてはこれまで予想されてきたよりも急速に換気系フィルタの目詰まりが進行する可能性があるといえる。

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 66 rod bundle

Han, X.*; Shen, X.*; 山本 俊弘*; 中島 健*; 孫 昊旻; 日引 俊*

International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12

https://doi.org/10.1016/j.ijheatmasstransfer.2019.118696

This paper presents a database of local flow parameters for upward adiabatic air-water two-phase flows in a vertical 66 rod bundle flow channel. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity vector were measured by using a four-sensor optical probe. Based on an existing state-of-the-art four-sensor probe methodology with the characteristic to count small bubbles, IAC in this study was derived more reliably than those in the existing studies. In addition, bubble velocity vector could be measured by the methodology. Based on this database, flow characteristics were investigated. The area-averaged void fraction and IAC were compared with the predictions from the drift-flux model and the IAC correlations, respectively. The applicability of those to the rod bundle flow channel was evaluated.

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

河 侑成; 高見澤 悠; 勝山 仁哉; 塙 悟史; 西山 裕孝

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

https://doi.org/10.1016/j.nimb.2019.10.018

The microstructural distribution in the heat-affected zone (HAZ) under the stainless overlay cladding of low alloy steel was investigated by metallurgical observation and finite element analysis (FEA). The distribution of the coarse-grain (CG) HAZ and the fine-grain (FG) HAZ in low alloy steel by metallurgical observation agreed well with the FEA results. Base metal contained mixed bainite with ferrite, whereas the CGHAZ and the FGHAZ contained mixed lower bainite with martensite and mixed upper bainite with ferrite, respectively. Ferrite fraction in FGHAZ was higher than those other areas. After ion irradiation at a fluence of 0.5 dpa, irradiation hardening and the formation of solute clusters were observed at the base metal, FGHAZ, and the CGHAZ. Atom probe tomography analysis revealed that irradiation hardening increased with increasing volume fraction of clusters, although irradiation hardening at the FGHAZ was greater than that at the CGHAZ, which contained more clusters than the FGHAZ. This difference in irradiation hardening may be due to the differences in the amount of ferrite, carbide precipitates and so on in the different microstructures.