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長期透水試験に対する再現シミュレーション末武 航弥*; 緒方 奨*; 安原 英明*; 青柳 和平; 乾 徹*; 岸田 潔*
第16回岩の力学国内シンポジウム講演論文集(インターネット), p.304 - 309, 2025/01
地層処分の安全性評価において、廃棄体処分坑道の掘削に伴うEDZ(掘削損傷領域)の進展範囲や、掘削後の岩盤の透水性変化挙動を予測することは非常に重要である。本研究では、三次元坑道掘削シミュレーターを用いて、幌延深地層研究センターで実施されている原位置坑道掘削とその後の透水試験を対象とした再現解析を試みた。その結果、掘削によるEDZの進展範囲と透水試験の結果について、原位置試験と類似する結果が得られた。このことから、本シミュレーターがわが国の大深度泥岩帯においての掘削に伴う力学的影響や、掘削後の岩盤変形-浸透といった連成現象とそれによる透水性変化などの予測評価に関して、有効であることが確認された。
宮川 和也; 石井 英一; 今井 久*; 平井 哲*; 大野 宏和; 中田 弘太郎*; 長谷川 琢磨*
原子力バックエンド研究(CD-ROM), 31(2), p.82 - 95, 2024/12
高レベル放射性廃棄物の処分地の選定過程における概要調査では、地下水の涵養域から流出域までを包含する数km-数十kmの広域を対象とした地下水流動解析により、地下水の移行時間・経路が評価される。亀裂の発達する岩盤中の地下水の移行時間を解析的に求める上で、岩盤の水理学的有効間隙率は感度の高いパラメータである。堆積岩ではボーリング調査における原位置水理試験から得られた亀裂の透水性を岩盤の透水性として扱う一方で、コア試料を用いて浮力法などの室内試験から得られた健岩部の間隙率を水理学的有効間隙率として扱うなど、有効間隙率の考え方が明確ではない。本研究では亀裂の発達する堆積岩である声問層および稚内層浅部を例として、亀裂の開口幅を基に推定した有効間隙率を用いた場合の移行時間を、観測結果と比較することで、有効間隙率の推定方法を検討した。その結果、亀裂の開口幅を基に推定した有効間隙率を用いた場合、観測結果と整合的な移行時間が得られた。その時の有効間隙率は、健岩部の間隙率と比較して1-3桁小さい値であった。多孔質な健岩部と割れ目からなる水みちネットワークを有する堆積岩の場合、亀裂の開口幅を基に有効間隙率を推定することが有効であることが示された。
広田 憲亮; 中野 寛子; 武田 遼真; 井手 広史; 土谷 邦彦; 小林 能直*
材料の科学と工学, 61(6), p.248 - 252, 2024/12
SUS304Lステンレス鋼の0.2%耐力に関する比較分析により、ひずみ速度が低下するほど、温度が上昇するほど、0.2%耐力は著しく低下することが明らかとなった。一方で結晶粒径を68.6
mから0.59mに微細化した場合における低ひずみ速度下での0.2%耐力への強度低下率の影響は小さかった。しかし、結晶粒微細化は、室温に比べて原子炉運転温度下での0.2%耐力低下には影響を及ぼした。粒内応力腐食割れ(SCC)を促進する条件下での低ひずみ速度引張試験では、28.4m以下の結晶粒径を持つSUS304Lに対しては、原子炉運転温度下での破断ひずみと同等の値を示したが、粗粒のSUS304Lでは破断ひずみが低下した。微細構造解析では、より結晶粒が微細な材料で87%以上の延性破面が観察され、特に0.59mの結晶粒径を持つSUS304Lでは{111}/
3粒界が数多く存在する一方で、結晶粒径が大きくなるにつれてその割合が減少していた。これらの結果は、結晶粒微細化により、{111}/3粒界の増加を通じて、腐食の進行が遅延し、粒内SCCが抑制されたことを示唆している。
Yuan, X.*; Hu, Q. H.*; Fang, X.*; Wang, Q. M.*; Ma, Y.*; 舘 幸男
Sedimentary Geology, 465, p.106633_1 - 106633_14, 2024/05
被引用回数：0 パーセンタイル：0.00(Geology)Archie's cementation factor, m, is a critical parameter for petrophysical studies, and the value is influenced by several factors such as the shape, type, and size of grains, degrees of diagenesis, and associated pore structure. Using integrated experimental and theoretical approaches, the goal of this study is to obtain the cementation factor of rocks (both reservoir rock and caprock) and assess the impact of diagenesis processes on the values of the cementation factor. Thirteen samples of geologically diverse rocks (six mudstones, four fossiliferous limestones, two marbles, and one sandstone) were selected to achieve these research objectives. Two approaches, the diffusion of gas tracers and the Bosanquet formula calculation using pore-throat sizes from mercury intrusion porosimetry analyses, were used to derive the cementation factors of these rock samples. These rocks were categorized into two groups based on the correlation between average pore-throat diameter and diffusivity, and an exponential-law relationship between the cementation factor and porosity was determined for these sample groups. In addition, thin-section petrography and field emission-scanning electron microscopy observations were utilized to investigate diagenetic processes, with four diagenetic patterns being established: (1) strong compaction, strong cementation, and weak dissolution-diagenesis pattern; (2) weak compaction, medium cementation, and weak dissolution-diagenesis pattern; (3) weak compaction, medium cementation, and strong dissolution-diagenesis pattern; and (4) fracture-matrix pattern. The results indicated that diagenetic processes and microfractures contribute to the variability in the cementation factors in these rock samples.
成川 隆文; 宇田川 豊
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
Information criteria such as a widely applicable information criterion (WAIC) and a widely applicable Bayesian information criterion (WBIC) enable the selection of models with high predictive accuracy and data fit, yet these criteria come with inherent uncertainties as they are statistical measures. To evaluate the uncertainty in model selection based on these information criteria, we performed numerical experiments using the bootstrap method, which is a resampling technique, on models for estimating the fracture probability of fuel cladding tubes during loss-of-coolant accidents (LOCAs). By calculating WAIC and WBIC for each of 10,000 bootstrap samples, we evaluated the dependency of model selection on these samples. Our key findings reveal that: (1) Sample-derived variation in information criteria was significantly greater than variability between models, underscoring the importance of assessing uncertainty from samples. (2) The Log-probit model, developed in our previous study, was selected as the optimal model for its superior predictive performance and data fit, despite the inherent uncertainties associated with WAIC and WBIC. (3) The presence of outliers at the fracture/non-fracture boundary of fuel cladding tubes may negatively impact the information criteria, suggesting the need for careful consideration when including such data in model parameter estimation.
成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊
Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09
被引用回数：1 パーセンタイル：27.70(Nuclear Science & Technology)For realizing a highly reliable fracture limit evaluation of fuel cladding tubes during loss-of-coolant accidents (LOCAs) in light-water reactors, we developed a method to quantify the fracture limit uncertainty of high-burnup advanced fuel cladding tubes. This method employs a hierarchical Bayesian model that can quantify uncertainty even with limited experimental data. The fracture limit uncertainty was quantified as a probability using the amount of oxidation (Equivalent cladding reacted: ECR) and the initial hydrogen concentration (the hydrogen concentration in the fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. We divided the regression coefficients of this model into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences among various types of fuel cladding tubes. This hierarchical structure enabled us to quantify the fracture limit uncertainty through the effective use of prior knowledge and data, even for high-burnup advanced fuel cladding tubes with a small number of data points. The fracture limits representing a 5% fracture probability with 95% confidence of the high-burnup advanced fuel cladding tubes evaluated by the hierarchical Bayesian model were higher than 15% ECR for the initial hydrogen concentrations of up to 700-900 wtppm and restraint loads below 535 N. These fracture limits were comparable to the limit of the unirradiated Zircaloy-4 cladding tube, indicating that the burnup extension and use of the advanced fuel cladding tubes do not significantly lower the fracture limit of fuel cladding tubes. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data, instead of the binary data, depending on the condition of the fuel cladding tube specimens after performing the LOCA-simulated test, thereby increasing the amount of information in the data.
成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之
Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08
被引用回数：3 パーセンタイル：65.16(Materials Science, Multidisciplinary)To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200-300 K higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of 5000 N, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at 1473 K for 1 h. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at 1673 K and fractured after abnormal oxidation at 1573 K for 1 h. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below 1473 K, where no melting or abnormal oxidation occurs.
Lu, K.; 高見澤 悠; Li, Y.; 眞崎 浩一*; 高越 大輝*; 永井 政貴*; 南日 卓*; 村上 健太*; 関東 康祐*; 八代醍 健志*; et al.
Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08
A probabilistic fracture mechanics (PFM) analysis code, PASCAL, has been developed by Japan Atomic Energy Agency for failure probability and failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. To strengthen the applicability of PASCAL, considerable efforts on verifications of the PASCAL code have been made in the past years. As a part of the verification activities, a working group consisted of different organizations from industry, universities and institutes, was established in Japan. In the early phase, the working group focused on verifying the PFM analysis functions for RPVs in pressurized water reactors (PWRs) subjected to pressurized thermal shock (PTS) events. Recently, the PASCAL code has been improved in order to run PFM analyses for both RPVs in PWRs and boiling water reactors (BWRs) subjected to a broad range of transients. Simultaneously, the working group initiated a verification plan for the improved PASCAL through independent PFM analyses by different organizations. Concretely, verification analyses for a PWR-type RPV subjected to PTS transients and a BWR-type RPV subjected to a low-temperature over pressure transient were performed using PASCAL. This paper summarizes those verification activities, including the verification plan, analysis conditions and results. Based on the verification studies, the reliability of PASCAL for probabilistic integrity assessments of Japanese RPVs was confirmed with confidence.
sp
HRL澤田 淳; 坂本 和彦*; 綿引 孝宜*; 今井 久*
SKB P-17-06, 154 Pages, 2023/08
An aim of Task 8, which was 8th modeling task of the SKB Task Forces on Groundwater Flow and Transport of Solutes, was to improve the knowledge of the bedrock-bentonite interface with regard to groundwater flow, mainly based on a set of data obtained by Bentonite Rock Interaction Experiment (BRIE) at sp. JAEA had developed an approach to Task 8 assuming that the discrete features dominate the delivery of groundwater to the bentonite columns emplaced into the vertically drilled boreholes from TASO tunnel floor, resulting in heterogeneous bentonite wetting behavior. This assumption was implemented as a FracMan Discrete Fracture Network (DFN) model for groundwater flow. Due to the assumption, no permeable rock matrix was implemented. The variability and uncertainty of this stochastic "HydroDFN" model was constrained by conditioning the model to match measured fracture location and orientation, and specific capacity (transmissivity) data observed at five probe boreholes. Groundwater from the HydroDFN being delivered to the bentonite columns, was modeled using Thames code with implementing a specific feature at the interface between the fractured rock mass and the bentonite. This modeling approach and the assumption of fracture dominated bentonite wetting appears to be able to provide a reasonable approximation to the observed heterogeneous bentonite wetting behavior of BRIE. We would suggest that a systematic investigation at pilot holes, including both geological mapping of the fractures and also testing of the hydraulic properties, might be required to get more practical prediction of heterogeneous wetting behavior in bentonite, as observed in BRIE.
成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊
Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12
To realize a more reliable safety evaluation of loss-of-coolant accidents (LOCAs) in light-water-reactors, we developed a quantification method of the fracture limit uncertainty of high-burnup advanced fuel cladding tubes using a hierarchical Bayes model that can quantify uncertainty even when experimental data are limited. The fracture limit uncertainty was quantified as a probability using the amount of oxidation and the initial hydrogen concentration (the hydrogen concentration in fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. The hierarchical Bayes model was developed by dividing the regression coefficients into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences between types of fuel cladding tubes. Using the developed model, we showed that the fracture limits of the high-burnup advanced fuel cladding tubes tended to be on average equal to or higher than that of an unirradiated conventional fuel cladding tube. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data depending on the condition of the fuel cladding tube specimens after the LOCA-simulated test instead of the binary data, thereby increasing the amount of information in each data.
Lu, K.; 高見澤 悠; 勝山 仁哉; Li, Y.
International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10
被引用回数：5 パーセンタイル：56.05(Engineering, Multidisciplinary)A probabilistic fracture mechanics (PFM) analysis code PASCAL was developed in Japan for probabilistic integrity assessment of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. To strengthen the practical applications of PFM methodology in Japan, PASCAL has been upgraded to a new version, PASCAL5, which enables PFM analyses of RPVs in both PWRs and boiling water reactors (BWRs) subjected to a broad range of transients, including PTS and normal operational transients. In this paper, the recent improvements in PASCAL5 are described such as the incorporated stress intensity factor solutions and corresponding calculation methods for external surface cracks and embedded cracks near the RPV outer surface. In addition, the analysis conditions and evaluation models recommended for PFM analyses of Japanese RPVs in BWRs are investigated. Finally, PFM analysis examples for core region of a Japanese BWR-type model RPV subjected to two transients (i.e., low-temperature over pressure and heat-up transients) are presented using PASCAL5.
鐵 桂一*; 高山 裕介; 澤田 淳
JAEA-Research 2022-005, 84 Pages, 2022/08
JAEA-Research-2022-005.pdf:10.32MB
JAEA-Research-2022-005-appendix(CD-ROM).zip:35.68MB
高レベル放射性廃棄物の地層処分における安全性の評価として、核種移行の評価が行われている。亀裂性岩盤を対象とした核種移行評価では、亀裂を平行平板で近似した解析モデルが主に使われている。しかし実際の岩盤中の亀裂は平行平板とは異なり、亀裂表面の粗さや亀裂内の充填物等の存在などの複雑な形状を呈している。このように不均質な形状を持つ亀裂を平行平板のモデルに近似するにあたり、透水量係数や亀裂開口幅などのパラメータ値をどのように設定するかが課題となっている。このような課題の解決策のひとつとして、実際の岩盤中の亀裂の幾何学的な特徴を調査することが挙げられる。本研究では、交差亀裂を含む亀裂の幾何学的な特徴の理解を目的とし、天然の交差亀裂を有する50cmスケールの花崗閃緑岩に対して平面研削する手法を用い、内部の亀裂形状を詳細に測定した。これより亀裂の亀裂幅(亀裂内の充填物を含む亀裂表面間の距離)、亀裂開口幅(亀裂内の充填物を含まない空隙のみの幅)、亀裂表面の形状(亀裂幅と亀裂幅の位置座標より算出した亀裂表面の位置座標)を取得した。また得られたデータより、亀裂幅の平均値や亀裂表面の粗さ、亀裂開口幅の分布などの特性を評価した。その結果、亀裂交差部近傍の亀裂の一つが、亀裂交差部を含む他の亀裂部に比べ亀裂幅、亀裂開口幅が特に大きいことが確認できた。これらのことから、本研究で用いた岩体試料では、最も透水性の高い透水経路は亀裂交差部そのものではなく、亀裂交差部近傍の亀裂開口幅が特に大きい特定の亀裂であると推測した。
笠原 直人*; 山野 秀将; 中村 いずみ*; 出町 和之*; 佐藤 拓哉*; 一宮 正和*
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07
破壊制御を利用して、設計想定を超える事象によって破損が生じた場合に、その拡大を抑制する技術の開発を進めている。開発課題として、(1)超高温時の破損拡大抑制技術、(2)課題地震時の破損拡大抑制技術、(3)原子炉構造レジリエンス向上手法の3つの計画を立てた。
Soler, J. M.*; Neretnieks, I.*; Moreno, L.*; Liu, L.*; Meng, S.*; Svensson, U.*; Iraola, A.*; Ebrahimi, K.*; Trinchero, P.*; Molinero, J.*; et al.
Nuclear Technology, 208(6), p.1059 - 1073, 2022/06
被引用回数：7 パーセンタイル：57.70(Nuclear Science & Technology)SKBタスクフォースは、亀裂性岩石中の地下水流動と物質移行のモデル化に関する国際フォーラムである。WPDE試験はフィンランドのオンカロ地下施設において実施された片麻岩中のマトリクス拡散試験である。複数の非収着性及び収着性のトレーサーを含む模擬地下水が試錐孔の試験区間に沿って注入された。タスク9Aは、WPDE試験で得られたトレーサー破過曲線に対する予測モデリングを行うことを目的とした。複数のチームが本タスクに参加し、異なるモデル化手法とコードを用いた予測解析を行った。この予測解析の重要な結論は、試錐孔の開口部における地下水流動に関連する分散パラメータにモデル化結果が大きく影響されることである。マトリクス拡散及び収着に関連する破過曲線のテール部に着目すると、異なるチーム間の解析結果の差異は相対的に小さい結果となった。
Park, Y.-J.*; 澤田 淳; 小堤 健紀*; 田中 達也*; 橋本 秀爾*; 森田 豊*
Proceedings of 3rd International Conference on Discrete Fracture Network Engineering (DFNE 2022) (Internet), 8 Pages, 2022/00
高レベル放射性廃棄物の地層処分の安全評価には地層中の長期にわたる地下水流動と核種移行プロセスの把握が求められる。沿岸部地下環境において、地下水流動は海水起源の塩水と陸水起源の淡水の密度差による複雑な相互作用の影響を受ける。加えて、数百万年の長期においては、海進・海退に伴う海水準変動の影響を受ける。本研究では、そのような沿岸域における亀裂性の結晶質岩を対象とした処分場の地下水流動と核種移行を評価するため、塩分濃度と地下水流速などの地下水環境の長期的な変遷を評価するための広域スケールとブロックスケールを組み合わせた評価フレームを構築した。
成川 隆文
日本原子力学会誌ATOMO, 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。
sp
Hard Rock Laboratory in Sweden舘 幸男; 伊藤 剛志*; Gylling, B.*
Water Resources Research, 57(11), p.e2020WR029335_1 - e2020WR029335_20, 2021/11
被引用回数：2 パーセンタイル：10.97(Environmental Sciences)本論文では、エスポ岩盤研究所で実施された原位置長期収着・拡散試験(LTDE-SD)のデータセットを用いて、収着及び拡散パラメータを実験室から原位置条件へと条件変換する手法を構築した。亀裂表面と岩石マトリクスの表面近傍の不均質性は、表面近傍での高い間隙率,拡散及び収着特性と、その段階的な変化を仮定した概念モデルによって評価された。非収着性のCl-36と低収着性のNa-22のモデル化結果によって、表面近傍の5mmの擾乱領域における間隙率と拡散係数の変化を考慮した概念モデルの妥当性が確認された。また、これらの陽イオンと陰イオンの拡散係数は、典型的な陽イオン加速と陰イオン排除の傾向を示した。一方で、収着メカニズムの異なる高収着性トレーサー(Cs-137, Ra-226, Ni-63, Np-237)のモデル化結果から、粒径サイズと収着分配係数との相関関係と、その表面近傍の擾乱との関係から条件変換する手法の有効性が確認された。
廣田 貴俊*; 名越 康人*; 北条 公伸*; 岡田 裕*; 高橋 昭如*; 勝山 仁哉; 上田 貴志*; 小川 琢矢*; 八代醍 健志*; 大畑 充*; et al.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
In order to establish a guideline for fracture evaluation by considering plastic constraint in the ductile-brittle transition temperature (DBTT) region, the CAF (Constraint-Based Assessment of Fracture in Ductile-Brittle Transition Temperature Region) subcommittee has been launched in 2018 in the Japan Welding Engineering Society. In the committee, fracture tests are conducted using C(T), SE(B), and 50mm-thick flat plate with a surface flaw subjected to bending load or tensile load to verify fracture evaluation methods. Since simulation results are easily affected by analysis conditions, benchmark analysis is essential for the potential users of the guideline. Therefore, benchmark analyses are executed on brittle and ductile damages by Beremin and Gurson-Tvergaard-Needleman (GTN) models implemented in the finite element (FE) codes. The benchmark analyses are carried out in four steps; Step 0 is to confirm the output of FE codes in each member with the same input data and the same FE model. Step 1 is to confirm the result of Weibull stress analysis for C(T) specimens tested at -125
C. The Weibull parameter, m, was fixed in this step. At step 2, sensitivity analyses are conducted on Weibull stresses in different conditions. The outputs by the GTN model are also confirmed. At the final step, the fracture simulation will be run for flat plate specimens with less plastic constraint than the standard fracture toughness specimen. As the results of the benchmark analyses up to step 2, a significant difference is not observed in the Weibull stress computed by committee members with the same input data and FE model and it is confirmed that the effects of element type, nonlinear deformation theory employed in FE analysis. For the calculation of the Weibull parameter m by using the fracture toughness test results and the developed programs by committee members, the converged values of m show good agreement among them.
Lu, K.; 勝山 仁哉; Li, Y.; 吉村 忍*
Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04
被引用回数：1 パーセンタイル：6.39(Engineering, Mechanical)Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.
鎌田 健人*; 奈良 禎太*; 松井 裕哉; 尾崎 裕介
第15回岩の力学国内シンポジウム講演論文集(インターネット), p.205 - 209, 2021/01
放射性廃棄物処分のようなプロジェクトを考える場合には、岩盤が有する物質の閉じ込め性能を評価することが重要であり、それに関して、岩盤内のき裂が透水性に及ぼす影響を調べることが必要不可欠である。しかし、き裂を含む泥岩の透水性の変化については未だ十分に研究されていない。そこで本研究では、北海道幌延地域に分布する泥岩の円柱形供試体に巨視き裂を導入し、透水性への影響を調べた。まず、圧裂引張試験によりき裂を導入した供試体に対して変水位透水試験を行い、透水係数を測定した。その後、トランジェントパルス法により測定したインタクトな供試体の透水係数と比較した。その結果、き裂導入により1オーダー程度の透水係数の上昇が確認された。本研究の結果から得られた透水係数の上昇の程度は、花崗岩や玄武岩などを用いた先行研究と比較すると小さい値であった。
