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岡藤 孝史*; 三浦 一浩*; 佐郷 ひろみ*; 村上 久友*; 安藤 勝訓; 岡島 智史
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07
高速炉の容器を大型化することに伴う直径と板厚の比の増加や免震設計への適用に対応するため、軸圧縮,曲げ、およびせん断荷重下で弾塑性座屈を評価できる座屈評価方法を開発している。本研究では、検討中の評価手法の適用性を検証するために、オーステナイト系鋼の円筒試験体に対して水平荷重に加えて繰返し鉛直荷重を負荷する条件等を付した座屈試験と有限要素法解析を実施した。その結果、提案式による評価は、試験の座屈荷重に対して保守的であることが確認できた。また、有限要素法解析により2.25Cr-1Mo鋼や550度を超える温度における適用性についても検証した。
Lacroix, V.*; 長谷川 邦夫; Li, Y.; 山口 義仁
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
The structural integrity assessment of pipes with circumferential surface flaw under the plastic collapse regime consists of net section collapse analysis. In recent years various researchers showed that this analysis, which has been developed based on classic beam theory, suffers from certain inaccuracies. As such, assessment purely based on net-section collapse and beam theory can reveal both conservative and unconservative results. To address those inaccuracies, in this paper authors introduced a correction factor which aims to mitigate the difference between the ASME B&PV Code Section XI equations and the experimental results. This correction factor is calculated using an empirical formula developed on the basis of a large experimental database of pipe collapse bending tests containing variety of diameter, thickness, flaw depth and flaw length values. Within this work, authors took a systematic approach to identify the most influencing factors on such a correction factor and showed that by applying this correction factor to the current solution of ASME B&PV Code Section XI, this solution becomes more accurate. This corrected approach also is in line with ASME B&PV Code Section XI Appendix C practice for axial flaw in pipes, where a semi-empirical correction factor has been considered as well.
山口 義仁; 長谷川 邦夫; Li, Y.; Lacroix, V.*
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 4 Pages, 2022/07
Four-point bending tests without internal pressure were performed for Type 304 stainless steel pipes with circumferential flaws at room temperature. The specimens are 1-inch diameter (33.7 mm) Schedule 160 pipes (6.3 mm wall thickness). The flaws were part-through flaws located at the pipe external side. The flaw angles are from 120
to 240, and the flaw depths are two cases of 50% and 75% of the wall thickness. Plastic collapse stresses obtained from experiments were compared with those calculated using Limit Load Criteria from Appendix C of the American Society of Mechanical Engineers Code Section XI. Limit Load Criteria were developed using flow stress at flawed section of the pipe. The plastic collapse stress test results were larger than those of the calculation results. For flaws with flaw depths less than 50% of the wall thickness, the experimental stresses were significantly large. The Limit Load Criteria given by Section XI provide conservative collapse stresses and could be improved.
山口 義仁; 西田 明美; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
原子力発電所おける配管減肉は重要な経年劣化事象の一つである。また近年、日本のいくつかの原子力発電所は大きな地震を経験している。そのため、長期間運転した原子力発電所を対象とした地震を起因とした確率論的リスク評価において、減肉と地震による応答応力の両方を考慮した配管系の地震フラジリティ評価が、重要となっている。本研究では、原子力機構が開発した減肉配管を対象とした破損確率解析コードPASCAL-ECに、減肉配管の地震応答応力の算出モデルや減肉配管の破壊評価法など、減肉配管の地震フラジリティを評価可能とする機能を導入した。また、整備した解析コードを用いて、地震フラジリティの評価事例を整備した。
山口 義仁; 真野 晃宏; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
蒸気発生器(SG)伝熱管は、加圧水型原子炉の重要な機器の一つである。SG伝熱管では、減肉や応力腐食割れなどの検出が報告されている。SG伝熱管の構造健全性を評価するためには、一次冷却系からの内圧と二次冷却系からの外圧の両方を考慮して破裂による破壊を評価する必要がある。そこで、様々な機関においてSG伝熱管を対象とした破裂試験が実施され、それに基づき破壊評価法が提案されている。本研究では、減肉又は亀裂を有するSG伝熱管を対象に、既存の破裂試験データを調査し、それを踏まえて、亀裂又は局所的な減肉を有するSG伝熱管に適した破壊評価法を新たに提案した。また、これらの破壊評価法の適用性を、予測結果と実機SG伝熱管の破裂試験データを比較することにより確認した。
勝山 仁哉; 山口 義仁; 根本 義之; 古田 琢哉; 加治 芳行
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07
To assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi, we have been developing an analysis method based on coupled analysis of three-dimensional multi-physics simulations composed of radiation transport, thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this simulation, Monte Carlo radiation transport calculation is firstly performed by using PHITS code to compute proton dose distribution considering molten conditions of core materials. Then the deposit energies at each location is imported into TH analysis code ANSYS Fluent with the same geometry and temperature distribution is simulated by thermal-fluid dynamics. Finally, temperature distribution obtained from TH analysis is applied to thermal-elastic-plastic-creep analyses using FINAS-STAR and then damage evaluation is carried out based on several criterions such as Kachanov, Larson-Miller-parameter, melting point. To conduct such analyses, we also have continued to obtain experimental data on creep deformation in high temperature range. In this study, to predict time and location of reactor pressure vessel (RPV) lower head rupture of boiling water reactors (BWRs) considering creep damage mechanisms, we performed creep damage evaluations based on developing analysis method by using detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes and welds. From the detailed analysis results, it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.
Zarazovski, M.*; Pistra, V.*; Lauerova, D.*; Obermeier, F.*; Mora, D.*; Dubyk, Y.*; Bolinder, T.*; Cueto-Felgueroso, C.*; Szavai, S.*; Dudra, J.*; et al.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 11 Pages, 2022/07
The APAL (Advanced Pressurized Thermal Shock (PTS) Analysis for Long-Term Operation (LTO)) project was launched in October 2020 for four years with funding from the European Union's HORIZON 2020 program. Within APAL, an extensive literature review was performed and experience with defining the state-of-the-art of the Warm Pre-Stress (WPS) effect, which has an impact on the reactor pressure vessel (RPV) brittle fracture margin in both deterministic and probabilistic terms, was collected. To gather the worldwide experience of the WPS approaches and models, a comprehensive questionnaire was developed followed by each APAL partner response. It mainly focused on the following aspects: collection of existing WPS approaches and models implemented in standards and rules for RPV brittle fracture assessment; identification of the WPS issues; collection and analysis of the existing experimental data. This work describes worldwide experience and best practice of the WPS and its application for the RPV integrity assessment. The paper's conclusions are also focused on the recommendations for dealing with WPS issues.
Lacroix, V.*; Dulieu, P.*; 長谷川 邦夫; Mares, V.*
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
現行のASME Code Section XIでは、非平面状欠陥に対して簡易的な手法が提案されている。すなわち、非平面状欠陥面を最大主応力方向に投影し2つの平面状欠陥に分解する。しかしながら、この非平面状欠陥の簡易的な投影は荷重に対して、あるいは非平面状欠陥の傾きに対して常に保守的であると言えない場合がある。本論文は、ASME Codeのアプローチに保守性のある包括的な評価を実施し、現行のASME Code Section XIの不適正を克服した非平面状欠陥の代替アプローチを提案する。