Relevant bibliographies by topics / Fracture mechanics. Materials Nuclear power plants

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers

Academic literature on the topic 'Fracture mechanics. Materials Nuclear power plants'

Author: Grafiati
Published: 4 June 2021
Last updated: 15 February 2022

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Journal articles on the topic "Fracture mechanics. Materials Nuclear power plants"

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Yoshida, K., M. Kojima, M. Iida, and I. Takahashi. "Fracture toughness of weld metals in steel piping for nuclear power plants." International Journal of Pressure Vessels and Piping 43, no. 1-3 (1990): 273–84. http://dx.doi.org/10.1016/0308-0161(90)90107-s.

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Timofeev, B. T., R. P. Vinogradov, S. P. Generalova, and T. A. Chernaenko. "Fracture-resistance evaluation for piping materials of BWR nuclear power plants of the RBMK type." International Journal of Pressure Vessels and Piping 52, no. 3 (1992): 303–11. http://dx.doi.org/10.1016/0308-0161(92)90088-w.

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Yagawa, G., Y. Ando, K. Ishihara, T. Iwadate, and Y. Tanaka. "Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension." Journal of Pressure Vessel Technology 111, no. 3 (1989): 234–40. http://dx.doi.org/10.1115/1.3265669.

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An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ int
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Yoon, Ji-Hyun, Bong-Sang Lee, and Jun-Hwa Hong. "J-R fracture characteristics of ferritic steels for RPVs and RCS piping of nuclear power plants." Metals and Materials International 7, no. 5 (2001): 505–12. http://dx.doi.org/10.1007/bf03027094.

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Siegl, Jan, Petr Haušild, Adam Janča, Radim Kopřiva, and Miloš Kytka. "Characterisation of Mechanical Properties by Small Punch Test." Key Engineering Materials 606 (March 2014): 15–18. http://dx.doi.org/10.4028/www.scientific.net/kem.606.15.

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The specific desired properties for structures and components working in critical environments (e.g. different structure parts of power plants) require current information about degradation processes coming out in materials. Obtaining of this information by the help of the classical tests of mechanical properties (tensile test, Charpy test, fracture toughness test, creep test etc.) is very limited namely in the case of nuclear power plants pressure vessel. Hence, the new innovative techniques based on miniaturized specimens have been developed for evaluation of mechanical properties and their
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Naizabekov, Abdrakhman, Alexandr Arbuz, Sergey Lezhnev, Evgeniy Panin, and Marcin Knapinski. "Study of Technology for Ultrafine-Grained Materials for Usage as Materials in Nuclear Power." New Trends in Production Engineering 2, no. 2 (2019): 114–25. http://dx.doi.org/10.2478/ntpe-2019-0077.

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Abstract Nuclear power is associated with great environmental risks. In many cases, the problem of accidents of nuclear power plants is related to the use of materials that do not fully meet the following requirements: high corrosion resistance; high temperature resistance; creep resistance; fracture toughness; stability of structure and properties under irradiation. Therefore, studies aimed at finding materials that can withstand long-term loads at high temperatures, aggressive environment and gradual structural degradation under the influence of radiation are relevant. One of the structural
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Jundong, Lu, Jiang Xiaobin, Sun Ke, Liu Bin, Li Xinmin, and Ni Qinwen. "Stress Corrosion Cracking Behavior of TP 439 and 690 TT under Film-Forming Amine Environment." Scanning 2021 (June 8, 2021): 1–8. http://dx.doi.org/10.1155/2021/6668537.

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Film-forming amines have been widely used in thermal power plants for maintenance after shutdown, and there are more and more applications and researches in nuclear power secondary circuits for this purpose. However, in the direction of stress corrosion cracking, there is not much research on the influence of film-forming amines on metal materials. This article uses the high temperature slow strain rate test (SSRT) method to evaluate the influence of a commercial film-forming amine on the stress corrosion cracking behavior of two conventional island materials for PWR nuclear power plants. Thes
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Torop, V. M., M. D. Rabkina, O. O. Shtofel’, V. V. Usov, N. M. Shkatulyak, and O. S. Savchuk. "On the Causes of Fractures of Reinforcing Ropes of the Protective Shells of Power-Generating Units of Nuclear Power Plants." Materials Science 54, no. 2 (2018): 240–49. http://dx.doi.org/10.1007/s11003-018-0179-y.

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Kim, J. C., Sang Min Lee, Yoon Suk Chang, Jae Boong Choi, Young Jin Kim, and Young Hwan Choi. "Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant." Solid State Phenomena 120 (February 2007): 157–62. http://dx.doi.org/10.4028/www.scientific.net/ssp.120.157.

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Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an
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Wang, Wei Bing, He Xue, Fu Qiang Yang, and Kun Liu. "Effects of Grain Size on Crack Tip Mechanical Fields of Intergranular Cracking." Advanced Materials Research 1004-1005 (August 2014): 1147–51. http://dx.doi.org/10.4028/www.scientific.net/amr.1004-1005.1147.

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Intergranular stress corrosion cracking is one of important failure form of structural materials in nuclear power plants, and the initiation and development of crack at grain boundary are affected by the grain size of materials. The macroscopic model and mesoscopic model of crack propagation was established by using finite element method, and the effects of grain size on fracture parameters such as Mises stress, the maximum principal stress and equivalent plastic strain nearby crack tip were studied. The results indicate that the distribution of Mises stress and equivalent plastic strain are d
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Dissertations / Theses on the topic "Fracture mechanics. Materials Nuclear power plants"

1

Rintamaa, Rauno. "Single specimen fracture toughness determination procedure using instrumented impact test /." Espoo [Finland] : Technical Research Centre of Finland, 1993. http://bibpurl.oclc.org/web/30658.

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CUNTO, GABRIEL G. de. "Aplicação do conceito "vazamento antes da falha" (LEAK BEFORE BREAK) em tubulações de aço 316LN soldado com metal de adição 316L." reponame:Repositório Institucional do IPEN, 2017. http://repositorio.ipen.br:8080/xmlui/handle/123456789/27963.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2017-11-01T17:02:12Z No. of bitstreams: 0<br>Made available in DSpace on 2017-11-01T17:02:12Z (GMT). No. of bitstreams: 0<br>Este trabalho apresenta um estudo prático da aplicação do conceito Leak Before Break (LBB), usualmente aplicado em usinas nucleares, em uma tubulação fabricada a partir de aço AISI 316LN soldada com a utilização de eletrodo revestido AISI 316L. O LBB é um critério fundamentado em análises de mecânica da fratura, que considera que um vazamento proveniente de uma trinca, presente em uma tubulação, possa se
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McMinn, Andrew. "Environmentally assisted cracking of nickel-base alloys in nuclear power plant components." Thesis, London South Bank University, 1993. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.240289.

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CRUZ, JULIO R. B. "Procedimento analitico para previsao do comportamento estrutural de componentes truncados." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10665.

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Made available in DSpace on 2014-10-09T12:42:51Z (GMT). No. of bitstreams: 0<br>Made available in DSpace on 2014-10-09T14:07:05Z (GMT). No. of bitstreams: 1 06110.pdf: 5239500 bytes, checksum: 175d6a6c784cd8fbadb485e4c6d90285 (MD5)<br>Tese(Doutoramento)<br>IPEN/T<br>Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Oliveira, Heloisa Maria Santos. "Avaliação numérica do comportamento à fratura de um protótipo de vaso de pressão de reator PWR submetido a choque térmico pressurizado." CNEN - Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, 2005. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=41.

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Nenhuma<br>No circuito primário de uma usina nuclear do tipo PWR (Pressurized Water Reactor), o refrigerante do reator é mantido a uma temperatura interna por volta de 300 C e pressão interna da ordem de 15,0 MPa, durante operação normal. O Vaso de Pressão do Reator (VPR) contém os elementos combustíveis e é considerado o componente mais importante do circuito primário. A integridade do VPR deve ser assegurada durante toda a vida útil da usina, de forma a proteger os trabalhadores da usina e o público em geral dos danos decorrentes da liberação de material radioativo.Uma das condições de carre
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Books on the topic "Fracture mechanics. Materials Nuclear power plants"

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Rintamaa, Rauno. Single specimen fracture toughness determination procedure using instrumented impact test. Technical Research Centre of Finland, 1993.

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Wilkowski, G. M. State-of-the-art report on piping fracture mechanics. Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

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Wilkowski, G. M. State-of-the-art report on piping fracture mechanics. Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

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Wilkowski, G. M. State-of-the-art report on piping fracture mechanics. Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

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CSNI/NRC, Workshop on Ductile Piping Fracture Mechanics (1984 San Antonio Tex ). Proceedings of the CSNI/NRC Workshop on Ductile Piping Fracture Mechanics: Held at San Antonio, Texas, June 21-22, 1984. The Commission, 1988.

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Hiser, A. L. A user's guide to the NRC's piping fracture mechanics data base (PIFRAC). Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1987.

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Hiser, A. L. A user's guide to the NRC's piping fracture mechanics data base (PIFRAC). Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1987.

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Pressure, Vessels and Piping Conference (2001 Atlanta Ga ). Fracture and fitness: Presented at the 2001 ASME Pressure Vessels and Piping Conference, Atlanta, Georgia, July 23-26, 2001. American Society of Mechanical Engineers, 2001.

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1932-, Zamrik S. Y., Perez E. H, American Society of Mechanical Engineers. Pressure Vessels and Piping Division., and Pressure Vessels and Piping Conference (1990 : Nashville, Tenn.), eds. High pressure technology, fracture mechanics, and service experience in operating power plants: Presented at the 1990 Pressure Vessels and Piping Conference, Nashville, Tennessee, June 17-21, 1990. American Society of Mechanical Engineers, 1990.

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S, Rahman, Battelle Memorial Institute, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., eds. Probabilistic pipe fracture evaluations for leak-rate-detection applications. U.S. Nuclear Regulatory Commission, 1995.

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Book chapters on the topic "Fracture mechanics. Materials Nuclear power plants"

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Hiser, Allen L., Simon C. F. Sheng, and Shah N. Malik. "Observations on Sensitivity of RPV Integrity Probabilistic Fracture Mechanics Evaluations to Input Parameters." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch84.

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Gamble, R. M. "Probabilistic fracture mechanics risk analysis of reactor pressure vessel (RPV) integrity." In Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) in Nuclear Power Plants. Elsevier, 2015. http://dx.doi.org/10.1533/9780857096470.3.378.

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Mueller, F., and M. Oechsner. "Fracture mechanics and testing for crack initiation and growth assessment in coal power plants." In Coal Power Plant Materials and Life Assessment. Elsevier, 2014. http://dx.doi.org/10.1533/9780857097323.2.229.

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Nanstad, R. K., W. L. Server, M. A. Sokolov, and M. Brumovský. "Evaluating the fracture toughness of reactor pressure vessel (RPV) materials subject to embrittlement**Some portions of this chapter have been gleaned from Chapter 3 of: International Atomic Energy Agency, Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels, IAEA Nuclear Energy Series NP-T-3.11, IAEA, Vienna (2009), a chapter authored by the first author of this chapter (no attribution in the IAEA document)Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the US Department of Energy. The United States Government retains and the publisher by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes." In Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) in Nuclear Power Plants. Elsevier, 2015. http://dx.doi.org/10.1533/9780857096470.3.295.

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Conference papers on the topic "Fracture mechanics. Materials Nuclear power plants"

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Simonen, F. A. "Uncertainties in Probabilistic Fracture Mechanics Calculations." In ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference. ASMEDC, 2010. http://dx.doi.org/10.1115/pvp2010-25231.

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This paper addresses uncertainties in probabilistic fracture mechanics (PFM) calculations for pressure boundary components at commercial nuclear power plants. Such calculations can predict the probability that a component will have failed after a specified period of operation, but with large uncertainties that are difficult to quantify. PFM models only approximate details of as-built components as well as actual operating conditions over the lifetime of the component. Statistical distributions used as inputs to the calculations are subject to uncertainties, which also results in large uncertai
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Hayashi, Shotaro, Mayumi Ochi, Kiminobu Hojo, Takahisa Yamane, and Wataru Nishi. "Allowable Flaw Size of Japanese Cast Stainless Steel Pipe Using Probabilistic Fracture Mechanics Method." In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-65633.

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The cast austenitic stainless steel (CASS) that is used for the primary loop pipes of nuclear power plants is susceptible to thermal ageing during plant operation. The Japanese JSME rules on fitness-for-service (JSME rules on FFS)[1] for nuclear power plants specify the allowable flaw depths. However, some of these allowable flaw sizes are small compared with the smallest flaw sizes, which can be detected by nondestructive testing. ASME Section XI Code Case N-838[2] recently specified the maximum tolerable flaw depths for CASS pipes determined by probabilistic fracture mechanics (PFM). In a si
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Hong, Jong-Dae, and Changheui Jang. "Probabilistic Fracture Mechanics Application for Alloy 82/182 Welds in PWRS." In ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference. ASMEDC, 2010. http://dx.doi.org/10.1115/pvp2010-25176.

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In operating PWRs (Pressurized Water Reactors), incidents of Alloy 82/182 cracking increased the concern for structural integrity of butt weld locations recently, because of high weld residual stresses. Studies on PWSCC (Primary Water Stress Corrosion Cracking) have been mainly performed using deterministic approaches by controlling parameters, but a quantitative evaluation is difficult because of large uncertainties in each parameter and test results. The purposes of this paper are to provide a probabilistic fracture mechanics (PFM) analysis methodology and quantify failure probabilities for
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Brust, Frederick W., R. E. Kurth, D. J. Shim, and David Rudland. "Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes." In ASME 2011 Pressure Vessels and Piping Conference. ASMEDC, 2011. http://dx.doi.org/10.1115/pvp2011-57937.

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Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, of
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Lidbury, David, Ste´phane Bugat, Olivier Diard, et al. "PERFECT — Prediction of Irradiation Damage Effects in Reactor Components: Overview of RPV Mechanics Sub-Project." In ASME 2005 Pressure Vessels and Piping Conference. ASMEDC, 2005. http://dx.doi.org/10.1115/pvp2005-71558.

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The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This state-of-the-art approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, this 48-month project is focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel, and the austenitic steel internals. I
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Lidbury, David, Ste´phane Bugat, Olivier Diard, et al. "PERFECT—Prediction of Irradiation Damage Effects in Reactor Components: Update of Progress in RPV Mechanics Sub-Project." In ASME 2007 Pressure Vessels and Piping Conference. ASMEDC, 2007. http://dx.doi.org/10.1115/pvp2007-26076.

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The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, PERFECT is a 48-month project focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel (RPV), and the austenitic steel internals. It is t
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Ilg, Ulf, Gerhard Nagel, Gu¨nter Ko¨nig, Dirk Schu¨mann, Wolfgang Mayinger, and Martin Widera. "Integrity Concept for Piping Systems With Corresponding Leak and Break Postulates in German Nuclear Power Plants." In ASME 2011 Pressure Vessels and Piping Conference. ASMEDC, 2011. http://dx.doi.org/10.1115/pvp2011-57810.

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In order to exclude the possibility of catastrophic failure of safety relevant pressure-retaining components in nuclear power plants during operation, the “integrity concept” is applied in Germany. It has been developed over the past 30 years on the basis of the safety criteria of the guidelines for damage precautions, as set by the German Advisory Committee on Reactor Safeguards (RSK-LL) and the basic safety concept. The integrity concept is based on the requirements of proven basic safety characteristics: design, construction, material, and manufacturing. Complementary elements (so-called re
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Ogawa, Takeshi, Motoki Nakane, Kiyotaka Masaki, Shota Hashimoto, Yasuo Ochi, and Kyoichi Asano. "Investigation of Effect of Pre-Strain on Very High-Cycle Fatigue Strength of Austenitic Stainless Steels." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48811.

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The austenitic stainless steels have excellent mechanical and chemical characteristics and these materials are widely used for the main structural components in the nuclear power plants. A part of structural components using these materials is considered to have strain-history by machining, welding and etc in the process of manufacturing and these parts would be hardened because these materials have a remarkable work-hardening property. On the other hand, conventional studies for the fatigue strength used to be investigated by the results of fatigue tests applying normal specimens without the
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Hayashi, Takahiro, Shigeaki Tanaka, Tomonori Abe, Seiji Sakuraya, Suguru Ooki, and Takayuki Kaminaga. "Fracture Toughness Criteria of Irradiated Austenitic Stainless Steels for Structural Integrity Evaluation of BWR Internal Components." In ASME 2019 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/pvp2019-93441.

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Abstract Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Me
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Takakura, Kenichi, Shigeaki Tanaka, Tomomi Nakamura, Kazuhiro Chatani, and Yoshiyuki Kaji. "IASCC Evaluation Method for Irradiated Core Internal Structures in BWR Power Plants." In ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference. ASMEDC, 2010. http://dx.doi.org/10.1115/pvp2010-25293.

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Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research &amp; development study for the aging management &amp; maintenance of the nuclear power plants. Based on the JNES project results, JNES proposed “IASCC evaluation guide for BWR core internals”. The purpose of this paper is to describe the background of the guide, especially crack growth rate (CGR) tests for irradi
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