원자로 내부구조물 균열개시 민감도에 미치는 영향인자 고찰
Review of Factors Affecting IASCC Initiation of Stainless Steel in PWRs
0
조회
0
다운로드
3
인용
0
참고문헌
키워드
초록
To safely operate domestic nuclear power plants approaching the end of their design life, the material degradationvmanagement strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.
Abstract
To safely operate domestic nuclear power plants approaching the end of their design life, the material degradationvmanagement strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.
참고문헌 (41)
- R. Gerard, "Situation of the baffle-former bolts in Belgian units", Proceedings of the 17th International Conference on Nuclear Engineering, ICONE17 (2009), ASME
- 임연수, "양성자 조사가 316 스테인리스강의 미세조직과 표면산화 특성에 미치는 영향", CORROSION SCIENCE AND TECHNOLOGY, 20(3), pp.158~168 (2021), 한국부식방식학회[DOI]
- Technical report on IASCC models of PWR reactor internals, "S.W. Kim", ~, pp.KAERI (2016), TR-6404
- D. Féron, "Behavior of stainless steels in pressurized water reactor primary circuits", Journal of Nuclear Materials, 427(1-3), pp.364~377 (2012), Elsevier BV[DOI]
- K. Arioka, , 11th International Symposium on Environmental of Materials in Nuclear Power Systems – Water Reactors (2003), NACE
- K. Arioka, "Effect of Temperature, Hydrogen and Boric Acid Concentration on IGSCC Susceptibility of Annealed 316 Stainless Steel", Colloque International Fontevraud 5 (2002), SFEN
- Gen Furutani, "Stress corrosion cracking on irradiated 316 stainless steel", Journal of Nuclear Materials, 288(2-3), pp.179~186 (2001), Elsevier BV[DOI]
- K. Arioka, "Influence of Boric Acid, Hydrogen Concentration and Grain Boundary Carbide on IGSCC Behaviors of SUS 316 under PWR Primary Water", 11th Int. Conf. Environmental Degradation of Materials in Nuclear Systems (2003), NACE Stevenson
- G. Furutani, , Journal of Nuclear Materials, 288, pp.279~ (2001)[DOI]
- K. Arioka, , Corrosion ’83 (1983)
- W. E. Ruther, "Effect of Temperature and Ionic Impurities at Very Low Concentrations on Stress Corrosion Cracking of AISI 304 Stainless Steel", CORROSION, 44(11), pp.791~799 (1988), Association for Materials Protection and Performance (AMPP)[DOI]
- S. Zhang, "Inhibition Effect of the Borate Ion on Intergranular Stress Corrosion Cracking of Sensitized Type 304 Stainless Steel", CORROSION, 54(6), pp.428~434 (1998), Association for Materials Protection and Performance (AMPP)[DOI]
- Kale J. Stephenson, "Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water", Journal of Nuclear Materials, 444(1-3), pp.331~341 (2014), Elsevier BV[DOI]
- Gary S. Was, "Irradiation-assisted stress corrosion cracking", Corrosion Reviews, 29(1-2), pp.7~ (2011), Walter de Gruyter GmbH[DOI]
- Irradiation Degradation Evaluation Technology Development of Nuclear Reactor Internals for Long Term Operation, "S. S Hwang", ~, pp.KAERI (2016), RR-4096
- K. Takakura, "Lifetime Evaluation for IASCC Initiation of Cold Worked 316 Stainless Steel’s BFB in PWR Primary Water", 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (2007), NACE
- P. D. Freyer, "Hot cell crack initiation testing of various heats of highly irradiated 316 stainless steel components obtained from three commercial PWRs", 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (2007), NACE
- Hiromasa NISHIOKA, "IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions", Journal of Nuclear Science and Technology, 45(10), pp.1072~1077 (2008), Informa UK Limited[DOI]
- Koji Fukuya, "A prediction model of IASCC initiation stress for bolts in PWR core internals", Nuclear Engineering and Design, 240(3), pp.473~481 (2010), Elsevier BV[DOI]
- R.W. Bosch, "Crack initiation testing of thimble tube material under PWR conditions to determine a stress threshold for IASCC", Journal of Nuclear Materials, 461, pp.112~121 (2015), Elsevier BV[DOI]
- P. L. Andresen, "Emerging Issues and Fundamental Processes in Environmental Cracking in Hot Water", CORROSION, 64(5), pp.439~464 (2008), Association for Materials Protection and Performance (AMPP)[DOI]
- G. S. Was, "Stress Corrosion Cracking Behavior of Alloys in Aggressive Nuclear Reactor Core Environments", CORROSION, 63(1), pp.19~45 (2007), Association for Materials Protection and Performance (AMPP)[DOI]
- R.W. Staehle, "Introduction to Initiation", Workshop on Detection, Avoidance, Mechanisms, Modeling, and Prediction of Stress Corrosion Cracking Initiation in Water-Cooled Nuclear Plants (2008)
- J. Conermann, "Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR", Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System – Water Reactors (2005), TMS
- O.K. Chopra, "A review of irradiation effects on LWR core internal materials – IASCC susceptibility and crack growth rates of austenitic stainless steels", Journal of Nuclear Materials, 409(3), pp.235~256 (2011), Elsevier BV[DOI]
- Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening, and Threshold Values, MRP-175, , ~, pp.EPRI Report 1012081 (2005)
- R. P. Shogan, , Proc. 10th Intl. Conf. on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (2001), NACE[DOI]
- H. Nishioka, , Proc. 13th Intl. Conf. on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (2007)
- S. Fyfitch, , Proc. 14th Intl. Conf. on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (2009), American Nuclear Society
- Koji Fukuya, "Current understanding of radiation-induced degradation in light water reactor structural materials", Journal of Nuclear Science and Technology, 50(3), pp.213~254 (2013), Informa UK Limited[DOI]
- R. Katsura, "Effect of stress on IASCC in irradiated austenitic stainless steels", Proceedings of the 6th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (1993), TMS
- A. J. Jacobs, "Use of a Constant Deflection Test to Evaluate Susceptibility to Irradiation-Assisted Stress Corrosion Cracking", CORROSION, 49(8), pp.650~655 (1993), Association for Materials Protection and Performance (AMPP)[DOI]
- A. Toivonen, "Post- irradiation SCC investigations on highly-irradiated core internals component materials", Proceedings of the 6th International Symposium Fontevraud 6, Contribution of Materials Investigations to Improve the Safety and Performance of LWRs (2006), SFEN
- K. Fujii, "Effects of dissolved hydrogen on IASCC behavior", Proceedings of the 7th International Symposium Fontevraud 7, Contribution of Materials Investigations to Improve the Safety and Performance of LWRs (2010), SFEN Fontevraud
- C. Pokor, "Determination of the time to failure curves as a function of stress for a highly irradiated AISI 304 stainless steel after constant load tests in simulated PWR water environment", Proceedings of the 7th International Symposium Fontevraud 7, Contribution of Materials Investigations to Improve the Safety and Performance of LWRs (2010), SFEN Fontevraud
- Boris Margolin, "Analysis of mechanisms inducing corrosion cracking of irradiated austenitic steels and development of a model for prediction of crack initiation", Engineering Failure Analysis, 107, pp.104235~ (2020), Elsevier BV[DOI]
- 김성우, "가압형 경수로 스테인리스강 내부 구조물의 조사유기 응력부식균열에 대한 통계적 수명 예측", 대한금속·재료학회지, 50(8), pp.583~589 (2012), 대한금속·재료학회[DOI]
- Y. Chimi, "Effects of locally deformed structure on oxide film properties in neutron-irradiated austenitic stainless steel", 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors (2015), NACE
- S. S. Raiman, "Composition and morphology of the oxide film formed on 316l stainless steel during in-situ proton irradiation in PWR conditions", 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors (2015), NACE
- T. Fukumura, "Grain Boundary Oxidation of Neutron Irradiated Stainless Steels in Simulated PWR Water", 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors (2015), NACE
- W.J. Phythian, "Microstructural evolution in reactor pressure vessel steels", Journal of Nuclear Materials, 205, pp.162~177 (1993), Elsevier BV[DOI]