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http://dx.doi.org/10.1016/j.net.2019.05.008

Vessel failure sensitivities of an advanced reactor for SBLOCA  

Jhung, Myung Jo (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety)
Oh, Chang-Sik (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety)
Choi, Youngin (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety)
Kang, Sung-Sik (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety)
Publication Information
Nuclear Engineering and Technology / v.52, no.1, 2020 , pp. 185-191 More about this Journal
Abstract
Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×1019 n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.
Keywords
Reactor pressure vessel(RPV); Pressurized thermal shock(PTS); Failure probability; Small break loss of coolant; accident(SBLOCA);
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  • Reference
1 N.R.C. US, Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events, 10 CFR Part 50.61, U.S. Nuclear Regulatory Commission, Washington, DC, 1996.
2 N.R.C. US, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Rev.2, U.S. Nuclear Regulatory Commission, Washington, DC, 1988.
3 KINS, Reactor Probabilistic Integrity Evaluation (R-PIE) Code: User's Guide, KINS/RR-545, Korea Institute of Nuclear Safety, Daejeon, 2008.
4 C. Jang, "Treatment of the thermal-hydraulic uncertainties in the pressurized thermal shock analysis, Nucl. Eng. Des. 237 (2007) 143-152.   DOI
5 ASME, Boiler and Pressure Vessel Code, Section XI, The American Society of Mechanical Engineers, New York, 2004.
6 N.R.C. US, Format And Content of Plant-specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactor, Regulatory Guide 1.154, US Nuclear Regulatory Commission, Washington, DC, 1987.
7 N.R.C. US, Rates of Initiating Events at U.S. Nuclear Power Plants: 1987-1995, NUREG/CR-5750, U.S. Nuclear Regulatory Commission, Washington, DC, 1999.
8 H.G. Kim, K.S. Choi, Design characteristics of the advanced power reactor 1400, in: International Conference on Opportunities and Challenges for Water Cooled Reactors in the 21 Century, IAEA-CN-164, International Atomic Energy Agency, Vienna, 2009, pp. 96-98.
9 OECD/NEA, in: Proceedings of the CSNI Workshop on Structural Reliability Evaluation and Mechanical Probabilistic Approaches of NPP Components, NEA/CSNI/R(2007)18, NEA Committee on the Safety of Nuclear Installations, Paris, 2008.
10 IAEA, Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment, IAEA-TECDOC-1627, International Atomic Energy Agency, Vienna, 2010.
11 KINS, in: A-Pro2: Phase 2 ASINCO Project for Probabilistic Fracture Mechanics Analysis - Korea Results, KINS/RR-1483, Korea Institute of Nuclear Safety, Daejeon, 2016.
12 UKAEA, An Assessment of the Integrity of the PWR Pressure Vessels, Second Report by a Study Group under the Chairmanship of Dr. W. Marshall, United Kingdom Atomic Energy Authority, 1982.