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http://dx.doi.org/10.12989/sem.2011.37.3.257

Comparison of vessel failure probabilities during PTS for Korean nuclear power plants  

Jhung, M.J. (Korea Institute of Nuclear Safety)
Choi, Y.H. (Korea Institute of Nuclear Safety)
Chang, Y.S. (Kyung Hee University)
Publication Information
Structural Engineering and Mechanics / v.37, no.3, 2011 , pp. 257-265 More about this Journal
Abstract
Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.
Keywords
pressurized thermal shock; reactor pressure vessel; structural integrity; stress intensity factor; failure probability; through-wall cracking;
Citations & Related Records
Times Cited By KSCI : 2  (Citation Analysis)
Times Cited By Web Of Science : 0  (Related Records In Web of Science)
Times Cited By SCOPUS : 0
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1 ASME (2004), Boiler and Pressure Vessel Code, Section XI, The American Society of Mechanical Engineers, New York.
2 Baek, W.P., Yang, J.E. and Ha, J.J. (2009), "Safety assessment of Korean nuclear facilities: current status and future", Nucl. Eng. Technol., 41(4), 391-402.   DOI
3 Dickson, T.L. (1994), FAVOR: A Fracture Analysis Code for Nuclear Reactor Pressure Vessels, Release 9401, ORNL/NRC/LTR/94/1, Marietta Energy Systems, Inc., Oak Ridge National Laboratory, February.
4 EPRI (1996), RETRAN-3D, A Program for Transient Thermal Hydraulic Analysis of Complex Fluid Flow System, NP-7450, Electric Power Research Institute.
5 Jang, C. (2007), "Treatment of the thermal-hydraulic uncertainties in the pressurized thermal shock analysis", Nucl. Eng. Des., 237, 143-152.   DOI   ScienceOn
6 Jang, C., Kang, S.C., Moonn, H.R., Jeong, I.S. and Kim, T.R. (2003), "The effects of the stainless steel cladding in pressurized thermal shock evaluation", Nucl. Eng. Des., 226, 127-140.   DOI   ScienceOn
7 Jhung, M.J. (2008), Reactor Probabilistic Integrity Evaluation (R-PIE) Code: User's Guide, KINS/RR-545, Korea Institute of Nuclear Safety, Daejeon.
8 Jhung, M.J., Choi, Y.H. and Jang, C. (2009), "Structural integrity of reactor pressure vessel for small break loss of coolant accident", J. Nuclear Sci. Technol., 46(3), 310-315.   DOI   ScienceOn
9 Raju, I.S. and Newman Jr., J.C. (1982), "Stress intensity factors for internal and external surface cracks in cylindrical vessels", J. Pres. Ves. Tech., 104, 293-298.   DOI
10 Ryu, Y.H. (2009), "Nuclear regulatory research in Korea: achievements and future direction", Nucl. Eng. Technol., 41(4), 403-412.   DOI
11 UKAEA (1982), An Assessment of the Integrity of the PWR Vessels, Second report by a study group under the chairmanship of D.W. Marshall, United Kingdom Atomic Energy Authority.
12 USNRC (1987), Format and Content of Plant-specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactor, Regulatory Guide 1.154, US Nuclear Regulatory Commission.
13 USNRC (1988), Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Rev.2, US Nuclear Regulatory Commission, Washington, DC.
14 USNRC (1996), Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10CFR50.61, US Nuclear Regulatory Commission, Washington, DC.