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http://dx.doi.org/10.3795/KSME-A.2002.26.3.505

Construction of the P-T Limit Curve for the Nuclear Reactor Pressure Vessel Using Influence Coefficient Methods : Cooldown Curve  

Jang, Chang-Hui (Korea Electric Power Corporation)
Publication Information
Transactions of the Korean Society of Mechanical Engineers A / v.26, no.3, 2002 , pp. 505-513 More about this Journal
Abstract
During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.
Keywords
Reactor Pressure Vessel; P-T Limit Curve; Linear Elastic Fracture Mechanics; Influence Coefficient; Sensitivity Analysis;
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  • Reference
1 PVRC Ad Hoc Group on Toughness Requrements, 1972, 'PVRC Recommendations on Toughness Requirements for Ferritic Materials,' WRC BULLETIN 175
2 USNRC, 1995, 'Fracture Toughness Requirements,' 10CFR50 App. G.
3 ASME Boiler and Pressure Vessel Code Sec. XI, 1998, 'Fracture Toughness Criteria for Protection Against Failure,' Appendix G.
4 Jang, C. H., 2000, 'The Effect of Reference Flaw Size on P-T Limit Curves for Pressurized Water Reactor,' Proceeding of PVP2000 Conference, July 23-27, Seattle, USA
5 이택진, 최재붕, 김영진, 박윤원, 정명조, 2000, '원자로용기 건전성평가를 위한 RVIES 시스템의 개발,' 대한기계학회논문집 (A) 제 24 권, 제 8 호, pp. 2083-2090   과학기술학회마을
6 ASME Boiler and Pressure Vessel Code Sec. II, 1995, 'Materials,' Part D
7 장창희, 문호림, 정일석, 홍승열, 2001, '개선된 확률론적 파괴역학해석 전산코드개발 : VINTIN,' 한국원자력학회 2001 춘계학술발표회논문집, 2001. 5. 24-25, 제주   과학기술학회마을
8 KAERI, 2000, 'The Final Report for the 5-th Surveillance Test of the Reactor Pressure Vessel Material (Capsule P) of Kori Nuclear Power Plant Unit 1,' KAERI-ST-K1-003/00
9 Marshall, W., 1982, 'An assessment of the integrity of PWR pressure vessels,' Second Report by a Study Group under the chairmanship of Dr. W. Marshall, UKAEA
10 ASME, 1997, ASME Boiler and Pressure Vessel Code Sec. XI, 'Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,' Code Case N-588
11 ASME, 1999, ASME Boiler and Pressure Vessel Code Sec. XI, 'Alternative Reference Fracture Toughness for Development of P-T Limit Curves,' Code Case N-640
12 USNRC, 1996, 'Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,' 10 CFR 50 50.61
13 USNRC, 1988, 'Radiation Embrittlement of Reactor Vessel Materials,' Regulatory Guide 1.99, Rev. 2
14 EPRI, 1993, 'Reactor Coolant System Heatup/Cooldown Curve Calculator,' EPRI TR-102552
15 ASME Boiler and Pressure Vessel Code Sec. III, 1989, 'Protection Against Nonductile Failure,' Appendix G.