Round Robin Analysis of Pressure-Temperature Limit Curve for Reactor Vessel

원자로 용기의 압력-온도 한계곡선 Round Robin 해석

  • Published : 2003.06.01

Abstract

Performed here is a comparative assessment study for the generation of the pressure-temperature limit curve of the reactor vessel. A round robin problem is proposed using the data available in Korea and all organizations interested in the generation of the pressure-temperature limit curve are invited. The problems consisting of 12 cases for cool-down are solved and their results are compared to generate a reference solution for the reference problem, which will be useful in the evaluation of the generation of the pressure-temperature limit curve in the future.

원자로 용기의 온도-압력 한계곡선을 위하여 국내공동비교연구를 수행하였다. 국내 원전의 데이터를 이용하여 국내 각 기관에서 온도-압력 한계곡선 작성에 사용하고 있는 방법 및 기법을 비교하기 위하여 round robin 해석을 제안하였고 주어진 문제에 대하여 각 기관이 문제를 해석한 후 결과를 제출하여 이들을 분석함으로써 온도-압력 한계곡선 작성에 대한 표준 해석 자료를 만들어 추후 평가에 이용할 수 있도록 하였다.

Keywords

References

  1. M. J. Jhung, Y. W. Park, 'Generation of Pressure/Temperature Limit Curve for Reactor Operation', Journal of the Computational Structural Engineering Institute of Korea, Vol.10, No.4, 1997, pp.155-164
  2. ASME, Fracture Toughness Criteria for Protection Against Failure, ASME Boiler and Pressure Vessel Code Section XI, Appendix G, The American Society of Mechanical Engineers, 1998
  3. W.Marshall, An Assessment of the Integrity of PWR Pressure Vessels, UKAEA, 1982
  4. Dickson, T. L. Bass B. R., and Williams, P. T., 'Validation of a Linear-Elastic Fracture Methodology for Postulated Flaws Embedded in the Wall of a Nuclear Reactor Pressure Vessel', The American Society of Mechanical Engineers, PVP-Vol.403, 2000
  5. ASME, 'Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel', ASME Boiler and Pressure Vessel Code Section XI, Code Case N-588, The American Society of Mechanical Engineers, 1998
  6. ASME, 'Alternative Reference Fracture Toughness for Development of P-T Limit Curves', ASME Boiler and Pressure Vessel Code Section XI, Code Case N-640, The American Society of Mechanical Engineers, 1998
  7. USNRC, Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events, 10 CFR 50 50.61, US Nuclear Regulatory Commission, 1998
  8. USNRC, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Rev.2, US Nuclear Regulatory Commission, 1988