한국안전학회지 (Journal of the Korean Society of Safety)

  • 제20권2호
  • /
  • Pages.32-37
  • /
  • 2005
  • /
  • 1738-3803(pISSN)
  • /
  • 2383-9953(eISSN)
한국안전학회 (The Korean Society of Safety)
권호 표지

SMART 연구로의 증기발생기 전열관 파열사고 민감도 분석

A Sensitivity Study of a Steam Generator Tube Rupture for the SMART-P

  • 김희경 (한국원자력연구소 신형원자로개발단) ;
  • 정영종 (한국원자력연구소 신형원자로개발단) ;
  • 양수형 (한국원자력연구소 신형원자로개발단) ;
  • 김희철 (한국원자력연구소 신형원자로개발단) ;
  • 지성균 (한국원자력연구소 신형원자로개발단)
  • Kim Hee-Kyung (Advanced Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Chung Young-Jong (Advanced Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Yang Soo-Hyung (Advanced Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Kim Hee-Cheol (Advanced Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Zee Sung Quun (Advanced Reactor Technology Development, Korea Atomic Energy Research Institute)
  • 발행 : 2005.06.01
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초록

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

키워드

참고문헌

  1. Young-Jong Chung et al., 'A Conservative Analysis Methodology for a Steamline Break Accident of the SMART-P Plant', Nuclear Technology, will be published
  2. 윤한영 외, 'TASS/SMR 열수력 모델 기술서', KAERI/TR-1835/2001
  3. 10CFR50, 'Domestic Licensing of Production and Utilization Facilities', Title 10, Code of Federal Regulations, Part 50, Appendix A, 'General Design Criteria for Nuclear Power Plant', April, 1993
  4. 10CFR100, 'Reactor Site Criteria', Part 100, April, 1962