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A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

참고문헌 (12)

  1. KEPCO, 2001, APR-1400 SSAR, Chapter 15 
  2. G. E. Wilson, B. E. Boyack, 1998, The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis, Nucl. Eng. Des. 186, 23-37 
  3. G. E. Wilson, C. D. Fletcher, C. B. Davis, J. D. Burtt, T. J. Boucher, 1997, Phenomena Identification and Ranking Tables for Westinghouse AP600 SBLOCA, MSLB, and SGTR Scenarios, NUREG/CR-6451 
  4. J. N. Reyes Jr., L. Hochreiter, 1998, Scaling analysis for the OSU AP600 test facility (APEX), Nucl. Eng. Des., 186, 53-109 
  5. B. D. Chung, J. H. Song, S. K. Sim, W. J. Lee, J. J. Jeong., 1997, Development of Preliminary PIRT of Thermal-Hydraulic Phenomena for 330MWt SMART Integral Reactor, KAERI/TR-912/97, KAERI 
  6. G. E. Wilson, B. E. Boyack, B. D. Cheung, L. E. Hochreiter, J. N. Reyes, J. M. Cozzol, 2001, Phenomena Identification and Ranking Tabulation, KNGR LBLOCA, INEEL, WFO861702 
  7. B. D. Chung et. al., Phenomena Identification and Ranking Tabulation for APR-1400 Direct Vessel Injection Line Break, C.D. Rom, F00211, The 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Seoul, Korea, October 5-9, 2003 
  8. ABB-CE, 1987, CESEC user^i manual, CEECES-78 Rev. O-P 
  9. C.S. Lee, 1996, Analysis methodology for the post-trip return to power steam line break event, KAERI/TR-698/96, KAERI 
  10. W.J. Lee et. al., 2002, Development of Realistic Thermal Hydraulic System Analysis Code, KAERI/RR-2235/2001, KAERI 
  11. H. Song, K. H. Bae, 2000, Evaluation of Analytically Scaled Model of a Pressurized Water Reactor Using the RELAP5/MOD3 Computer Code, Nucl. Eng. Des., 199, pp.215-225 
  12. W. P. Baek, C. H. Song, B. J. Yun, T. S. Kwon, S. K. Moon, S. J. Lee, KAERI Integral Effect Test Program and the ATLAS Design, F00201, The 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Seoul, Korea, October 5-9, (2003) 

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