Phenomena Identification and Ranking Table for the APR-1400 Main Steam Line Break

정재준; 정법동; 방영석; 송진호; 백원필

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자료유형: 학술저널

저자정보: 정재준 정법동 방영석 송진호 백원필

저널정보: 한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제36권 제5호

발행연도: 2004.1

수록면: 388 - 402 (15page)

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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 selected 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.

#PIRT #main steam line break #APR-1400 #MARS #rank #knowledge level

참고문헌 (10)

참고문헌 신청

/ 1996 / Analysis methodology for thepost-trip return to power steam line break event

/ 1997 / Development of a Preliminary PIRT of Thermal-Hydraulic Phenomena for 330MWt SMART Integral Reactor

/ 1997 / Phenomena Identification and Ranking Tables for Westinghouse AP600 SBLOCA, MSLB, andSGTR Scenarios

/ 1998 / Scalinganalysis for the OSU AP600 test facility (APEX) 186 : 53 ~ 109

/ 1998 / The role ofthe PIRT process in experiments, codedevelopment and code applications associatedwith reactor safety analysis 186 : 23 ~ 37

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