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      KCI등재 SCIE SCOPUS

      Thermal and Structural Analysis of Calandria Vessel of a PHWR during a Severe Accident

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      https://www.riss.kr/link?id=A104143196

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      다국어 초록 (Multilingual Abstract) kakao i 다국어 번역

      In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed.
      The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost?In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.
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      In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. T...

      In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed.
      The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost?In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

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      참고문헌 (Reference)

      1 Ambroziak A, "The elsato- Viscoplastic Chaboche Model" 49-61, 2006

      2 Chen J., "Stress–strain curves for stainless steel at elevated temperatures" 28 : 229-239, 2006

      3 Mladin M., "SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions" 239 : 353-364, 2009

      4 Fletcher, C.D., "RELAP5/Mod3.2 code manual" Idaho Falls 1995

      5 Dupleac. D, "Generic CANDU 6 plant severe accident analysis employing SCAPSIM/ RELAP5 code" 239 : 2093-2103, 2009

      6 Chellapandi P., "Development of noniterative self correcting Solution (NONSS) method for the viscoplastic Analysis with the chaboche model" 43 : 621-654, 1998

      7 "CAST3M code"

      8 "ASME Section II,Materials, Part D - Properties"

      1 Ambroziak A, "The elsato- Viscoplastic Chaboche Model" 49-61, 2006

      2 Chen J., "Stress–strain curves for stainless steel at elevated temperatures" 28 : 229-239, 2006

      3 Mladin M., "SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions" 239 : 353-364, 2009

      4 Fletcher, C.D., "RELAP5/Mod3.2 code manual" Idaho Falls 1995

      5 Dupleac. D, "Generic CANDU 6 plant severe accident analysis employing SCAPSIM/ RELAP5 code" 239 : 2093-2103, 2009

      6 Chellapandi P., "Development of noniterative self correcting Solution (NONSS) method for the viscoplastic Analysis with the chaboche model" 43 : 621-654, 1998

      7 "CAST3M code"

      8 "ASME Section II,Materials, Part D - Properties"

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      학술지 이력

      학술지 이력
      연월일 이력구분 이력상세 등재구분
      2023 평가예정 해외DB학술지평가 신청대상 (해외등재 학술지 평가)
      2020-01-01 평가 등재학술지 유지 (해외등재 학술지 평가) KCI등재
      2014-01-01 평가 SCIE 등재 (등재유지) KCI등재
      2014-01-01 평가 SCOPUS 등재 (등재유지) KCI등재
      2011-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2009-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2007-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2006-07-31 학술지명변경 한글명 : Jorunal of the Korean Nuclear Society -> Nuclear Engineering and Technology
      외국어명 : 미등록 -> Nuclear Engineering and Technology
      KCI등재후보
      2004-01-01 평가 등재후보학술지 선정 (신규평가) KCI등재후보
      2003-01-01 평가 등재후보 1차 PASS (등재후보1차) KCI등재후보
      2002-01-01 평가 등재후보학술지 유지 (등재후보1차) KCI등재후보
      1999-01-01 평가 등재후보학술지 선정 (신규평가) KCI등재후보
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      학술지 인용정보

      학술지 인용정보
      기준연도 WOS-KCI 통합IF(2년) KCIF(2년) KCIF(3년)
      2016 1.04 0.17 0.77
      KCIF(4년) KCIF(5년) 중심성지수(3년) 즉시성지수
      0.63 0.56 0.343 0.11
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