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Volumn 34, Issue 4, 2007, Pages 288-296

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

Author keywords

[No Author keywords available]

Indexed keywords

FORTRAN (PROGRAMMING LANGUAGE); INTEGRAL EQUATIONS; MATHEMATICAL MODELS; NUCLEAR REACTORS; SAFETY FACTOR; TRANSIENT ANALYSIS;

EID: 34147185179     PISSN: 03064549     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.anucene.2007.01.010     Document Type: Article
Times cited : (23)

References (16)
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    • Chen, Y.Z.1
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    • Conduction in nuclear fuel rods
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    • Ghiaasiaan, S.M.1
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    • A look-up table for fully developed film-boiling heat transfer
    • Groeneveld D.C. A look-up table for fully developed film-boiling heat transfer. Nucl. Eng. Des. 225 1 (2003) 83-97
    • (2003) Nucl. Eng. Des. , vol.225 , Issue.1 , pp. 83-97
    • Groeneveld, D.C.1
  • 8
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    • The transient flow model for pumps of reactor system
    • (in Chinese)
    • Guo Y.J., et al. The transient flow model for pumps of reactor system. Nucl. Sci. Eng. 15 3 (1995) 221-226 (in Chinese)
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    • Guo, Y.J.1
  • 9
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    • COOLOD-N: a computer code for the analysis of steady-state thermal-hydraulics in plate type research reactor
    • Kanminaga M. COOLOD-N: a computer code for the analysis of steady-state thermal-hydraulics in plate type research reactor. JAERI-M 90 21 (1990) 2201-2213
    • (1990) JAERI-M , vol.90 , Issue.21 , pp. 2201-2213
    • Kanminaga, M.1
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    • A new CHF correlation scheme proposed for vertical rectangular channels heated from both sides in nuclear research reactors
    • Sudo Y., and Kaminaga M. A new CHF correlation scheme proposed for vertical rectangular channels heated from both sides in nuclear research reactors. Trans. ASME J. Heat Transfer (1993) 115-126
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    • Development of a thermal-hydraulic analysis code for CARR
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    • Tian, W.X.1


* 이 정보는 Elsevier사의 SCOPUS DB에서 KISTI가 분석하여 추출한 것입니다.