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Volumn 2008, Issue , 2008, Pages

An overview of westinghouse realistic large break Loca evaluation model

Author keywords

[No Author keywords available]

Indexed keywords

CODE SCALING; EVALUATION MODELING; INPUT UNCERTAINTY; NON-PARAMETRIC STATISTICS; SAFETY ANALYSIS; THERMAL-HYDRAULIC CODES; TRANSIENT SIMULATION; WESTINGHOUSE;

EID: 70349870406     PISSN: 16876075     EISSN: 16876083     Source Type: Journal    
DOI: 10.1155/2008/498737     Document Type: Review
Times cited : (80)

References (46)
  • 2
    • 0346332422 scopus 로고    scopus 로고
    • Code qualification document for best estimate LOCA analysis
    • Pittsburgh, Pa, USA Westinghouse Electric Company
    • Bajorek S. M., Code qualification document for best estimate LOCA analysis. Tech. Rep. WCAP 12945 NP-A 1998 Pittsburgh, Pa, USA Westinghouse Electric Company
    • (1998) Tech. Rep. WCAP 12945 NP-A
    • Bajorek, S.M.1
  • 3
    • 0032203387 scopus 로고    scopus 로고
    • Application of code scaling applicability and uncertainty methodology to the large break loss of coolant
    • PII S0029549398002179
    • Young M. Y., Bajorek S. M., Nissley M. E., Hochreiter L. E., Application of code scaling applicability and uncertainty methodology to the large break loss of coolant. Nuclear Engineering and Design 1998 186 1-2 39 52 10.1016/S0029-5493(98)00217-9 (Pubitemid 128421051)
    • (1998) Nuclear Engineering and Design , vol.186 , Issue.1-2 , pp. 39-52
    • Young, M.Y.1    Bajorek, S.M.2    Nissley, M.E.3    Hochreiter, L.E.4
  • 4
    • 0346307940 scopus 로고    scopus 로고
    • Assessment of flooding in a best estimate thermal hydraulic code (W + (combining low line)COBRA/TRAC)
    • PII S0029549398002246
    • Takeuchi K., Young M. Y., Assessment of flooding in a best estimate thermal hydraulic code (WCOBRA/TRAC). Nuclear Engineering and Design 1998 186 1-2 225 255 10.1016/S0029-5493(98)00224-6 (Pubitemid 128421848)
    • (1998) Nuclear Engineering and Design , vol.186 , Issue.1-2 , pp. 225-255
    • Takeuchi, K.1    Young, M.Y.2
  • 5
    • 0347568942 scopus 로고    scopus 로고
    • Scaling effects predicted by W + (combining low line)COBRA/TRAC for UPI plant best estimate LOCA
    • PII S0029549398002362
    • Takeuchi K., Nissley M. E., Spaargaren J. S., Dederer S. I., Scaling effects predicted by WCOBRA/TRAC for UPI plant best estimate LOCA. Nuclear Engineering and Design 1998 186 1-2 257 278 10.1016/S0029-5493(98)00236-2 (Pubitemid 128421849)
    • (1998) Nuclear Engineering and Design , vol.186 , Issue.1-2 , pp. 257-278
    • Takeuchi, K.1    Nissley, M.E.2    Spaargaren, J.S.3    Dederer, S.I.4
  • 8
    • 10644233900 scopus 로고    scopus 로고
    • Comparison of realistic large break LOCA analyses of a 3-loop westinghouse plant using response surface and statistical sampling techniques
    • April 2004 Arlington, Va, USA
    • Muftuoglu K., Ohkawa K., Frepoli C., Nissley M. E., Comparison of realistic large break LOCA analyses of a 3-loop westinghouse plant using response surface and statistical sampling techniques. 3 Proceedings of the 12th International Conference on Nuclear Engineering (ICONE '04) April 2004 Arlington, Va, USA 811 818
    • Proceedings of the 12th International Conference on Nuclear Engineering (ICONE '04) , vol.3 , pp. 811-818
    • Muftuoglu, K.1    Ohkawa, K.2    Frepoli, C.3    Nissley, M.E.4
  • 12
    • 10644279332 scopus 로고    scopus 로고
    • Realistic large-break LOCA evaluation methodology using the automated statistical treatment of uncertainty method (ASTRUM)
    • Pittsburgh, Pa, USA Westinghouse Electric Company
    • Nissley M. E., Frepoli C., Ohkawa K., Muftuoglu K., Realistic large-break LOCA evaluation methodology using the automated statistical treatment of uncertainty method (ASTRUM). Tech. Rep. WCAP-16009-NP 2003 Pittsburgh, Pa, USA Westinghouse Electric Company
    • (2003) Tech. Rep. WCAP-16009-NP
    • Nissley, M.E.1    Frepoli, C.2    Ohkawa, K.3    Muftuoglu, K.4
  • 13
    • 33845742122 scopus 로고    scopus 로고
    • Notes on the implementation of non-parametric statistics within the westinghouse realistic large break LOCA evaluation model (ASTRUM)
    • June 2006 Reno, Nev, USA Paper 625
    • Frepoli C., Oriani L., Notes on the implementation of non-parametric statistics within the westinghouse realistic large break LOCA evaluation model (ASTRUM). Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP '06) June 2006 Reno, Nev, USA 1059 1065. Paper 625
    • Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP '06) , pp. 1059-1065
    • Frepoli, C.1    Oriani, L.2
  • 16
    • 11144237000 scopus 로고
    • Quantifying reactor safety margins - 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology
    • 10.1016/0029-5493(90)90071-5
    • Boyack B. E., Catton I., Duffey R. B., Quantifying reactor safety margins-1: an overview of the code scaling, applicability, and uncertainty evaluation methodology. Nuclear Engineering and Design 1990 119 1 1 15 10.1016/0029-5493(90)90071-5
    • (1990) Nuclear Engineering and Design , vol.119 , Issue.1 , pp. 1-15
    • Boyack, B.E.1    Catton, I.2    Duffey, R.B.3
  • 17
    • 0344528431 scopus 로고
    • TPG response to the foregoing letters-to-the-editor
    • 10.1016/0029-5493(92)90243-O
    • Boyack B. E., Wilson G. E., Catton I., TPG response to the foregoing letters-to-the-editor. Nuclear Engineering and Design 1992 132 3 431 436 10.1016/0029-5493(92)90243-O
    • (1992) Nuclear Engineering and Design , vol.132 , Issue.3 , pp. 431-436
    • Boyack, B.E.1    Wilson, G.E.2    Catton, I.3
  • 18
    • 38249013813 scopus 로고
    • Comments on 'quantifying reactor safety margins'
    • 10.1016/0029-5493(92)90236-O
    • Hochreiter L. E., Comments on 'quantifying reactor safety margins'. Nuclear Engineering and Design 1992 132 3 409 410 10.1016/0029-5493(92)90236-O
    • (1992) Nuclear Engineering and Design , vol.132 , Issue.3 , pp. 409-410
    • Hochreiter, L.E.1
  • 21
    • 9344262986 scopus 로고    scopus 로고
    • Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios
    • Washington, DC, USA USNRC
    • Wilson G. E., Fletcher C. D., Davis C. B., Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios. NUREG/CR-6541 1997 Washington, DC, USA USNRC
    • (1997) NUREG/CR-6541
    • Wilson, G.E.1    Fletcher, C.D.2    Davis, C.B.3
  • 26
    • 0346962280 scopus 로고
    • 8th Water Reactor Safety Information Meetin
    • Thurgood M. J., COBRA-TF development. 1980. 8th Water Reactor Safety Information Meetin
    • (1980) COBRA-TF Development
    • Thurgood, M.J.1
  • 27
    • 0343059282 scopus 로고
    • TRAC-PD2: Advanced best-estimate computer program for pressurized water reactor loss-of-coolant accident analysis
    • Los Alamos, NM, USA Los Alamos National Laboratory
    • Liles D. R., TRAC-PD2: advanced best-estimate computer program for pressurized water reactor loss-of-coolant accident analysis. NUREG/CR-2054 1981 Los Alamos, NM, USA Los Alamos National Laboratory
    • (1981) NUREG/CR-2054
    • Liles, D.R.1
  • 30
    • 0343345570 scopus 로고
    • User effects on the thermal-hydraulic transient system code calculations
    • 10.1016/0029-5493(93)90065-H
    • Aksan S. N., D'Auria F., Städtke H., User effects on the thermal-hydraulic transient system code calculations. Nuclear Engineering and Design 1993 145 1-2 159 174 10.1016/0029-5493(93)90065-H
    • (1993) Nuclear Engineering and Design , vol.145 , Issue.1-2 , pp. 159-174
    • Aksan, S.N.1    D'Auria, F.2    Städtke, H.3
  • 35
    • 20444455903 scopus 로고    scopus 로고
    • AREVA's realistic large break LOCA analysis methodology
    • DOI 10.1016/j.nucengdes.2005.02.004, PII S0029549305000762
    • Martin R. P., O'Dell L. D., AREVA's realistic large break LOCA analysis methodology. Nuclear Engineering and Design 2005 235 16 1713 1725 10.1016/j.nucengdes.2005.02.004 (Pubitemid 40819652)
    • (2005) Nuclear Engineering and Design , vol.235 , Issue.16 , pp. 1713-1725
    • Martin, R.P.1    O'Dell, L.D.2
  • 38
    • 0002765765 scopus 로고
    • Determination of sample sizes for setting tolerance limits
    • Wilks S. S., Determination of sample sizes for setting tolerance limits. The Annals of Mathematical Statistics 1941 12 1 91 96
    • (1941) The Annals of Mathematical Statistics , vol.12 , Issue.1 , pp. 91-96
    • Wilks, S.S.1
  • 39
    • 0037409959 scopus 로고    scopus 로고
    • Statistical aspects of best estimate method-I
    • 10.1016/S0951-8320(03)00022-X
    • Guba A., Makai M., Pál L., Statistical aspects of best estimate method-I. Reliability Engineering & System Safety 2003 80 3 217 232 10.1016/S0951-8320(03)00022-X
    • (2003) Reliability Engineering & System Safety , vol.80 , Issue.3 , pp. 217-232
    • Guba, A.1    Makai, M.2    Pál, L.3
  • 40
    • 33845784707 scopus 로고    scopus 로고
    • Reply to contribution of Graham B. Wallis
    • 10.1016/S0951-8320(03)00035-8
    • Makai M., Pál L., Reply to contribution of Graham B. Wallis. Reliability Engineering & System Safety 2003 80 3 313 317 10.1016/S0951-8320(03)00035-8
    • (2003) Reliability Engineering & System Safety , vol.80 , Issue.3 , pp. 313-317
    • Makai, M.1    Pál, L.2
  • 41
    • 0037409943 scopus 로고    scopus 로고
    • Contribution to the paper 'Statistical aspects of best estimate method-1' by A. Guba, M. Makai, L. Pál
    • 10.1016/S0951-8320(03)00034-6
    • Wallis G. B., Contribution to the paper 'Statistical aspects of best estimate method-1' by A. Guba, M. Makai, L. Pál. Reliability Engineering & System Safety 2003 80 3 309 317 10.1016/S0951-8320(03)00034-6
    • (2003) Reliability Engineering & System Safety , vol.80 , Issue.3 , pp. 309-317
    • Wallis, G.B.1
  • 42
    • 4544345621 scopus 로고    scopus 로고
    • Reply to "comments on 'Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties' by W.T. Nutt and G.B. Wallis," by Y. Orechwa
    • 10.1016/j.ress.2004.04.007
    • Wallis G. B., Nutt W. T., Reply to "Comments on 'Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties' by W.T. Nutt and G.B. Wallis," by Y. Orechwa. Reliability Engineering & System Safety 2005 87 1 137 145 10.1016/j.ress.2004.04.007
    • (2005) Reliability Engineering & System Safety , vol.87 , Issue.1 , pp. 137-145
    • Wallis, G.B.1    Nutt, W.T.2
  • 45
    • 4544362349 scopus 로고    scopus 로고
    • Comments on 'Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties' by W.T. Nutt and G.B. Wallis
    • 10.1016/j.ress.2004.04.002
    • Orechwa Y., Comments on 'Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties' by W.T. Nutt and G.B. Wallis. Reliability Engineering & System Safety 2005 87 1 133 135 10.1016/j.ress.2004.04.002
    • (2005) Reliability Engineering & System Safety , vol.87 , Issue.1 , pp. 133-135
    • Orechwa, Y.1
  • 46
    • 0242691795 scopus 로고    scopus 로고
    • Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties
    • 10.1016/j.ress.2003.08.008
    • Nutt W. T., Wallis G. B., Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties. Reliability Engineering & System Safety 2004 83 1 57 77 10.1016/j.ress.2003.08.008
    • (2004) Reliability Engineering & System Safety , vol.83 , Issue.1 , pp. 57-77
    • Nutt, W.T.1    Wallis, G.B.2


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