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0012482956
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NIREG/CR-5249
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Boyack, B., Duffey, R., Griffith, P., Lellouche, G., Rohatgi, U., Wilson, G., Wulff, W., Zuber, N., 1989. Quantifying reactor safety margins - application of code scaling, applicability, and uncertainty evaluation methodology to a large-break loss-of-coolant accident, NIREG/CR-5249.
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Quantifying Reactor Safety Margins - Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large-break Loss-of-coolant Accident
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Boyack, B.1
Duffey, R.2
Griffith, P.3
Lellouche, G.4
Rohatgi, U.5
Wilson, G.6
Wulff, W.7
Zuber, N.8
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3
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0346332435
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Application of Westinghouse large break LOCA SECY UPI best estimate methodology to Kori Unit 1
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Dederer, S.I., Spaargaren, J.S., Lee, J.H., 1996. Application of Westinghouse large break LOCA SECY UPI best estimate methodology to Kori Unit 1, The 11th KAIF/KNS Annual Conference, pp. 811.
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The 11th KAIF/KNS Annual Conference
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Dederer, S.I.1
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0347593367
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JAERI-Memo
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Iguchi, T., Oyama, T., Arase, T., Niitsuma, Y., Nakajima, K., Chiba, T., Okabe, K., Sugimoto, J., Akimoto, H., Okubo, T., Komori, K., Sonobe, H., Hojo, T., Owada, A., Winsel, C.E., Murao, Y., 1985a. Data report on large scale reflood test-96, CCTF core-II test C2-13 (Run 072), JAERI-Memo 60-157.
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Data Report on Large Scale Reflood Test-96, CCTF Core-II Test C2-13 (Run 072)
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Iguchi, T.1
Oyama, T.2
Arase, T.3
Niitsuma, Y.4
Nakajima, K.5
Chiba, T.6
Okabe, K.7
Sugimoto, J.8
Akimoto, H.9
Okubo, T.10
Komori, K.11
Sonobe, H.12
Hojo, T.13
Owada, A.14
Winsel, C.E.15
Murao, Y.16
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0346332426
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CCTF core-II test C2-16 (Run 076)
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JAERI-Memo
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Iguchi, T., Oyama, T., Arase, T., Niitsuma, Y., Nakajima, K., Chiba, T., Sugimoto, J., Akimoto, H., Okubo, T., Komori, K., Sonobe, H., Hojo, T., Owada, A., Feldman, E.M., Murao, Y., 1985b. Data report on large scale reflood test-99, CCTF core-II test C2-16 (Run 076), JAERI-Memo 60-158.
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Data Report on Large Scale Reflood Test-99
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Iguchi, T.1
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Niitsuma, Y.4
Nakajima, K.5
Chiba, T.6
Sugimoto, J.7
Akimoto, H.8
Okubo, T.9
Komori, K.10
Sonobe, H.11
Hojo, T.12
Owada, A.13
Feldman, E.M.14
Murao, Y.15
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7
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0347593366
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2D/3D program UPTF quick look report, test number 20 upper plenum injection simulation test
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Siemens (UB KWU), 1988a. 2D/3D program UPTF quick look report, test number 20 upper plenum injection simulation test, Siemens Report U9 316/88/07.
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Siemens Report U9 316/88/07
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8
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0347593366
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2D/3D program UPTF experimental data report, test number 20 upper plenum injection simulation test
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Siemens (UB KWU), 1988b. 2D/3D program UPTF experimental data report, test number 20 upper plenum injection simulation test", Siemens Report U9 316/88/08.
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Siemens Report U9 316/88/08
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9
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0346332427
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Simulation and analysis of cylindrical core test facility upper plenum injection experiments
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Takeuchi, K., Nissley, M.E., Bajorek, S.M., Cho, C.S., 1998. Simulation and analysis of cylindrical core test facility upper plenum injection experiments, to be published in the 11th International Heat Transfer Conference.
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11th International Heat Transfer Conference
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Takeuchi, K.1
Nissley, M.E.2
Bajorek, S.M.3
Cho, C.S.4
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0032178688
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Analyses of sub-cooled CCFL tests for evaluation of WCOBRA/TRAC applicability
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to be published
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Takeuchi, K., Nissley, M.E., Young, M.Y. Analyses of sub-cooled CCFL tests for evaluation of WCOBRA/TRAC applicability, Nucl. Eng. Des., to be published.
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Takeuchi, K.1
Nissley, M.E.2
Young, M.Y.3
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11
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0003599718
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NUREG/CR-3046
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Thurgood, M.J., Kelly, J.M., Guidotti, T.E., Koht, R.J., Crowell, K.R., 1982. COBRA/TRAC-a thermal-hydraulics code for transient analysis of nuclear reactor vessels and primary coolant systems, NUREG/CR-3046.
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COBRA/TRAC-a Thermal-hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems
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Thurgood, M.J.1
Kelly, J.M.2
Guidotti, T.E.3
Koht, R.J.4
Crowell, K.R.5
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12
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0346962287
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Development of a best estimate LOCa methodology
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Beijing, China, Paper 03
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Young, M.Y., Ginsgerg, A., Nissley, M.E., Bajorek, S.M., 1997. Development of a best estimate LOCA methodology, Proceedings of the Fifth International Topical Meeting on Nuclear Thermal Hydraulics, Operations, and Safety (NUTHOS-5), Beijing, China, Paper 03.
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Proceedings of the Fifth International Topical Meeting on Nuclear Thermal Hydraulics, Operations, and Safety (NUTHOS-5)
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Young, M.Y.1
Ginsgerg, A.2
Nissley, M.E.3
Bajorek, S.M.4
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