Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Nuclear Engineering and Design

Volumn 234, Issue 1-3, 2004, Pages 129-145

  Vierow, K ^a   Liao, Y ^a   Johnson, J ^b   Kenton, M ^c   Gauntt, R ^d  

^a PURDUE UNIVERSITY   (United States)

^b Exelon Corporation   (United States)

^c CREARE INC   (United States)

^d SANDIA NATIONAL LABORATORIES   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

DEGRADATION; HYDRAULICS; NUCLEAR POWER PLANTS; RADIOISOTOPES; RISK ASSESSMENT; STEAM GENERATORS;

CONSERVATION EQUATIONS; MECHANISTIC PHYSICS MODEL; REACTOR SYSTEMS;

PRESSURIZED WATER REACTORS;

EID: 9944238708     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2004.09.001     Document Type: Article

References (18)

1. Bayless, P.D., Brownson, D.A., Dobbe, C.A., Jones, K.R., O'Brien, J.E., Pafford, D.J., Schlenker, L.D., Tung, F.X., 1995. Severe Accident Natural Circulation Studies at the INEL. NUREG/CR-6285.
2. Bergeron, K.D., et al., 1985. User's Manual for CONTAIN 1.0, A Computer Code for Severe Nuclear Reactor Accident Containment Analysis. NUREG/CR-4085, SAND84-1204.
3. Bixler, N., Schaperow, J., 1998. VICTORIA 2.0: A Mechanistic Model for Radionuclide Behavior in a Nuclear Reactor Coolant System under Severe Accident Conditions. NUREG/CR-6131, SAND93-2301.
4. Boyd, C.F., et al., 2003. CFD Analysis of l/7th Scale Steam Generator Inlet Plenum Mixing During a PWR Severe Accident. NUREG-1781.
5. Coryell, E.W., Siefken, L.J., Harvego, E.A., 2002. Development and application of SCDAP-3D. In: Proceedings of 10th International Conference on Nuclear Engineering, Arlington, VA, paper ICONE10-22638.
6. EPRI, 2002. Steam Generator Tube Integrity Risk Assessment: vol. 1: General Methodology, Revision 1, Report No. TR-107623-V1.
7. Fauske and Associates, Inc., 1994a. MAAP4 - Modular Accident Analysis Program for LWR Power Plants, vol. 2, Part 1: Code Structure and Theory, prepared for Electric Power Research Institute.
8. Fauske and Associates, Inc., 1994b. MAAP4 - Modular Accident Analysis Program for LWR Power Plants, vol. 2, Part 2 (description of each subroutine), prepared for Electric Power Research Institute.
9. Fauske & Associates, Inc., 1994c. MAAP4 - Modular Accident Analysis Program for LWR Power Plants vol. 3: Benchmarking, prepared for Electric Power Research Institute.
10. Gauntt, R.O., et al., 2002. MELCOR 1.8.5 Simulation of TMI-2 Phase 2 with an Enhanced 2-Dimensional In-Vessel Natural Circulation Model, ICONE10-22321, Proceedinds of the ICONE-10.
11. Hohorst, J.K., et al., 1994. TMI-2 Analysis using SCDAP/RELAP5/MOD3.1, INEL-94/0157.
12. Leonard, M.T., et al., 1996. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR&BWR accident sequences. Proceedings of the International Topical Meeting on PSA'96.
13. Paik, C.Y., et al., 1995. Proceedings of the International Conference on Probabilistic Safety Assessment Methodology and Applications, PSA 95, Seoul.
14. Sandia National Laboratories, 2000a. NUREG/CR-6119, Rev. 2, MELCOR Computer Code Manuals, vol. 1: Primer and User's Guide, Version 1.8.5, Prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research.
15. Sandia National Laboratories, 2000b. NUREG/CR-6119, Rev. 2, MELCOR Computer Code Manuals, vol. 2: Reference Manuals, Version 1.8.5, Prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research.
16. Siefken, L.J., Coryell, E.W., Harvego, H.A., Hohorst, J.K., 2001a. SCDAP/RELAP5/MOD 3.3 Code Manual, Code Architecture and Interface of Thermal-Hydraulic and Core Behavior Models, NUREG/CR-6150, vol. 1, Rev. 2, INEL-96/0422, Idaho National Engineering and Environmental Laboratory.
17. Siefken, L.J., Coryell, E.W., Harvego, E.A., Hohorst, J.K., 2001b. SCDAP/RELAP5/MOD 3.3 Code Manual, Modeling of Reactor Core and Vessel Behavior During Severe Accidents, NUREG/CR-6150, vol. 2, Rev. 2, INEL-96/0422, Idaho National Engineering and Environmental Laboratory.
18. U.S. Nuclear Regulatory Commission SGTR Severe Accident orking Group, 1998. Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture. NUREG-1570.