AREVA's realistic large break LOCA analysis methodology

Nuclear Engineering and Design

Volumn 235, Issue 16, 2005, Pages 1713-1725

  Martin, Robert P ^a   O'Dell, Larry D ^b  

^a AREVA NP INC   (United States)

^b AREVA NP   (France)

Author keywords

[No Author keywords available]

Indexed keywords

CODES (SYMBOLS); COMPUTATIONAL METHODS; COOLING SYSTEMS; EVALUATION; NUCLEAR ENGINEERING; NUCLEAR POWER PLANTS; PRESSURIZED WATER REACTORS; THERMAL EFFECTS;

COMPUTER CODES; EMERGENCY CORE COOLING SYSTEMS (ECCS); EVALUATION METHODOLIGIES (EM); PEAK CLADDING TEMPERATURE (PCT);

LOSS OF COOLANT ACCIDENTS;

EID: 20444455903     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2005.02.004     Document Type: Article

References (53)

1. Adams, J.P., et al., 1982. Quick-Look Report on LOFT Experiment L2-5, EGG-LOFT-5921.
2. Adams, J.P., et al., 1983. Quick-Look Report on OECD LOFT Experiment LP-02-6, OECD LOFT-T-3404, EG&G Idaho, Inc.
3. Adams, J.P., et al., 1984. Quick-Look Report on OECD LOFT Experiment LP-LB-1, OECD LOFT-T-3504. EG&G Idaho, Inc.
4. Akimoto, et al., 1987. Evaluation Report on CCTF Core-II Reflood Test C2-8 (Run 67) - Effect of System Pressure, JAERI-M-87-001.
5. Akimoto, et al., 987. Evaluation Report on CCTF Core-II Reflood Test C2-9 (Run 68) - Effect of LPCI Flow Rate, JAERI-M-87-002.
6. Anklam, T.M., Miller, R.J., White, M.D., 1982. Experimental Investigations of Uncovered Bundle Heat Transfer and Two-Phase Mixture Level Swell Under High Pressure Low Heat Flux Conditions, NUREG/CR-2456, ORNL-5848.
7. AREVA, 2001. Realistic Large Break LOCA Methodology, AREVA Document EMF-2103(NP) Revision.
8. AREVA, 2002. RODEX3A Code Manual, AREVA Document EMF-1557.
9. Bennett, A.W., Hewitt, G.F., Kearsey, H.A., Keeys, R.K.F., 1967. Heat Transfer to Steam-Water Mixtures Flowing in Uniformly Heated Tubes in Which the Critical Heat Flux has been Exceeded. UKAEA Research Group Report, AERE-R 5373.
10. Berkow, H.N., 2003. Safety Evaluation on Framatome ANP Topical Report Accident Methodology for Pressurized Water Reactors, EMF-2103 (P) Revision 0, Realistic Large Break Loss-of-Coolant (Letter to J.F. Mallay).
11. Cadek, F.F., et al., 1971. PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report, WCAP-7665.
12. Carlson, K.E., et al., 1990. RELAP5/MOD3 Code Manual, NUREG/CR-5535, EGG-2596.
13. Chelemer, H., Chu, P.T., Hochreiter, L.E., 1973. THINC-IV - An Improved Program for Thermal-Hydraulic Analysis of Rod Bundle Cores. Westinghouse Electric Corporation, WCAP 7956.
14. Findlay, J.A., 1981. BWR Refill-Reflood Program Task 4.8 - Model Qualification Task Plan, NUREG/CR-1899, EPRI NP-1527, GEAP-24898.
15. FLECHT SEASET Program, 1980. PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report, vols. 1 and 2. NUREG/CR-1532, EPRI NP-1459, WCAP-9699.
16. E. Hofer The GRS programme package for uncertainty and sensitivity analysis Proceedings of the Seminar on Methods and Codes for Assessing the Off-site Consequences of Nuclear Accidents, EUR 13013 1990 Commission of the European Communities
17. Holmes, B.J., 1991. I25 Comparison Report, NEA/CSNI/R(91)1, AEA-TRS-1043.
18. Iguchi, et al., 1983. Data Report on Large Scale Reflood Test-43 - CCTF Core Shakedown Test C2-SH2 (Run 54), JAERI-M-58-155.
19. R.T. Lahey, and F.J. Moody The Thermal-Hydraulics of a Boiling Water Nuclear Reactor second ed. 1993 American Nuclear Society La Grange Park, Illinois
20. Lilly, G.P., Hochreiter, L.E., 1975. Mixing of ECC water with steam: 1/3 scale test and summary. EPRI Report, EPRI-294-2.
21. MPR Associates, 1989. Research Information Report of the Slab Core Test Facility (SCTF) Core II Test Series, MPR-1115.
22. Nylund, O., et al., 1968. Hydrodynamic and Heat Transfer Measurements on a Full-Scale Simulated 36-Rod Marviken Fuel Element with Uniform Heat Flux Distribution. ASEA and AB Atomenergi, Frigg-2, R4-447/RTL-1007.
23. Oak Ridge National Laboratory (ORNL), 1981. ORNL Small-Break LOCA Heat Transfer Test Series I: High-Pressure Reflood Analysis, NUREG/CR-2114, ORNL/NUREG/TM-446.
24. Oak Ridge National Laboratory (ORNL), 1982a. Dispersed Flow Film Boiling in Rod Bundle Geometry: Steady State Heat Transfer Data and Correlation Comparisons, NUREG/CR-2435, ORNL-5822.
25. Oak Ridge National Laboratory (ORNL), 1982b. An Analysis of Transient Film Boiling of High-Pressure Water in a Rod Bundle, NUREG/CR-2469, ORNL/NUREG-85.
26. Oak Ridge National Laboratory (ORNL), 1982. Experimental Investigations of Bundle Boiloff and Reflood Under High-Pressure Low Heat Flux Conditions, NUREG/CR-2455, ORNL-5846.
27. Ohnuki, A., et al, 1991. Study on ECC Injection Modes in Reflood Tests with SCTF Core II ΠComparison between Gravity and Forced Feeds. JAERI-M-91-01.
28. Okubo, et al., 1985. Evaluation Report on CCTF Core-II Reflood Test C2-4 (Run 62) - Investigation of Reproducibility, JAERI-M-85-026.
29. Pressinos, P.G., et al., 1979. Experimental Data Report for LOFT Power Ascension Experiment L2-3, NUGEG/CR-0792.
30. Ransom, V.H., et al., 1985. RELAP5/MOD2 Code Manual, NUREG/CR-4312, EGG-2396.
31. Reeder, D.L., 1978. LOFT System and Test Description (5.5 ft Nuclear Core 1 LOCEs), NUREG/CR-TREE-1208.
32. Siemens AG UB KWU, 1986. Upper Plenum Test Facility, Test No. 12 Tie Plate Countercurrent Flow Testle, R515/86/13.
33. Siemens AG UB KWU, 1988. Upper Plenum Test Facility, Test No. 8 Cold/Hot Leg Flow Pattern Test Experimental Data Report, U9 316/88/12.
34. Siemens AG UB KWU, 1988. Experimental Data Report - UPTF - Test No. 10, Tie Plate Countercurrent Flow Test, U9 316/88/1.
35. Siemens AG UB KWU, 1989. Experimental Data Report - UPTF - Test No. 6, Downcomer Countercurrent Flow Test, U9 316/88/18.
36. Siemens AG UB KWU, 1989. Experimental Data Report - UPTF - Test No. 7, Downcomer Countercurrent Flow Test, U9 316/89/14.
37. Siemens AG UB KWU, 1990. Experimental Data Report - UPTF - Test No. 29, Entrainment/Deentrainment Test, U9 314/90/05.
38. Siemens Power Corporation, 1998. HTP Reflood Test Characterization Report, EMF-P60,149 (Proprietary).
39. P.N. Somerville Tables for obtaining non-parametric tolerance limits Ann. Math. Stat. 29 2 1958 599 601
40. Studsvik Eco & Safety AB, 1982. The Marviken Full-Scale Critical Flow Tests, Summary Report, NUREG/CR-2671, MXC-301.
41. Technical Program Group (TPG), 1989. Quantifying Reactor Safety Margins, NUREG/CR-5249, EGG-2552.
42. T.G. Theofanous Topical issue: quantifying reactor safety margins Nucl. Eng. Des. ix 1990 119
43. T.G. Theofanus Topical issue: discussion on quantifying reactor safety margins Nucl. Eng. Des. 1992 132
44. Tien, C.L., Chung, K.S., Liu, C.P., 1979. Flooding in Two-Phase Countercurrent Flows, EPRI NP-1283.
45. USNRC, 1978a. Experiment Data Report for Semiscale Mod-1 Test S-06-3 (LOFT Counterpart Test), NUREG/CR-0251, TREE-1123.
46. USNRC, 1978b. Semiscale MOD-3 Test Program and System Description. NUREG/CR-0239, TREE-1212.
47. USNRC, 1978. Experiment Data Report for Semiscale MOD-3 Blowdown Heat Transfer Test S-07-1 (Baseline Test Series), NUREG/CR-0281, TREE-1221.
48. USNRC, 1988. Compendium of ECCS Research for Realistic LOCA Analysis, NUREG/CR-1230.
49. USNRC, 1989. Best-Estimate Calculations of Emergency Core Cooling System Performance, USNRC Regulatory Guide 1.157.
50. Wheat, L.L., 1975. CONTEMPT-LT A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-Of-Coolant-Accident. Aerojet Nuclear Company, TID-4500, ANCR-1219.
51. S.S. Wilks Determination of sample sizes for setting tolerance limits Ann. Math. Stat. 12 1941 91 96
52. M.Y. Young, S.M. Bajorek, M.E. Nissley, and L.E. Hochreiter Application of code scaling, applicability, and uncertainty methodology to the large break loss of coolant Nucl. Eng. Des. 186 1998 39 52
53. E. Weiss, R.A. Markley, and A. Battacharyya Open duct cooling-concept for the radial blanket region of a fast breeder reactor Nucl. Eng. Des. 16 1971 175 386