A general evaluation methodology development and application process (EMDAP) paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management guidelines. The basic framework described in this paper extends the top-down, bottom-up strategy described in the U.S Nuclear Regulatory Commission Regulatory Guide 1.203 to severe accident evaluations addressing U.S. NRC expectation for plant design certification applications. 1. Introduction Associated with rare, hazardous events, such as nuclear power plant (NPP) severe accidents, is a degree of uncertainty that provides a significant challenge to the evaluation and resolution of related design and analysis methods issues. For events occurring at some sufficiently observable frequency, design improvements can evolve through the understanding gained from such events and applicable test programs, leading to long-term acceptance. The broad uncertainties associated with severe accident initiators and event progression impose inherent limits on the benefits of this approach for severe accident design. As such, there is greater reliance on analysis and emphasis on the better understood severe accident phenomena. The engineering design process for an NPP’s severe accident response strategy has evolved to a process that relies on the(i)establishment of safety goals,(ii)identification of processes and phenomena,(iii)iterative design processes focused on risk reduction,(iv)test programs,(v)expert elicitation on important severe accident safety issues,(vi)analysis methods development,(vii)process studies. Consistent with current US regulatory requirements and guidance, final acceptance and resolution of relevant beyond-design-basis events is demonstrated through detailed deterministic studies and probabilistic risk assessment (PRA). The unique characteristic of this process for severe accidents is the consideration of risk in the resolution of severe accident safety issues. Generation III and IV advanced reactors designs incorporate features that significantly reduced risk relative to current-generation light water reactors (LWRs). Practical consideration of this reduced risk requires that this information be incorporated into measures not only of acceptable performance,
References
[1]
U.S. Nuclear Regulatory Commission and Regulatory Guide 1.203, “Transient and Accident Analysis Methods,” Revision 0, December 2005.
[2]
R. P. Martin, M. W. Bingham, C. A. Bonilla et al., “AREVA NP's severe accident safety issue resolution methodology for the U.S. EPR,” in Proceedings of the International Conference on Advances in Nuclear Power Plants (ICAPP'08), pp. 1122–1131, Anaheim, Calif, USA, June 2008.
[3]
USNRC and NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” most recent revision.
[4]
USNRC and Regulatory Guide 1.206, “Combined License Applications for Nuclear Power Plants,” most recent revision.
[5]
US NRC and SECY-93-087, “Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water (ALWR) Designs,” issued April, 1993 and the corresponding SRM, issued July, 1993.
[6]
US NRC, “Reactor Safety Study-An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,” WASH-1400 (NUREG-75/014), 1975.
[7]
M. Firnhaber, K. Trambauer, S. Hagen, et al., “CORA-13 Experiment on Severe Fuel Damage,” NEA/CSNI-(93) 17, GRS-106, KfK 5287, July, 1993.
[8]
Z. R. Martinson, Idaho National Engineering Laboratory, et al., PBF Severe Accident Fuel Damage Test 1–3 Test Results Report, 1989, NUREG/CR-5354.
[9]
M. Firnhaber, “ISP 28: Phebus-SFD B9+ Experiment on the Degradation of a PWR Type Core,” NEA/CSNI/R(92)17, Volumes 1 and 2, 1992.
[10]
B. Clément, N. Hanniet-Girault, et al., “The first international PHEBUS fission product program for investigation of phenomena of severe water reactor accidents,” Nuclear Engineering and Design, vol. 226, no. 1, pp. 5–28, 2003.
[11]
W. Hering, Ch. Homann, and J. S. Lamy, “Comparison report on the blind phase of the OECD International Standard Problem no.45, exercise (QUENCH-06),” OECD Document ISP45-OECD, 2002, Report FZKA-6677.
[12]
W. Hering, Ch. Homann, J. S. Lamy, et al., “Comparison and interpretation report of OECD International Standard Problem no.45 exercise (QUENCH-06),” ISP45-OECD, 2002, Report FZKA-6722.
[13]
E. W. Coryell, Summary of Important Results and SCDAP/RELAP5 Analysis for OECD LOFT Experiment LP-FP-2, 1994, NUREG/CR-6160, NEA/CSNI-(94) 3.
[14]
J. L. Rempe, S. A. Chavez, G. L. Thinnes, et al., Light Water Reactor Lower Head Failure Analysis, 1993, NUREG/CR-5642, EGG-2618.
[15]
J. L. Rempe, “MASCA and RASPLAV test program—description, status, and implications,” INEEL/EXT- 03-00224, 2003.
[16]
J. L. Rempe, K. G. Condie, and D. L. Knudson, “Thermal properties for candidate SCWR materials,” INL/EXT- 05-01030, 2005.
[17]
L. A. Stickler, J. L. Rempe, et al., “Calculations to estimate the margin-to-failure in the TMI-2 vessel,” NUREG/CR-6196, EGG-2733, 1994.
[18]
T. Y. Chu, U.S. Nuclear Regulatory Commission, et al., Lower Head Failure Experiments and Analyses, 1999, NUREG/CR-5582 (SAND98-2047).
[19]
J. R. Wolf, J. L. Rempe, et al., “TMI-2 vessel investigation project integration report,” NUREG/CR -6197, 1994, EGG-2734.
[20]
D. Magallon and I. Huhtiniemi, “Corium melt quenching tests at low pressure and subcooled water in FARO,” Nuclear Engineering and Design, vol. 204, no. 1–3, pp. 369–376, 2001.
[21]
I. Huhtiniemi and D. Magallon, “Insight into steam explosions with corium melts in KROTOS,” Nuclear Engineering and Design, vol. 204, no. 1–3, pp. 391–400, 2001.
[22]
J. H. Kim, I. K. Park, Y. S. Shin, et al., “A study on intermediate scale steam explosion experiments with zirconia and corium melts,” in Proceedings of the Transactions of the International Congress on Advanced Nuclear Power Plants (ICAPP'02), Hollywood, Fla, USA, June 2002.
[23]
W. W. Tarbell, Sandia National Laboratories, et al., Results from the DCH-1 Experiment, Sandia National Laboratory, 1987, SAND86-2483, NUREG/CR-4871.
[24]
M. D. Allen, R. T. Nichols, J. E. Brockmann, et al., DCH-2: Results from the Second Experiment Performed in the Surtsey Direct Heating Test Facility, Sandia National Laboratory, 1988, SAND87-0976, NUREG/CR-4917.
[25]
M. D. Allen, M. M. Pilch, R. O. Griffith, and R. T. Nichols, Experiments to Investigate the Effects of Water in the Cavity on Direct Containment Heating in Surtsey Test Facility- the WC-1 and WC-2 Tests, Sandia National Laboratories, 1992, SAND91-1173.
[26]
M. D. Allen, M. M. Pilch, T. K. Blanchat, et al., Experiments to Investigate to Direct Containment Heating Phenomena With Scaled Models of the Zion Nuclear Power Plant in the Surtsey Test Facility, 1994, NUREG/CR-6044.
[27]
T. K. Blanchat, U.S. Nuclear Regulatory Commission, et al., Experiments to Investigate Direct Containment Heating Phenomena With Scaled Models of a Surry Nuclear Power Plant, 1994, NUREG/CR- 6152.
[28]
M. Pilch, M. D. Allen, E. W. Klamerus, et al., Resolution of the Direct Containment Heating Issue for All Westinghouse Plants With Large Dry Containments or Subatmospheric Containments, 1996, NUREG/CR-6338, SAND95-2381.
[29]
M. M. Pilch and U.S. Nuclear Regulatory Commission, The Probability of Containment Failure by Direct Containment Heating in Surry, 1995, NUREG/CR-6109, SAND93-2078.
[30]
J. L. Binder, L. M. McUmber, and B. W. Spencer, Direct Containment Heating Integral Effects Tests at 1/40 Scale in Zion Nuclear Power Plant Geometry, Argonne National Laboratory, 1994, NUREG/CR-6168, ANL-94/18.
[31]
B. W. Spencer, J. J. Sienicki, L. M. McUmber, et al., Hydrodynamics and Heat Transfer Aspects of Corium-water Interactions, Argonne National Laboratory, 1987, EPRI NP-5127.
[32]
R. E. Henry, R. J. Hammersley, and G. T. Klopp, “Direct containment heating experiments in a zion-like geometry,” AIChE Symposium Series, vol. 87, no. 283, pp. 86–98, 1991.
[33]
H. Karwat, “SOAR on Containment Thermalhydraulics and Hydrogen Distribution,” NEA/CSNI (99), 1999.
[34]
J. Royen, “Final Comparison Report on ISP-35: NUPEC Hydrogen Mixing and Distribution Test (Test M-7-1),” NEA/CSNI (94), 1994.
[35]
M. Firnhaber, T. F. Kanzleiter, S. Schwarz, et al., “ISP37: VANAM M3—a multi compartment aerosol depletion test with hygroscopic aerosol material,” NEA/CSNI (96), 1996.
[36]
G. W. Koroll, W. A. Dewit, J. L. Sitar, et al., “Hydrogen recombiner development at AECL,” in Proceedings of the OECD/NEA/CSNI Workshop on the Implementation of Hydrogen Mitigation Techniques, Winnipeg, 1996, AECL-11762, NEA/CSNI/R(96)8.
[37]
K. Fischer, “Qualification of a passive catalytic module for hydrogen mitigation,” Nuclear Technology, vol. 112, no. 1, pp. 58–62, 1995.
[38]
J. Hosler and G. Sliter, “PARs for combustible gas control in advanced light water reactors,” in Proceedings of the OECD/NEA/CSNI Workshop on the Implementation of Hydrogen Mitigation Techniques, Winnipeg, 1996, AECL-11762, NEA/CSNI/R(96)8.
[39]
P. Rongier and E. Studer, “H2PAR facility,” in Proceedings of the OECD/NEA/CSNI Workshop on the Implementation of Hydrogen Mitigation Techniques, Winnipeg, 1996, AECL-11762, NEA/CSNI/R(96)8.
[40]
W. Breitung, C. Chan, S. B. Dorofeev, et al., “Flame acceleration and deflagration-to-detonation transition in nuclear safety,” NEA/CSNI (2000), 2000.
[41]
M. Firnhaber, “ISP 30: BETA V5.1 experiment on melt-concrete interaction,” NEA/CSNI/R (92)9, 1992.
[42]
D. G. Thompson, et al., “Compilation, analysis and interpretation of ACE phase C and MACE experimental data—MCCI thermohydraulic results,” ACEX TR-C 14, Argonne National Laboratory, 1997.
[43]
M. T. Farmer, S. Lomperski, and S. Basu, “Mace test M3b data report volume 1,” EPRI TR-108806, 1997, MACE-TR-D13.
[44]
Farmer M. T., et al., “Mace test M4 data report,” MACE-TR-D 16, 1999.
[45]
M. T. Farmer, S. Lomperski, D. J. Kilsdonk, and R. W. Aeschlimann, “OECD MCCI project 2-D core concrete interaction (CCI) tests: CCI-1 test data report-thermalhydraulic results,” OECD/MCCI 2004-TR01, 2004.
[46]
M. T. Farmer, S. Lomperski, D. J. Kilsdonk, and R. W. Aeschlimann, “OECD MCCI project 2-D core concrete interaction (CCI) tests: CCI-2 test data report-thermalhydraulic results,” OECD/MCCI 2004-TR05, 2004.
[47]
M. T. Farmer, S. Lomperski, D. J. Kilsdonk, and R. W. Aeschlimann, “OECD MCCI project 2-D core concrete interaction (CCI) tests: CCI-3 test data report-thermalhydraulic results,” OECD/MCCI 2005-TR04, 2005.
[48]
B. W. Spencer, M. Fischer, M. T. Farmer, et al., “MACE scoping test,” MACE-TR-D 03, Argonne National Laboratory, 1999.
[49]
J.-M. Bonnet, “Thermal hydraulic phenomena in corium pools for ex-vessel situations: the BALI experiment,” in Proceedings of the 8th International Conference on Nuclear Engineering (ICONE 8), Baltimore, Md, USA, April 2000.
[50]
B. Tourniaire and J.-M. Bonnet, “Study of the mixing of immiscible liquids by sparging gas results of the BALISE experiments,” in Proceedings of the 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), Seoul, Korea, October 2003.
[51]
M. Nie, M. Fischer, and W. Koller, “Status of interpretation of selected transient MCCI experiments conducted in the frame of the CORESA R&D project,” in Proceedings of the Jahrestagung Kerntechnik, Stuttgart, Germany, May 2002.
[52]
S. Hellmann, V. Lansmann, B. Friedrich, et al., “Physico-chemical and material aspects of the core melt retention concept of the EPR,” in Proceedings of the OECD Workshop on Ex-Vessel Debris Coolability, Karlsruhe, Germany, November 1999.
[53]
B. Adroguer, P. Chatelard, J. P. van Dorsselaere, et al., “Corium interactions and thermochemistry (CIT project),” INV-CIT (99)-P040, European Commission, 1999.
[54]
B. Eppinger, T. Cron, G. Stern, et al., “KAPOOL experiments on melt-through of a metal plate by an overlying melt pool,” FZKA 7024, Forschungszentrum Karlsruhe, August 2004.
[55]
M. T. Farmer, “Melt eruption test (MET-1) results,” in Proceedings of the 6th Program Review Meeting of the MCCI-OECD Project, Madrid, Spain, April 2005.
[56]
J. M. Veteau and R. Wittmaack, “CORINE experiments and theoretical modelling,” FISA-95-EU Research on Severe Accidents, EUR 16896 EN, 1996.
[57]
B. Eppinger, F. Fellmoser, G. Fieg, et al., “Simulationsexperimente zum Ausbreitungsverhalten von Kernschmelze: KATS 8-17,” Forschungszentrum, Karlsruhe, FZKA 6589, March 2001.
[58]
W. Steinwarz, et al., “COMAS spreading experiments with prototypic oxidic corium melts for optimization of spreading compartment designs,” ICONE7, Tokyo, Japan, April 1999.
[59]
R. Silverii and D. Magallon, “FARO LWR programme, test L26-S data report, technical note,” EXV-CSC (98)-D007, 1998, EU 4th FWP.
[60]
G. Cognet, J.-M. Seiler, I. Szabo, et al., “The VULCANO spreading programme article,” EU 4th FWP, EXV-CSC(98)-D018, SARJ-98, 1999.
[61]
M. J. Konovalikhin, T. N. Dinh, and B. R. Sehgal, “Experimental simulation of core melt spreading in two dimensions,” in Proceedings of the The 9th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9), San Francisco, Calif, USA, October 1999.
[62]
B. R. Sehgal and B. W. Spencer, “Ace program phase D: melt attack and coolability experiments (MACE) program,” in Proceedings of the Second OECD (NEA) CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions, Karlsruhe, Germany, April 1992, NEA/CSNI-(92) 10, KfK 5108.
[63]
H. Alsmeyer, et al., “Experiment ECOKATS-2 on melt spreading and subsequent top flooding test and data report,” Forschungszentrum, Karlsruhe, FZKA 7084, January 2005.
[64]
M. Fischer, O. Herbst, and H. Schmidt, “Demonstration of the heat removing capabilities of the EPR spreading compartment,” in Proceedings of the 3rd International Symposium on Two-Phase flow Modelling and Experimentation, Pisa, Italy, September 2004.
[65]
G. E. Wilson and B. E. Boyack, “The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis,” Nuclear Engineering and Design, vol. 186, no. 1-2, pp. 23–37, 1998.
[66]
D. Magallon, A. Mailliat, J. M. Seiler et al., “European expert network for the reduction of uncertainties in severe accident safety issues (EURSAFE),” Nuclear Engineering and Design, vol. 235, no. 2–4, pp. 309–346, 2005.
[67]
K. Nagashima, M. Alammar, and H. C. da Silva, “Application of uncertainty analyses with the MAAP4 code,” in Proceedings of the 5th International Conference on Simulation Methods in Nuclear Engineering, Canadian Nuclear Society, Montreal, Quebec, month year.
[68]
R. Ashley, M. El-Shanawany, F. Eltawila, et al., “Good practices for user effect reduction,” Status Report NEA/CSNI/R(98)22, 1998, Paris, France.
[69]
R. O. Gauntt, R. K. Cole, C. M. Erickson, et al., “MELCOR computer code manuals,” Version 1.8.6, NUREG/CR-6119, Rev. 3, SAND-2417/2, Sandia National Laboratory, September 2005.
[70]
SCDAP/RELAP5 Development Team, “SCDAP/RELAP5/MOD3.2 code manual, Vol. 1–5,” NUREG/CR-6150, INEL-96/0422, July 1998.
[71]
Fauske and Associates, Inc, “MAAP4—modular accident analysis program for LWR power plants, vol. 2, part 1: code structure and theory,” prepared for Electric Power Research Institute, May 1994.
[72]
J. S. Butler, D. Kapitz, R. K. Sundaram, and R. P. Martin, “MAAP4.0.7 analysis and justification for PRA level 1 mission success criteria,” in Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP '08), Anaheim, Calif, USA, June 2008, paper 8200.
[73]
S. S. Wilks, “Determination of sample sizes for setting tolerance limits,” The Annals of Mathematical Statistics, vol. 12, pp. 91–96, 1941.
[74]
R. P. Martin and W. T. Nutt, “Perspectives on the application of order-statistics in best-estimate plus uncertainty nuclear safety analysis,” Nuclear Engineering and Design, vol. 241, no. 1, pp. 274–284, 2011.
[75]
M. M. Pilch, M. D. Allen, E. W. Klamerus, et al., Resolution of the Direct Containment Heating Issue for All Westinghouse Plants With Large Dry Containments or Subatmospheric Containments, 1996, NUREG/CR-6338, SAND95-2381.
[76]
N. Zuber, G. E. Wilson, M. Ishii et al., “An integrated structure and scaling methodology for severe accident technical issue resolution: development of methodology,” Nuclear Engineering and Design, vol. 186, no. 1-2, pp. 1–21, 1998.
[77]
J. C. Helton, J. D. Johnson, C. J. Sallaberry, and C. B. Storlie, “Survey of sampling-based methods for uncertainty and sensitivity analysis,” Reliability Engineering and System Safety, vol. 91, no. 10-11, pp. 1175–1209, 2006.
[78]
K. Chan, A. Saltelli, and S. Tarantola, “Sensitivity analysis of model output: variance-based methods make the difference,” in Proceedings of the Winter Simulation Conference, pp. 261–268, Atlanta, Ga, USA, December 1997.
[79]
H. Glaeser, “Uncertainty evaluation of thermal-hydraulic code results,” in Proceedings of the International Mtg. Best-Estimate Methods in Nuclear Installation Safety Analysis (BE-2000), Washington, DC, USA, November 2000.
[80]
R. P. Martin, “Quantifying phenomenological importance in best-estimate plus uncertainty analyses,” Nuclear Technology, vol. 175, no. 3, pp. 652–662, 2011.
[81]
“U.S. EPR final safety analysis report,” Rev. 0, AREVA NP Inc., December 2007.
