Assessment and Uncertainty Analysis of ROSA/LSTF Test on Pressurized Water Reactor 1.9% Vessel Upper Head Small-Break Loss-of-Coolant Accident

Authors: Takeshi Takeda

Abstract:

An experiment utilizing the ROSA/LSTF (rig of safety assessment/large-scale test facility) simulated a 1.9% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under the total failure of high-pressure injection system of emergency core cooling system in a pressurized water reactor. Steam generator (SG) secondary-side depressurization on the AM measure was started by fully opening relief valves in both SGs when the maximum core exit temperature rose to 623 K. A large increase took place in the cladding surface temperature of simulated fuel rods on account of a late and slow response of core exit thermocouples during core boil-off. The author analyzed the LSTF test by reference to the matrix of an integral effect test for the validation of a thermal-hydraulic system code. Problems remained in predicting the primary coolant distribution and the core exit temperature with the RELAP5/MOD3.3 code. The uncertainty analysis results of the RELAP5 code confirmed that the sample size with respect to the order statistics influences the value of peak cladding temperature with a 95% probability at a 95% confidence level, and the Spearman’s rank correlation coefficient.

Keywords: LSTF, LOCA, uncertainty analysis, RELAP5.

Digital Object Identifier (DOI): doi.org/10.5281/zenodo.3455635

References:

[1] F. Menzel, G. Sabundjian, F. D’Auria, and A. Madeira, “Proposal for systematic application of BEPU in the licensing process of nuclear power plants,” Int. J. Nucl. Energy Sci. Technol., vol. 10, 2016, pp. 323–337.
[2] A. Annunziato, H. Glaeser, J. Lillington, P. Marsili, C. Renault, and A. Sjoberg, “CSNI Integral Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic Codes for LWR LOCA and Transients,” NEA/CSNI/R(96)17, 1996.
[3] N. Aksan, F. D’Auria, H. Glaeser, J. Lillington, R. Pochard, and A. Sjoberg, “Evaluation of the CSNI Separate Effects Tests (SET) Validation Matrix,” CSNI report OECD/GD(97)9, 1996.
[4] T. Skorek, “Input uncertainties in uncertainty analyses of system codes: Quantification of physical model uncertainties on the basis of CET (combined effect test),” Nucl. Eng. Des., vol. 321, 2017, pp. 301–317.
[5] D. Bestion, F. D’Auria, P. Lien, and H. Nakamura, “A state-of-the-art report on scaling in system thermal-hydraulics applications to nuclear reactor safety,” NEA/CSNI/R(2016)14, 2017.
[6] The ROSA-V Group, “ROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies,” JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.
[7] K. Umminger, L. Dennhardt, S. Schollenberger, and B. Schoen, “Integral Test Facility PKL: Experimental PWR Accident Investigation,” Sci. Technol. Nucl. Installations, Article ID 891056, vol. 2012, 2012, pp. 1–16.
[8] K.Y. Choi, Y.S. Kim, C.H. Song, and W.P. Baek, “Major Achievements and Prospect of the ATLAS Integral Effect Tests,” Sci. Technol. Nucl. Installations, Article ID 375070, vol. 2012, 2012, pp. 1–18.
[9] USNRC, “Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity,” Bulletin 2002-01, OMB Control No.: 3150-0012, U.S. Nuclear Regulatory Commission, Washington, DC, 2002.
[10] USNRC, “Davis-Besse Reactor Pressure Vessel Head Degradation: Overview, Lessons Learned, and NRC Actions Based on Lessons Learned,” NUREG/BR-0353, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, DC, 2008.
[11] H. Shiroyama, “Regulatory failures of nuclear safety in Japan –the case of Fukushima accident,” in: Proc. of the Earth System Governance Tokyo Conference: Complex Architectures, Multiple Agents, Earth System Governance, Tokyo, Japan, January 2013.
[12] Y. Kukita, K. Tasaka, H. Asaka, T. Yonomoto, and H. Nakamura, “The effects of break location on PWR small break LOCA: experimental study at the ROSA-IV LSTF,” Nucl. Eng. Des., vol. 122, 1990, pp. 255–262.
[13] T. Takeda, “ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature,” Nucl. Eng. Technol., vol. 50, 2018, pp. 1412–1420.
[14] USNRC Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual,” NUREG/CR-5535/Rev 1, Information Systems Laboratories, Inc., 2001.
[15] T. Takeda, “ROSA/LSTF test and RELAP5 code analyses on PWR hot leg small-break LOCA with accident management measure based on core exit temperature and PKL counterpart test,” Ann. Nucl. Energy, vol. 121, 2018, pp. 594–606.
[16] T. Takeda, “Uncertainty analysis of ROSA/LSTF test on pressurized water reactor cold leg small-break loss-of-coolant accident without scram,” Int. J. Nucl. Quantum Eng., vol. 13, 2019, pp. 82–90.
[17] N. Zuber, “Problems in Modeling Small Break LOCA,” NUREG-0724, USNRC, Washington, DC, 1980.
[18] NEA, “Final Integration Report of OECD/NEA ROSA Project 2005–2009,” NEA/CSNI/R(2013)1, 2013.
[19] H. Austregesilo, B. Krzykacz-hausmann, and T. Skorek, “Uncertainty and sensitivity analysis of results of the posttest calculation of a LSTF experiment with the code ATHLET,” GRS Annual Report 2008, 2008, pp. 35–41.
[20] J. Freixa and A. Manera, “Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE,” Nucl. Eng. Des., vol. 240, 2010, pp. 1779–1788.
[21] C. Queral, J. Gonzalez-cadelo, G. Jimenez, and E. Villalba, “Accident management actions in an upper-head small-break loss-of-coolant accident with high-pressure safety injection failed,” Nucl. Technol., vol. 175, 2011, pp. 572–593.
[22] H. Kumamaru and K. Tasaka, “Recalculation of Simulated Post-scram Core Power Decay Curve for Use in ROSA-IV/LSTF Experiments on PWR Small-break LOCAs and Transients,” JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990.
[23] H.K. Fauske, “The discharge of saturated water through tubes,” AlChE Symp. Ser., vol. 61, 1965, pp. 210–216.
[24] K.H. Ardron and R.A. Furness, “A study of the critical flow models used in reactor blowdown analysis,” Nucl. Eng. Des., vol. 39, 1976, pp. 257–266.
[25] D.W. Sallet, “Thermal hydraulics of valves for nuclear applications,” Nucl. Sci. Eng., vol. 88, 1984, pp. 220–244.
[26] T. Takeda, A. Ohnuki, and H. Nishi, “RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization,” J. Energy Power Eng., vol. 9, 2015, pp. 426–442.
[27] A. Guba, M. Makai, and L. Pál, “Statistical aspects of best estimate method–I,” Reliability Eng. Syst. Safety, vol. 80, 2003, pp. 217–232.
[28] IAEA, “Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation,” IAEA Safety Reports Series No. 52, Vienna, Austria, 2008.
[29] A. de Crécy, P. Bazin, H. Glaeser, H., et al., “Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme,” Nucl. Eng. Des., vol. 238, 2008, pp. 3561–3578.
[30] W.W. Daniel, “Spearman rank correlation coefficient,” Applied Nonparametric Statistics (second ed.), PWS-Kent Publishing, Boston, MA, 1990.