{"title":"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","volume":152,"journal":"International Journal of Nuclear and Quantum Engineering","pagesStart":415,"pagesEnd":427,"ISSN":"1307-6892","URL":"https:\/\/publications.waset.org\/pdf\/10010668","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.<\/p>\r\n","references":"[1]\tF. Menzel, G. Sabundjian, F. D\u2019Auria, and A. Madeira, \u201cProposal for systematic application of BEPU in the licensing process of nuclear power plants,\u201d Int. J. Nucl. Energy Sci. Technol., vol. 10, 2016, pp. 323\u2013337.\r\n[2]\tA. Annunziato, H. Glaeser, J. Lillington, P. Marsili, C. Renault, and A. Sjoberg, \u201cCSNI Integral Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic Codes for LWR LOCA and Transients,\u201d NEA\/CSNI\/R(96)17, 1996.\r\n[3]\tN. Aksan, F. D\u2019Auria, H. Glaeser, J. Lillington, R. Pochard, and A. Sjoberg, \u201cEvaluation of the CSNI Separate Effects Tests (SET) Validation Matrix,\u201d CSNI report OECD\/GD(97)9, 1996.\r\n[4]\tT. Skorek, \u201cInput uncertainties in uncertainty analyses of system codes: Quantification of physical model uncertainties on the basis of CET (combined effect test),\u201d Nucl. Eng. Des., vol. 321, 2017, pp. 301\u2013317.\r\n[5]\tD. Bestion, F. D\u2019Auria, P. Lien, and H. Nakamura, \u201cA state-of-the-art report on scaling in system thermal-hydraulics applications to nuclear reactor safety,\u201d NEA\/CSNI\/R(2016)14, 2017.\r\n[6]\tThe ROSA-V Group, \u201cROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies,\u201d JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.\r\n[7]\tK. Umminger, L. Dennhardt, S. Schollenberger, and B. Schoen, \u201cIntegral Test Facility PKL: Experimental PWR Accident Investigation,\u201d Sci. Technol. Nucl. Installations, Article ID 891056, vol. 2012, 2012, pp. 1\u201316.\r\n[8]\tK.Y. Choi, Y.S. Kim, C.H. Song, and W.P. Baek, \u201cMajor Achievements and Prospect of the ATLAS Integral Effect Tests,\u201d Sci. Technol. Nucl. Installations, Article ID 375070, vol. 2012, 2012, pp. 1\u201318.\r\n[9]\tUSNRC, \u201cReactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity,\u201d Bulletin 2002-01, OMB Control No.: 3150-0012, U.S. Nuclear Regulatory Commission, Washington, DC, 2002.\r\n[10]\tUSNRC, \u201cDavis-Besse Reactor Pressure Vessel Head Degradation: Overview, Lessons Learned, and NRC Actions Based on Lessons Learned,\u201d NUREG\/BR-0353, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, DC, 2008.\r\n[11]\tH. Shiroyama, \u201cRegulatory failures of nuclear safety in Japan \u2013the case of Fukushima accident,\u201d in: Proc. of the Earth System Governance Tokyo Conference: Complex Architectures, Multiple Agents, Earth System Governance, Tokyo, Japan, January 2013.\r\n[12]\tY. Kukita, K. Tasaka, H. Asaka, T. Yonomoto, and H. Nakamura, \u201cThe effects of break location on PWR small break LOCA: experimental study at the ROSA-IV LSTF,\u201d Nucl. Eng. Des., vol. 122, 1990, pp. 255\u2013262.\r\n[13]\tT. Takeda, \u201cROSA\/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature,\u201d Nucl. Eng. Technol., vol. 50, 2018, pp. 1412\u20131420.\r\n[14]\tUSNRC Nuclear Safety Analysis Division, \u201cRELAP5\/MOD3.3 Code Manual,\u201d NUREG\/CR-5535\/Rev 1, Information Systems Laboratories, Inc., 2001.\r\n[15]\tT. Takeda, \u201cROSA\/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,\u201d Ann. Nucl. Energy, vol. 121, 2018, pp. 594\u2013606.\r\n[16]\tT. Takeda, \u201cUncertainty analysis of ROSA\/LSTF test on pressurized water reactor cold leg small-break loss-of-coolant accident without scram,\u201d Int. J. Nucl. Quantum Eng., vol. 13, 2019, pp. 82\u201390.\r\n[17]\tN. Zuber, \u201cProblems in Modeling Small Break LOCA,\u201d NUREG-0724, USNRC, Washington, DC, 1980.\r\n[18]\tNEA, \u201cFinal Integration Report of OECD\/NEA ROSA Project 2005\u20132009,\u201d NEA\/CSNI\/R(2013)1, 2013.\r\n[19]\tH. Austregesilo, B. Krzykacz-hausmann, and T. Skorek, \u201cUncertainty and sensitivity analysis of results of the posttest calculation of a LSTF experiment with the code ATHLET,\u201d GRS Annual Report 2008, 2008, pp. 35\u201341.\r\n[20]\tJ. Freixa and A. Manera, \u201cAnalysis of an RPV upper head SBLOCA at the ROSA facility using TRACE,\u201d Nucl. Eng. Des., vol. 240, 2010, pp. 1779\u20131788.\r\n[21]\tC. Queral, J. Gonzalez-cadelo, G. Jimenez, and E. Villalba, \u201cAccident management actions in an upper-head small-break loss-of-coolant accident with high-pressure safety injection failed,\u201d Nucl. Technol., vol. 175, 2011, pp. 572\u2013593.\r\n[22]\tH. Kumamaru and K. Tasaka, \u201cRecalculation of Simulated Post-scram Core Power Decay Curve for Use in ROSA-IV\/LSTF Experiments on PWR Small-break LOCAs and Transients,\u201d JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990.\r\n[23]\tH.K. Fauske, \u201cThe discharge of saturated water through tubes,\u201d AlChE Symp. Ser., vol. 61, 1965, pp. 210\u2013216.\r\n[24]\tK.H. Ardron and R.A. Furness, \u201cA study of the critical flow models used in reactor blowdown analysis,\u201d Nucl. Eng. Des., vol. 39, 1976, pp. 257\u2013266.\r\n[25]\tD.W. Sallet, \u201cThermal hydraulics of valves for nuclear applications,\u201d Nucl. Sci. Eng., vol. 88, 1984, pp. 220\u2013244.\r\n[26]\tT. Takeda, A. Ohnuki, and H. Nishi, \u201cRELAP5 code study of ROSA\/LSTF experiments on PWR safety system using steam generator secondary-side depressurization,\u201d J. Energy Power Eng., vol. 9, 2015, pp. 426\u2013442.\r\n[27]\tA. Guba, M. Makai, and L. P\u00e1l, \u201cStatistical aspects of best estimate method\u2013I,\u201d Reliability Eng. Syst. Safety, vol. 80, 2003, pp. 217\u2013232.\r\n[28]\tIAEA, \u201cBest Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation,\u201d IAEA Safety Reports Series No. 52, Vienna, Austria, 2008.\r\n[29]\tA. de Cr\u00e9cy, P. Bazin, H. Glaeser, H., et al., \u201cUncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme,\u201d Nucl. Eng. Des., vol. 238, 2008, pp. 3561\u20133578.\r\n[30]\tW.W. Daniel, \u201cSpearman rank correlation coefficient,\u201d Applied Nonparametric Statistics (second ed.), PWS-Kent Publishing, Boston, MA, 1990.","publisher":"World Academy of Science, Engineering and Technology","index":"Open Science Index 152, 2019"}