Uncertainty analysis of large break LOCA for pressurized heavy water reactor

• Published in: Nuclear engineering and design Vol. 245; pp. 180 - 188
• Main Authors: Trivedi, A.K., Srivastava, A., Lele, H.G., Kalra, M.S., Munshi, P.
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 01.04.2012; Elsevier
• Subjects: Applied sciences; Controled nuclear fusion plants; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Installations for energy generation and conversion: thermal and electrical energy; Nuclear fuels; Monte Carlo method; LOCA; Statistical analysis; Heavy water reactor; Safety; Sampling; Clad; Pressurized water reactor; Nuclear reactor
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2012.01.009

► We have predicted uncertainty in RELAP5 simulations. ► Simulations of large-break LOCA are the starting point. ► Indian PHWR is analysed for the first time. ► Three statistical methods have been used. ► Recommendations have been made on the choice of method. Present study involves uncertainty analysis for a large break LOCA of pressurized heavy water reactor (PHWR). It is carried out for the primary safety criteria, peak clad temperature (PCT). Aim of this analysis is to determine the 95th percentile value for PCT during LOCA. Uncertainty quantification for PCT is done based on three different methods: Wilk's method, response surface method (RSM) followed by Monte Carlo simulation and direct Monte Carlo simulation using Latin Hypercube Sampling (LHS). Results obtained by these three methods are compared with acceptance criteria. A best estimate code, RELAP5 MOD3.2, is selected for simulating large break LOCA in 220MWe PHWR. Statistical analysis is carried out in MATLAB environment.