REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables

[EN] The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA at Hot Zero Power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS v2.7 a REA in Almaraz NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions.This work has been partially supported by the Spanish Ministerio de Educación y Ciencia under project PHB2007-0067-PC and by the Brazilian Coordenação de Aperfeiçoamento de Pessoal de Nível Superior (CAPES) under project CAPES/DGU 159/08. This work has also been partially supported by the Spanish Ministerio de Educación y Ciencia under project ENE2008-02669, the Generalitat Valenciana under project ACOMP/2009/058, and the Universitat Politècnica de València under project PAID-05-09-4285. The authors wish to acknowledge the technical and financial support of the Almaraz-Trillo AIE for this work.Barrachina Celda, TM.; Garcia-Fenoll, M.; Ánchel Añó, FC.; Miró Herrero, R.; Verdú Martín, GJ.; Pereira, C.; Da Silva, C.... (2011). REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables. Progress in Nuclear Energy. 53(8):1167-1180. https://doi.org/10.1016/j.pnucene.2011.07.012S1167118053

Uncertainty and sensitivity analysis in the results of coupled RELAP5/PARCS simulation for the Almaraz nuclear power plant resulting from uncertainties in the neutronic parameters

Proyecto ConfidencialÁnchel Añó, FC. (2010). Uncertainty and sensitivity analysis in the results of coupled RELAP5/PARCS simulation for the Almaraz nuclear power plant resulting from uncertainties in the neutronic parameters. http://hdl.handle.net/10251/31782.Archivo delegad

Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic-neutronic simulations

The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether¿Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.Ánchel Añó, FC.; Barrachina Celda, TM.; Miró Herrero, R.; Verdú Martín, GJ.; Juanas ., J.; Macián-Juan, R. (2012). Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic-neutronic simulations. Nuclear Engineering and Design. 246:98-106. doi:10.1016/j.nucengdes.2011.11.016S9810624