Coupled neutronic thermal–hydraulic transient analysis of accidents in PWRs

With the sustained development in computer technology, code capabilities have been expanding substantially. Of major interest to nuclear engineers is the use of best-estimate methods that can predict important safety margins and their associated uncertainties. This paper presents a package called IRTRAN which couples well known codes to analyze reactivity transients of PWRs along considering thermal–hydraulic (T–H) feedbacks. T–H feedbacks are included as an automatic built-in feature in IRTRAN. Availability of automated generation of T–H feedbacks along capabilities to analyze neutronics of the core and cross sections generation, make IRTRAN capable of accurate modeling of conventional as well as advanced future reactor systems. In this package three codes are linked together; WIMSD4, CITATION neutronic modules along with the RELAP5/3.2 thermal–hydraulic module constitute the main core of IRTRAN. WIMS is used to perform cell calculations in every time step through transients to obtain neutronic parameters. The CITATION code is implemented for core calculations to obtain effective multiplication factors as well as flux and power distributions. Furthermore, RELAP5 is employed to determine the temporal and spatial distributions of the flow field thermal–hydraulic conditions. The point kinetic solver routine is considered for including fuel Doppler and coolant density feedbacks. In order to link WIMS, CITATION and RELAP5 codes a FORTRAN 90 program is developed. The program may be used to perform detailed steady state and transient analyses for a PWR. The main points of strength of this package in transient simulations are reactivity calculation routine, possibility of core and loop transient analysis, modeling of various operating conditions of PWR reactors and updating neutronic cross sections during core operating condition (without interpolating traditional algorithms). To demonstrate the usefulness of the coupled codes, the drop of a control rod transient case is simulated. In addition, the reactivity coefficients of Boushehr NPP which is a VVER-1000 are calculated. The obtained transient results are in good agreement with published data.

► The three well known codes WIMSD4, CITATION, RELAP5 are coupled for accident analysis.
► T–H feedback as an automatic build-in feature is presented.
► Possibility of core and loop transient analysis with consideration of full scope PWR NPP is developed.
► Updating neutronic cross sections through transient interval is presented.