Analysis of a rod withdrawal accident in a BWR with the neutronic-thermalhydraulic coupled code TRAC-BF1/VALKIN and TRACE/PARCS

Conference · Sat Jul 01 04:00:00 EDT 2006

OSTI ID:22039640

Miro, R; Verdu, G; Sanchez, A M; Barrachina, T ^[1]; Gomez, A ^[2]

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1. Chemical and Nuclear Engineering Dept., Polytechnic Univ. of Valencia, Cami de Vera s/n, 46022 Valencia (Spain)
2. Iberinco, Avenida de Burgos, Madrid (Spain)

The control rod withdrawal accident at hot zero power (HZP) is characterized by a single rod withdrawal from a core position with high reactivity worth, starting at criticality with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main factor limiting the consequences of the accident is a mixed void-Doppler feedback in BWR. The peak power occurs while important power distribution changes take place in the core and also the rod extraction continues. To check the performance of the coupled codes TRAC-BF1/VALKIN and TRACE/PARCS against complex 3D neutronic transients, a rod withdrawal accident in COFRENTES NPP is simulated. This transient is a dynamically complex event, where neutron kinetics is coupled with thermal hydraulics in the reactor primary system, and reactor variables change very rapidly. TRAC-BF1/VALKIN code uses the best estimate TRAC-BF1 code to give account of the heat transfer and thermalhydraulic processes, and a 3D neutronic module. This module has two options, MODKIN that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. TRACE is a code to study also transients in LWR reactors. This code used as a neutronic module the PARCS code. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22039640
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICS AND COMPUTING
COFRENTES REACTOR
CONTROL ELEMENTS
CRITICALITY
DOPPLER COEFFICIENT
HEAT TRANSFER
NEUTRON DIFFUSION EQUATION
POWER DISTRIBUTION
REACTIVITY INSERTIONS
REACTIVITY WORTHS
REACTOR CORES
REACTOR KINETICS
ROD EJECTION ACCIDENTS
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS
VOID COEFFICIENT