Assessment of TRAC-PF1/MOD1 Against an Inadvertent Feedwater Line Isolation Transient in the Ringhals 4 Power Plant (NUREG/IA-0038)
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Date Published: March 1992
Swedish Nuclear Power Inspectorate
Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Thermal-Hydraulic Code Assessment
and Application Program (ICAP)
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555
An inadvertent feedwater line isolation transient in a three loop Westinghouse PWR has been simulated with the frozen version of the TRAC-PF1/MOD1 computer code. The results reveal the capacity of the code to quantitatively predict the different pertinent phenomena. For accurate predictions of the system response it was found that a careful nodalization of the steam generator downcomer was essential for accurate pressure distribution and associated level prediction. Also the core moderator temperature reactivity coefficient was recommended to be decreased somewhat (be less negative) whereas the fuel gap conductance should be rather low in order to increase the initial stored energy of the fuel rods. It was also found during the course of the calculations that some restrictions had to be imposed on the allowable maximum timestep size in order not to violate the convergence of the solution procedure.