Assessment of TRAC-PF1/MOD1 Against an Inadvertent Steam Line Isolation Valve Closure in the Ringhals 2 Power Plant (NUREG/IA-0041)
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Date Published: March 1992
Consejo de Seguridad Nuclear
c/Sor Angela De La Cruz, 3
Studsvik Energiteknik A. B.
S-61182 Nykoping, Sweden
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
A steam line isolation valve closure transient in a three loop Westinghouse PWR has been simulated with the frozen version of TRAC-PF1/MOD1 computer code. The results reveal the capability of the code to quantitatively predict the different pertinent phenomena. For accurate predictions of the system response it was realized that careful nodalization of the steam generator dome region and outlet nozzle was required as well as of the pressurizer walls and spray nozzle. The amount of initially stored energy in the fuel had an essential impact on the after scram short-term prediction. Proper control system behaviour was of major concern. Difficulties in adequate control system operation were encountered when large timestep sizes were used.