Assessment of computer codes for VVER-440/213-type nuclear power plants

Technical Report · Fri Sep 01 04:00:00 EDT 1995

DOI:https://doi.org/10.2172/115082· OSTI ID:115082

Szabados, L.; Ezsol, Gy.; Perneczky, ^[1]

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1. Atomic Energy Research Institute, Budapest (Hungary)

Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 115082
• Report Number(s): NUREG/CP--0142-Vol.3; CONF-950904--Vol.3; ON: TI95017079
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
99 GENERAL AND MISCELLANEOUS
COMPARATIVE EVALUATIONS
COMPUTER CODES
HYDRAULICS
IAEA
LOSS OF COOLANT
MATHEMATICAL MODELS
NUMERICAL DATA
PAKS-1 REACTOR
PAKS-2 REACTOR
PAKS-3 REACTOR
PAKS-4 REACTOR
R CODES
REACTOR SAFETY
SAFETY ANALYSIS
WWER TYPE REACTORS