Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

Technical Report · Thu Jul 01 04:00:00 EDT 1993

DOI:https://doi.org/10.2172/10186950· OSTI ID:10186950

Preece, R J; Putney, J M ^[1]

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1. National Power, Leatherhead (United Kingdom). Technology and Environment Centre

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; National Power, Leatherhead (United Kingdom). Technology and Environment Centre
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 10186950
• Report Number(s): NUREG/IA--0113; ON: TI94000498
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200
99 GENERAL AND MISCELLANEOUS
990200
COMPUTER CALCULATIONS
COMPUTERIZED SIMULATION
HEAT TRANSFER
HYDRAULICS
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PWR TYPE REACTORS
R CODES
STEAM GENERATORS
VOID FRACTION
WOLF CREEK-1 REACTOR