RELAP5 assessment: semiscale natural circulation tests S-NC-3, S-NC-4, and S-NC-8

Technical Report · Tue May 01 04:00:00 EDT 1984

DOI:https://doi.org/10.2172/6894374· OSTI ID:6894374

Wong, C N.C.; Kmetyk, L N

The RELAP5/MOD1 independent assessment project at Sandia National Laboratories (SNLA) is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The RELAP5 code is being assessed at SNLA against test data from a number of integral and separate effects test facilities. As part of this assessment matrix, natural circulation tests performed at the Semiscale facility were analyzed. Results for the single-loop and two-loop steady state basecase tests S-NC-2 and S-NC-7 have already been documented separately; this report gives the results of calculations for two single-loop degraded heat transfer tests, S-NC-3 and S-NC-4, and for the two-loop ultra-small break transient test S-NC-8. For tests S-NC-3 and S-NC-4, the analyses show that RELAP5/MOD1 describes correctly the qualitative influence of steam generator secondary side heat transfer degradation on both two-phase and reflux natural circulation. The results for test S-NC-8, an ultra-small (0.4%) cold leg break, also compare reasonably well with the outcome of that experiment. 17 references.

• Research Organization: Sandia National Labs., Albuquerque, NM (USA)
• DOE Contract Number: AC04-76DP00789
• OSTI ID: 6894374
• Report Number(s): NUREG/CR-3690; SAND-84-0402; ON: TI84016447
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT
AFTER-HEAT REMOVAL
BOILERS
COMPUTER CALCULATIONS
CONVECTION
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NATURAL CONVECTION
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
REMOVAL
SAFETY
STEAM GENERATORS
TEMPERATURE GRADIENTS
TEST FACILITIES
THEORETICAL DATA
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS