RELAP5 assessment: LOFT large break L2-5

Technical Report · Wed Feb 01 04:00:00 EST 1984

DOI:https://doi.org/10.2172/6815860· OSTI ID:6815860

Thompson, S L; Kmetyk, L N

RELAP5 is part of an effort 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 various integral and separate effects test facilities. As part of this assessment matrix, a large break transient performed at the LOFT facility has been analyzed. The results show that RELAP5/MOD1 correctly calculates many of the major system variables (i.e., pressure, break flows, peak clad temperature) early in a large break LOCA. The major problems encountered in the analyses were incorrect pump coastdown and loop seal clearing early in the calculation, excessive pump speedup later in the transient (probably due to too much condensation-induced pressure drop at the ECC injection point), and excess ECC bypass calculated throughout the later portions of the test; only the latter problem significantly affected the overall results. This excess ECC bypass through the downcomer and vessel-side break resulted in too-large late-time break flows and high system pressure due to prolonged choked flow conditions. It also resulted in a second core heatup being calculated after the accumulator emptied, since water was not being retained in the vessel. Analogous calculations with a split-downcomer nodalization delivered some ECC water to the lower plenum, which was then swept up the core and upper plenum and out the other (pump-side) break; thus no significant differences in long-term overall behavior were evident between the calculations.

• Research Organization: Sandia National Labs., Albuquerque, NM (USA)
• DOE Contract Number: AC04-76DP00789
• OSTI ID: 6815860
• Report Number(s): NUREG/CR-3608; SAND-83-2549; ON: DE84011863
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
ECCS
ENGINEERED SAFETY SYSTEMS
LOFT REACTOR
LOSS OF COOLANT
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
TANK TYPE REACTORS
TEST REACTORS
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS