A study of RELAP5/MOD2 and RELAP5/MOD3 predictions of a small-break loss-of-coolant accident simulation conducted at the ROSA-IV Large-Scale Test Facility

Journal Article · Thu Oct 01 00:00:00 EDT 1992 · Nuclear Technology; (United States)

OSTI ID:7186873

Sloan, S M ^[1]; Hassan, Y ^[2]

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1. EG and G Idaho, Inc., Idaho Falls, ID (United States)
2. Texas A and M Univ., College Station, TX (United States). Dept. of Nuclear Engineering

The thermal-hydraulics simulation codes RELAP5/MOD2 and RELAP5/MOD3 are utilized to calculate the phenomena that occurred during a small-break loss-of-coolant accident (LOCA) simulation conducted at the ROSA-IV Large-Scale Test Facility. In this paper the RELAP5/MOD2 and RELAP5/MOD3 predictions are compared with each other and assessed against the experimental results. The overall conclusion is that both code versions predict trends well, but each differs in the prediction of the magnitude and timing of concurrences. Specific areas of difference include primary system pressure, differential pressure in the upper plenum, core liquid level depression and subsequent heatup, core void fraction profile, and the differential pressure in the steam generator inlet plenum. All but the last of these differences are related to the RELAP5/MOD3 prediction of excessive liquid holdup in the upper plenum during the first core liquid depression, which is believed to lead to the prediction of water trickling into the upper core volumes and providing a cooling mechanism not present during the experiment. The liquid holdup is believed to be the result of an overprediction of interphase drag at the junctions between the upper plenum volumes.

• OSTI ID: 7186873
• Journal Information: Nuclear Technology; (United States), Vol. 100:1; ISSN 0029-5450
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
REACTOR ACCIDENTS
SIMULATION
REACTOR CORES
TEST FACILITIES
FORECASTING
HEAT TRANSFER
LOSS OF COOLANT
M CODES
PRESSURE GRADIENTS
R CODES
STEAM GENERATORS
VOID FRACTION
ACCIDENTS
BOILERS
COMPUTER CODES
ENERGY TRANSFER
REACTOR COMPONENTS
VAPOR GENERATORS
220900* - Nuclear Reactor Technology- Reactor Safety
420400 - Engineering- Heat Transfer & Fluid Flow
990200 - Mathematics & Computers