Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

Technical Report · Wed Jul 01 00:00:00 EDT 1992

DOI:https://doi.org/10.2172/7074036· OSTI ID:7074036

Roth, P A; Schultz, R R ^[1]; Choi, C J ^[2]

---

1. EG and G Idaho, Inc., Idaho Falls, ID (United States)
2. Korea Atomic Energy Research Inst., Daebuk (Korea, Republic of)

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

• Research Organization: EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: NRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 7074036
• Report Number(s): EGG-NE-10353; ON: DE92019689
• 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
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
DEPRESSURIZATION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HUMAN FACTORS
HYDRAULICS
LOSS OF COOLANT
MECHANICS
POWER REACTORS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SCALE MODELS
SIMULATION
STRUCTURAL MODELS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS