Analysis of the General Electric Company swell tests with RELAP4/MOD7. [BWR]

Conference · Mon Jan 01 00:00:00 EST 1979

OSTI ID:6270138

Fischer, S R; Hendrix, C E

The RELAP4/MOD7 nuclear reactor transient analysis code, presently being developed by EG and G Idaho, Inc., will incorporate several significant improvements over earlier versions of RELAP4. As part of the development of RELAP4/MOD7, a thorough assessment of the capability of the code to simulate water reactor LOCA phenomena is being made. This assessment is accomplished in part by comparing results from code calculations with test data from experimental facilities. Simulations of the General Electric Company (GE) level swell tests were performed as part of the code checkout. In these tests, a pressurized vessel partially filled with nearly saturated water was blown down through a simulated break located near the top of the vessel. Comparison of RELAP4 calculations with data from these experiments indicates that the code has the capability to model the unequal phase velocity flow and resulting density gradients that might occur in a BWR steam line break transient. Comparisons of RELAP4 calculations with data from two level swell experiments are presented.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 6270138
• Report Number(s): CONF-790602-13; TRN: 79-013720
• Resource Relation: Conference: ANS annual meeting, Atlanta, GA, USA, 3 Jun 1979
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BLOWDOWN
DYNAMIC LOADS
SIMULATION
BWR TYPE REACTORS
LOSS OF COOLANT
PRESSURE SUPPRESSION
TEST FACILITIES
ACCIDENTS
REACTOR ACCIDENTS
REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled