Benchmarking assessment of RELAP5/MOD3 for the low flow and natural circulation experiment

Conference · Sun Nov 01 00:00:00 EST 1992

OSTI ID:10103105

Martin, R P ^[1]; Taylor, B K ^[2]

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1. EG and G Idaho, Inc., Idaho Falls, ID (United States)
2. Westinghouse Savannah River Co., Aiken, SC (United States)

The RELAP5/MOD3 code was assessed against experimental thermal hydraulics data for a 12.5 foot test section comprised of two vertical concentric tubes with water flowing upward in the tubes. The inner tubewas stainless steel and uniformly heated. ne outer tube was transparent polycarbonate (lexan) and unheated. The experimental procedure incorporated a test matrix of 24 tests to address single- and two-phase flow, forced and natural circulation flow and heated and unheated fluid. The tests were conducted at system pressures of 14.7 and 17.0 psia. Nine of the tests representing the full range of test conditions were analyzed using RELAP5/MOD3. RELAP5/MOD3 analysis of the tests yielded general agreement with experiment with regard to the prediction of forced flow and natural circulation trends. However, a number of deficiencies were observed in the RELAP5/MOD3 treatment and these, along with recommendations for their resolution, are described in the paper.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 10103105
• Report Number(s): WSRC-TR-92-088; CONF-9209172-4; ON: DE93003180
• Resource Relation: Conference: 5. nuclear reactor thermal-hydraulics (NURETH) conference,Salt Lake City, UT (United States),21-24 Sep 1992; Other Information: PBD: [1992]
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
PRODUCTION REACTORS
REACTOR COOLING SYSTEMS
HEAT TRANSFER
R CODES
HYDRAULICS
CIRCULATING SYSTEMS
FLUID FLOW
FUEL ASSEMBLIES
REACTOR CHANNELS
220600
420400
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
HEAT TRANSFER AND FLUID FLOW