RELAP5/MOD2 assessment, OECD-LOFT small break experiment LP-SB-03

Technical Report · Sun Apr 01 00:00:00 EST 1990

DOI:https://doi.org/10.2172/6913283· OSTI ID:6913283

Guntay, S ^[1]

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1. Paul Scherrer Inst., Wuerenlingen (Switzerland)

An analysis of the experimental results and post-test calculations using RELAP5/MOD2 carried out for OECD-LOFT small break experiment LP-SB-3 are presented. Experiment LP-SB-3 was conducted on March 5, 1984 in Loss-of-Fluid Test (LOFT) facility located at the Idaho National Engineering Laboratory (INEL). The experiment simulated a small cold leg break, with concurrent loss of high pressure injection system, and cooldown and recovery by feed and bleed of the steam generator secondary side and accumulator injection, respectively. The analysis was under taken as a part of a program at EIR aimed at developing experience in using the latest generation of best estimate Loss of Coolant Accident (LOCA) analysis computer codes, and to improve understanding of Small Break LOCA transients and as well as a part of a program aimed at assessing the RELAP5/MOD2 code. The latest available version (Cycle 33 to 36.1) of the code was used. The particular test selected for the analysis included several phenomena potentially relevant to any PWR plant operation in Switzerland. This report documents a short post-test analysis of the experiment emphasizing the results of additional analysis performed during the course of this task. RELAP5/MOD2 input model and results of the post-test calculation are documented. Included in the report are the results of a sensitivity analysis which show the predicted thermal-hydraulic response to a different input model. 7 refs., 55 figs., 2 tabs.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Paul Scherrer Inst., Wuerenlingen (Switzerland)
• Sponsoring Organization: USNRC
• OSTI ID: 6913283
• Report Number(s): NUREG/IA-0018; ON: TI90010669; TRN: 90-016077
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
COMPUTERIZED SIMULATION
PWR TYPE REACTORS
CLADDING
COMPARATIVE EVALUATIONS
FAILURE MODE ANALYSIS
FLUID FLOW
FUEL ELEMENTS
HEAT TRANSFER
HOMOGENEOUS MIXTURES
HYDRAULICS
INTERNATIONAL COOPERATION
LEAKS
O CODES
PERFORMANCE TESTING
PIPES
PRESSURE EFFECTS
R CODES
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY
RISK ASSESSMENT
STEAM GENERATORS
SWITZERLAND
THERMODYNAMICS
TRANSIENTS
US DOE
VAPORS
ACCIDENTS
BOILERS
COMPUTER CODES
COOLING SYSTEMS
COOPERATION
DEPOSITION
DISPERSIONS
ENERGY SYSTEMS
ENERGY TRANSFER
EUROPE
FLUID MECHANICS
FLUIDS
GASES
MECHANICS
MIXTURES
NATIONAL ORGANIZATIONS
OPERATION
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
SIMULATION
SURFACE COATING
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
TESTING
US ORGANIZATIONS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WESTERN EUROPE
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled