Review and evaluation of the RELAP5YA computer code and the Vermont Yankee LOCA (Loss-of-Coolant Accident) licensing analysis model for use in small and large break BWR (Boiling Water Reactor) LOCAS
A review has been completed of the RELAP5YA computer code to determine its acceptability for performing licensing analyses. The review was limited to Boiling Water Reactor (BWR) reactor applications. In addition, a Loss-Of-Coolant Accident (LOCA) licensing analysis method, using the RELAP5YA computer code, has been reviewed. This method is applicable to the Vermont Yankee Nuclear Power Station to perform full break spectra LOCA and fuel cycle independent analyses. The review of the RELAP5YA code consisted of an evaluation of all Yankee Atomic Electric Company (YAEC) incorporated modifications to the RELAP5/MOD1 Cycle 18 computer code from which the licensing version of the code originated. Qualifying separate and integral effects assessment calculations were reviewed to evaluate the validity and proper implementation of the various added models. The LOCA licensing method was assessed by reviewing two RELAP5YA system input models and evaluating several small and large break qualifying transient calculations. A review of the RELAP5YA code modifications and their assessments, as well as the submitted LOCA licensing method, is given and the results of the review are provided.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6570634
- Report Number(s):
- EGG-RTH-7506; ON: DE87009083
- Resource Relation:
- Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
Similar Records
RELAP5YA simulation of large-break tests in the two-loop test apparatus
Small break LOCA analysis for Maanshan nuclear power plant
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
BWR TYPE REACTORS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
R CODES
FUEL CYCLE
HIGH PRESSURE COOLANT INJECTION
MATHEMATICAL MODELS
PRIMARY COOLANT CIRCUITS
RCIC SYSTEMS
REACTOR LICENSING
REACTOR SAFETY
TESTING
THERMODYNAMICS
TRANSIENTS
VERMONT YANKEE REACTOR
ACCIDENTS
COMPUTER CODES
COOLING SYSTEMS
ECCS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLUID MECHANICS
LICENSING
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
990220 - Computers
Computerized Models
& Computer Programs- (1987-1989)