Independent code assessment at BNL in FY 1982. [TRAC-PF1; RELAP5/MOD1; TRAC-BD1]
Conference
·
OSTI ID:6669708
Independent assessment of the advanced codes such as TRAC and RELAP5 has continued at BNL through the Fiscal Year 1982. The simulation tests can be grouped into the following five categories: critical flow, counter-current flow limiting (CCFL) or flooding, level swell, steam generator thermal performance, and natural circulation. TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes were assessed by simulating all of the above experiments, whereas the TRAC-BD1 (Version 12.0) code was applied only to the CCFL tests. Results and conclusions of the BNL code assessment activity of FY 1982 are summarized below.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 6669708
- Report Number(s):
- BNL-NUREG-32101; CONF-821037-38; ON: DE83003162
- Country of Publication:
- United States
- Language:
- English
Similar Records
Independent assessment of TRAC-PF1 (Version 7. 0), RELAP5/MOD1 (Cycle 14), and TRAC-BD1 (Version 12. 0) codes using separate-effects experiments
Independent assessment of TRAC and RELAP5 codes through separate effects tests
Comparative analysis of constitutive relations in TRAC-PF1 and RELAP5/MOD1
Technical Report
·
Thu Aug 01 00:00:00 EDT 1985
·
OSTI ID:6187959
Independent assessment of TRAC and RELAP5 codes through separate effects tests
Conference
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5388470
Comparative analysis of constitutive relations in TRAC-PF1 and RELAP5/MOD1
Technical Report
·
Sat Jun 01 00:00:00 EDT 1985
·
OSTI ID:6332609
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
ACCIDENTS
BOILERS
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONVECTION
CRITICAL FLOW
EVALUATION
FLUID FLOW
NATURAL CONVECTION
PRESSURE GRADIENTS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY EXPERIMENTS
REACTORS
SIMULATION
STEAM GENERATORS
T CODES
TEMPERATURE GRADIENTS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BOILERS
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONVECTION
CRITICAL FLOW
EVALUATION
FLUID FLOW
NATURAL CONVECTION
PRESSURE GRADIENTS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY EXPERIMENTS
REACTORS
SIMULATION
STEAM GENERATORS
T CODES
TEMPERATURE GRADIENTS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS