Development assessment of RELAP5/MOD3 using the Semiscale natural circulation tests
Conference
·
· Transactions of the American Nuclear Society; (USA)
OSTI ID:5786231
This paper documents the simulation of the Semiscale natural circulation (SNC) tests SNC-01, SNC-03, and SNC-04 using RELAP5/MOD3 for developmental assessment. The main purpose of applying MOD3 to these tests is to show the code's capability of single- and two-phase natural circulation, reflux heat transfer, and countercurrent flow with the improved models. A brief description of the Semiscale test facility and RELAP5/MOD3 system model is given, followed by a description of some code results and analysis of the phenomena simulated. The RELAP5/MOD3 systems analysis code has simulated the Semiscale natural circulation tests. In general, the code calculations are in good agreement with the measured data at the higher PCS and steam generator mass inventories. Additionally, the code performance at the higher PCS mass inventories is an improvement over previous RELAP5/MOD2 calculations of this problem.
- OSTI ID:
- 5786231
- Report Number(s):
- CONF-901101--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 62
- Country of Publication:
- United States
- Language:
- English
Similar Records
Developmental assessment of RELAP5/MOD3 using the semiscale natural circulation tests
Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)
Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)
Conference
·
Sun Dec 31 23:00:00 EST 1989
·
OSTI ID:5561078
Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)
Technical Report
·
Sat May 01 00:00:00 EDT 1993
·
OSTI ID:10163345
Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)
Technical Report
·
Sat May 01 00:00:00 EDT 1993
·
OSTI ID:6357917
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 CODES
COMPUTERIZED SIMULATION
CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MASS TRANSFER
MECHANICS
MODIFICATIONS
NATIONAL ORGANIZATIONS
NATURAL CONVECTION
PERFORMANCE
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
SAFETY
SIMULATION
STEAM GENERATORS
TEST FACILITIES
TRANSIENTS
TWO-PHASE FLOW
US NRC
US ORGANIZATIONS
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 CODES
COMPUTERIZED SIMULATION
CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MASS TRANSFER
MECHANICS
MODIFICATIONS
NATIONAL ORGANIZATIONS
NATURAL CONVECTION
PERFORMANCE
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
SAFETY
SIMULATION
STEAM GENERATORS
TEST FACILITIES
TRANSIENTS
TWO-PHASE FLOW
US NRC
US ORGANIZATIONS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS