Development assessment of RELAP5/MOD3 using the Semiscale natural circulation tests
Creators
Description
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
Additional details
Publishing Information
- Journal Title
- Transactions of the American Nuclear Society
- Journal Volume
- 62
- Series
- Trans. Am. Nucl. Soc.
- Journal Page Range
- 687-689
- ISSN
- 0003-018X
- CODEN
- TANSA
Conference
- Title
- American Nuclear Society (ANS) winter meeting.
- Dates
- 11-15 Nov 1990.
- Place
- Washington, DC (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 22060713
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPUTER CODES; COMPUTERIZED SIMULATION; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; MODIFICATIONS; NATURAL CONVECTION; PERFORMANCE; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; R CODES; REACTOR ACCIDENTS; REACTOR CORES; REACTOR SAFETY; REACTOR SAFETY EXPERIMENTS; STEAM GENERATORS; TEST FACILITIES; TRANSIENTS; TWO-PHASE FLOW; US NRC
- Descriptors DEC
- ACCIDENTS; BOILERS; CONVECTION; COOLING SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID FLOW; NATIONAL ORGANIZATIONS; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; US ORGANIZATIONS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-901101--.