CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core
The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power.
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6547432
- Report Number(s):
- LA-UR-82-2504; CONF-830301-2; ON: DE82021764
- Resource Relation:
- Conference: ASME-JSME thermal engineering joint conference, Honolulu, HI, USA, 20 Mar 1983; Other Information: Portions are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
Similar Records
Fort St. Vrain proof test element number two design, fabrication, and assembly report
Safety analysis report use of H-451 graphite in Fort St. Vrain fuel elements
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
REACTOR ACCIDENTS
HEAT TRANSFER
VRAIN REACTOR
C CODES
COMPUTER CODES
FINITE ELEMENT METHOD
ACCIDENTS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
NUMERICAL SOLUTION
POWER REACTORS
REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210300 - Power Reactors
Nonbreeding
Graphite Moderated