COSTAX-BOIL, Transient Dynamic Analysis of BWR and PWR in Axial Geometry
Creators
- 1. J.R.C. EURATOM, Ispra Establishment (Italy)
Description
1 - Nature of physical problem solved: Transient analysis of BWR are performed in the time domain. The code couples the two velocity groups neutron diffusion kinetic calculations (up to 10 delayed neutron groups) in one dimension (axial) with the heat conduction equations in the fuel rods and a two phase flow dynamics calculation for the typical coolant channel. The program is intended for the analysis of intermediate accidents or severe operational transients in which a detailed spatial representation is essential. 2 - Method of solution: Doppler and void reactivity feedback are accounted for pointwise. A flexible input allows representation of reactivity transients, flow recirculation, or inlet enthalpy transients, or an arbitrary mixture of them, including control rod bank movement. No pressure variation is possible in the present version
Availability note (English)
Available on-line: http://www.nea.fr/abs/html/nea-0398.htmlAdditional details
Identifiers
Publishing Information
- Imprint Pagination
- [html]
INIS
- Country of Publication
- Nuclear Energy Agency of the OECD (NEA)
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 40085980
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Computer Program Description, Non-conventional Literature
- Descriptors DEI
- BWR TYPE REACTORS; C CODES; COMPUTER PROGRAM DOCUMENTATION; COMPUTERIZED SIMULATION; CONTROL ELEMENTS; COOLANTS; DELAYED NEUTRONS; DOPPLER COEFFICIENT; ENTHALPY; FUEL RODS; HYDRODYNAMICS; LOSS OF COOLANT; NEUTRON DIFFUSION EQUATION; PWR TYPE REACTORS; REACTIVITY; REACTOR ACCIDENT SIMULATION; THERMAL CONDUCTION; TRANSIENTS; TWO-PHASE FLOW; WEBSITES
- Descriptors DEC
- ACCIDENTS; BARYONS; COMPUTER CODES; DIFFERENTIAL EQUATIONS; DIFFUSION EQUATIONS; DOCUMENT TYPES; ELEMENTARY PARTICLES; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EQUATIONS; FERMIONS; FISSION NEUTRONS; FLUID FLOW; FLUID MECHANICS; FUEL ELEMENTS; HADRONS; HEAT TRANSFER; MECHANICS; NEUTRONS; NUCLEONS; PARTIAL DIFFERENTIAL EQUATIONS; PHYSICAL PROPERTIES; POWER REACTORS; REACTIVITY COEFFICIENTS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SIMULATION; THERMAL REACTORS; THERMODYNAMIC PROPERTIES; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 3 refs.