Published 1996
| Version v1
Book
Modeling of fuel-rod behavior during reactor power cycling and ramping experiments with computer code FRAPCON-2
Description
Modelling of fuel-rod behavior during reactor power cycling and ramping (including power-cooling mismatch experiments) with the computer code FRAPCON-2 is discussed. FRAPCON-2 computer calculations, using different mechanical models (Rigid Pellet, Deformable Pellet and Finite Element Mechanical Models) are compared with experimental results. The range of conditions over which FRAPCON-2 may be applied for PWR fuel rod behavior modelling during reactor power cycling and ramping are illustrated
Additional details
Publishing Information
- Publisher
- American Society of Mechanical Engineers.
- Imprint Place
- New York, NY (United States)
- ISBN
- 0-7918-1226-X
- Imprint Title
- ICONE-4: Proceedings. Volume 3: Safety and reliability
- Imprint Pagination
- 595 p.
- Journal Page Range
- p. 87-90.
Conference
- Title
- ASME/JSME international conference on nuclear engineering.
- Acronym
- ICONE 4
- Dates
- 10-13 Mar 1996.
- Place
- New Orleans, LA (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 27063082
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S99: GENERAL AND MISCELLANEOUS;
- Resource subtype / Literary indicator
- Conference, Numerical Data
- Descriptors DEI
- COMPARATIVE EVALUATIONS; COMPUTER CODES; COMPUTERIZED SIMULATION; EVALUATION; F CODES; FUEL RODS; POWER-COOLING-MISMATCH ACCIDENTS; PWR TYPE REACTORS; REACTOR SAFETY; REACTOR SAFETY EXPERIMENTS; THEORETICAL DATA
- Descriptors DEC
- ACCIDENTS; DATA; ENRICHED URANIUM REACTORS; FUEL ELEMENTS; INFORMATION; NUMERICAL DATA; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-960306--.