Published December 1980
| Version v1
Report
Restricted
The effect of hot spots upon swelling of Zircaloy cladding as modelled by the code CANSWEL-2
- 1. UKAEA Springfields Nuclear Power Development Labs.
Description
The code CANSWEL-2 models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised-water reactor (PWR). It can treat azimuthal non-uniformities in cladding thickness and temperature, and model the mechanical restraint imposed by the nearest neighbouring rods, including situations where cladding is forced into non-circular shapes. The physical and mechanical models used in the code are presented. Applications of the code are described, both as a stand-alone version and as part of the PWR LOCA code MABEL-2. Comparison with a limited number of relevant out-of-reactor creep strain experiments has generally shown encouraging agreement with the data. (author)
Availability note (English)
MF available from INIS under the Report Number.
Files
Additional details
Publishing Information
- Imprint Pagination
- 22 p.
- Report number
- ND-R--529(S)
Conference
- Title
- Specialists' meeting on the safety aspects of fuel behaviour in off-normal and accident conditions.
- Dates
- 1 - 4 Sep 1980.
- Place
- Espoo, Finland.
INIS
- Country of Publication
- United Kingdom
- Country of Input or Organization
- United Kingdom
- INIS RN
- 13713616
- Subject category
- S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- C CODES; COMPUTER CODES; CREEP; DEFORMATION; FUEL CANS; FUEL ELEMENT CLUSTERS; LOSS OF COOLANT; MATHEMATICAL MODELS; PERFORMANCE; PWR TYPE REACTORS; ZIRCALOY
- Descriptors DEC
- ACCIDENTS; ALLOYS; FUEL ASSEMBLIES; MECHANICAL PROPERTIES; REACTOR ACCIDENTS; REACTORS; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS