Cladding axial elongation models for FRAP-T6. [PWR; BWR]

Conference · Sat Jan 01 00:00:00 EST 1983

OSTI ID:5539450

Shah, V N; Carlson, E R; Berna, G A

This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5539450
• Report Number(s): EGG-M-17682; CONF-830805-59; ON: DE84001049
• Resource Relation: Conference: 7. international conference on structural mechanics in reactor technology, Chicago, IL, USA, 22 Aug 1983
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
FUEL CANS
TRANSIENTS
DEFORMATION
FUEL-CLADDING INTERACTIONS
PWR TYPE REACTORS
COMPUTER CALCULATIONS
FUEL RODS
REACTOR SAFETY
STRAINS
STRESS ANALYSIS
THERMAL EXPANSION
EXPANSION
FUEL ELEMENTS
REACTOR COMPONENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled