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Title: Fuel Performance Calculations for FeCrAl Cladding in BWRs

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.
Authors:
 [1] ;  [1] ;  [1] ;  [1] ;  [2] ;  [2]
  1. Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
OSTI Identifier:
1237610
DOE Contract Number:
AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: ANS Winter 2015, Washington, DC (United States), 8-12 Nov 2015
Research Org:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ACCIDENT-TOLEANT NUCLEAR FUELS; BWR TYPE REACTORS; FUEL CANS; ZIRCALOY; PERFORMANCE; FUEL RODS; REACTIVITY; STRAINS; TEMPERATURE DISTRIBUTION; IRON BASE ALLOYS; CHROMIUM ALLOYS; ALUMINIUM ALLOYS; TERNARY ALLOY SYSTEMS; COMPARATIVE EVALUATIONS; B CODES nuclear; nuclear fuel; accident tolerant fuel; ATF; boiling water reactor; BWR; fuel performance; BISON; NESTLE; FeCrAl; cladding; alternate cladding material