PCI analysis of a commercial PWR using BISON fuel performance code

Conference · Fri Jul 01 04:00:00 EDT 2016

DOI:https://doi.org/10.1016/j.nucengdes.2017.09.004· OSTI ID:22764089

Capps, N.; Wirth, B. D. ^[1]; Montgomery, R.; Sunderland, D. ^[2]

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1. Department of Nuclear Engineering, University of Tennessee, 1412 Circle Dr, Knoxville, TN 37916 (United States)
2. Pacific Northwest National Lab, Richland, WA 99354 (United States)

Through the early 2000's, cladding failure by Pellet-Cladding Interaction (PCI) processes in commercial PWR fuel has been managed successfully using power ramp restrictions. However, more demanding fuel duty requirements in modern reactor operating strategies combined with missing pellet surface (MPS) defects have resulted in a small number of fuel rod failures in the last decade. In response to these events, efforts are underway in fuel behavior modeling and simulation to improve our understanding of the conditions that may lead to cladding failure with MPS defects. This paper will evaluate a methodology, using the BISON fuel performance code (formally Peregrine), to model PCI related failures in the ramp test experiments using 2-D and 3-D geometric representation. This methodology consists of four main steps that together are used to understand the PCI behavior of irradiated fuel. The first step consists of a steady state R-Z depletion analysis of the limiting rod or one with a known failure, is performed to establish the fuel rod conditions, e.g. pellet-cladding gap, plenum pressure, and released fission gas, following the first cycle of operation. The results of the steady state R-Z analysis provides the initial fuel rod conditions used in the third and fourth steps, which consist of analyzing the startup power ramp or a mid cycle power maneuver. The second step of the analysis consists of a full length R-Z analysis of the startup ramp. The purpose is to locate the region in the cladding were the maximum hoop stress is identified, using the R-Z power ramp analysis. In the third step, the local cladding stresses and PCI damage index response are calculated using the R-θ local effects model at the peak stress location from the R-Z analysis. Similar to the third step, step four will evaluate the 3-D geometric effects on the local stress concentration, PCI damage index response, as well as critical strain energy density, and evaluate the failure potential for each rod in the assembly. The purpose of the 3-D analysis will be to determine what the critical length and width of MPS will lead to a through wall failure.

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22764089
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
B CODES
CLADDING
COMPUTERIZED SIMULATION
CONCENTRATION RATIO
DAMAGE
DEFECTS
ENERGY DENSITY
FAILURES
FISSION PRODUCT RELEASE
FISSION PRODUCTS
FUEL PELLETS
FUEL RODS
GEOMETRY
PWR TYPE REACTORS
SPENT FUELS
STEADY-STATE CONDITIONS
THREE-DIMENSIONAL CALCULATIONS
TWO-DIMENSIONAL CALCULATIONS