Assessing pellet-clad interaction with VERA: WBN1, cycles 6-7

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

OSTI ID:22764072

Stimpson, S. G.; Clarno, K. T.; Powers, J. J. ^[1]; Pawlowski, R. P. ^[2]

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1. Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)
2. multi-physics Applications Department, Sandia National Laboratories, Albuquerque, NM 87185 (United States)

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high fidelity simulations of nuclear reactors through the development of the Virtual Environment for Reactor Applications (VERA). One of the primary goals of CASL is to estimate local cladding failure through pellet-clad interaction (PCI). While this capability is being pursued through several different approaches, the procedure presented here focuses on running independent fuel performance calculations with the BISON fuel performance code being developed by Idaho National Laboratory using the multicycle output data from coupled neutron transport/thermal-hydraulics simulations generated by MPACT, CTF, and MAMBA. This type of approach is consistent with traditional fuel performance analysis methods, where the fuel performance analyses are typically separate from core simulation analyses. Recent work simulating 12 cycles of Watts Bar Unit 1 (WBN1) with VERA are capitalized upon, and stand-alone quarter-core BISON results can be obtained for several cycles, assessing parameters of interest to PCI (e.g., maximum centerline fuel temperature, maximum clad hoop stress, minimum gap thickness, and cumulative damage index). This is accomplished by using the coupled neutronics/thermal-hydraulics output power and temperature distributions as inputs for each BISON fuel rod case. Building on previous work which demonstrated this capability for Cycles 1 through 3, this paper presents the results for Cycles 6 and 7. Cycle 7 is of particular interest because significant CRUD-induced power shift was observed. While the current results do not immediately indicate that any rods are at an extreme risk of failure, further analysis on rods highlighted by this work may shed additional light on the likelihood of failure. (authors)

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22764072
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
CLADDING
COMPUTERIZED SIMULATION
DAMAGE
FAILURES
FUEL PELLETS
FUEL RODS
IDAHO NATIONAL LABORATORY
NEUTRON TRANSPORT
NUCLEAR FUELS
PERFORMANCE
TEMPERATURE DISTRIBUTION
THERMAL HYDRAULICS
THICKNESS
WATER COOLED REACTORS
WATER MODERATED REACTORS