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Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

Journal Article · · Nuclear Engineering and Design
 [1];  [1];  [1];  [2];  [3];  [3];  [3];  [3]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
  3. Idaho National Lab. (INL), Idaho Falls, ID (United States)
+ Show Author Affiliations
The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.
Research Organization:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility (OLCF)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE); USDOE Office of Science (SC)
Grant/Contract Number:
AC05-00OR22725; AC07-05ID14517
OSTI ID:
1429216
Alternate ID(s):
OSTI ID: 1549000
Journal Information:
Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Journal Issue: C Vol. 327; ISSN 0029-5493
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English

References (7)

A new fission-gas release model journal August 1983
Pellet-clad interaction (PCI) failures of zirconium alloy fuel cladding — A review journal August 1990
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT journal December 2016
The Virtual Environment for Reactor Applications (VERA): Design and architecture journal December 2016
Multidimensional multiphysics simulation of nuclear fuel behavior journal April 2012
MOOSE: A parallel computational framework for coupled systems of nonlinear equations journal October 2009
VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion journal January 2017

Cited By (1)

The technological challenge for current generation nuclear reactors journal September 2019

Figures / Tables (14)

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Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
BISON
CASL
Fuel performance
PCMI
VERA