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Title: Behavior of U 3Si 2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

Abstract

As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

Authors:
 [1];  [1];  [1];  [1];  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1364507
Report Number(s):
INL/EXT-16-40059
TRN: US1703365
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ACCIDENT-TOLERANT NUCLEAR FUELS; CLADDING; FUKUSHIMA DAIICHI NUCLEAR POWER STATION; ACCIDENTS; Accident Tolerant Fuel; BISON; LOCA; Sensitivity Analysis

Citation Formats

Gamble, Kyle Allan Lawrence, Hales, Jason Dean, Barani, Tommaso, Pizzocri, Davide, and Pastore, Giovanni. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions. United States: N. p., 2016. Web. doi:10.2172/1364507.
Gamble, Kyle Allan Lawrence, Hales, Jason Dean, Barani, Tommaso, Pizzocri, Davide, & Pastore, Giovanni. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions. United States. doi:10.2172/1364507.
Gamble, Kyle Allan Lawrence, Hales, Jason Dean, Barani, Tommaso, Pizzocri, Davide, and Pastore, Giovanni. 2016. "Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions". United States. doi:10.2172/1364507. https://www.osti.gov/servlets/purl/1364507.
@article{osti_1364507,
title = {Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions},
author = {Gamble, Kyle Allan Lawrence and Hales, Jason Dean and Barani, Tommaso and Pizzocri, Davide and Pastore, Giovanni},
abstractNote = {As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.},
doi = {10.2172/1364507},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less
  • Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less
  • The annular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain time-resolved data on the release of fission products from irradiated fuels under well-controlled light water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST-1 was conducted in a highly reducing environment at a system pressure of approximately 0.19 MPa, and at maximum fuel temperatures of about 2490 K. The data will be used for the development and validation of mechanistic fission product release computer codes such as VICTORIA.
  • The annular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain time-resolved data on the release of fission products from irradiated fuels under well-controlled light water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST-1 was conducted in a highly reducing environment at a system pressure of approximately 0.19 MPa, and at maximum fuel temperatures of about 2490 K. The data will be used for the development and validation of mechanistic fission product release computer codes such as VICTORIA.