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Title: Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor

Abstract

An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin's temperature, fission power, and fission product ad actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in US reactors. The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently beingmore » irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX plus via this new methodology can provide insights into MOX fuel performance.« less

Authors:
;
Publication Date:
Research Org.:
Idaho National Engineering and Environmental Lab., Idaho Falls, ID (US)
Sponsoring Org.:
USDOE
OSTI Identifier:
20015667
DOE Contract Number:  
AC07-76ID01570
Resource Type:
Journal Article
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 129; Journal Issue: 3; Other Information: PBD: Mar 2000; Journal ID: ISSN 0029-5450
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; MIXED OXIDE FUELS; BURNUP; FUEL PINS; PWR TYPE REACTORS; ATR REACTOR; PERFORMANCE; M CODES; POWER DISTRIBUTION

Citation Formats

Chang, G S, and Ryskamp, J M. Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor. United States: N. p., 2000. Web.
Chang, G S, & Ryskamp, J M. Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor. United States.
Chang, G S, and Ryskamp, J M. Wed . "Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor". United States.
@article{osti_20015667,
title = {Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor},
author = {Chang, G S and Ryskamp, J M},
abstractNote = {An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin's temperature, fission power, and fission product ad actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in US reactors. The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently being irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX plus via this new methodology can provide insights into MOX fuel performance.},
doi = {},
url = {https://www.osti.gov/biblio/20015667}, journal = {Nuclear Technology},
issn = {0029-5450},
number = 3,
volume = 129,
place = {United States},
year = {2000},
month = {3}
}