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Title: Application of the DARTart code for the assessment of advanced fuel behavior.

Abstract

No abstract prepared.

Authors:
; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
OSTI Identifier:
971143
Report Number(s):
ANL/NE/CP-59034
TRN: US1001038
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Conference
Resource Relation:
Conference: GLOBAL '07 - Advanced Nuclear Fuel Cycles and Systems; Sep. 9, 2007 - Sep. 132, 2007; Boise, Idaho
Country of Publication:
United States
Language:
ENGLISH
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; NUCLEAR FUELS; ANL

Citation Formats

Rest, J., Totev, T., and Nuclear Engineering Division. Application of the DARTart code for the assessment of advanced fuel behavior.. United States: N. p., 2007. Web.
Rest, J., Totev, T., & Nuclear Engineering Division. Application of the DARTart code for the assessment of advanced fuel behavior.. United States.
Rest, J., Totev, T., and Nuclear Engineering Division. Mon . "Application of the DARTart code for the assessment of advanced fuel behavior.". United States. doi:.
@article{osti_971143,
title = {Application of the DARTart code for the assessment of advanced fuel behavior.},
author = {Rest, J. and Totev, T. and Nuclear Engineering Division},
abstractNote = {No abstract prepared.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

Conference:
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  • The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less
  • The combination of new fuel compositions and higher burn-ups envisioned for the future means that representing fuel properties will be much more important, and yet more complex. Behavior within the oxide fuel rods will be difficult to model owing to the high temperatures, and the large number of elements generated and their significant concentrations that are a result of fuels taken to high burn-up. This unprecedented complexity offers an enormous challenge to the thermochemical understanding of these systems and opportunities to advance solid solution models to describe these materials. This paper attempts to model and simulate that behavior using anmore » oxide fuels thermochemical description to compute the equilibrium phase state and oxygen potential of LWR fuel under irradiation.« less
  • Verification of the SUHAM-U code has been carried out by the calculation of two-dimensional benchmark-experiment on critical light-water facility VENUS-2. Comparisons with experimental data and calculations by Monte-Carlo code UNK with the same nuclear data library B645 for basic isotopes have been fulfilled. Calculations of two-dimensional facility were carried out with using experimentally measured buckling values. Possibility of SUHAM code application for computations of PWR reactor with uranium and MOX fuel has been demonstrated. (authors)