Heat transfer simulations of the UO2 particle–graphite system in TREAT fuel

Abstract In this study, a heat transfer simulation of a UO 2 particle–graphite system in highly enriched nuclear fuel at the Transient Reactor Test Facility (TREAT) was performed using the finite element method. Different factors that can impact fuel performance were modeled and implemented in the simulated micro-scale UO 2 particle–graphite system. The fission fragments caused an irradiation-induced degradation of the thermal conductivity of the graphite, which added major heat resistance to the irradiated system. The effect of graphite quality and irradiation on the UO 2 particles has also been evaluated, but neither has an impact as pronounced as the fission fragment damage to the graphite. By combining these factors, the dynamic temperature profiles were obtained, and the limitations on particle size in the irradiated and unirradiated UO 2 particle–graphite systems have been determined.

[1]  S. Zinkle,et al.  Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation , 2005 .

[2]  Steven J. Zinkle,et al.  Advanced Oxidation Resistant Iron-Based Alloys for LWR Fuel Cladding , 2014 .

[3]  L. P. Hunter.  Effect of Fission Recoil Fragments on the Thermal Conductivity of Graphite , 1959 .

[4]  J. K. Fink,et al.  Thermophysical properties of uranium dioxide , 2000 .

[5]  K. Terrani,et al.  Silicon carbide composite for light water reactor fuel assembly applications , 2014 .

[6]  John H. Harding,et al.  A recommendation for the thermal conductivity of UO2 , 1989 .

[7]  R. A. Verrall,et al.  Thermal conductivity of hyperstoichiometric SIMFUEL , 1995 .

[8]  Kevin R Robb,et al.  Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions , 2014 .

[9]  C. Ronchi,et al.  Thermal Conductivity of Uranium Dioxide up to 2900 K from Simultaneous Measurement of the Heat Capacity and Thermal Diffusivity. , 1999 .

[10]  J. Biersack,et al.  A Monte Carlo computer program for the transport of energetic ions in amorphous targets , 1980 .

[11]  Hansoo Lee,et al.  Computer-assisted design and experimental validation of multielectrode electrorefiner for spent nuclear fuel treatment using a tertiary model , 2013 .

[12]  Christian Duriez,et al.  Thermal conductivity of hypostoichiometric low Pu content (U,Pu)O2−x mixed oxide , 2000 .

[13]  T. Burchell.  Radiation Effects in Graphite , 2020, Comprehensive Nuclear Materials.

[14]  Juan J. Carbajo,et al.  A review of the thermophysical properties of MOX and UO , 2001 .

[15]  J. H. W. Simmons,et al.  RADIATION DAMAGE IN GRAPHITE , 1965 .

[16]  L. Snead.  Accumulation of thermal resistance in neutron irradiated graphite materials , 2008 .

[17]  A. Kohyama,et al.  Graphite, Ceramics, and Ceramic Composites for High-Temperature Nuclear Power Systems , 2009 .

[18]  B. T. Kelly,et al.  Physics of Graphite , 1981 .

[19]  G. Padmakumar,et al.  Development of active magnetic bearings and ferrofluid seals toward oil free sodium pumps , 2013 .

[20]  José Sánchez,et al.  Theoretical considerations about a membrane process for helium purification in multichannel monoliths for high temperature nuclear reactors , 2012 .

[21]  Y. Philipponneau,et al.  Thermal conductivity of (U, Pu)O2−x mixed oxide fuel , 1992 .

[22]  D. G. Martin.  A re-appraisal of the thermal conductivity of UO2 and mixed (U, Pu) oxide fuels , 1982 .

[23]  Chinthaka M. Silva,et al.  Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments☆ , 2014 .

[24]  M. Straumanis.  Absorption Correction in Precision Determination of Lattice Parameters , 1959 .

[25]  L. W. Deitrich.  Experiments on transient fuel failure mechanisms: selected ANL programs , 1975 .

[26]  Steven J. Zinkle,et al.  Accident tolerant fuels for LWRs: A perspective , 2014 .

[27]  Carlo Fiorina,et al.  Dimensional Effects in the Modelling of MSR Dynamics: Moving on from Simplified Schemes of Analysis to a Multi-Physics Modelling Approach , 2012 .

[28]  K. Terrani,et al.  Scoping assessments of ATF impact on late-stage accident progression including molten core–concrete interaction , 2014 .

[29]  J. S. Bell,et al.  CANDU fuel bundle deformation modelling with COMSOL multiphysics , 2012 .

[30]  Y. J. Kim,et al.  High temperature oxidation of molybdenum in water vapor environments , 2014 .

[31]  S. Nakagawa,et al.  Development of an Evaluation Model for the Thermal Annealing Effect on Thermal Conductivity of IG-110 Graphite for High-Temperature Gas-Cooled Reactors , 2009 .

[32]  B. Kelly.  The effect of defects on the basal plane thermal conductivity of a graphite crystal , 1967 .

[33]  J. Ziegler.  The stopping and range of ions in solids vol 1 : The stopping and ranges of ions in matter , 2013 .

[34]  Daniel M. Wachs.  Transient Testing of Nuclear Fuels and Materials in the United States , 2012 .

[35]  Steven J. Zinkle,et al.  Materials Challenges in Nuclear Energy , 2013 .

[36]  H. Matzke,et al.  Modelling of UO2-based SIMFUEL thermal conductivity The effect of the burnup , 1994 .

[37]  H. Matzke,et al.  A Pragmatic Approach to Modelling Thermal Conductivity of Irradiated UO2 Fuel. Review and Recommendations , 1996 .

[38]  Valentino Di Marcello,et al.  Extension of the TRANSURANUS plutonium redistribution model for fast reactor performance analysis , 2012 .

[39]  A. Cammi,et al.  A multi-physics time-dependent model for the Lead Fast Reactor single-channel analysis , 2013 .

[40]  Timothy D. Burchell,et al.  Thermal conductivity degradation of graphites due to nuetron irradiation at low temperature , 1995 .