Separate effects irradiation testing of miniature fuel specimens

Abstract Qualification of new nuclear fuels is necessary for their deployment and requires a thorough understanding of fuel behavior under irradiation. Traditionally, nuclear fuels have been qualified by performing exhaustive integral tests under a limited range of prototypic conditions designed for their specific reactor application. While some integral fuel testing is essential, basic data on behavior and property evolution under irradiation can be obtained from separate effects tests. These irradiations could offer reduced cost, reduced complexity, and in the case of accelerated testing, reduced time to achieve a given burnup. Furthermore, it may be desirable to design test irradiations capable of deconvoluting the myriad effects of burnup, temperature gradients, and other factors inherent to integral irradiation tests. Oak Ridge National Laboratory has developed an experimental capability to perform separate effects irradiation testing of miniature fuel specimens in the High Flux Isotope Reactor (HFIR): the “MiniFuel” irradiation vehicle. The small size ( 1200 °C), compositions, enrichments, and even geometries without requiring detailed analyses for each fuel variant. This paper summarizes the experiment design concept, evaluates potential applications for specific fuel forms, and briefly describes the first set of experiments on uranium nitride kernels that have been assembled and are currently being irradiated in the HFIR.

[1]  J. Oxley,et al.  CERAMIC-COATED-PARTICLE NUCLEAR FUELS , 1964 .

[2]  Y. Katoh,et al.  Method for analyzing passive silicon carbide thermometry with a continuous dilatometer to determine irradiation temperature , 2016 .

[3]  Timm Preusser,et al.  Modeling of carbide fuel rods , 1982 .

[4]  R. Manzel,et al.  The radial distribution of fission gases and other fission products in irradiated PWR fuel , 1984 .

[5]  Joel Lee McDuffee Heat Transfer Through Small Moveable Gas Gaps in a Multi-Body System Using the ANSYS Finite Element Software , 2013 .

[6]  Mohamed S. El-Genk,et al.  Uranium nitride fuel swelling correlation , 1990 .

[7]  Kenneth L. Peddicord,et al.  Material property correlations for uranium mononitride: IV. Thermodynamic properties , 1990 .

[8]  Chinthaka M. Silva,et al.  Investigation of sol-gel feedstock additions and process variables on the density and microstructure of UN microspheres , 2019, Journal of Nuclear Materials.

[9]  Irradiation of Miniature Fuel Specimens in the High Flux Isotope Reactor , 2018 .

[10]  Boris Wilthan,et al.  Thermophysical Properties of Solid and Liquid Ti-6Al-4V (TA6V) Alloy , 2006 .

[11]  K. Terrani,et al.  Production of near‐full density uranium nitride microspheres with a hot isostatic press , 2018, Journal of the American Ceramic Society.

[12]  R. T. Primm,et al.  Modeling of the High Flux Isotope Reactor Cycle 400 , 2004 .

[13]  J. Snelgrove,et al.  Irradiation behaviour of uranium silicide compounds , 2004 .

[14]  D. Prieur,et al.  Melting point determination of uranium nitride and uranium plutonium nitride: A laser heating study , 2014 .

[15]  Y. Katoh,et al.  Handbook of SiC properties for fuel performance modeling , 2007 .

[16]  Nicholas R. Brown,et al.  Neutronic performance of uranium nitride composite fuels in a PWR , 2014 .

[17]  Gregory K. Miller,et al.  The challenges associated with high burnup, high temperature and accelerated irradiation for TRISO-coated particle fuel , 2007 .

[18]  J. E. Palentine The development of silicon carbide as a routine irradiation temperature monitor, and its calibration in a thermal reactor , 1976 .

[19]  David A. Petti,et al.  An approach to fuel development and qualification , 2007 .

[20]  Kenneth L. Peddicord,et al.  Material property correlations for uranium mononitride , 1990 .

[21]  B.R.T. Frost,et al.  The carbides of uranium , 1963 .

[22]  B. Betzler,et al.  Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading , 2016 .

[23]  H. Matzke,et al.  Oxygen potential measurements in high burnup LWR U02 fuel , 1995 .

[24]  K. M. Chidester,et al.  Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test) , 1993 .

[25]  R. Bruce Matthews,et al.  Uranium-plutonium carbide fuel for fast breeder reactors , 1983 .

[26]  Stewart L Voit,et al.  Quantification of process variables for carbothermic synthesis of UC 1-x N x fuel microspheres , 2017 .

[27]  O. Bersillon,et al.  Fission Product Yields of 233 U, 235 U, 238 U and 239 Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons , 2010 .

[28]  R. L. Petty,et al.  Fabrication and testing of uranium nitride fuel for space power reactors , 1988 .

[29]  Kurt A. Terrani,et al.  Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions , 2014 .

[30]  L. Snead,et al.  Microencapsulated fuel technology for commercial light water and advanced reactor application , 2012 .

[31]  A. A. Proshkin,et al.  Mononitride Fuel for Fast Reactors , 2003 .

[32]  C. Walker Measurement of retained xenon in advanced fuels by microprobe analysis , 1979 .

[33]  K. Mcclellan,et al.  Thermophysical properties of U3Si2 to 1773K , 2015 .

[34]  Y. Katoh,et al.  Dimensional stability and anisotropy of SiC and SiC-based composites in transition swelling regime , 2018 .

[35]  John Saurwein,et al.  The DOE advanced gas reactor fuel development and qualification program , 2005 .

[36]  A. Nelson,et al.  Fabrication and thermophysical property characterization of UN/U3Si2 composite fuel forms , 2017 .

[37]  K. Mcclellan,et al.  Corrigendum to “Thermophysical properties of U3Si2 to 1773 K” [J. Nucl. Mater. 464 (2015) 275–280] , 2016 .

[38]  Gregory K. Miller,et al.  Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance , 2003 .

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

[40]  Kurt A. Terrani,et al.  Fabrication and characterization of fully ceramic microencapsulated fuels , 2012 .

[41]  E. Storms An equation which describes fission gas release from UN reactor fuel , 1988 .

[42]  L. Borst,et al.  Nuclear Engineering Handbook , 1958 .

[43]  B. Lewis Fission product release from nuclear fuel by recoil and knockout , 1987 .

[44]  G. Subhash,et al.  Enhanced thermal conductivity of uranium dioxide–silicon carbide composite fuel pellets prepared by Spark Plasma Sintering (SPS) , 2013 .

[45]  Mohamed S. El-Genk,et al.  Thermal conductivity correlation for uranium nitride fuel between 10 and 1923 K , 1988 .

[46]  C. Parish,et al.  Restructuring in high burnup UO2 studied using modern electron microscopy , 2018, Journal of Nuclear Materials.

[47]  Paul A. Lessing,et al.  Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation , 2015 .

[48]  Alvin M. Weinberg,et al.  Molten Fluorides as Power Reactor Fuels , 1957 .

[49]  Douglas C. Crawford,et al.  Fuels for sodium-cooled fast reactors: US perspective , 2007 .

[50]  G. Vasudevamurthy,et al.  Production of High-Density Uranium Carbide Compacts for Use in Composite Nuclear Fuels , 2008 .

[51]  Yang-Hyun Koo,et al.  Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea , 2016 .

[52]  F. Ingold,et al.  Nitrides as a nuclear fuel option , 2005 .

[53]  Donald R. Olander,et al.  Nuclear fuels - Present and future , 2009 .

[54]  Y. Rhee,et al.  Fabrication of micro-cell UO2–Mo pellet with enhanced thermal conductivity , 2015 .