Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO₂ fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor k_inf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO₂ fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO₂, UO₂ with 4.0 vol.% BeO, and UO₂ with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO₂ fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the k_inf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO₂.     

Radial redistribution of actinides in irradiated FR-MOX fuels

The redistributions of neptunium, plutonium and americium during two kinds of short-term irradiation tests for 10 min and 24 h at high linear heating rate around 430 W cm⁻¹ were studied in the uranium and plutonium mixed oxide fuel containing Am and/or Np. It was found in the irradiation test for 24 h that the concentrations of Pu and Am increased toward the central void, but there was no change in the concentration of Np. The obtained experimental redistributions of Am and Pu were analyzed, based on both pore migration and thermal diffusion models. As a result, the calculated redistributions of Pu and Am showed good agreements with the experimentally obtained ones.     

Behavior of Si impurity in Np-Am-MOX fuel irradiated in the experimental fast reactor Joyo

The irradiation behavior of uranium-plutonium mixed oxide fuels containing a large amount of silicon impurity was examined by post-irradiation examination. Influences of Si impurity on fuel restructuring and cladding attack were investigated in detail. Si impurity, along with Am, Pu and O were transported by spherical pores and cylindrical tubular pores to the fuel center during fuel restructuring of the Np-Am-MOX fuel, where a eutectic reaction of fuel and Si-rich inclusions occurred. After fuel restructuring of the Np-Am-MOX fuel, Si-rich inclusions without fuel constituents were agglomerated at fuel crack openings where shallow attacks on the inner wall of the cladding were seen. Such shallow attacks on the inner wall of the cladding were likewise observed near the location of fuel cracks in long-term steady-state irradiated MOX fuels. Evidence of these shallow attacks on the inner wall of the cladding remained after fuel restructuring in normal MOX fuel. However, grain boundary corrosion of the cladding inner wall at the opening of the fuel cracks was selective and was marked in MOX fuel at higher oxygen potential by the release of reactive fission products such as Cs and Te in comparison with other regions of cladding wall.     

Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor

In order to investigate the effect of americium addition in MOX fuel on the irradiation behavior, the ‘Am-1’ program is being conducted in the experimental fast reactor Joyo. The Am-1 program consists of two short-term irradiation tests of 10 min and 24 h irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. This paper reports on the results of PIEs for Am-containing MOX fuel irradiated for 10 min. MOX fuel pellets containing 3% or 5% Am were fabricated in a shielded air-tight hot cell using a remote handling technique. The oxygen to metal ratio (O/M) of these fuel pellets was 1.98. They were irradiated at peak linear heating rate of about 43 kW m⁻¹. Focus was being placed on migration behavior of Am during the irradiation. The ceramography results showed that structural changes such as lenticular pores and a central void occurred early, within the brief 10 min of irradiation. The results of electron probe microanalysis revealed that the concentration of Am increased in the vicinity of the central void.     

Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor

A fuel irradiation program is being conducted using the experimental fast reactor ‘Joyo’. Two short-term irradiation tests in the program were completed in 2006 using a uranium and plutonium mixed oxide fuel which contains minor actinides (MA-MOX fuel). The objective of the tests is the investigation of early thermal behavior of MA-MOX fuel such as fuel restructuring and redistribution of minor actinides. Three fuel pins which contained MA-MOX: 2% neptunium and 2% americium doped uranium plutonium mixed oxide (Am,Pu,Np,U)O_2−x fuel were supplied for testing. The first test was conducted with high-linear heating rate of approximately 430 W cm⁻¹ for only 10 min. After the first test, one fuel pin was removed for examinations. Then the second test was conducted with the remaining two pins at nearly the same linear power for 24 h. In these tests, two oxygen-to-metal molar ratios were used for fuel pellets as a test parameter. Non-destructive and destructive post-irradiation examinations results are discussed with early on the behavior of the fuel during irradiation.     

Oxygen potential of hypo-stoichiometric Lu-doped UO2

Oxygen potentials of hypo-stoichiometric Lu-doped UO₂, (U_0.80Lu_0.20)O_2−x, were experimentally investigated by thermogravimetric analysis using H₂O/H₂ gas equilibria at 1173, 1273 and 1473 K. The oxygen potentials of (U,Lu)O_2−x were higher than those of other forms of rare earth-doped UO₂, specifically (U,Nd)O_2−x, (U,Gd)O_2−x, and (U,Er)O_2−x. Slope analyses for plots of oxygen potential versus deviation from stoichiometry indicated that (U_0.80Lu_0.20)O_2−x had a similar defect structure to that of the other forms of rare earth-doped UO₂. A relationship between the effective ionic radii and oxygen potentials was found for the hypo-stoichiometric rare earth-doped UO₂.     

Recovery and restructuring induced by fission energy ions in high burnup nuclear fuel

In light water commercial reactors, extensive change of grain structure was found at high burnup ceramic fuels. The mechanism is driven by bombardment of fission energy fragments and studies were conducted by combining accelerator based experiments and computer-science. Specimen of CeO₂ was used as simulation material of fuel ceramics. With swift heavy ion (Xe) irradiation on CeO₂, with 210 MeV, change of valence charge and lattice deviation of cations were observed by XPS and XRD. Combined irradiations of Xe implantation and swift heavy ion irradiation successfully produced sub-micrometer sized sub-grains, similar as that observed in commercial fuels. Studying components of mechanism scenarios, with first principle calculations using the VASP code, we found stable hyper-stoichiometric defect structures of UO_2+x. Molecular dynamics studies revealed stability of Xe planar defects and also found rapid transport mode of oxygen-vacancy clusters.     

A method for the quantification of total xenon concentration in irradiated nuclear fuel with SIMS and EPMA

A new method for the quantitative determination of the total xenon concentration in irradiated nuclear fuel is presented. The SIMS measurement of xenon enables the detection of the gas filling bubbles which are not detected with EPMA. The quantification is achieved using the EPMA data as reference at position where no or nearly no bubbles are detected. A new approach using the complementary information given by EPMA, SEM and SIMS is proposed, it opens new horizons for the characterisation of fission gases in irradiated nuclear fuel.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Oxygen potentials of mixed oxide fuels for fast reactors

Oxygen potentials of homogenous (Pu_0.2U_0.8)O_2−x and (Am_0.02Pu_0.30Np_0.02U_0.66)O_2−x which have been developed as fuels for fast breeder reactors were measured at temperatures of 1473-1623 K by a gas equilibrium method using an (Ar, H₂, H₂O) gas mixture. The measured oxygen potentials of (Pu_0.2U_0.8)O_2−x were about 25 kJ mol⁻¹ lower than those of (Pu_0.3U_0.7)O_2−x measured previously and were consistent with the values calculated by Besmann and Lindemer’s model. The measured oxygen potentials of (Am_0.02Pu_0.30Np_0.02U_0.66)O_2−x were slightly higher than those of MOX without minor actinides. No fuel-cladding chemical interaction is affected significantly by adding their minor actinides.     

Alloying and microstructure stability in the high-temperature Mo-Si-B system

Multi-phase alloys in the Mo-Si-B system are identified as high-temperature structural materials due to their high melting points (above 2000 °C) and excellent oxidation resistance attributed to the self-healing characteristics of borosilica layer up to 1400 °C. In the current study, the effect of alloying additions to achieve a reduced weight density has been examined in terms of changes in the microstructure and phase stability. The critical factor underlying the microstructural changes is related to the influence of the alloying additions on the stability of the high melting temperature ternary-based Mo₅SiB₂ (T₂) borosilicide phase.     

Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO₂ as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO₂ fuel particles, use of oxide stabilizers in the UO₂, minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet.     

Optimization of resistance spot welding on the assembly of refractory alloy 50Mo-50Re thin sheet

Resistance spot welding (RSW) was employed to pre-join refractory alloy 50Mo-50Re (wt%) sheet with a 0.127 mm gage. Five important welding parameters (hold time, electrode, ramp time, weld current and electrode force) were adjusted in an attempt to optimize the welding quality. It was found that increasing the hold time from 50 ms to 999 ms improved the weld strength. Use of rod-shaped electrodes produced symmetric nugget and enhanced the weld strength. Use of a ramp time of 8 ms minimized electrode sticking and molten metal expulsion. The weld strength continuously increased with increasing the weld current up to 1100 A, but the probabilities of occurrence of electrode sticking and molten metal expulsion were also increased. Electrode force was increased from 4.44 N to 17.8 N, in order to reduce the inconsistency of the welding quality. Welding defects including porosities, columnar grains and composition segregation were also studied.     

Microstructural and mechanical property changes in the Ta-base T-111 alloy following thermal aging

The microstructural changes occurring in the Ta-base T-111 (Ta-8W-2Hf) alloy during 1100 h thermal aging at 1098, 1248 and 1398 K under inert atmosphere and the influence on mechanical properties are reported. Electrical resistivity, hardness and tensile properties are compared between the as-annealed and aged conditions. Microstructural evaluations were performed by optical, scanning electron microscopy and transmission electron microscopy. An increase in the amount of grain boundary precipitation with increasing aging temperature was found to decrease the electrical resistivity and material strength. Precipitation at the grain boundaries was found to be a mixture of monoclinic and cubic structures, suggesting the development of mixed Hf oxides, carbides and nitrides. Precipitate development caused pronounced embrittlement of the alloy following aging at 1398 K.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

Increase in pyrolytic carbon optical anisotropy and density during processing of coated particle fuel due to heat treatment

Measurements of pyrolytic carbon optical anisotropy and density have been made on a series of tri-isotropic (TRISO) coated particles prepared for the United States Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification (AGR) program. These measurements show the effect of varying the deposition conditions, especially the deposition temperature, on the density and optical anisotropy of the carbon layers. Additional heat treatment studies of the coated particles at various stages illustrate the strong effect of post-deposition thermal processing on these two pyrolytic carbon properties. Such post-deposition heat treatment occurs during SiC deposition and fuel compact firing, resulting in increased anisotropy and density of the pyrolytic carbon layers.     

Equilibrium and nonequilibrium molecular dynamics simulations of heat conduction in uranium oxide and mixed uranium-plutonium oxide

The thermal conductivity of nuclear fuels such as UO_2+x and (U,Pu)O_2−x has been calculated by the molecular dynamics (MD) simulation in terms of oxygen stoichiometric parameter x, temperature and Pu content. In the present study, the MD calculations were carried out in both equilibrium (EMD) and nonequilibrium (NEMD) systems. In the EMD simulation, the thermal conductivity was defined as the time-integral of the correlation function of heat fluxes according to the Green-Kubo relationship. Meanwhile, in the homogeneous NEMD, it was given by the ratio of the time-averaged heat flux to the perturbed external force subjected to each particle in the simulated cell. NEMD, as compared with EMD, gave somewhat precise results efficiently. Furthermore, both MD calculations showed that the thermal conductivity of these oxide fuels decreased with increase of temperature and defects, i.e. excess oxygen or vacancy, and was rather insensitive to Pu content for the stoichiometric fuel.     

Analytical investigations of irradiated inert matrix fuel

Yttria stabilised zirconia (YSZ) inert matrix fuel (IMF) fabricated at PSI and irradiated 3 years in the Halden Material Test Reactor (HBWR) since 2000, has been examined by Electron Probe Microanalysis (EPMA) and Secondary Ion Mass Spectroscopy (SIMS) after irradiation and compared with data gained for the unirradiated material. The examined pellet cross-section was estimated to have an equivalent burn-up of 22 MW d kg⁻¹. EPMA measurements demonstrate that the burn-up was rather flat over more than the half pellet radius. A Pu consumption of about 2.5 wt% has been measured with a higher rate in the fuel border zone. The high fuel temperature is responsible for a certain homogenisation of the mineral phases in the fuel centre region whereas the border zone has remained rather with an as-fabricated phase distribution. The central part was also characterised by a dense porosity distribution as well as a temperature and relocation driven depletion of the volatile fission products Xe and Cs. In addition, SIMS has been realised on the same specimen in order to determine the semi-quantitative distribution of different isotopes in the pellet.     

Optical anisotropy measurements of TRISO nuclear fuel particle cross-sections: The method

The analysis of two-modulator generalized ellipsometry microscope (2-MGEM) data to extract information on the optical anisotropy of coated particle fuel layers is discussed. Using a high resolution modification to the 2-MGEM, it is possible to obtain generalized ellipsometry images of coating layer cross-sections with a pixel size of 2.5 μm and an optical resolution of ∼4 μm. The most important parameter that can be extracted from these ellipsometry images is the diattenuation, which can be directly related to the optical anisotropy factor (OAF or OPTAF) used in previous characterization studies of tristructural isotropic (TRISO) coated particles. Because high resolution images can be obtained, the data for each coating layer contains >6000 points, allowing considerable statistical analysis. This analysis has revealed that the diattenuation of the inner pyrocarbon (IPyC) and outer pyrocarbon (OPyC) coatings varies significantly throughout the layer. The 2-MGEM data can also be used to determine the principal axis angle of the pyrocarbon layers, which is nearly perpendicular to the TRISO radius (i.e., growth direction) and corresponds to the average orientation of the graphene planes.     

Evaluation of melting point of UO2 by molecular dynamics simulation

The melting point of UO₂ has been evaluated by molecular dynamics simulation (MD) in terms of interatomic potential, pressure and Schottky defect concentration. The Born-Mayer-Huggins potentials with or without a Morse potential were explored in the present study. Two-phase simulation whose supercell at the initial state consisted of solid and liquid phases gave the melting point comparable to the experimental data using the potential proposed by Yakub. The heat of fusion was determined by the difference in enthalpy at the melting point. In addition, MD calculations showed that the melting point increased with pressure applied to the system. Thus, the Clausius-Clapeyron equation was verified. Furthermore, MD calculations clarified that an addition of Schottky defects, which generated the local disorder in the UO₂ crystal, lowered the melting point.