Creep behavior of ceramic nuclear fuels under neutron irradiation

A theoretical estimation of the irradiation-induced creep rate of UO₂ resulted in a creep rate range between about 6 × 10⁻⁶/h and 8 × 10⁻⁵/h for a fission rate of 1 × 10¹⁴ f/cm³·s and a stress of 2 kgf/mm². It is essentially due to the “thermal rods” along the fission fragment tracks. Therefore, creep rates should only weakly depend on temperature (below 1000–1200°C) and must be markedly lower for carbide and nitride fuel.     

Observations on the low-temperature creep properties of uranium carbide under irradiation

The creep behaviour of 97% dense hyperstoichiometric UC has been examined during irradiation in three-point bend tests carried out at 450°C up to a dose of $\text{1.65 × 10}{}^{26}\text{fissions/m}{}^3$. A rapid decrease in measured strain rate with dose was observed at each stress level, nominally steady-state creep being established above $\text{≈ 1 × 10}{}^{26}\text{fissions/m}{}^3$ when the creep rate was a factor of 8 lower than that observed in UO₂ irradiated under identical conditions. Creep rates were found to be directly proportional to stress at high doses. Comparison of results from this test with data from other experiments up to $\text{2 × 10}{}^{25}\text{fissions/m}{}^3$ in compression and tension indicates little variation in the radiation-creep constant between 450°C and 800°C. The creep rate for UC, much lower than that observed in UO₂, is consistent with recently reported determinations of the effective uranium self-diffusion coefficients under irradiation in those materials.     

Elastic,diffusional, and mechanical properties of carbide and nitride nuclear fuels — a review

The elastic, diffusional, and mechanical-properties data for carbide and nitride nuclear fuels, including UC, UN, (U, Pu)C, (U, Pu)N, and some mixed systems are summarized and critically reviewed. The review was written in an attempt to unify our understanding of these important materials as well as to provide an objective source of information for fuel-element designers. Important phenomena that have not been explored or are inadequately understood are cited, and recommendations for additional studies are made.     

Final-stage densification of mixed oxide fuels

The final-stage densification of hypostoichiometric (U, Pu)O_2−x has been studied using a double-action punch and die, in the temperature range of 1325 to 1550°C between stresses of 7.6 and 76 MPa. At low stresses the densification rate is nearly proportional to stress and has an activation energy comparable to that measured for steady-state creep. The stress exponent increases as stress is increased. The data compare tolerably well with the predictions of theoretical pressures-intering maps once grain growth (which was found to occur at the higher temperatures) is accounted for. The controlling mechanism throughout the experiments is shown to be lattice diffusion.     

Swelling under irradiation of MgO pellets containing americium oxide: The ECRIX-H irradiation experiment

The ECRIX-H irradiation experiment studied the behaviour of pellets containing americium dispersed in MgO. The purpose of the irradiation was to demonstrate the capacity of magnesia to provide an efficient support matrix. After fabrication, the sintered pellets contained 16.65 wt.% of Am microdispersed in the inert matrix. The ECRIX-H pellets were irradiated under a locally moderated neutron flux in the Phénix sodium-cooled fast reactor (SFR) for 318 Effective Full Power Days (EFPD). Post-test calculations indicated that the fission and transmutation rates of americium at the maximum flux plane reached 33.9% and 92.6% respectively at the end of the irradiation phase. The results of the post-irradiation examinations - both non-destructive and destructive - are discussed in this paper. These results indicate a satisfactory behaviour of the MgO matrix. Particularly, a moderate swelling occurs in the pellets under irradiation even with significant quantities of helium generated and at high transmutation rate.     

Post-irradiation examination of uranium-doped rock-like oxide fuels

Post-irradiation examinations of rock-like oxide fuels were performed in JAERI to evaluate irradiation behavior and geochemical stability. Five kinds of fuels were prepared using 20% enriched U instead of Pu; a single-phase fuel of an yttria-stabilized zirconia containing UO₂ (U-YSZ), two particle-dispersed type fuels of U-YSZ particles + MgAl₂O₄/Al₂O₃ powder, two homogeneously blended type fuels of U-YSZ powder + MgAl₂O₄/Al₂O₃ powder. The fuels were irradiated in JRR-3 for about 100 days and estimated irradiation conditions were as follows; linear power was 15 kW m⁻¹, thermal neutron fluence was 7 x 10²⁴ m⁻² and fuel temperatures at the surface were 740–1130 K. From the results of non-destructive examinations, the stainless steel cladding surfaces were partially discolored by oxidation and no remarkable deformation of the pins was observed. Significant pellet fragmentation was not observed in spite of the crack formation as observed in irradiated LWR UO₂ fuels. Nonvolatile FPs were observed only in pellets but volatile Cs moved partly to the plenum. From these examinations, no significant difference in macroscopic irradiation behavior was distinguished among 5 fuels.     

Computer modeling of steady state fission gas behavior in carbide fuels

A model for the simulation of long-term, steady-state fission gas behavior in carbide fuels is formulated. It is assumed that fission gas release occurs entirely through gas atom diffusion to grain boundaries and cracks. Fission gas bubbles are assumed to remain stationary and to grow as the net result of gas atom precipitation into the bubbles from the matrix solid and gas atom re-solution from the bubbles into the matrix. Furthermore, assuming that local gas atom redistribution process in the immediate neighborhood of a bubble is very rapid, the bubble size is assumed to correspond to the equilibrium size that maintains exact balance between the rate of gas atom re-solution and that of gas atom precipitation.The model also treats the effect of attachment between bubbles and second-phase precipitates; the experimentally observed faster growth rate of precipitate bubbles is simulated using a reduced re-solution parameter for precipitate bubbles. With the grain matrix assumed to be spherical, the model allows the computation of the radial distribution of the intragranular bubbles and the gas atom concentration in the matrix.The flux of gas atoms arriving at the grain boundary is computed. The continual growth of grain boundary bubbles, resulting from the accumulation of gas atoms on the grain boundary, leads to grain boundary interlinkage and all gas atoms that subsequently reach the grain boundary are assumed to be released. Similarly, all gas atoms generated following the interlinkage of intragranular bubbles are also assumed to be immediately released.Application of the model indicates that fission gas swelling is largely due to intragranular bubbles. Grain boundary bubbles, although very large in size, contribute little to fission gas swelling and the contribution from gas atoms in solid solution in the matrix is even less significant.Physical parameters entering the model were assigned numerical values that closely represent the physical characteristics of the irradiation samples. Careful comparisons between the results of sensitivity studies and the experimental data readily identify the re-solution parameter to have the strongest influence on the results predicted by the code and that the grain size, and not the temperature, is the dominant factor affecting gas release.When allowance is made for the uncertainties of the experimental data, the predicted fission gas swelling also correlates well with experiment. The spread in the fuel swelling data, however, indicates that fuel cracking, and not fission gas swelling alone, very often contributes significantly to the fuel external dimensional changes. The linear fission gas swelling rate prediceted by the model exhibits almost a linear variation with temperature. This result correlates well with the linear swelling rate obtained from experimental swelling data if immersion density data alone are used, in order to eliminate the sources of uncertainties associated with fuel cracking.     

Fuel performance evaluation of rock-like oxide fuels

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ–spinel–corundum system, the single-phase fuels of YSZ, the particle-dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation testings. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO₂ fuel at the similar fuel temperatures. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.     

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.     

Ion irradiation effects on the optical properties of tungsten oxide films

The optical changes in amorphous WO₃ film prepared by reactive RF sputtering and irradiated by 200-800 keV oxygen ions were measured to study the relationship between coloration and energy deposition. The color centers were effectively created by ion irradiation with contributions from nuclear collisions and electronic energy loss. The increase in the absorption coefficient was reasonably explained by a first order reaction, whose production rate depended roughly on the total deposited energy. During heat treatment in air atmosphere, transmittance recovery started at 400 K and completed at 550 K. No significant difference was found among films irradiated by different incident energies; therefore indicating that the ion-induced damage structure is not strongly influenced by the type of energy loss.     

Creep anisotropy of Zircaloy-2 cladding during irradiation

A procedure is described to determine the axial-radial and circumferential-radial contractile strain ratios (the R and P factors in the Backofen modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 of boiling water reactor fuel rods, tie rods and water rods. Values for R and P have been determined for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding. A sensitivity study was performed to evaluate the effects of measurement uncertainties on the derived values of R and P and on the engineering application of the model to predict the in-reactor deformation of CWSR Zircaloy-2 cladding.     

PN结型器件在氚钛片辐照下电输出性能

用多片具有不同金属钛膜厚度和充氚量的氚钛片对两种单晶硅基PN结型器件进行了辐照,在线测量了它们的电输出性能并进行了定性分析.结果表明,在本文所采用的钛膜厚度和氚量级下,器件输出短路电流等随充氚量增加而小幅增大,但不成正比关系;器件的掺杂浓度、结深等结构参数对器件电输出性能影响较大.