Recommendations for calculating the characteristics of thermal creep of mixed uranium–plutonium oxide fuel when analyzing fuel-element serviceability

Recommendations for calculating the characteristics of thermal creep of mixed uranium–plutonium oxide fuel when analyzing the serviceability of fuel elements are developed on the basis of a physical model of deformation. The deformation processes in the model include diffusion and diffusion-controlled motion of dislocations. It is shown on the basis of an analysis of the thermodynamics of point defects in ionic crystals that the diffusion of ions in the cationic sublattice is controlled by the vacancy and displacement mechanisms and the coefficient of diffusion depends on the temperature and the oxygen coefficient. The model takes account of the effect of temperature, stress, fuel density, plutonium content, grain size, and oxygen coefficient on the creep rate. The application of physical ideas to obtain computational relations made it possible to improve by more than a factor of 10 the agreement between the calculations and the experimental data as compared with the previously used empirical relations to describe the characteristics of creep.     

Preparation and characterization of the simulated burnup americium-containing uranium–plutonium mixed oxide fuel

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium–plutonium mixed oxide (MOX) fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements and melting temperature measurement. Elastic moduli for the simulated low decontamination MOX fuel were almost the same level as fuel without americium and fission products and decrease in the moduli was slight with increasing simulated burn-up. The melting temperature of high burn-up, low decontamination MOX fuel may be estimated by using the findings on the effect of americium, plutonium addition and fission products accumulation.     

Supercritical-pressure water cooled thermal reactor with uranium–plutonium–thorium fuel cycle and bidirectional coolant flow articles

Computational results obtained for fuel assemblies and a core with uni- and bidirectional coolant flow are presented for a water-cooled reactor with a thermal neutron spectrum. It is shown that a bidirectional scheme for cooling fuel assemblies has advantages over a unidirectional scheme and holds promise for Gen IV water-cooled reactors with supercritical coolant pressure, which make it possible to perfect the technology for closing and drawing thorium into the fuel cycle.     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

A BWR fuel assembly design for efficient use of plutonium in thorium–plutonium fuel

The objective of this study is to develop an optimized BWR fuel assembly design for thorium–plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel.     

Effects of molybdenum on microstructural evolution and mechanical properties in Zr–Nb alloys as nuclear fuel cladding materials

The Zr–Nb alloys were modified by doping of Mo as a minor alloying element to seek for the nuclear fuel cladding materials with better characteristics. The effects of Mo on microstructural evolution and mechanical properties in Zr–Nb alloys were systematically investigated and elucidated. Results showed that the martensitic microstructure, a mixture of lath martensites and lens martensites with internal twins, was observed in the alloys quenched from β-phase. Width of the lath martensite reduced with the increasing Mo concentration, and the volume fraction of lens martensite increased with increase in the Mo concentration. After final annealing, a new kind of precipitate, namely β-(Nb, Mo, Zr), was identified in the Mo-containing alloys. It was also found that Mo reduced the growth of the precipitates but increased their number density. Furthermore, Mo addition retarded the recrystallization process strongly and reduced the grain size significantly. In terms of the mechanical properties, Mo addition enhanced the yield strength and the ultimate tensile strength at room temperature, however decreased the ductility. The grain size strengthening was presumed as the greatest contributor in this system.     

Compatibility of erbium oxide coating with liquid lithium–lead alloy and corrosion protection effect of iron layer

Li–Pb compatibility of Er₂O₃ and Er₂O₃-Fe two-layer coatings has been explored for an understanding of corrosion behaviors and effects of the protection layer. The coatings were peeled off after static Li–Pb immersion test at 600 °C due to the degradation of adhesion between the coating–substrate interface. A loss of Er and then subsequent corrosion of Er₂O₃ were shown after immersion at 500 °C for 500 and 1505 h. However, the outer Fe layer played a role to decrease corrosion rate of the coatings by comparing with the results of Er₂O₃ single layer coatings. Deuterium permeation measurements after corrosion tests at 500 °C showed that the Er₂O₃ coatings kept permeation reduction factors of 10²–10³ after 500 h immersion, but seriously degraded after 1505 h immersion. Corrosion mechanisms suggest that corrosion protection properties will be modified by an optimization of the outer Fe layer and a control of oxygen concentration in Li–Pb.     

Thermodynamic analysis of uranium–gadolinium fuel stability at high temperatures

Using the most accurate measurements of the liquidus temperature in the UO₂–Gd₂O₃ system up to 30 mol.% of Gd₂O₃, thermodynamic models of the melt and cubic solution GdO_1.5 in UO₂ are constructed. The equilibrium phase diagram of the system UO₂–GdO_1.5 in the interval 1900–3200 K is calculated in the entire composition range and the metastable diagram is calculated assuming that no cubic solid solutions are formed. The upper and lower boundaries of the melting onset temperature (solidus) of uraniumgadolinium fuel are presented. The phase composition of the pellets made from such fuel and, ultimately, the technology determine the melting onset temperature uniquely.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Fabrication of SiC–SiC composites for fuel cladding in advanced reactor designs

Both advanced fission reactor concepts and fusion energy systems demand materials that can survive extremely harsh operating environments having persistent high temperature and high neutron flux conditions. Silicon carbide fiber/silicon carbide matrix (SiC–SiC) composites have shown promise for these applications, which include fuel cladding and reactor structural components. However, the composite fabrication process is time consuming and the fabrication of complicated geometries can be difficult.In this work, SiC–SiC and carbon fiber–SiC composite samples were fabricated using chemical vapor infiltration (CVI), and the mechanical and thermal properties of samples with a range of densities and total infiltration times were characterized and compared. Both sample density and the reinforcing fiber material were found to have a very significant influence on the composite mechanical and thermal material properties. In particular, internal porosity is found to have a significant effect on the mechanical response, as can be observed in the crack propagation in low density samples. In order to better understand the densification of the composites, a computer model is being developed to simulate the diffusion of reactants through the fiber preform, and SiC deposition on the fiber surfaces. Preliminary modeling has been correlated with experimental results and shows promising results.     

Nuclear and thermal–hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.     

Measurement and analysis of fission rates in depleted uranium and depleted uranium–polyethylene shells with neutrons from D-T reaction

To validate the concept design of a novel fusion–fission hybrid energy reactor, a depleted uranium assembly and a combined assembly of uranium and polyethylene were designed and assembled based on a depleted uranium spherical shell and a polyethylene spherical shell. The distribution of the fission rates for the depleted uranium and enriched uranium in the two assemblies, as a function of the distance of the detection position to the centre, was measured using a plate fission chamber bombarded by D-T neutrons. The addition of a polyethylene shell significantly changed the neutron spectrum; in particular, the neutron fluxes with energies of 1 MeV and lower were changed. Using MCNP5 and the attached libraries, the fission rate experiments were simulated, and the experimental configuration, including the wall of the experimental hall, was described in detail in the model. The fission rate distributions for depleted uranium and enriched uranium in the two assemblies were reproducible. The difference between the calculated results with different libraries and different tallies is as small as 1.0%. By considering the neutron flux, the fission rate and the C/E values, it is concluded that the fission rates of depleted uranium and enriched uranium induced by the fast neutrons were overestimated, and it is proposed that the fission parameters of uranium for fast neutrons should be re-evaluated, or the margin of the concept design should be enlarged, to make the concept effective.     

Kinetics of chlorination of uranium–antimony composite oxide for uranium removal from waste catalyst used for acrylonitrile synthesis

For uranium removal from waste catalyst used for acrylonitrile synthesis, kinetics of chlorination of uranium–antimony composite oxide was studied. During the chlorination treatment with hydrogen chloride gas at a partial pressure of 0.6–6.7 kPa and 873–1173 K, the uranium–antimony composite oxide, USb₃O_10, which was contained in the waste catalyst converted to another composite oxide, USbO₅, then changed to uranium oxide. Both reaction rates of the conversions, from USb₃O₁₀ to USbO₅ and from USbO₅ to U₃O₈, were described by a first order function of the fraction of USb₃O₁₀ and USbO₅, and their activation energies under the condition at 1.0 kPa hydrogen chloride gas were almost same values at (8.0 ± 0.4) × 10⁴ J mol^?1.     

Oxidation kinetics of Zry-4 fuel cladding in mixed steam–air atmospheres at temperatures of 1273–1473 K

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam–air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0% up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation, applicable for severe accident computer codes of nuclear power reactors. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transition and post-transition regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transition regime.     

The chemical durability of glass and graphite–glass composite doped with cesium oxide

The role of temperature in determining the chemical stability of a waste form, as well as its leach rate, is very complex. This is because the dissolution kinetics is dependent both on temperature and possibility of different rate-controlling mechanisms that appear at different temperature regions. The chemical durability of Alumina-Borosilicate Glass (ABG) and Glass–Graphite Composite (GGC), bearing Tristructural Isotropic (TRISO) fuel particles impregnated with cesium oxide, were compared using a static leach test. The purpose of this study is to examine the chemical durability of glass–graphite composite to encapsulate coated fuel particles, and as a possible alternative for recycling of irradiated graphite. The test was based on the ASTM C1220-98 methodology, where the leaching condition was set at a temperature varying from 298 K to 363 K for 28 days. The release of cesium from ABG was in the permissible limit and followed the Arrhenius’s law of a surface controlled reaction; its activation energy (E_a) was 65.6 ± 0.5 kJ/mol. Similar values of Ea were obtained for Boron (64.3 ± 0.5) and Silicon (69.6 ± 0.5 kJ/mol) as the main glass network formers. In contrast, the dissolution mechanism of cesium from GGC was a rapid release, with increasing temperature, and the activation energy of Cs (91.0 ± 5 kJ/mol) did not follow any model related to carbon kinetic dissolution in water. Microstructure analysis confirmed the formation of Crystobalite SiO₂ as a gel layer and Cs⁺¹ valence state on the ABG surface.     

Irradiation of a thorium–plutonium rodlet: Experiment and benchmark calculations

A benchmark exercise for thorium–plutonium fuel, based on experimental data, has been carried out. A thorium–plutonium oxide fuel rodlet was irradiated in a PWR for four consecutive cycles, to a burnup of about 37 MWd/kgHM. During the irradiation, the rodlet was inserted into a guide tube of a standard MOX fuel assembly. After the irradiation, the rod was subjected to several PIE measurements, including radiochemical analysis. Element concentrations and radial distributions in the rodlet, multiplication factors and distributions within the carrier assembly of burnup and power were calculated. Four participants in the study simulated the irradiation of the MOX fuel assemblies including the thorium–plutonium rodlet using their respective code systems; MCBurn, HELIOS, CASMO-5 and ECCO/ERANOS combined with TRAIN. The results of the simulations and the measured results of the radiochemical analysis were compared and found to be in fairly good agreement when the calculated results were calibrated to give the same burnup of the thorium–plutonium rodlet as that experimentally measured. Average concentrations of several minor actinides and fission products were well reproduced by all codes, to the extent that can be expected based on known uncertainties in the experimental setup and the cross section libraries. Calculated results which could not be confirmed by experimental measurement were compared and only two significant anomalies were found, which can probably be addressed by limited modifications of the codes.     

Change of chemical states of niobium in the oxide layer of zirconium–niobium alloys with oxide growth

The change of chemical states of niobium with oxide growth was examined in the oxide layers of Zr–2.5Nb around the first kinetic transition by the conversion electron yield – X-ray absorption near-edge structure measurements. The detailed depth profiles of niobium chemical states were obtained in both the pre- and the post-transition oxide layers of Zr–2.5Nb formed in water at 663 K for 40–280 d. The depth profiling revealed that the inner oxide layer remained protective to oxidizing species even though in the post-transition region and this excellent stability of barrierness would be attributed the suppression of hydrogen pickup.     

Compatibility of reduced activation ferritic martensitic steel JLF-1 with liquid metals Li and Pb–17Li

The corrosion of reduced activation ferritic martensitic steel, JLF-1 (Fe–9Cr–2W–0.1C), in high-purity Li was quite small. However, carbon in the steel matrix was depleted by the immersion to the Li. The depletion caused the phase transformation of the steel surface in which the morphology of the steel surface changed to ferrite structure from initial martensite structure. The phase transformation degraded the mechanical property of the steel. However, the carbon depletion and the phase transformation of the steel were suppressed in carbon doped Li. The carbon in the steel was chemically stable and did not dissolve into the Li when the carbon potential in the Li was high. The concentration of nitrogen and oxygen must be kept as low as possible because the corrosion was larger when the concentration of oxygen or nitrogen in the Li was higher. The chemical reaction between the steel and the Li compounds of Li₃N and Li₂O was also investigated. The corrosion of the JLF-1 steel in Pb–17Li was summarized as the dissolution of Fe and Cr from the steel into the melt. The corrosion of the specimen with Er₂O₃ coating fabricated by metal organic decomposition process in the Li and the Pb–17Li was investigated. The coating was deformed, cracked and partially exfoliated in the liquid metals, though the oxide itself was chemically stable in the liquid breeders. The damage was probably made by the stress, which was generated by a large difference of the thermal expansion ratio between the solidified Li or Pb–17Li and the coating during a heat up and a cool down process of the corrosion test.     

A novel concept of QUADRISO particles – Part III: Applications to the plutonium–thorium fuel cycle

In the present study, a plutonium–thorium fuel cycle is investigated including the ²³³U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing ²³³U. The above configuration has been compared with a configuration where fissile (neptunium–plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the ²³³U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.     

Injection casting of U–Zr and U–Zr–RE fuel slugs and their characterization

U–Zr fuel slugs containing rare-earth elements can be difficult to cast because of the high reactivity of rare-earth elements. In this study, U–Zr and U–Zr–RE (RE: a rare-earth alloy comprising 53% Nd, 25% Ce, 16% Pr, and 6% La by weight) fuel slugs were prepared by injection casting, and their characteristics were evaluated. The as-cast fuel slugs were fabricated to the full length of the mold, and they showed no thin sections or cracks. Compared to the theoretical density, the measured density of the U–Zr and U–Zr–RE fuel slug was lower and higher, respectively. Chemical analysis revealed that the Zr and RE compositions in the U–10Zr and U–10Zr–3RE fuel slugs matched the target composition within 1.0 wt%. However, the RE composition in the U–10Zr–7RE fuel slug differed from the target composition by over 4 wt%. The melting crucible was further deteriorated and the casting yield was lower for the casting of a high rare-earth bearing fuel slug.