Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO₂ and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO₂ fueled PWR under both reactivity initiated accidents and loss of coolant accidents.     

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.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Oxygen potential and the chemical state of the fission products ruthenium,rhodium and palladium in irradiated oxide fuels

Thermodynamic properties of intermetallic compounds of the type M₃U and M₃Pu where $\text{M = Ru}$, Rh or Pd, are reviewed. The critical oxygen potential of the oxide fuel matrix necessary for their formation is suggested.     

Doppler effect experiment of resonance materials for rock like oxide fuels

A Doppler effect experiment of resonance materials such as erbium, tungsten, thorium and uranium was carried out in the Fast Critical Assembly of Japan Atomic Energy Research Institute. Cylindrical shaped samples of 150 mm in stack length and 23 mm in diameter were fabricated. The sample was contained in capsules and placed at the center of the core. Temperature of the sample was raised up to 1073 K. Measured Doppler reactivities of erbium, thorium and tungsten are comparable with those of uranium. Analysis was performed using the SRAC code system. Effective absorption cross sections of the samples were generated by two different methods. One is the f-table method based on the NR approximation and the other is the PEACO method that performs direct calculations of resonance absorption with an ultra fine energy group structure. Calculated results were compared with the measured values. For all samples except the tungsten, the f-Table method gives 17 % smaller reactivity than the PEACO method. Both methods predict the measurements within the error of 6 %. For the tungsten sample, the calculations underestimate the measurement by about 10 %.     

碳化铀核燃料缺陷结构的研究现状

与传统的氧化型核燃料相比,碳化铀因其有潜力用于第四代反应堆而受到广泛关注。为了安全有效地利用这种先进核燃料,有必要对它的结构、体性质和缺陷形成机制充分认识,从而了解随着裂变反应的进行裂变产物对核燃料性能的影响规律。本文对铀碳系列化合物(UC、UC_2和U_2C_3)实验和理论研究进行了综述,重点分析其缺陷结构下的化学和力学性能,同时对进一步的研究提出了展望。目的在于帮助核能领域的研究者厘清碳化铀的服役行为,同时为燃料设计者提供有价值的线索。     

Evaluation of a pellet-clad mechanical interaction in mixed oxide fuels during power transients by using axisymmetric finite element modeling

MOX fuel rod behavior due to PCMI during power transients was evaluated using a finite element code, ABAQUS. Clad elongation is calculated through a coupled temperature–displacement analysis where a half-pellet is axisymmetrically modeled. Parametric study for the PCMI model is preliminary performed to identify the dominant factors and examine the applicable range of the model. The comparison of the predicted results with recent MOX in-pile data shows that the centerline temperature and clad elongation are evaluated within an acceptable range.     

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.     

Application of chemical interaction between (Fe,Cr) oxides and Mo oxide at high temperature for self-healing intelligence on nuclear fuel cladding in LWRs

A novel scheme for a bilayer coating with self-healing ability is proposed in this study. The candidate materials for the coatings and the potential self-healing reaction are assessed in high-temperature aqueous environments and high-temperature air. The pure Cr₂O₃ layer and the composite of Cr₂O₃ and MoO₃ are the candidate materials for the outer layer and inner layer, respectively, due to their compatibility under normal condition and fabricability. Fe₂O₃–MoO₃ reactions exhibit a potential ability to heal the cracks because of a high reaction rate under normal condition. The self-healing process proceeds via the following mechanism under normal condition: Fe₂O₃ (a corrosion product in the coolant) diffuses into the cracks on the coating and reacts with MoO₃ (inner layer) to produce the insoluble Fe₂(MoO₄)₃, which deposits and repairs the cracks. In the loss-of-coolant accident (LOCA) situation, Cr₂O₃–MoO₃ reaction is expected to strengthen the adhesion of the coating.     

Optimization of nanopores obtained by chemical etching on swift-ion irradiated lithium niobate

The morphology of the nanopores obtained by chemical etching on ion-beam irradiated LiNbO₃ has been investigated for a variety of ions (F, Br, Kr, Cu, Pb), energies (up to 2300 MeV), and stopping powers (up to 35 keV/nm) in the electronic energy loss regime. The role of etching time and etching agent on the pore morphology, diameter, depth, and shape has also been studied. The transversal and depth profiles of the pore have been found to be quite sensitive to both irradiation and etching parameters. Moreover, two etching regimes with different morphologies and etching rates have been identified.     

AFM characterization of model nuclear fuel oxide multilayer structures modified by heavy ion beam irradiation

This work explored a potential new model dispersion fuel form consisting of an actinide material embedded in a radiation tolerant matrix that captures fission products (FPs) and is easily separated chemically as waste from the fuel material. To understand the stability of this proposed dispersion fuel form design, an idealized model system composed of a multilayer film was studied. This system consisted of a tri-layer structure of an MgO layer sandwiched between two HfO₂ layers. HfO₂ served as a surrogate fissile material for UO₂ while MgO represented a stable, fissile product (FP) getter that is easily separated from the fissile material. This type of multilayer film structure allowed us to control the size of and spacing between each layer. The films were grown at room temperature by e-beam deposition on a Si(1 1 1) substrate and post-annealed annealing at a range of temperatures to crystallize the HfO₂ layers. The 550 °C annealed sample was subsequently irradiated with 10 MeV Au³⁺ ions at a range of fluences from 5 × 10¹³ to 3.74 × 10¹⁶ ions/cm². Separate single layer constituent films and the substrate were also irradiated at 5 × 10¹⁵ and 8 × 10¹⁴ and 2 × 10¹⁶, respectively. After annealing and irradiation, the samples were characterized using atomic force imaging techniques to determine local changes in microstructure and mechanical properties. All samples annealed above 550 °C cracked. From the AFM results we observed both crack healing and significant modification of the surface at higher fluences.     

Investigation on the feasibility of the interaction between NiO and Mo oxide for self-healing intelligence on the nuclear fuel cladding

ABSTRACT

The feasibility of NiO-MoO₃ reaction as a complement for the self-healing reaction proposed in our previous study is assessed in high-temperature aqueous environment. NiMoO₄·nH₂O and crystal α-NiMoO₄ are produced in an aqueous environment by Ni(OH)₂-MoO₃ and NiO-MoO₃ reactions, of which the starting temperature is about 100°C. NiMoO₄·nH₂O is completely dehydrated into β-NiMoO₄ when annealed at about 600°C in air. Therefore, NiO-MoO₃ reaction is expected to self-heal the cracks as a supplement of Fe₂O₃-MoO₃ reaction in the high-temperature aqueous environment.     

压水堆核燃料组件价格形成及对核发电成本的影响初探

从分析核燃料组件价格的形成入手,讨论了核燃料组件价格的形成特点及其对核发电成本的影响。     

Thermal Expansion and Thermal Conductivity of Cesium Uranates

Two kinds of cesium uranates, which are often predicted by thermochemical calculations to be formed in irradiated oxide fuels with high oxygen potentials, were prepared from U₃O₈ and Cs₂CO₃ to determine the thermal expansions and the thermal conductivities. The lattice parameters of tetragonal Cs₂UO₄ and monoclinic Cs₂U₂O₇ were measured by the high-temperature X-ray diffraction method as a function of temperature. The linear thermal expansions of Cs₂UO₄ and Cs₂U₂O₇ obtained from the temperature dependencies of the lattice parameters were 1.2% and 1.1% from room temperature to 973 and 1,073K, respectively. The thermal diffusivities of Cs₂UO₄ and Cs₂U₂O₇ were measured on the disk-shaped samples by the laser flash method as a function of temperature. The thermal conductivities were evaluated from the measured thermal diffusivities and the bulk densities, and the specific heat capacity available in literature. The thermal conductivity of Cs₂UO₄ corrected for 100%TD was 1.2W/m·K at 980K and that of Cs₂U₂O₇ was 0.94W/m·K at 1,093K, which are about 30% and 27% of that of UO₂, respectively.     

Categorisation and priorities for future research on inert matrix fuels: An extended synthesis of the panel discussions

There are three categories of basic fuel cycle needs, which are being addressed by the different types of inert matrix fuel (IMF) concepts currently under development. These are: plutonium burning in existing LWRs, plutonium burning in fast reactors and minor actinide transmutation — corresponding to three distinct timescales for perceived IMF implementation, viz. short, medium and long term, respectively. The current paper, based partly on the two panel discussions organised at the 6^th IMF workshop, presents viewpoints and priorities for each of the three categories of IMF applications, both in terms of the fuel concepts to be pursued and the corresponding R&D requirements.     

Evaluation Methods of Resonance Absorption for System with Pellet Surrounded by Fuel Solution

Abstract

The reaction cross sections of ²⁷Al(n, p)²⁷Mg, ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn, ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb have been measured by the activation method in an energy range of 13.3–14.9 MeV using the intense D-T neutron source, FNS. Absolute flux was determined by the associated α-particle counting method incorporated with neutron spectra obtained from both a Monte Carlo calculation and a time-of-flight measurement. Corrections were extensively performed not only for the neutron flux determination, but also for the low energy neutron contribution to the reaction rates. The present data were compared with comprehensive evaluations as well as recent experimental data. The measured cross sections of ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn and ⁹⁰Zr(n, 2n)^89m+gZr are generally in good agreement within experimental errors with the values in both the JENDL Dosimetry File and IRDF-90. It is also shown that there are the overestimation of the cross sections for ⁹³Nb(n, 2n)^92mNb in the JENDL Dosimetry File, and the over- estimation and underestimation of the cross section for ²⁷Al(n, p)²⁷Mg in the JENDL Dosimetry File and IRDF-90, respectively.     

中国实验快堆蒸汽发生器首次进水试验分析研究

中国实验快堆(CEFR)作为中国第一座钠冷快中子反应堆,蒸汽发生器作为分隔二回路和三回路器的重要设备,其运行的稳定性、可用性对于中国实验快堆的稳定运行具有重要意义。中国实验快堆蒸汽发生器首次进水试验验证了蒸汽发生器具备了在低功率下运行的完整性、可用性、稳定性和良好的传热性能。根据记录数据就钠侧和水侧的对流换热进行了热平衡校核计算,计算结果表明了本次试验钠侧和水侧换热量平衡,计算结果表明在低流量、低钠温的运行工况下水侧为主要换热热阻,该热阻值可以由格尼林斯基(Gnielinski)公式确定。     

Evaluation of sex specificity on oxidative stress induced in lungs of mice irradiated by ^12C^6＋ ions

The aim of this work is to identify if there is sex specificity on ^12C^6＋ ion-induced oxidative damage in mouse lung at different time points. Kun-Ming mice were divided into two groups, each composed of six males and six females： control group and irradiation group with a single acute dose of 4 Gy. Animals were sacrificed at 2, 4 and 12 h respectively, there lungs were removed immediately, and the oxidative stress-related biomarkers were measured by Diagnostic Reagent Kits. The results showed that the relative activities of superoxide dismutase （4 h）, catalase （2 h） and Se-dependent glutathione peroxidase （12 h） have significant changes （P〈0.05） between male groups and female groups, suggesting that the lungs of male mice are more sensitive to counteracting the oxidative challenge. Moreover, higher levels of malondiadehyde and lower contents of glutathione were also found in males, indicating that oxidative stress induced by ^12C^6＋ ion is pronounced in the lungs of males. We thought that these sex-responded differences may be attributed to the influence of sex hormones.     

Thermodynamic Analyses on Oxygen and Carbon Potentials of Impure Helium

Equilibrium oxygen and carbon potentials of impure He containing small amounts of impurities such as CO, CO₂, CH₄, H₂, H₂O and O₂ at 1,073 and 1,273 K were studied. The calculation of equilibrium composition of impurities was carried out assuming the gas-gas and gas-metal reactions. The diagrams, expressed with atomic oxygen fraction ō/S and atomic carbon fraction [Cbar]/S were represented and showed a wide range of equilibrium oxygen and carbon potentials of impure He. A parameter, (ō-[Cbar])/S showing the difference between atomic oxygen and carbon fraction, was found to give a measure of both oxygen and carbon potentials of gas mixtures. The results of the calculation showed that the oxygen and carbon potentials of the impure He with (ō-[Cbar])/S value of around zero was easily affected by the small variation of the gas composition. The corrosion behaviors of Inconel 617 in various impure He gases at 1,273 K could be explained by (ō-[Cbar])/S values.