Search Results (95)

This investigation employs focused ion beam (FIB) and transmission electron microscopy (TEM) techniques to systematically analyze the distribution characteristics of fission products in medium-burnup (40.6 GWd/tU) UO₂ fuel and their impact on fuel cracking behavior. The findings indicate that grain boundary embrittlement is predominantly attributed to the accumulation of spherical particles of solid fission products, including Mo, Ru, Rh, and Pd, which preferentially segregate around impurity particles, leading to localized stress concentration. Intragranular cracks are associated with the strip-like segregation of fission elements and the amorphization process. It also reveals that the size and number density of intragranular Xe bubbles are ~6.24 ± 0.24 nm and 5.2 × 10²² m⁻³, respectively, while Xe did not, under the analyzed conditions, significantly influence crack nucleation. This research elucidates the correlation mechanism between fission product distribution and fuel cracking behavior at medium burn up, offering experimental evidence to enhance the reliability and safety of nuclear fuel assemblies. Full article

Pebble Bed Reactors (PBRs) represent a new generation of nuclear reactors. However, modeling TRi-structural ISOtropic (TRISO) fuel particles employed in PBRs presents a unique challenge in comparison to most conventional reactor designs. Rapid generation of different possible fuel particle configurations for Monte-Carlo simulations provides improved insights into the effects of particle distribution irregularities on the neutron economy. Defective pebbles could cause changes in the neutron flux in a nuclear reactor due to increased or decreased moderating effects. Different configurations of particle fuel also impact isotope production within the nuclear reactor. This study simulates several TRISO configurations representing limited capabilities of randomization algorithms, manufacturing defects configurations and/or special pebble design. All predictions are compared to an equivalent homogenized model used as baseline. The results show that the TRISO configuration has a non-negligible impact on the parameters under consideration. To explain these results, the ratio of the thermal flux of each model to the thermal flux of the homogeneous model is calculated. A clear pattern is observed in the data: as irregularities in the moderator medium emerge due to the distribution of TRISO particles, the neutron spectrum softens, leading to higher values of k_∞ and better fuel utilization. This dependence of the spectrum on the TRISO configuration is used to explain the pattern observed in the depletion calculation. The results open the possibility of optimizing the TRISO configuration in manufactured pebbles for fuel utilization and safeguards. Future work should focus on full core simulations to determine the extent of these findings. Full article

Burnable poisoning and fuel enrichment zoning are two techniques often combined in order to optimize the fuel assembly behavior during the burnup cycle. In the present work, these two techniques will be applied to the 2D optimization of the fuel assembly conceptual design for the supercritical water-cooled reactor developed in the framework of the Joint European Canadian Chinese development of Small Modular Reactor Technology project, funded within the Euratom Research and Training programme 2019–2020. The initial configuration of the fuel assembly does not include any burnable absorbers and uses a homogeneous fuel enrichment of 7.5% in ²³⁵U. The infinite multiplication factor, $k_{\infty }$, starts from approximately 1.32 and drops, almost linearly, to 1.0 after a burnup of 40.0 MWd·kg⁻¹. The uniform enrichment is, however, responsible for a pin-power peaking factor that with fresh fuel starts from 1.32 and reduces to 1.08 at the end of the burnup cycle. A simplified analytical model is developed to assess the effect of different lumped burnable absorbers on the time dependence of the assembly $k_{\infty }$. It is shown that using an adequate number of B₄C rods, positioned in the outer wall of the fuel assembly, together with a suitable distribution of six different ²³⁵U enrichments, it allows for obtaining an assembly $k_{\infty }$ factor that starts from 1.11 at the beginning of the cycle and remains quite constant over a large fraction of the burnup cycle. Moreover, the pin-power peaking factor is reduced to 1.03 at the beginning of the cycle and remains almost unchanged until the end of the burnup cycle. Full article

As a promising nuclear technology, molten salt reactors (MSRs) have a bright future in the energy sector due to their unique advantages such as high efficiency, safety, and fuel flexibility. However, the accurate analysis of fission products in MSRs requires reliable fission yield data. Current reactor burnup analysis often uses the ORIGEN2 code, whose fission yield libraries mainly originate from the outdated 1970s ENDF/B-VI nuclear database, thus risking data obsolescence. This study evaluates ORIGEN2’s fission yield libraries (THERMAL, PWRU, PWRU50) against the modern ENDF/B-VIII.0 library. Through a comprehensive comparative analysis of Oak Ridge National Laboratory’s Molten Salt Reactor Experiment (MSRE) model, numerical simulations reveal library-dependent differences in MSR burnup characteristics. The PWRU library best matches ENDF/B-VIII.0 for U-235-fueled cases in keff results, while the PWRU50 library has minimal keff deviation in U-233-fueled setups. Moreover, in both fuel cases, the fission yield library was found to significantly affect the activity of key radionuclides, including Kr-85, Kr-85m, I-133m, Cs-136, Sn-123, Sn-125, Sn-127, Sb-124, Sb-125, Cd-115m, Te-125m, Te-129m, etc. Additionally, the fission gas decay heat power calculated via the ORIGEN2 library is over 20% lower than that from the ENDF/B-VIII.0 library tens of days after shutdown, mainly due to differences in long-lived Kr-85 production. These findings highlight the need to update traditional fission yield libraries in burnup codes. For next-generation MSR designs, this is crucial to ensure accurate safety assessments and the effective development of this promising energy technology. Full article

Motivated by the growth of global energy demand and the goal of carbon neutrality, lead-cooled fast reactors, which are core reactor types of fourth-generation nuclear energy systems, have become a global research hotspot due to their advantages of high safety, nuclear fuel breeding capability, and economic efficiency. However, its engineering implementation faces key challenges, such as material compatibility, closed fuel cycles, and irradiation performance of structures. This paper comprehensively reviews the latest progress in the core design of lead-cooled fast reactors in terms of the innovation of nuclear fuel, optimization of coolant, material adaptability, and design of assemblies and core structures. The research findings indicate remarkable innovation trends in the field of lead-cooled fast reactor core design, including optimizing the utilization efficiency of nuclear fuel based on the nitride fuel system and the traveling wave burnup theory, effectively suppressing the corrosion effect of liquid metal through surface modification technology and the development of ceramic matrix composites; replacing the lead-bismuth eutectic system with pure lead coolant to enhance economic efficiency and safety; and significantly enhancing the neutron economy and system integration degree by combining the collaborative design strategy of the open-type assembly structure and control drums. In the future, efforts should be made to overcome the radiation resistance of materials and liquid metal corrosion technology, develop closed fuel cycle systems, and accelerate the commercialization process through international standardization cooperation to provide sustainable clean energy solutions for basic load power supply, high-temperature hydrogen production, ship propulsion, and other fields. Full article

Due to its advantages of simple structure and high inherent safety, the heat pipe cooled reactor (HPR) could be widely applied in deep-sea navigation, deep-space exploration and land-based power supply as a promising advanced special nuclear power equipment option. In HPRs, the space between the components (fuel rods and heat pipes) is filled with solid matrix material, forming a continuous solid reactor core. Thermo-mechanical response of the solid core is a special issue for HPRs and has great impacts on reactor safety. Considering the irradiation and burnup effects, the thermal and mechanical modeling of an HPR was conducted with ABAQUS-2021 in this study. The thermo-mechanical response under long-term normal operation, accident transients and single heat pipe failed conditions was simulated and analyzed. The whole core presents relatively good isothermality due to the high thermal conductivity of the solid matrix. As for the mechanical performance, the maximum stress was about 300 MPa, and the maximum displacement of the matrix could be as high as 3.7 mm. It could lead to significant variation of the reactor physical parameters, which warrants further attention in reactor design and safety analysis. Reactivity insertion accidents or single heat pipe failure has obvious influence on the thermo-mechanical performance of the local matrix, but they did not cause any failure risks, because the HPR design eliminates the dramatic power flash-up and the solid-state core avoids the heat transfer crisis caused by the coolant phase transition. A quantitative evaluation of thermo-mechanical performance was completed by this research, which is of great value for reactor design and safety evaluation of HPRs. Full article

This study presents a novel multiphysics phase-field fracture model to analyze high-burnup uranium dioxide (UO₂) fuel behavior under transient reactor conditions. Fracture is treated as a stochastic phase transition, which inherently accounts for the random microstructural effects that lead to variations in the value of fracture strength. Moreover, the model takes into consideration the effects of temperature and burnup on thermal conductivity. Therefore, the model is able to predict crack initiation, propagation, and complex morphologies in response to thermal gradients and stress distributions. Several simulations were conducted to investigate the effects of operational and transient conditions on fracture behavior and the resulting cracking patterns. High-burnup fuels exhibit reduced thermal conductivity, elevating temperature gradients and resulting in extensive radial and circumferential cracks. Transient heating rates and temperatures significantly affect fracture patterns, with higher heating rates generating steeper gradients and more irregular crack trajectories. This approach provides critical insights into fuel integrity during accident scenarios and supports the safety evaluation of extended burnup limits. Full article

In the rim zone of UO₂ nuclear fuel pellets, high burn-up and low temperatures drive changes in the microstructure, leading to the formation of high burn-up structures (HBS). This review focuses on the formation of HBS, beginning with a description of the two contentious mechanisms—recrystallization and polygonization—that are believed to be the primary controlling factors. We discuss experimental and simulation studies that support both mechanisms, emphasizing that although each mechanism can explain certain aspects of HBS formation, neither recrystallization nor polygonization alone is sufficient to fully explain the observed phenomena. Furthermore, we emphasize the intrinsic relationship between these two mechanisms, suggesting that they represent different manifestations of the same underlying process under varying conditions, and we reference relevant studies that support this perspective. Lastly, we underline the significance of investigating the formation processes of HBS and provide an outlook on future research directions based on the current state of knowledge. Full article

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with ²³⁵U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies. Full article

This paper analyzes the behavior of the entropy of a nuclear fuel rod under deep burnup conditions, beyond standard operational ranges, reaching up to 60 years. The evolution of the neutron source distribution in a pressurized water reactor (PWR) fuel pin was analyzed using the Monte Carlo method and Shannon information entropy. To maintain proper statistics, a novel scaling method was developed, adjusting the neutron population based on the fission rate. By integrating reactor physics with information theory, this work aimed at the deeper understanding of nuclear fuel behavior under extreme burnup conditions. The results show a “U-shaped” entropy evolution: an initial decrease due to self-organization, followed by stabilization and eventual increase due to degradation. A minimum entropy state is reached after approximately 45 years of pin operation, showing a steady-state condition with no entropy change. This point may indicate a physical limit for fuel utilization. Beyond this point, entropy rises, reflecting system degradation and lower energy efficiency. The results show that entropy analysis can provide valuable insights into fuel behavior and operational limits. The proposed scaling method may also serve to control a Monte Carlo simulation, especially for the analysis of long-life reactors. Full article

This work presents the synthesis, purification, and characterization of a molten salt fuel for the irradiation experiment SALIENT-03 (SALt Irradiation ExperimeNT), a collaborative effort between the Nuclear Research and Consultancy Group and the Joint Research Centre, European Commission. The primary objective of the project is to investigate the corrosion behavior of selected Ni-alloy based structural materials which are being considered for the construction of fluoride molten salt reactors. During the test, these materials will be exposed to selected liquid molten fuel salts under irradiation in the High Flux Reactor in Petten, the Netherlands. In addition, the properties and distribution of the fission products formed in the fuel salt during burn-up will be studied by various post irradiation examinations. In the SALIENT-03 fuel, U and Pu fluorides, as fissile material, are dissolved in a carrier melt based on a 78⁷LiF-22ThF₄ eutectic mixture to form fuel salts with four different compositions, containing PuF₃, UF₄, UF₃, and CrF₃. This article comprehensively describes all the steps of this fuel synthesis process: the synthesis of the required pure fluoride powders (⁷LiF, ThF₄, UF₄, UF₃, and PuF₃); the mixing, melting, and purification of the different fuel salt compositions; and the fabrication of solid ingots to be loaded into the irradiation capsules. The characterization of the intermediate and final products is also carried out, following a rigorous quality assurance protocol. The quality assurance is achieved using an analytical scheme consisting of mass balance-based conversion efficiency evaluation, X-ray diffraction, and differential scanning calorimetry analyses. All experimental goals were successfully achieved, highlighting promising prospects for advancing future research and development regarding fuel production methods for fluoride-based molten salt reactors. Full article

Traditional computational methods for pressurized water reactors are unable to handle dispersed fuel particles as the double heterogeneity and the direct volumetric homogenization can result in significant errors. In contrast, reactivity-equivalent physical transformation techniques offer high precision for addressing the double heterogeneity introduced by dispersed fuel particles. This approach converts the double heterogeneity problem into a single heterogeneity problem, which is then subsequently investigated by using the conventional pressurized water reactor computational procedure. However, it is currently empirical and takes a lot of time to obtain the right k_∞. In this paper, we train the RPT model by using the existing dataset of plate-dispersed fuel and rod-dispersed fuel by a machine learning method based on a linear regression model, and we then use the new data to make predictions and derive the corresponding similarity ratios. The burnup verification, density verification, fission rate verification, and neutron energy spectrum analysis are calculated through the OpenMC program. For plate-type fuel elements, the method maintains an accuracy within 200 pcm during depletion, with deviations in the ²³⁵U density and ²³⁵U fission rate within 0.1% and neutron energy spectrum errors within 6%. For rod-type fuel elements, the method maintains an accuracy within 100 pcm during depletion, with deviations in ²³⁵U and ²³⁹Pu density within 1.5% and neutron energy spectrum errors within 1%. The numerical validation indicates that the reactivity-equivalent physical transformation method based on the linear regression model not only greatly improves the computational efficiency, but also ensures a very high accuracy to deal with double heterogeneity in nuclear reactors. Full article

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10⁻⁴); the splitting method could reach a precision of 10⁻² level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors. Full article

The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. The three-dimensional numerical model consisting of the reactor pressure vessel with core internals was developed using available geometrical and material data as well as the reactor’s operating conditions. The reactor core was divided into 11 axial and 22 radial burnup zones in order to recreate the spatial distribution of the fuel burnup. The isotopic changes due to the nuclear transmutations and decays were calculated in each burnup zone until the desired average burnup of 17.571 GWd/tHM_int was reached. The calculations include changes in the boric acid concentration at defined time steps and the burnout of the gadolinia burnable absorber embedded in the nuclear fuel. This study shows the spatial distribution of minor and major actinides at the end of the reactor cycle as well as the depletion of uranium, the build-up of plutonium, and the formation of neptunium, americium, and curium during the reactor’s operation. Full article

A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for refueling and fuel shuffling operations. The incorporation of 0.075 wt.% ²³⁷Np provides favorable plutonium isotopic vectors throughout an operational lifetime of 5.5 years. providing 35 MWe. Core performance was assessed using a verification-by-comparison approach for core designs with or without ²³⁷Np and/or gadolinia burnable absorber. Burnup Monte Carlo calculations were performed via MCOS coupling of MCNP and ORIGEN to an achievable burnup of ~62.5 GWd/t. The results demonstrate a minimal penalty to reactor performance due to the addition of these materials as compared against the reference design. Coupling of a proliferation-resistant fuel concept with a uniform loading of natural gadolinia burnable absorber for LEU+ fuel (7.5 wt.% ²³⁵U/U UO₂) provides favorable excess reactivity considerations with minimized concerns for additional residual waste and more uniform distribution of un-depleted ²³⁵U in discharged fuel assemblies. Full article