The Method of Characteristics (MOC) has been widely used for high-fidelity reactor physics analysis, due to its numerical accuracy and geometrical flexibility. At the same time, the MOC method has a high demand for computing resources. The selection of Flat Source Region (FSR) discretization and track generation parameters will have a great impact on MOC calculation time and accuracy. Therefore, appropriate parameter selection and optimization are necessary. This work applies a multi-objective genetic algorithm to optimize the parameters of OpenMOC for the calculation of the 2D C5G7 benchmark problem, aiming at reducing the error of keff and power distribution and the calculation time. After optimization, the parameters solution with the highest calculation efficiency can be selected in the Pareto-optimal set according to the required calculation accuracy in the practical engineering application. The results of this work demonstrate that genetic algorithms can effectively optimize the selection of MOC parameters. When using OpenMOC to solve the 2D C5G7 benchmark problem, the optimized parameters can improve the solving speed by up to several tens of times compared to the default parameters while achieving the same accuracy. At the same time, these optimized MOC parameters also show applicability in 2D NuScale simulation.

A research project on the aging concrete of Hamaoka nuclear power plant (NPP) Unit 1―it is currently decommissioning―, was started in 2016. The objective of this project is to create a concrete database of material properties, and employ non-destructive evaluation (NDE) method to construct a better soundness method. For this purpose, approximately 200 core samples were obtained from various concrete structural members, exceed 1.0m owing to seismic performance requirements. It was that the compressive strength of thick members showed a convex-shaped trend, indicating that the center of the cross-section was larger than the surface (Ref. [1]). In ordinary environmental conditions, the average compressive strength of all specimens obtained from seismicresistant 1.0 ~ 1.5m thick walls is approximately 10% larger than the average compressive strength of surface element regardless of these surface coatings.

HPR1000’s symptom-based emergency operating procedure (SEOP) system contains optimal recovery procedures and functional recovery procedures, which can independently handle design basis accidents and most superimposed accidents. In order to solve the problem in lack of safety engineer procedure in the HPR1000 symptom-based procedure system and improve the HPR1000 SEOP system, this paper proposes an overall program for the improvement of the SEOP system structure includes extending critical safety function status diagnosis, adding support functional recovery procedures and adding safety engineer procedure. The research of superimposed accident supervision technology was carried out, effective supervision for superimposed accident was achieved after design of ESF status monitoring module, important function failure monitoring module and support system monitoring module. A framework for independent supervision of safety engineer has been established. A preliminary verification of the completeness of conditions coverage and of the effectiveness of independent supervision for improved SEOP system was carried out based on Zhangzhou NPP Unit1/2 design extension conditions (DEC-A), the result of verification shows that the improved SEOP system has distinctive features and can better handle superimposed accident conditions compared with other symptom-based procedure systems at home and abroad.

As the fourth-generation advanced reactor, the air-cooled micro reactor has the characteristics of modularity, high inherent safety, high level of automatic control, and non-dynamic safety design. The origin of the accident is clear, and the accident handling methods are concentrated, which greatly simplifies the emergency operation strategy. In order to further reduce the burden on operators and improve the automation level of the reactor, the research is develop on the automated implementation of emergency operation procedures. On the basis of ensuring nuclear safety, based on the characteristics of the core and auxiliary system of the air-cooled micro-reactor, the functional requirements of the automated execution system of accident procedures are put forward from the design principles, interface display, basic functions, etc.. In this research, the requirements for the entry method of the procedures, the display of the procedures, the rights management of the procedures, and the operation of the procedures are given, and relevant examples are given for some of the requirements. According to the functional requirements of the research, the current emergency operation procedure strategy can be transformed to have the ability of automatic implement. The transformed procedure system can meet the requirements of regulations, has friendly man-machine interface, and has relatively high automatic execution ability.

In order to realize the retention of core melt, some nuclear power plant has set up a Cavity Injection and Cooling System. The system includes two parts, active injection and passive injection, of which the passive part adopts high-level water source injection to meet the flow demand, and the active injection part is supplied by the fire pump, and a flow control valve is set on the cooling water supply pipeline. After the serious accident of core melting, the pressure in the containment has a large range of change, and the flow control valve is used to adjust the flow of cooling water injected into the pit, and then take away the core melt heat through the external cooling of the pressure vessel, reduce its temperature, and maintain the integrity of the pressure vessel. Finally, the retention of the core melt in the pressure vessel is realized. Among them, the active injection part, according to the accident analysis, is divided into the initial stage of large-flow injection, and when the reactor cavity is filled with water, it enters the long-term low-flow water injection stage, which meets water-supply need for a certain period of time. By using commercial software “FLOWMASTER” to model the water injection cooling system of some nuclear power reactor cavity, based on different pressure in the containment vessel and the injection flow demand at different stages, the calculation, analysis and research are carried out. The operating point of the control valve suitable for some nuclear power plant was determined.

The raft foundation at the lower part of VVER-1200 reactor building is a cylindrical concrete structure, which is cast in one piece and belongs to the category of mass concrete construction.Because the temperature change and shrinkage caused by the hydration of the cementitious material in the mass concrete are easy to cause harmful temperature cracks in the concrete during the construction pouring to the curing age, it is necessary to carry out the research on the construction method and hydration heat of VVER-1200 nuclear island raft foundation concrete.Taking an actual VVER-1200 nuclear island raft foundation as the research object, the concrete pouring method was innovated, the finite element model of the lower raft foundation of the reactor building was established, the temperature field distribution of the lower raft foundation of the reactor building was analyzed in combination with the thermal test parameters of the project, and the change rule of the temperature field and temperature time history curve of the lower raft foundation of the reactor building with time was studied.The results show that the simulation analysis results of the transient temperature field of the raft foundation under the VVER-1200 reactor building are basically consistent with the temperature data measured in the project. The concrete construction method and the concrete simulation analysis method can effectively control the generation of harmful temperature cracks in the raft foundation of the nuclear island, which is of guiding significance to the construction of the raft foundation of the nuclear island of the same type of cylindrical nuclear power plant.

Risk-Informed management is an effective approach to improve safety and economy of Nuclear Power Plants(NPPs). Living PSA method and Risk Monitor are basic methodology and tool to implement risk management in NPPs. There are some deficiencies for Risk Monitors currently in use, including: (1) Insufficient specificity of risk assessment; (2) Insufficient ability to predict risks in advance. To solve the issues above, technique of Performance Based Risk Monitoring and Predicting(PBRMP) is proposed in this paper, which is based on the combination of Prognostics and Health Management(PHM) and Risk Monitor. And a case study is carried out to verify the proposed technology, based on Level 1 PSA model of Fuqing 5/6 unit. The case study indicates that: (1) The proposed methodology is feasible; (2) By use of the proposed technology, the risk of NPPs can be monitored and predicted based on the performance of components; and (3) Compared with the existing risk monitor technology, it can improve the pertinence and authenticity of risk assessment, and also the ability to predict risks in advance. It is an innovative application of intelligent technology in the field of nuclear safety, leading to complementary improvement of PHM and Risk Monitor technologies.

This article summarized the test results of water level fluctuation characteristics of several typical domestic coastal NPPs in the water intake channel and pump house channel. From the current specification requirements, open channel entrance orientation, open channel length, internal boundary, pump house flow channel and long-term and short-period water level fluctuations, the impact of water level fluctuations in water intake channels and control standards are discussed and demonstrated. It is proposed that the filtering efficiency of the inlet channel, coarse grid, steel gate and fine grid in the pump room should be considered in the design, and the water level fluctuation limit of the water intake front pool can be appropriately relaxed. Regarding the long-period water level fluctuations of the open channel, short-period water level fluctuations mainly affect the liquid level difference before and after the rotating filter in the pump house channel, which may cause false alarms and normal operation of the pump and should be paid attention to in the specific design.

A self-analysis of the accident at the Fukushima Dai-ichi Nuclear Power Station concluded that one of the reasons for the accident was the inability to utilize OE: Operating Experience data.

For this reason, we have implemented various OE initiatives such as utilizing OE information calendars, providing new OE information to contractors, defining highlevel priority information from the viewpoint of a graded approach, and improving OE information processing rate.

Currently, the process of OE information continues to be improved. At present, all staff in the nuclear power division including contractors, have been able to obtain OE data as soon as possible and utilize these information to their own work situations. In addition, the speed of information sharing with other companies, as well as within the company, has been improved with regard to our OE information.

Nuclear graphite has several levels of structural complexity at different scales. The effect of temperature and irradiation on graphite properties depends not only on the current structure but also on smaller scales. Therefore, some mechanism issues need to be studied across different scales. To investigate the effect of changes in microscopic properties on polycrystalline and bulk, a cross-scale modeling approach connecting molecular dynamics model and finite element model has been constructed in this work. The molecular dynamics method was used to evaluate the effects of temperature and irradiation on the structures and properties of graphite crystallite. The finite element method was then used to develop a continuous medium model for microcrack free graphite polycrystalline, assigning the properties of the crystallite to the polycrystalline model. In the microcrack free polycrystalline model, the effects of polycrystalline texture and crystallite properties are coupled. Therefore, the model response to external environmental factors may be of great complexity. In this way, the effects of temperature and irradiation on the properties of microcrack free polycrystalline can be evaluated in terms of crystallite properties.

Japan Atomic Energy Agency is developing extraction chromatography technology for MA(III) recovery from spent nuclear fuel. Developments in the extraction chromatography system especially focusing on safety and stable operation are required for practical application of the technology. In this paper, main tasks which have to be challenged preferentially are discussed based on achievements obtained by previous studies and potential MA(III) recovery process flow. For the safety assessment, removal of fine particles, optimization in adsorbent for reduction in pressure drop of the column, degradation behavior of the adsorbents and recycle of the adsorbents are topics for development. Enhancement in resistance against radiation and acid of a material in pump and development of automated operation system are required to maintain product quality during long time operation. Those studies are carried out simultaneously in a research project of 2019-2024, and demonstration with a large scale apparatus will be carried out at end of the project.

Aiming at the challenging problems of low utilization efficiency of uranium resources and difficult safe disposal of spent fuel, the Chinese Academy of Sciences has proposed the concept of Accelerator Driven Advanced Nuclear Energy System (ADANES), which is composed of ADS burner and a fuel recycle system. ADS burner with subcritical operation can burn the extensive fuel processed through a fuel recycle system, so that ADANES can realize power generation, transmutation and breeding. Accelerator driven recycle of used fuel is mainly to eliminate part of fission products (gas, volatile fission products, lanthanide nuclides) in spent fuel through high-temperature evaporation, recrystallization and other processes to form regenerated spent fuel. At present, the fuel recycle of ADANES is still under study. This study mainly studies the reactor core long-term burnup under different refueling modes. Two coolants Al₂O₃ coolant and ZrO₂ granule are simulated for fuel cycle. The different removed rate of gas and volatile substances, lanthanide nuclides were removed from the SNF at each refueling. At the 100% removal of gas and volatile substances and > 50% removal of rare-earth elements, the ceramic reactor core can burn for 60 years between 7 refueling. The results show high-temperature evaporation and recrystallization need to be used together to achieve the longcycle fuel cycle.

In the Fukushima Daiichi Nuclear Power Plant, a meltthrough severe accident happened in 2011. The fuel debris of various compositions, such as the core elements, the control rods, and the cladding tube, was solidified at the pressure vessel. Yusufu et al. made the simulated fuel debris of B₂O₃-UO₂ and investigated the relationship between the microstructure and the mechanical properties of the fractural strength. It revealed that the simulated fuel debris was a porous medium, which depended on the cooling rates. The B₂O₃-UO₂ debris at high cooling rates showed low fracture strength, compared with that at the low cooling rates due to the porosity density.

In this study, we have performed the phase-field and multiphase-field simulations in the pseudo-binary B₂O₃-UO₂ system. The phase-field simulation focused on the formation process of the liquid-solid structure at a wide range of temperatures and concentrations. This model implicitly treated monotectic reaction using the Gibbs free energy. Randomly nucleated UO₂ particles grew up and combined with each other, depending on the cooling rate. The cooling rates affected the concentration at the liquid phase and the growth rates of UO₂ particles. On the other hand, the multiphase-field simulation focused on the monotectic reaction around the monotectic temperature ( T_mono ~ 1841K). The monotectic reaction is L#2 →L#1 FCC : “L#1” and “L#2”are liquid phases, “FCC” is solid phase of UO₂. We investigated the formation of the solid-liquid microstructures (“FCC” and “L#1”) by the monotectic reaction. This model explicitly treated the reaction. After the formation, the liquid phase of “L#1” could vaporized because of high B₂O₃ concentration (x 0.96).

In this study, ballooning and rupture behaviors of the FeCrAl cladding tube have been investigated experimentally, which is one of the alternative materials of accident tolerant fuel (ATF) cladding. An experiment facility has been built, and severe accident conditions of the cladding can be physically simulated, including high temperature, high internal pressure, and oxidation with steam outside. Temperature, pressure, and diameter of the cladding can be measured online. Image of the ballooning and rupture process can also be recorded through sight glass.

With this facility, behaviors of the Zr-4 cladding tubes have been tested firstly. Experimental data of the cladding are compared with existing results, and credibility of the experimental results are verified. Ballooning and rupture performance of the FeCrAl cladding tubes are tested. The tube is pressurized to be 2-8 MPa with argon. Then it is heated until rupture, and ballooning occurs before the rupture.

Transient performance of the cladding temperature, pressure and diameter are presented, and rupture temperature in different pressure conditions are summarized. Photographs of the cladding and crevasses are displayed. Experiment results of the FeCrAl and Zr-4 claddings are compared, including rupture temperature, and morphology of the crevasse. Potential influence of the claddings behaviors on scenario of severe accidents are discussed.

The Nuclear Regulation Authority (NRA) has tackled the experimental approach for determining the criticality of pseudo-fuel debris plausibly simulating actual fuel debris since 2014, collaborating with the Japan Atomic Energy Agency. To elucidate the characteristics of the pseudo-fuel debris, the Japan Atomic Energy Agency modified the STACY (STAtic experiment Critical facilitY) to conduct critical experiments simulating fuel debris. Thus, we proposed three types of modified STACY core configurations.

In critical experiments in the modified STACY core, it is important to judge whether the proposed modified STACY core configurations are representative of pseudo-molten core– concrete interaction debris or not. In this study, we built pseudo-fuel debris models considering a volume ratio of pseudo-fuel debris to moderation (V_m/V_f) and calculated uncertainty-based similarity values (C_k) between the modified STACY core configurations and pseudo-fuel debris models using Tools for Sensitivity and Uncertainty Analysis Methodology Implementation-Indices and Parameters (TSUNAMI-IP) in SCALE 6.2. Consequently, the modified STACY core configuration loading structure rods we proposed completely resulted in high similarity to the pseudo-fuel debris models through Vm/Vf values. The main contributions to C_k values were ²³⁵U nu-bar, ²³⁵U chi, and ⁵⁶Fe (n,γ), except for the pseudo-fuel debris model, including extremely high concrete components.

In molten salt reactors, a nickel-based alloy (NA) named GH3535 is used as structural material such as the vessel, claddings of control rods and the support material of the reactor core. The thermal neutron scattering (TNS) effect of the GH3535 alloy should be considered in the design of the reactor neutronics due to the safety requirements of the Generation IV advanced nuclear power systems. However, there is no TNS data for the NA in the major evaluated nuclear data files (ENDF). Considering the neutron scattering mechanism of the NA is complicated owing to the long-range disordered lattice structure, this study is focused on the lattice structure approximation of the NA in the calculation of TNS data and the data process of the TNS cross section of nickel-based materials. To investigate the lattice structure of the NA, the neutron total cross section (TCS) of the GH3535 alloy is measured using the transmission method based on the TMSR-PNS in this study. An approximation to simplify the lattice structure of the alloy in the calculation of the coherent elastic scattering cross section is proposed according to the conformance of the Bragg edges (BEs) shown in the TCS results between the NA and natural nickel. Meanwhile, the characteristics of the TNS data for nickelbased materials are investigated by generating TNS cross section for natural nickel. To validate the TNS data calculated, the theoretical TCS of natural nickel, which is obtained by combining the calculated scattering cross section in this work and the absorption cross section from the ENDF/B-V library, is compared with experiment results. Fine agreement between the theoretical TCS for natural nickel and experiment results has been achieved.

High level liquid waste generated from reprocessing of spent nuclear fuel contains various metal ions and is possible to form precipitations so called as sludge. The sludge will significantly influence on extraction chromatography technology for trivalent minor actinide (MA(III)) recovery. Applicability of a pre-column as a filter for removal of the sludge was experimentally evaluated. In this study, remote handling media exchanging experiments were carried out, and a concept of the pre-column system was proposed based on the results.

Three types of column were made, and 5 experiments were carried out. The media of the pre-column was porous silica particle. Bed of the particle was successfully made by supplying slurry of porous silica particle with water from upper inlet of the column through a tube pump. Since the media formed block, it interfered water current on ejecting porous silica particle. When water flow to the column was slow, porous silica particle settled down even if water was fed from lower inlet. Trickling basis system with upper outlet was the most suitable system, but the operation to control the media block behavior and optimization of water flow is needed.

In order to economically increase cycle lengths of pressurized water reactor, current regulatory burnup limit of 62 GWd/tU is expected to extend to ~75 GWd/tU. According to the experimental observation in the Halden IFA-650 tests, high burnup fuel seems more susceptible to turn into fine fragments (pulvers) in loss-of-coolant accidents (LOCAs). Due to the cladding ballooning and burst under temperature transients, fuel relocation and dispersal may occur, raising the licensing concerns related to core coolability as described in 10 CFR 50.46. The fuel relocation has been widely observed in various experimental programs and numerically investigated by onedimensional algorithm and three-dimensional simulation. However, there is few effort devoted to numerical investigation on fuel dispersal. This paper proposed a three-dimensional simulation framework with the coupled finite element method and discrete element method for fuel dispersal, in which coarse fuel fragments were modelled with Voronoi cells and the high burnup structure was established through small isolated particles. Besides, a method based on spline projection was used to build the geometry of cladding with burst opening. Three tests (191, 192, 193) conducted in the Studsvik program were chosen to validate the simulation results, in which the predicted dispersed mass of fragments fitted well with the experimental observation, indicating the viability of the simulation framework. Based on the validated simulation results, the mass fraction and filling ratio of fuel fragments along the axial direction were discussed to investigate the effects of burst opening size on fuel dispersal. Finally, the verification against the fuel dispersal model proposed by U.S. Nuclear Regulatory Commission was conducted to verify the high fidelity features of the DEM model in simulating fuel dispersal under LOCA.

Research on Cr-coated Zircaloy (Zry) as the most near-term solution for Accident-Tolerant-Fuel (ATF) concept has been boosted in the last decade, while a great concern is addressed on the interfacial failure between coating and substrate. In this study, Cr-Zry bilayer plates were firstly prepared using the diffusion bonding method at different temperatures (1073 and 1273K), subsequently the tensile tests were performed at RT and 573K. The finite element method (FEM) was further applied to simulate the stress distribution and evolution in the Cr-Zry specimen during the tensile test by ABAQUS. The results show that the Cr-Zry bilayer specimens exhibit distinctly different deformation behaviors between RT and 573 K: a two-stage deformation is noted for RT test and no sudden mechanical response for 573 K test, which is attributed to the ductile-to-brittle-transition of Cr. Besides, the fractography reveals delamination and cracks near the Cr/Zry interface at 1273 K-bonded specimens, indicating weak bonding with a higher presence of ZrCr2. Cracks initiating at the interface propagate to the two layers when plastic deformation occurs, caused by the local stress concentration and the followed prefracture at the interface. The presence of interlaminar shear stress is confirmed by FEM simulation, which is due to the transverse stress difference between Cr and Zry layers in the vicinity of the interface, which is probably responsible for the delamination of the Cr-layer. The obtained insights from this study indicate that the Cr-Zircaloy-4 interface is a weak point for Cr-Zry system materials, which will be essential to understand the mechanical behavior of Cr-coated Zr-based ATF claddings in reactors.

Iterated fission probability (IFP), which is proportional to a fundamental mode of adjoint angular neutron flux, has increasingly been used as a weighting function in Monte Carlo calculations. The Monte Carlo IFP methods stochastically estimate IFP for a given phase-space location. In this work, we investigated the applicability of a deep neural network for approximating an unknown underlying function, which maps from a phase-space location to an IFP in a given fissile system, from a dataset produced by a Monte Carlo IFP method. The preliminary application has been performed for the Godiva core and the comparison showed a varying degree of agreement and discrepancy between the estimated IFPs by the DNN and the reference adjoint angular neutron flux by a deterministic neutron transport code PARTISN.

Due to its outstanding performance, uranium zirconium (U-Zr) metallic fuel has been considered as a candidate fuel in fast reactor. However, the serious swelling and decrease in thermal conductivity caused by fission gas bubble is a severe challenge for its application. In the present work, a phase-field model considering the resolution of gas atom was proposed to investigate the distribution of bubble size and the decrease of effective thermal conductivity due to irradiation-induced bubble in U-Zr metallic fuel. The evolution of microstructures of U-Zr fuel with intragranular gas bubbles was obtained by solving a set of modified Cahn-Hilliard and Allen-Cahn equations in this model. The bubble evolution in U-Zr during irradiation was simulated. The distribution of bubble size is bimodal and agree well with the experimental data when the spatial distributed resolution rate is used. The resolution of gas atom is deduced to be the primary reason for bimodal bubble distribution in nuclear fuel. Furthermore, the variation in effective thermal conductivity of U-Zr under different bubble distributions was calculated quantitatively based on the microstructures provided by the phase-field model. Results showed that the effect of bubble distribution on the effective thermal conductivity was not significant at the same porosity. Overall, this research is helpful in understanding the decrease in thermal conductivity caused by fission gas bubble in U-Zr metallic fuel.

Thermophysical properties of corium are required in models and computer codes to predict the severe accident progression in light water reactors. However, the measurement of molten corium properties is challenging due to high melting points. This paper presents a surface tension measurement system for molten zirconia based on techniques of aerodynamic levitation and laser heating. Zirconia is one of the main components in corium and aerodynamic levitation is a contactless method to avoid interactions between the sample and container wall at high temperatures. A sample of zirconia was levitated by argon gas flow above a conical converging-diverging nozzle and then melted into a droplet by laser beams. The oscillation of molten zirconia was imaged by a high-speed camera. The resonant frequency was then obtained through image processing. Finally, the surface tension was derived according to the Rayleigh formula.

Silicon carbide (SiC) has been considered as one of the promising Accident Tolerant Fuel (ATF) cladding materials. The multi-layer cladding designs consisting of both monolithic SiC(mSiC) and SiC fiber/SiC matrix composites (SiCf/SiC) have been considered as a candidate for light water reactors, meeting the engineering requirements of safety improvement. A twodimensional axisymmetric mechanical model with anisotropic material properties was developed and integrated into the fuel performance code FROBA. The developed mechanical model was solved through the finite difference method and can be employed to analyze both inter-layer interaction and pelletcladding mechanical interaction (PCMI) of anisotropic multilayer claddings. Mechanical performance and failure probability of SiC cladding fuel during normal operation was then simulated. The results suggest that during steady-state operation, SiC cladding show a higher fuel temperature, no PCMI occurs during operation, the hoop stress is in safety level, and the cladding remains integrity.

Nuclear fuel cladding is subjected to neutron irradiation in a high-temperature stress environment, and the structural integrity of the cladding is very important for safe operation. The irradiation performance of Fe-Cr-Al alloy is significant for the development of multi-scale modeling and a deeper understanding of defect production. In this work, displacement cascade behaviors of Fe-Cr-Al crystal with a grain boundary (GB) are investigated by molecular dynamics (MD) simulations. Four grain boundaries, namely ∑3{111}, ∑5{120}, ∑9{221}, ∑11{113}, with misorientation angle varying from 38.9° to 129.5° were considered. It is found that the formation energy of point defects close to the boundary is much lower than that in the grain according to molecular static simulation results. The presence of GB reduces the maximum temperature of the thermal spike in the displacement cascade, which leads to the increase of FPs in the displacement spike and the delay of the time. The absorption bias of GBs on interstitials and vacancies results in more surviving FPs, but the majority of the defects are in the GB, and FPs in the bulk region are significantly lower. Of the four types of GBs studied, ∑3{111} has a high attraction to interstitials and vacancies, but essentially does not affect vacancy formation energy, leading to the most surviving FPs.

In a block-type High-Temperature Gas-Cooled Reactor (HTGR), a fixed number of burnable poisons are loaded at the beginning of operation time to control a large amount of excess reactivity. Therefore, it is necessary to evaluate the burnable poison (BP) reactivity worth to achieve the optimum design of the HTGR. In this study, an experiment to measure the burnable poison worth reactivity was conducted at the B-Core of Kyoto University Critical Assembly (KUCA). B-core is solid moderator materials such as polyethylene and graphite combined with fuel plates to form the fuel element.

The experiment was performed to measure the reactivity worth of a small cadmium plate (15 × 15 × 0.5 mm) at the B-Core of KUCA. In the experiment, there are 8-unit cells in a fuel element. In this study, the unit cell position of cadmium is called the cadmium unit cell. The experiments were carried out by varying the cadmium plate position inside the cadmium unit cell.

This study evaluated the cadmium reactivity worth using the Monte Carlo MVP3. The objective of this study was to evaluate the appropriate results between the measured and the calculation values. The simulation using MVP3 code was conducted by varying the number of batches in the calculation. The results showed that the maximum discrepancy between experimental and calculated results was 24% for 5,000 batches. However, the discrepancy decreased when the number of batches increased to 50,000. The cadmium reactivity worth difference between the experiment and simulation was approximately 18 % depending on the cadmium plate position in the cadmium unit cell.

Fission gas will induce a series of changes in fuel rod performance after the accident. The amount of fission gas will increase dramatically under accidental conditions. Fuel rod internal pressure and temperature will rise subsequently owing to the gap conductance between the fuel pellet and cladding aggravating. And then fuel rod performance will degrade. Therefore, the investigation of fission gas behavior is worthwhile and significant for assessing the accidents and evaluating the fuel rod performance. Furthermore, the effect of fission gas on the fuel rod is supposed to be minimized. Therefore, predicting fission gas release (FGR) is essential for the integrity and safety of nuclear reactors. The fission gas release in a Light Water Reactor (LWR) is modeled by COMSOL Multiphysics with finite element method under normal or abnormal conditions, and the latter is classified as the Class-II type incidents. This model assumes gas diffusion varies with time inside a spherical grain. Perfect sinks with gas production, perfect sinks without gas production, and imperfect sinks under steady-state conditions are studied by setting different initial and boundary conditions. Inputting parameters from other models and experiments and comparing their results have validated the accuracy and universality of expressions possible.

JAEA has been working on development of extraction chromatography technology for recovery of trivalent minor actinides (MA(III): Am, Cm) from high-level radioactive waste generated in reprocessing of spent fuel. The technology utilizes porous silica particles with about 50μm diameter for support of adsorbents. The particles are coated by styrene-divinylbenzene copolymer, and an extractant for MA recovery is impregnated into the polymer. Pressure drop of the packed column depends on characteristics of the particle (diameter, uniformity and pore size). Large pressure drop of the column is not favorable for safety assessment of the technology although a certain level of the pressure drop is indispensable for excellent separation performance. In this study, spray drying granulation experiments and fundamental characterization of the product particle were carried out to find optimal specs of the particle and conditions of the granulation operation.

Fission gas release and swelling were modeled by Comsol Multiphysics in Mixed Oxide Fuel (MOX). This model is mainly based on the Operational Gas RElease and Swelling (OGRES) model. In previous studies, the fission gas release and swelling are primarily related to burnup, time, and temperature. In this model it is assumed a spherical grain at the fuel pellet inner radius. The grain is 2D axisymmetric geometry in Comsol Multiphysics. Fission gas atom generation, gas atom diffusion, gas atom diffusion into the bubble, bubble resolution, gas bubble nucleation, gas bubble migration to the grain boundary, bubble pressure, gas bubble swelling, grain growth, oxygen to metal atomic ratio, fuel porosity, fuel thermal conductivity, heat generation rate, pore velocity, pore concentration, plutonium migration, and fuel material properties are considered into this model. All relevant formulations are input into the model. The results include grain size, gas bubble diameter, gas bubble concentration and the number of gas atoms per bubble, gas atom diffusion coefficient, bubble diffusion coefficient and surface diffusion coefficient with temperature, fission gas release with time, as well as swelling with time and radius. These results will finally be analyzed and compared with other modeling outcomes and experiments.

TRISO coated fuel particle has been adopted in the design of various reactors and Accident Tolerant Fuel (ATF) due to its excellent accident tolerant performance under high temperatures and other severe conditions. Accurately estimating fission product release from TRISO-coated fuel particles is indispensable to calculate the mechanistic source term. In most models, a bare kernel is substituted conservatively for failed particle. Thus, a theoretical method was proposed to describe the transport of fission products in intact and failed fuel particles.

In this paper, to analyze the diffusion and release characteristics of intact TRISO-coated fuel particles’ fission products, the fission products diffusion model of TRISO-coated fuel particles with thermal-mechanical coupling was established based on the COMSOL Multiphysics software and verified by the IAEA CRP-6 benchmark. The results indicate that the predictions by the developed fission product transport model is reasonable. Moreover, to analyze fission product release for failed partials, a conceptual theoretical method was proposed. The adsorption and desorption constraint model for fission products at the gas-solid interface and the gas phase diffusion model were added to the original model to analyze. Subsequently, predictions comparison for fission products' fractional release in failed particles between the proposed method and the Settings adopted in BISON were completed. The result suggests that the gas-phase diffusion model is feasible and seems favorable to simulate the diffusion of fission products in the failed coating layers.

The National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) is one of the primary U.S. national research facilities that hosts thousands of visiting domestic and international scientists/researchers for various research projects. The National Bureau of Standards Reactor (NBSR) has been operational more than 50 years, but due to increasing outage times and costly maintenance, a new replacement reactor, namely the NIST Neutron Source (NNS), is being planned by the NCNR to replace the aging NBSR. The preliminary conceptual design of the NNS core was based on Material Test Reactor (MTR) plate type U-10Mo fuel; however, due to recent difficulties and unknown timeline for the certification of U-10Mo fuel in the U.S., NCNR is investigating the possibility of using low-enriched uranium silicide dispersion (U₃Si₂/Al) as alternative fuel material. In this paper, the neutronics performance of low-enriched U₃Si₂/Al fuel and U10Mo fuel are compared with the objective of identifying the dimensions of U₃Si₂/Al fuel plates that can yield similar neutronics behavior to the current U-10Mo fuel plates without modifying overall assembly size. Hence, the main objective of the NNS is to provide neutrons for the two cold sources located around the core, and the current compact core design delivers more neutron intensity and brightness for the cold neutron sources. Neutron transport analyses follow an optimization process for minimizing the difference between the k_∞ of the silicide plates to the U-10Mo, maximizing the coolant channel thickness and ²³⁵U content inside the assembly. Results show correlations between the fuel-to-coolant area ratio and the reactivity worth characteristics of the U-10Mo and U₃Si₂/Al plates in the NNS.

As a severe consequence of the Fukushima Daiichi Nuclear Accident in 2011, the hydrogen explosions are attributed to zirconium oxidation with steam surrounding the fuel claddings at high temperatures. This event encouraged the development of accident-tolerant fuels (ATF) worldwide. Active ATF research studies currently focus on using a Cr-based metal coating to protect the current Zr alloy tubing. However, concerning the properties of Cr like brittleness, materials modifications could be desirable that offer good performance in fabrication and reactor operations. Alloying of Cr is expected for possible improvements, and the desirable alloying elements are expected to maximize corrosion resistance along with sufficient mechanical properties, acceptable irradiation resistance, and sufficient protection at high temperatures. The objectives of this work are to synthesize Cr-based alloys and to clarify corrosion kinetics and related microstructure evolutions. Cr-based binary alloys with Fe, Al, and Sn were fabricated using vacuum arc melting, followed by heat treatment and surface preparation. The specimens were subjected to batch autoclave corrosion in distilled water at 360°C and 17.8 MPa for duration up to 56 days. The corrosion rates were determined by measuring the weight gains of the samples and fitted to the oxidation law. It was observed that Cr-0.5Sn and Cr-2Al have similar kinetic corrosion performance compared to pure Cr. Cr-2Fe and Cr-7Fe showed similar kinetic corrosion performance but slightly higher corrosion rates. Kinetic behaviors of Cr-3Sn and Cr-9.5Al did not follow the expected exponential law. Microscopic observations also indicate the uniform corrosion in the most of samples. Exceptions are Cr-3Sn and Cr-9.5Al which exhibited Sn dissolution resulting in surface void formation, and redistribution of Al, respectively. TEM observations and GIXRD analysis indicated formation of oxide layers such as Cr₂O₃, Fe₃O₄, Fe₂O₃, SnO, SnO₂ and Al₂O₃ in the specimens. By comparing with Zr alloys, significant improvement in corrosion behavior was evident in Cr and its binary alloys.

Helium is generally used as a coolant in Very-High-Temperature Reactors (VHTRs), but a small amount of impurity gas will inevitably be mixed in the primary coolant during construction, operation, and maintenance. At high temperature, these impurity gases will corrode with the alloy materials of the intermediate heat exchanger, resulting in the decline of the properties of the superalloy materials. This paper mainly studied the effect of methane in helium on the tensile property of Alloy 617 and Alloy 800H. Two impure helium environments are designed, one with low methane content (10 ppm), called He1, and the other with high methane content (350 ppm), called He2. The methane content of six typical international hightemperature reactors under normal operating conditions is generally not higher than 20 ppm. The two alloys were corroded at 950°C for 50 h in two different environments and the analysis showed that in He-2, Alloy 617 underwent severe carburization and its strength and plasticity were drastically reduced. Alloy 800H had no significant change in tensile properties due to its excellent resistance to carburization.

This paper aims to establish a general resonance calculation method under various neutron spectrum characteristics. The subgroup method with the narrow approximation is chosen as resonance calculation method based on the ultra-fine group structure. In order to consider the heterogeneous resonance effect, the heterogeneous resonance calculation is performed with the method of characteristics. Besides, the embedded self-shielding method is applied to handle the interference effect between different resonance nuclides. Serval typical problems with different neutron spectrum characteristics, including thermal spectrum, intermediate spectrum and fast spectrum problems, are designed to verify the effectiveness of resonance calculation. The numerical results show that the chosen methods are applicable to all problems. It also revealed that the mesh grid of resonance calculation could be coarser than that of transport calculation.

Stability analysis can be performed by transient analysis using different system codes which model the phenomena in detail. These system codes can capture the intrinsic nonlinearities in the system but the simulations have to be performed for different values of parameters and initial perturbations. This becomes a computationally intensive work and the estimation of the stability characteristics for the entire domain is difficult using this method. In this study, the effect of Core length (L)/Core diameter (D) ratio on stability characteristics of different modes of spatial Xenon oscillations in a large Pressurized Heavy Water Reactor has been analyzed using multipoint kinetics coupled with Xenon and Iodine equations.

This research aims to increase the efficient usage of nuclear fuel. It analyses the possibility of axial optimization of a fuel pin containing a Gd₂O₃ burnable absorber. The work deals with the effect of shortening (one-sided and double-sided) the fuel pin with Gd₂O₃ burnable absorber. Analysis of the effect on the power distribution was presented. Furthermore, it investigates the effect on stability during xenon oscillations and selected reactivity coefficients. The study performed a set of variations with different axial shortening of the Gd₂O₃ fuel and replacement of the fuel parts without the burnout absorber. For the evaluation, a special reference fuel loading pattern was designed and calculated by MOBY-DICK diffusion macro-code.

The online Burnup Measurement System (BUMS) of High Temperature Gas Cooled Reactor Pebble Bed Module (HTR-PM) is used to measure the gamma spectra of the fuel spheres during the fuel cycle. However, it focuses on only part of gamma rays and the corresponding nuclide activity, which are used to calculate the burnup. Therefore, the acquired data has not been sufficiently and explicitly utilized, which could have been used for validation of nuclide inventory analysis codes, estimation of irradiation history of fuel spheres and nuclear material accounting.

In this paper, relative detection efficiency curves and attenuation corrections of the components of BUMS were obtained from Monte Carlo photon transport code. Combined with NUclear Inventory Tool (NUIT, a code system for burnup calculation developed by the Institute of Nuclear and New Energy Technology, Tsinghua University), this paper calculates the activity of each nuclide in the fuel sphere based on anticipated power history of HTR-PM and gamma spectrum counting information. From these counts, a list of nuclides that is measurable by BUMS is screened.

NUIT and genetic algorithm (GA) are combined to find the possible irradiation history, which is an unknown quantity as in actual operating condition, recorded power history is always subject to large systematic errors because of unlocatable characteristic of fuel sphere flow movements. The nuclide densities simulated by the genetic algorithm are taken as predicted values, while the nuclide densities calculated by NUIT are taken as target values. The mean absolute percentage error between the two is the objective function. The genetic algorithm performs iterations over predicted irradiation history until nuclide densities match targets. The irradiation history calculated by this method is compared with that at target value of power. This study helps to refine BUMS, increase the accuracy of estimating the irradiation history of fuel spheres and improve the operation strategy of the pebble bed high-temperature gas-cooled reactors (HTGR).

Compared with the standard UO₂-Zr system, accident-tolerant fuel (ATF) can improve the margin of safety in extreme conditions, and therefore it is one of the most promising nuclear fuels for future reactors. Combined with technological and commercial information, patents are an important source of knowledge, and provide essential information relating to the policies and development focus within the technology subject. This paper studied the globally published patents relating to ATF. The leading applicants, their geographic distribution, and the technology focus were extensively investigated. The results showed that China and the USA were both the primary target markets and the primary technology origin countries. Approximately 40% of ATF patents were related to fuel cladding, and the remainder were related to fuel pellets. In terms of fuel cladding, zirconium (Zr)-based alloys with protective coatings were the predominant technological focus. For fuel pellets, fully ceramic microencapsulated (FCM) fuel was the predominant research topic. Each patent applicant utilized alternative approaches to patenting strategies. This paper significantly enhances the understanding of advanced developments in ATF-related technology by analyzing patents from an alternative perspective. These results will assist with decisions relating to the future of ATF.

In nature, each element corresponds to multiple isotopes. Different isotopes have the same arrangement of extranuclear electrons, and thus they are supposed to have the same chemical properties. Obviously, isotopes cannot be separated by chemical methods. However, there are still differences between isotopes. The isotope effect can be defined as the difference of various isotopes on chemical equilibrium, chemical reaction rates, and extranuclear electron energy levels. It reveals the effects on the behavior of atoms due to the mass and volume effects of the nucleus, but does not involve the effects caused by the radioactivity of the nucleus. For radionuclides, radioactive decay leads to ionization or excitation of daughter atoms, which in turn may arise the rupture or formation of chemical bonds. In this paper, by focusing on β decay, we explained the mechanism of the atomic-level changes caused by self-decay, and briefly summarized the unique behaviors of radionuclides in four cases: diffusion, adsorption, bond rupture and oxidation, respectively. A possible application of the β-decay in adsorption and diffusion process and synthesizing new crystalline compounds has been discussed. Current research can shed a light on the further study of the self-decay effect of radionuclides.

This work establishes a finite-difference lattice Boltzmann method (FDLBM) to solve the Boltzmann transport equation (BTE) to improve the accuracy and stability of the standard lattice Boltzmann method (LBM). The BTE is discretized using the FDLBM equation with adjustable spatial and time discretization, and the relationship between the FDLBM equation and the BTE is obtained using the Chapman-Enskog expansion. Since the FDLBM has the same data structure as the LBM, it can be easily combined with the standard LBM processes under the LBM framework. Numerical results show that the proposed FDLBM has higher accuracy and stability than traditional methods. This work can provide some ideas for improving the LBM technique under the unified LBM framework. For further researches, this work can provide some new perspectives of higher-accurate neutral particle transport calculations and an optional multi-physics technique for largescale engineering calculations.

The main reason that restricts the increase of fast reactor burn up is the radiation swelling of the cladding. The reported research shows that the main elements and trace alloy/impurity elements have a great influence on the radiation swelling. However, the behavior of the above elements in the process of coupling with radiation defects is complex, and it is difficult to directly measure the relationship between these elements, radiation defects and microstructure evolution in experiments. The emergence of machine learning and big data mining technology will help to gain new understanding of the impact of radiation swelling on austenitic stainless steel, so as to find a new type of austenitic stainless steel cladding material resistant to radiation swelling. Therefore, in this work, about 1000 groups of data such as composition, irradiation conditions and radiation swelling of austenitic stainless steel are collected, and the data are cleaned and screened for modelling by machine learning. The deep neural network with back propagation is used in this work, and the correlation between alloy composition such as Cr, Ni, Ti and C, irradiation dose and temperature and radiation swelling of austenitic stainless steel is established. The results show that the addition of a certain amount of Ti and Si can effectively inhibit the radiation swelling of austenitic stainless steel, but the addition of Ni will aggravate the swelling effect. The addition of Cr, Ni, Ti and Cr will increase the swelling inflection point dose, while the addition of C and P will reduce the swelling inflection point dose. Besides, the influence of multi factor coupling such as composition on radiation swelling of austenitic stainless steel will help to promote the material optimization design of austenitic stainless steel cladding material.

In a space nuclear reactor with a heat pipe cooling core, the heat transfer limit of the heat pipe is an important factor affecting the reactor power. In heat pipes with wicks, the capillary limit is one of the most common limits. The capillary capacity of the wick is the key factor affecting the capillary limit of the heat pipe. The wettability of the working fluid and the wick material greatly affects the capillary capacity of the wick in the heat pipe. In this paper, the wettability of liquid sodium on 304 stainless steel was studied, and the change of contact angle of sodium on 304 stainless steel plate with temperature was measured. The experiment was carried out in a high-temperature vacuum contact angle measuring instrument. The vacuum degree of the experiment was 0.06 MPa. The experiment found that before 300 °C, the contact angle is maintained above 140 °, showing a non-wetting state. This is due to the stainless steel plate surface with a dense layer of Cr₂O₃ oxide film, which hinders the wetting between liquid sodium and stainless steel. After 300 °C as the temperature increases, the contact angle decreases. The overall change trend of contact angle shows a process of rapid decline first, then slow decline, and finally rapid decline. With the increase of temperature, Cr₂O₃ is gradually reduced by liquid sodium to form stable NaCrO₂. The wettability between liquid sodium and a stainless steel plate was improved. At 420 °C, the contact angle has reached 50 °. When the temperature reaches 530 °C, the contact angle of sodium and stainless steel plate size changes to 0°, reaching a state of complete wetting.

Effects of burnable poison (BP) addition on radiotoxicity by TRU were investigated for FORSETI fuel, which is assumed to be loaded into a BWR plant. Since the purpose of this study was to confirm the relative change in radiotoxicity due to TRU, the weight ratio of TRU was used as an index of the relative magnitude of radiotoxicity. The weight fraction of TRU nuclides with ²³⁵U enrichment from 3.8 wt% to 20 wt% was examined with and without BP. The weight fraction of TRU nuclides in the spent fuel decreased by about 60% for ²³⁸Pu, ²⁴⁰Pu, and ²⁴¹Pu. The results of these nuclides are almost the same between with and without BP. On the other hand, ²³⁹Pu, which is not dominant in the toxicity of TRU, is slightly increased by the addition of BP. The relative difference in TRU weight fraction between with-BP and without-BP by increase of ²³⁵U enrichment does not exceed 20wt% except for ²³⁹Pu.It is desirable to consider the error in the amount of TRU generation when evaluating without BP for analytical simplicity.

After the fuel element is damaged, the delayed neutron precursor flows into the coolant, which will break the inherent balance of the normal core, thus causing the transient change of neutron noise. Based on the neutron noise technology, it is expected to break through the key technology of online monitoring of fuel element damage in the reactor. Based on this idea, this work plans to investigate the fuel failure conditions based on the frequency domain finite element method, focusing on the neutron noise dynamics calculation under the flow scenario of delayed neutron precursor, it can provide some reference and technical support for related work.

Models using physical computing models and global search algorithms are well used in the nuclear core inverse problems. In the framework of such a problem, the 3D core physics analysis code CPACT and observation data from detector readings are used to infer and renew the model parameters by global optimization methods. In this way, the operating state of the core can be obtained.

However, uncertainties in parameters, inputs, and observation readings have undermined its application value. The objective of this work is to develop a Bayesian framework of core calculation models to estimate and diagnose uncertainties. Besides, taking into account the impact of uncertainty, more convincing objective functions are proposed to evaluate the problem. We test this uncertainty in this way: use the optimization method to directly solve the model parameters that best match under the maximum likelihood estimation; All the methods were tested on the numerical calculation examples of a pressurized water reactor case.

Dispersed pellet fuels and plate fuel elements in new pressurized water reactors have good prospects for application. Dispersed fuel has the advantages of extremely infusible character, strong fission product retention, high oxidation resistance, and mainly exists in the form of pellets. Meanwhile, reactivity is a crucial issue during the operation of the reactor, which needs to be well controlled throughout its life cycle, and the implementation of burnable poison is one of the most effective measures. Therefore, in this paper, based on the Monte Carlo method program OpenMC, a two-dimensional plate fuel grid element is constructed with a single plate in the MTR assembly as the study object , and the dispersed particles are set up and then the volume is homogenized. As a comparison, both two methods have the same material with the same mass, and the conventional volume homogenization method will overestimate the neutron absorption effect, which will lead to a large deviation in the reactivity calculation. At the same time for different setting conditions deviation is different, with the increase of enrichment, particle size increases, packing factor decreases its deviation will gradually increase; the increase of the temperature of the moderator affects the temperature coefficient of the moderator and thus makes the reactivity deviation change but the trend is not obvious, at the same time the dispersed particles with poison will also cause deviation, and different poison affects differently.

The XML-based GNDS has been emerging as the state-of-art format for nuclear data. However, the XML file format has complicated structures, and the encoding and decoding of which bring a handful of computing burdens. In this work, the so-called GNDS-JSON format is proposed using JSON instead of XML as a variant data format to the coming GNDS 2.0 standard. With reduced data structural complexity, the data structure of this new format matches naturally with the syntax of most programming languages. Without providing users vendor APIs, the data infrastructure is thinned and entitles users to a nonintrusive integration of data access programming modules, while not sacrificing user experience. Moreover, compared to existing JSON schema for the GNDS data in the XML format, a new JSON schema is proposed for direct validation of data in the GNDS-JSON format. As an initial attempt, a set of minimal JSON schema has been demonstrated, which is enough for the description of the neutron reaction data of 364 (out of 372) materials from BROND 3.1, 272 materials from the CENDL 3.2, 557 materials from the ENDF/B-VIII.0, 550 (out of 562) materials from JEFF 3.3, 406 materials from JENDL 4.0, 795 materials from JENDL 5.0, and 628 (out of 2,813) materials from TENDL 2021. The set of new JSON schema, converted data files in GNDS-JSON and example data accessing codes are opensourced for community feedback. A resonance cross section reconstruction code is in-house developed and verified again Fudge to demonstrate the application of the GNDS-JSON format in nuclear data processing.

Open-source repository: jihulab.com/newcomputelab/gndsjson

Accurate cognition of the neutron reaction rule is an important guarantee of the safety and economy of nuclear power plant. Since the material composition and geometry configuration of the new type reactor has gradually become more complex, it brings severe challenges to the high-fidelity reactor physics calculation. The resonance self-shielding calculation provides effective cross section for the neutron transport and depletion calculation, so it is a key point in the reactor physics calculation. The precision of resonance selfshielding cross section is an important basis for analyzing neutron behaviors in the reactor. The subgroup method has relatively high efficiency and good geometry applicability, so it is commonly applied to commercial core analyzing programs. The accuracy of the subgroup method depends on the subgroup parameters, which are generated by fitting the multivariate equations based on the resonance integrals. Therefore, the definition method of resonance integrals greatly influences the accuracy of the subgroup method.

Conventional resonance integral is generated based on a series of homogenous systems consisting of resonant nuclide and moderator nuclide, which is called the homogenous integral. Heterogeneous resonance integral is another type resonance integral which is generated by typical heterogeneous pin or slab problems. The subgroup parameters could be calculated either by homogenous or heterogeneous resonance integrals. However, the influence of these two kinds of resonance integrals for subgroup resonance calculation is still not clear. Since the new type of reactor is getting more and more complex, it is necessary to carry out the research to evaluate this influence. In this work, both the homogenous and heterogeneous resonance integrals are generated to fit the subgroup parameters, and a series of problems are calculated to analyze the accuracy. The calculating results display the application range of each kind of resonance integral. Finally, the potential improving method is also discussed. Better accuracy of subgroup parameters could be achieved by adopting the improved fine group structure.

Tritium is one of the most concerned nuclides in nuclear power plants because of its high mobility and circulation in the environment. The high-temperature gas-cooled reactor pebblebed module (HTR-PM) constructed in Rongcheng City, Shandong Province, China, is the first stage of the very hightemperature gas-cooled reactor (VHTR), which has been identified as one of the six Generation IV nuclear energy systems by Generation IV International Forum (GIF). Since it can diffuse from the primary loop to the secondary loop through heat exchanger tube, the control of tritium is significant for the high temperature thermal application of high-temperature gas-cooled reactors (HTGRs). Graphite is regarded as tritium sink to decrease the tritium activity in the primary loop due to its strong adsorption capability. Some researches on interaction between tritium and graphite indicate that tritium distributes mainly on the graphite surface layer, while in the interior the specific activity of tritium is almost constant. After oxidation, graphite can trap more tritium. However, the ratio of tritium in the surface to bulk decreased because of more tritium access from graphite surface to the bulk. In this paper, first-principles density functional theory was implemented to study the effect of oxidation on tritium adsorption at the graphite zigzag edge. The adsorption energy of oxygen and tritium at the zigzag edge of single graphene sheet was calculated. The influence of adsorbed oxygen on tritium adsorption was investigated. The electronic structure and charge density were presented to illustrate the bonding properties between zigzag edge carbon, oxygen, and tritium. Current study can provide a preliminary investigation on the mechanism of oxidation effect on tritium adsorption at graphite zigzag edge from an atomic scale.

In this paper, we investigated the feasibility of producing Mo-99 in a LEU-fueled molten-salt reactor. The reactor was based on the ORNL’s MSRE (Molten-Salt Reactor Experiment) prototype, which was originally designed to operate with HEU fuel. In order to compensate for the loss of reactivity as a result of the fuel conversion, necessary modifications to the core design were made in the aspects of fuel content, fuel-to-graphite ratio, core temperature, and the use of reflectors. A multi-irradiation scheme was proposed for the purpose of Mo-99 production. The neutronics calculation was performed using the SERPENT, which was benchmarked against JMCT and historical MSRE design reports by ORNL.

In the current work, several newly developed austenitic stainless steel with minor additives Sc have been developed based on quantum chemical molecular dynamics method and density functional theory. The oxide films formed on the alloys surface in BWR and PWR environment are characterized by SEM, Raman spectroscopy, and XPS. The effect of Sc addition on the SCC behaviour of modified 304L in PWR and BWR environment were investigated by using SSRT technique, and the SCC susceptibility of the Sc added new alloys in high temperature water are evaluated. It is found that the Sc additive can reduce the thickness and defect density of the oxide film compared with the reference alloys. SCC tests show that the additives can also decrease the SCC susceptibility in terms of lower SCC fracture area ratios and higher strength. The SCC resistance decreased with Sc addition increasing. Sc seems to be most effective with the amount of 0.05%-0.1% for 304L. The new alloys seem promising for retarding both oxidation and SCC in high temperature water environment.