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Since the war in Gaza began, there has been a growing chorus calling for using nuclear weapons. This raises the question: Should those individuals calling for the use of nuclear weapons in Gaza be prosecuted for violating the statute of incitement to genocide under Article III of the Genocide Convention of 1948 or are these statements simply a case of free speech? Are these individuals advocating for nuclear war or inciting? Is calling for the use of nuclear weapons, the same as calling for genocide? This article will attempt to answer these questions focusing on the relationship between nuclear weapons and incitement to genocide.

BackgroundThe 1.2 MeV/10 mA electron accelerator, as one of electron irradiation sources for comprehensive irradiation test chamber in Space Environment Simulation and Research Infrastructure (SESRI), can provide electron beams on the millimeter scale. However, the electron beam, as the electron irradiation source in space environment ground simulation experiment for aerospace, must be uniformly irradiated on large objects. Therefore, a well-suited irradiated technique is significant.PurposeThis study aims to obtain beam scanning with less than 10% of inhomogeneity for 0.6 MeV, 1.0 MeV, 1.2 MeV electron beam for SESRI.MethodsBased on the overall irradiation requirements, a specific beam-scanning system, including the scanning magnet with customized design, digital power supply and a dedicated apparatus for beam uniformity measurement, was developed. Particularly, in order to eliminate the influence of 45° incidence of electron beam upon scanning uniformity, an asymmetric and non-standard triangular waveform for the magnetic field excitation current was employed and implemented. The measurements for the beam non-uniformity were carried out on field experiments.ResultsExperimental results show that the scanning area of this electron accelerator reaches 1 000 mm×1 000 mm, and the scanning nonuniformity is less than 10% for variable beam energy from 0.6 MeV to 1.2 MeV, achieving the design goal and satisfying the irradiation requirements of SESRI.ConclusionsA specific beam-scanning system developed for SESRI is verified in this study, offering a good reference for any similar beam-scanning scenarios.

Pakistan has a long history of social movements, some of which have challenged the military’s domination. However, apart from a short period after 1998 nuclear weapon tests by India and Pakistan, there has been limited opposition to nuclear weapons owned by the military. This paper examines one potential factor for this near absence: the intellectual arguments deployed by the main proponents of nuclear disarmament in Pakistan over the last 25 years. The post-1998 anti-nuclear critique in Pakistan has overwhelmingly focused on challenges pertaining safety and security, command and control, geopolitical imperatives of arms races, the operational limits of delivery systems like the short-range ballistic missiles and the risk of a nuclear war. We argue that such arguments inadvertently normalize nuclear weapons. Other arguments based on geopolitics make nuclear weapons appear inevitable in an anarchic world. Highlighting the consequences of a nuclear war has limited purchase in Pakistan where nuclear weapons are seen as the ultimate defense against all large-scale wars, nuclear and conventional. The alternative, we suggest, is to engage a broader critique focussed on the many ways in which nuclear weapons are implicated in everyday social and political life.

This study examines the effect of perceived military threat on Japanese people’s opinion about Japan’s accession to the Treaty on the Prohibition of Nuclear Weapons. While nuclear aversion has been historically strong and deep-rooted in Japan, we expected that people would hesitate to support the treaty when reminded of the deteriorating security environment, considering recent experimental evidence suggesting that nuclear aversion is conditional and tends to fluctuate often and that Japanese citizens tend to be apathetic towards international laws. Moreover, we examine the possibility that pressure from foreign countries and international organizations reinforces people’s support for the treaty. Our survey experiment reveals that Japanese people are less likely to support Japan’s accedence to the treaty when they are primed with security threats, although the difference is small, and that this negative effect is limited to centrist and conservative citizens. However, we find no evidence that international pressure has the power to promote people’s support for the treaty.

When solving several research problems, the need arises for reactor testing of non-standard objects. The IVV-2M reactor is a pool, heterogeneous, light-water reactor. The 500 mm-high reactor core comprises six sections of six fuel assemblies (FA) with central water traps and a water trap in the center of the core. The personnel of INM JSC were tasked with testing an object whose length exceeded the height of the IVV-2M reactor core. A non-standard core layout was developed with six sections of seven fuel assemblies to accomplish this task.. The test object is located in the center of the core, surrounded by special aluminum blocks. Several fuel assemblies are installed on special supports, which allow the fuel assemblies to be installed 200 mm higher. Shim rods have been moved to the center of the sections. The safety of the developed arrangement was confirmed by neutron-physical calculations using the certified MCU-PTR software tool. Calculations have shown that the maximum maximum power in the fuel assembly is 236 kW, which is 50% less than the permissible value. To confirm the safety of the layout and ensure the required test conditions, tests of a mock-up of the object were carried out. The efficiency of the control members was experimentally determined. The overall efficiency of the control members was 4.85% ∆k/k, the physical weight of the object was 1.46% ∆k/k. Based on the requirements of the Nuclear Safety Rules for Research Reactors, the test duration is no more than three days. The distribution of neutron flux density over the height of the core was determined using indium neutron activation detectors. The developed design solutions made it possible to form a unique IVV-2M reactor core layout for testing objects exceeding its height.. Neutron-physical calculations and tests of the irradiated device mock-up confirmed the safety and performance of this core arrangement.

This article investigates alternate fuel options for Pressurized Water Reactors (PWRs), focusing on thorium use to address safety, efficiency, and waste issues associated with standard UO2 fuel. Challenges in thorium utilization, such as the lack of a fissile isotope, are handled using approaches such as homogeneous mixtures and heterogeneous arrangements, promoting the exploration of (Th-233U-235U)O2 fuel in various assembly configurations. According to recent research, the annular dual-cooled assembly design has promising results in terms of fuel efficiency and safety while lowering the requirement for higher fissile enrichment levels. Studies additionally demonstrate that annular dual-cooled duplex fuel configurations can produce higher discharge burnup and lower power peaking factors than traditional UO2 fuel. The purpose of this work is to analyze and compare the performance of (Th-233U-235U)O2 fuel in various configurations against conventional UO2 fuel, focusing on key characteristics such as reactivity change, criticality, discharge burnups, and reactivity feedback coefficients.

The successful conclusion of the Treaty on the Prohibition of Nuclear Weapons (TPNW) has challenged the global nuclear order, but was dependent on bypassing the nuclear weapon states (NWS) and their allies. This has led critics to dismiss its practical value for nuclear disarmament. Proponents of the ban treaty, however, hope for a “transformative political dynamic” that delegitimizes nuclear weapons and leads to NWS and their allies joining the TPNW. This article contributes both conceptually and empirically to the discussion of this “transformative political dynamic”. It deals with the role of nuclear weapons for NATO membership and the strategies for legitimizing and delegitimizing nuclear weapons. The case study on Belgium analyzes the processes of coalition building and the discursive strategies related to Belgium’s nuclear role in NATO. The study shows that the exchange of symbolic arms control and disarmament initiatives for the approval of military budgets and the acquiescence to the claimed nuclear obligations of NATO membership stands on shaky ground.

Japan, the only country to have experienced atomic bombs during wartime, is the only one among the US allies to have positively evaluated the Treaty on the Prohibition of Nuclear Weapons (TPNW) at the highest political level, with Prime Minister Kishida saying that as “a very important treaty that can serve an exit as we seek to realize a world without nuclear weapons”. Yet Japan has distanced itself from the TPNW, not participating even as an observer in the TPNW’s Meeting of States Parties. This paper attempts to explain Japan’s stance toward the TPNW process, before and after the formation of the treaty, based on a hypothesis that Japan’s nuclear policy is formed by a balancing act between two factors: humanitarian and security concerns. It then attempts to chart a path forward for any possible engagement between Japan and the TPNW. It argues that a substance-over-formality approach, not sticking to the TPNW formalities but in a framework separate from the TPNW, may open a window of opportunity for cooperation. Japan can also flexibly consider security benefits in engaging with the TPNW and comprehensively calculate possible security benefits and drawbacks without a predetermined conclusion. Since the NPT may not be sustainable unless there is a significant progress in nuclear disarmament, which itself is unlikely, like it or not, the TPNW cannot be ignored indefinitely and the time may come in the not-too-distant future when Japan will eventually have to consider engagement with the TPNW.

High-order harmonic techniques can be used to recreate neutron flux distributions in reactor cores using the neutron diffusion equation. However, traditional source iteration and source correction iteration techniques have sluggish convergence rates and protracted calculation periods. The correctness of the implicitly restarted Arnoldi method (IRAM) in resolving the eigenvalue problems of the one-dimensional and two-dimensional neutron diffusion equations was confirmed by computing the benchmark problems SLAB_1D_1G and two-dimensional steady-state TWIGL using IRAM. By integrating Galerkin projection with Proper Orthogonal Decomposition (POD) techniques, a POD-Galerkin reduced-order model was developed and the IRAM model was used as the full-order model. For 14 macroscopic cross-section values, the TWIGL benchmark problem was perturbed within a 20% range. We extracted 100 sample points using the Latin hypercube sampling method, and 70% of the samples were used as the testing set to assess the performance of the reduced-order model The remaining 30% were utilized as the training set to develop the reduced-order model, which was employed to rebuild the TWIGL benchmark problem. The reduced-order model demonstrates good flexibility and can efficiently and accurately forecast the effective multiplication factor and neutron flux distribution in the core. The reduced-order model predicts keff and neutron flux distribution with a high degree of agreement compared to the full-order model. Additionally, the reduced-order model's computation time is only 10.18% of that required by the full-order model.The neutron flux distribution of the steady-state TWIGL benchmark was recreated using the reduced-order model. The obtained results indicate that the reduced-order model can accurately predict the keff and neutron flux distribution of the steady-state TWIGL benchmark.Overall, the proposed technique not only has the potential to accurately project neutron flux distributions in transient settings, but is also relevant for reconstructing neutron flux distributions in steady-state conditions; thus, its applicability is bound to increase in the future.

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

Abstrat: Modern risk management philosophy emphasizes the invulnerability of human beings to cope with all kinds of emergencies. The Nuclear Accidents Emergency Response Organization (NAERO) of Nuclear Power Plant (NPP) is the primary body responsible for nuclear accidents emergency response. The invulnerability of the organization to disturbance or attack from internal and external sources is crucial in the completion of its response missions, reduction of severity of accidents, and assurance of public and environmental safety. This paper focused on the NAERO of a certain NPP in China, and applied the complex network theory to construct the network model of the organization. The topological characteristics of the network were analyzed. Four importance evaluation indexes of network nodes including Degree Centrality (DC), Betweeness Centrality (BC), Closeness Centrality (CC) and Eigenvector Centrality (EC), along with Pearson coefficient correlation among the indexes were calculated and analyzed. Size of the Largest Connected Component (LCC) and Network Efficiency were used as measures regarding the invulnerability of the network. Simulation experiments were conducted to assess the invulnerability of network against various attack strategies. These experiments were conducted both in the absence of node protection measures and under protection measures with different node protection rates. This study evaluated the invulnerability of the NAERO network, and provided significant decision-making basis for the enhancement of the network's invulnerability.

Radiocarbon (14C), with a radioactive half-life of approximately 5730 years, poses a long-term environmental contamination risk when released into the atmosphere. The quantification analysis of its release estimates plant-specific generation rates based on factors such as plant power, core neutron flux distribution, and the volume of water exposed to this flux. Utilizing the improved estimation method, the 14C production rate for several Korean Pressurized Water Reactors (PWRs) was calculated. Also, improvements in measurement methods through sampling have also been made. These enhancements include the verification of the absorption method versus the mixing method.The results of this study indicate that plant-specific 14C production rates range from 0.213 to 0.317 TBq/yr, which are comparable to the global range observed in PWRs. Furthermore, the study evaluated a quenching correction curve for a liquid scintillation counter using two quenching correction methods: the external standard method and the internal standard method. The accuracy of these methods with 72 samples was validated with an average relative error within ±2.5%. The relative error of the mixing method, when compared to the direct absorption method, was found to be within ±20%. This finding underscores the validity of the improved measurement technique.

The Republic of Korea is performing independent national inspections under the IAEA's State System of Accounting for and Control (SSAC), and developing an evaluation methodology for the material unaccounted for (MUF) to reinforce capabilities with the purpose of assessment for the accounting system of the facility handling bulk nuclear materials. In relation to this, a new approach for MUF evaluation was proposed in this study based on the guide to the expression of uncertainty in measurement (GUM). Both the conventional MUF evaluation method and the GUM method were applied to a hypothetical list of inventory items including material balance. Considering the ease of uncertainty propagation according to the GUM, it was assumed that independent uncertainty factors correspond to random factors, while correlated uncertainty factors correspond to systematic factors. The total MUF uncertainties were similar for both methods; however, it was verified that some uncertainties were affected by the measurement procedure in the GUM method. Furthermore, the GUM method was found to be more conducive to conducting a factor analysis for the MUF uncertainty. It was therefore concluded that application of the GUM approach could be beneficial in cases of national safeguard inspections where factor analysis is required for MUF assessment.

The one-dimensional Two-Fluid concept uses an area-average approach to simplify the time and phase-averaged Two-Fluid conservation equations, making it more suitable for addressing difficulties at an industrial scale. Nevertheless, the mathematical framework has inherent weaknesses due to the loss of details throughout the averaging procedures. This limitation makes the conventional model inappropriate for some flow regimes, where short-wavelength perturbations experience uncontrolled amplification, leading to solutions that need to be physically accurate. The critical factor in resolving this problem is the integration of closure relations. These relationships play a crucial function in reintroducing essential physical characteristics, thus correcting the loss that occurs during averaging and guaranteeing the stability of the model. To improve the accuracy of predictions, it is essential to assess the stability and grid dependence of one-dimensional formulations, which are particularly affected by closure relations and numerical schemes. The current research presented in the text focuses on improving the well-posedness of the TFM, specifically within the TRACE code, which is widely utilized for nuclear reactor safety assessments. Incorporating a bubble collision model in the momentum equations is demonstrated to enhance the TFM's resilience, especially in scenarios with high void fractions where conventional TFMs may face challenges. The analysis presents a linear stability analysis performed for the transient one-dimensional Two-Fluid Model of system code TRACE within the framework of vertically dispersed flows. The main emphasis is on evaluating the stability characteristics of the model while also acknowledging its susceptibility to closure relations and numerical techniques.

In ITER, the evaluation of the activated water radiation source and its impact on the radiological levels is necessary to demonstrate compliance with the safety requirements. The use of simplified or conservative approaches often results in the application of expensive constraints on the installation that impact its economics, operations, and construction schedule. In this work, we propose a novel methodology to calculate the activated water source term with a higher degree of realism. The methodology is based on the coupling of a system-level code with a Computational Fluid Dynamics (CFD) code in an explicit, one-way approach. We apply this methodology to the evaluation of the 16N radioisotope within the ITER Vacuum Vessel Primary Heat Transfer System (VV-PHTS) cooling circuit in a steady-state and transient scenarios. We chose this system since previous analyses of the VV-PHTS were done with simple, ad-hoc calculations that yielded results that differed by up to a factor of five, underscoring a higher level of uncertainty. As a result, we generate a computational model of the source term that can be used to evaluate the radiological condition surrounding the cooling systems during the operations.

In memory semiconductors such as a static random access memory (SRAM), a common problem is soft errors under radiation environment. These soft errors cause bit flips, which are referred to as single event upsets (SEUs). Some radiation-hardened SRAM cells such as a Quatro SRAM, we-Quatro SRAM, and DICE SRAM cells have been reported for years. However, these designs have the disadvantage of taking up more area than a conventional 6T SRAM cell. Thus, we propose a radiation-hardened SRAM cell design that we named capacitor-static random access memory (C-SRAM) without area overhead. The C-SRAM is formed by simply adding a capacitor to the conventional 6T SRAM. It was designed to mitigate the radiation effect using the conservation law of electrical charge. Moreover, it has the same cell size as the conventional 6T SRAM cell. Its static noise margins (SNMs), which are indicators of operational stability, are equal to the conventional 6T SRAM values of 530 mV, 220 mV, and 860 mV in hold, read, and write modes, respectively. The results of the SEU simulation test showed that it had 4.761 times better flipping tolerance than the conventional 6T SRAM with a charge value of 247.494 fC. In addition, irradiation experiments also confirmed that the C-SRAM cell was more tolerant than the 6T SRAM cell. The conventional 6T SRAM and C-SRAM were fabricated using a standard 0.18 μm CMOS process.

Steel containment vessels for nuclear power plants can experience gas leakage due to minute defects such as cracks, corrosion, and aging, leading to gas leakage. A gas leakage model for microchannels is established to elucidate the mechanism underlying gas leakage within microchannels caused by these defects, specifically addressing the issue of unidirectional gas flow. Computational Fluid Dynamics (CFD) and the UK R6 method are employed to calculate the gas leakage rate within microchannels. Furthermore, the characteristics of gas flow within microchannels are explored, including the factors affecting the gas leakage rate. Validation of the calculation results is verified experimentally. The results indicate that the gas mass flow rate exhibits a linear decrease with decreasing internal pressure and a non-linear decline as temperature increases. Additionally, the gas mass flow rate demonstrates a negative correlation with the microchannel length but a positive association to its hydraulic diameter. The primary influencing factors on gas leakage rates are hierarchically ranked as follows: pressure difference, microchannel cross-sectional area, temperature, microchannel length, and microchannel hydraulic diameter.

Sliding bearings are crucial rotating mechanical components in nuclear power plants, and their failures can result in severe economic losses and human casualties. Deep learning provides a new approach to bearing fault diagnosis, but there is currently a lack of a universal fault diagnosis model for studying bearing-rotor systems under various operating conditions, speeds and faults. Research on bearing-rotor systems supported by sliding bearings is limited, leading to insufficient fault data. To address these issues, this paper proposes a fault diagnosis model framework for bearing-rotor systems based on a deep convolutional generative adversarial network (TF-DLGAN). This model not only exhibits outstanding fault diagnosis performance but also addresses the issue of insufficient fault data. An experimental platform is constructed to conduct fault experiments under various operating conditions, speeds and faults, establishing a dataset for sliding bearing-rotor system faults. Finally, the model's effectiveness is validated using this dataset.

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48–97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56–70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

Crevice corrosion is a localized attack of metal that occurs in occluded areas of materials as a result of a degradation of the oxide passivity on the metal surface in contact with stagnant environments. Materials suffer crevice corrosion when generally the crevice opening gap is so narrow that the migration or diffusion of ionic species into the crevice can be restricted and consequently results in the production of aggressive crevice solutions and differential aeration conditions over time. Among several factors affecting the crevice corrosion, differential aeration causing oxygen depletion associated with the geometry of components, acidification, and accumulation of aggressive species (e.g., Cl−, SO4−2, NO3−) in the crevice solution become main aspects of the mechanism of the crevice corrosion. Thus, controlling such factors is most critically necessary to either prevents or terminates the crevice corrosion. This paper covers electrochemical aspects of the crevice corrosion, roles of critical factors affecting the crevice corrosion, and electrochemical processes of impurity species in the crevice in high temperature water. A better and clear understanding of mechanisms of the crevice corrosion is important to develop the protection and mitigation technology against the crevice corrosion in order for maintaining the integrity and longevity of structural components at various industries

To improve the predictive capability of a nuclear thermal hydraulic safety analysis code by developing a better constitutive equation for individual phenomenon has been the general research direction until now. This paper proposes a new method to directly use complex experimental data obtained from integral effect test (IET) to improve constitutive models holistically and simultaneously. The method relies on the sensitivity of a simulation result of IET data to the multiple constitutive equations utilized during the simulation, and the sensitivity of individual model determines the direction of modification for the constitutive model. To develop a robust and generalized method, a clustering algorithm using an artificial neural network, sample space size determination using non-parametric statistics, and sampling method of Latin hypercube sampling are used in a combined manner. The value of the proposed methodology is demonstrated by applying the method to the ATLAS DSP-05 IET experiment. A sensitivity of each observation parameter to the constitutive models is analyzed. The new methodology suggested in the study can be used to improve the code prediction results of complex IET data by identifying the direction for constitutive equations to be modified.

In this work, single photon avalanche diodes (SPADs) were fabricated using the standard 180 nm complementary metal-oxide semiconductor process. Their small size of 15–16 μ m and low operating voltage made it possible to easily integrate them with readout circuits for compact on-chip sensors, particularly those used in the radiation sensor network of a nuclear plant. Four architectures were proposed for the SPADs, with a shallow trench isolation (STI) guard ring and different depletion regions designed to demonstrate the main performance parameters in each experimental configuration. The wide absorption region structure with PSD and a deep N-well could achieve a uniform electric field, resulting in a stable dark count rate (DCR). Additionally, the STI guard ring was implanted to mitigate the premature edge breakdown. A breakdown voltage was achieved for a low operating voltage of 10.75 V. The DCR results showed 286.3 Hz per μm2 at an excess voltage of 0.04 V. A photon detection probability of 21.48% was obtained at 405 nm.

An electron beam may be used to generate bremsstrahlung photons that go on to create photoneutrons within metals. This serves as a low-energy neutron source for irradiation experiments. In this article, we present simulation results for optimizing photoneutron yield for a 10-MeV electron beam on tantalum foil and erbium deuteride (ErD3). The thickness of the metal layers was varied. A tantalum foil thickness of 1.5 mm resulted in the most photons reaching the second metal layer. When a second metal layer of ErD3 was included, the photoneutron yield increased with the thickness of the secondary layer. When the electron beam was directly incident upon a layer of ErD3, the photoneutron yield did not differ significantly from the yield when a layer of tantalum was included. The directional photoneutron yield reached a maximum level when the thickness of the ErD3 layer was around 12 cm. About 1 neutron was generated per 104 source electrons. When using a 2-mA beam current, it is possible to generate up to 1012 neutrons per second, making this combination a relatively-inexpensive neutron generator.

Static corrosion tests in LBE under oxygen-saturated and –depleted conditions at 500 °C for 1000 h were conducted to investigate the corrosion-induced diffusion layers on T91 and SS316L steels subject to dissolution and oxidation corrosion. Following nano-indentation experiments examined the variations of local mechanical properties within the corrosion layers. Under the present conditions, T91 steel showed better resistance to dissolution corrosion while higher susceptibility to oxidation corrosion against SS316L. For both steels, nano-indentation results showed a gradual decrease in mechanical properties from steel matrix to the corrosion side after dissolution corrosion. When subject to oxidation corrosion, the oxide scale formed on T91 steel showed lower Er-modulus and enhanced hardness in comparison with those of the steel matrix. An obvious decrease in both modulus and hardness was found at the interface between steel matrix and oxide scale.

Since the level 2 PSA of OPR-1000 was the requirement for regulatory purposes, Cs-137 release estimation was contained as the Nuclear Safety Act of ROK in which the Cs-137 release frequency exceeding 100 TBq was determined to happen less than 1.0E-6 per year after the Fukushima Daiichi Accident. However, Cs-137 release estimation from the conventional level 2 PSA of OPR-1000 provided uncertainty due to dominant accident sequence consideration. Thus, this study aimed to develop systematic methods through the overall framework to quantify realistic uncertainty concerns of radioactive material release using sensitivity and uncertainty analysis methods and apply them to OPR-1000. This framework helped to quantify confidential value for the Cs-137 release under the BEPU approach using both parametric and non-parametric methods to cover both realistic and conservative points. Uncertainty propagation analysis showed the unexpected uncertainty increase of Cs-137 release exceeding 100 TBq. The non-parametric uncertainty analysis provided higher conservative concerns for safety than the realistic concerns in terms of economics when compared with the parametric uncertainty analysis. Wilks’ uncertainty analysis showed the importance to consider conservative Cs-137 release in order to reach the higher safety need. Sensitivity analysis showed reasonable relationships between engineering safety parameters with the Cs-137 release.

The dicentric chromosome assay is a gold standard method to estimate radiation exposure by calculating the ratio of dicentric chromosomes existing in cells. The objective of this study was to propose an automatic dicentric chromosome discrimination method based on deep convolutional neural networks using radiation exposure patient data. From 45 patients with radiation exposure, conventional Giemsa-stained images of 116,258 normal and 2800 dicentric chromosomes were confirmed. ImageNet was used to pre-train VGG19, which was modified and fine-tuned. The proposed modified VGG19 demonstrated dicentric chromosome discrimination performance, with a true positive rate of 0.927, a true negative rate of 0.997, a positive predictive value of 0.882, a negative predictive value of 0.998, and an area under the receiver operating characteristic curve of 0.997.

Stratified flow in horizontal tubes is frequently observed in gas-liquid two-phase flow system. In the two-fluid modeling, it is important to define the interface shape in solving the balance equations to determine the key parameters such as the interfacial transfer terms, void fraction, and pressure drop. A double-circle model is usually introduced to depict the concave-down interface in a horizontal circular tube under the stratified-wavy flow condition. However, calculation of the central angle in the double-circle model, which represents the interfacial curvature, requires an appropriate iterative numerical root-finding scheme to solve the implicit transcendental equation. In this study, an explicit approximate equation has been proposed without requirement of the iterative scheme and numerical instability, which is expected to improve the coding process and computation efficiency in the analysis code with the two-fluid model.

The in-situ TEM irradiation experiments of zirconium alloy were conducted at 573 K, 673 K, and 773 K utilizing a 400 keV Kr + single beam and a 400 keV Kr+ and 30 keV He + dual beam. The results show that a large number of dislocation loops have been characterized in the matrix of the zirconium alloy under irradiation. With increasing the irradiation damage dose, some dislocation loops have reacted with one another to form a larger dislocation loop, which has finally formed dislocation lines or other defect structures. In zirconium alloys irradiated with Kr + single beam and Kr+ and He + dual-beam radiation, the proportion of type dislocation loops with different Burgers vectors is essentially the same at low damage doses, but the proportion of interstitial type dislocation loops with the same Burgers vectors is obviously different. The amorphization of the second phase and the dissolution of the small-sized second phase were also pointed out. With the increase in temperature, the density of the dislocation loop in zirconium alloy gradually decreases, and the size of dislocation loop first increases and then decreases. Kr+ and He + dual beam irradiation increases the size of dislocation loops but decreases their density as compared with Kr + single beam irradiation.

In abnormal states of nuclear power plants (NPPs), operators undertake mitigation actions to restore a normal state and prevent reactor trips. However, in abnormal states, the NPP condition fluctuates rapidly, which can lead to human error. If human error occurs, the condition of an NPP can deteriorate, leading to reactor trips. Sudden shutdowns, such as reactor trips, can result in the failure of numerous NPP facilities and economic losses. This study develops a remaining trip time (RTT) prediction system as part of an operator support system to reduce possible human errors and improve the safety of NPPs. The RTT prediction system consists of an algorithm that utilizes artificial intelligence (AI) and explainable AI (XAI) methods, such as autoencoders, light gradient-boosting machines, and Shapley additive explanations. AI methods provide diagnostic information about the abnormal states that occur and predict the remaining time until a reactor trip occurs. The XAI method improves the reliability of AI by providing a rationale for RTT prediction results and information on the main variables of the status of NPPs. The RTT prediction system includes an interface that can effectively provide the results of the system.

In prompt fission neutron uranium logging, borehole environmental parameters affect the measured results and must be corrected. In order to explore the influence of borehole parameters on the interpretation of logging results, this paper builds a sandstone type uranium ore block model to simulate the field production drilling device based on the “Epithermal/Thermal neutron counting rate ratio” (E/T) theory. The effects of borehole diameter, thickness of iron tube and well fluid on the decay rate of epithermal and thermal neutrons and their uncertainty correction methods were investigated. The results show that the effect of borehole diameter on E/T is negligible. The iron tube thickness has a certain effect on the moderation and absorption of epithermal and thermal neutrons, and its E/T increases slightly with increasing thickness. The influence of iron tube thickness on E/T is corrected and the relative uncertainty is less than 5%. The well fluid thickness also affects the decay rate of epithermal and thermal neutrons, and its E/T follows the law of negative exponential attenuation. The influence of well fluid thickness on E/T is corrected and the relative uncertainty is less than 5%. This study provides technical guidance for field well survey of uranium deposit.

During the decommissioning of nuclear and particle accelerator facilities, a considerable amount of large-scale radioactive waste may be generated. Accurately defining the activation level of the waste is crucial for proper disposal. However, directly measuring the internal radioactivity distribution poses challenges. This study introduced a novel technology employing machine learning to assess the internal radioactivity distribution based on external measurements. Random radioactivity distribution within a structure were established, and the photon spectrum measured by detectors from outside the structure was simulated using the FLUKA Monte-Carlo code. Through training with spectrum data corresponding to various radioactivity distributions, an evaluation model for radioactivity using simulated data was developed by above Monte-Carlo simulation. Convolutional Neural Network and Transformer methods were utilized to establish the evaluation model. The machine learning construction involves 5425 simulation datasets, and 603 datasets, which were used to obtain the evaluated results. Preprocessing was applied to the datasets, but the evaluation model using raw spectrum data showed the best evaluation results. The estimation of the intensity and shape of the radioactivity distribution inside the structure was achieved with a relative error of 10%. Additionally, the evaluation based on the constructed model takes only a few seconds to complete the process.

Efficient and secure decommissioning of nuclear facilities demands advanced technologies. In this context, gamma-ray detection and imaging are crucial in identifying radioactive hotspots and monitoring radiation levels. Our study is dedicated to developing a gamma-ray detection system tailored for integration into robotic platforms for nuclear decommissioning, offering a safe and automated solution for this intricate task and ensuring the safety of human operators by mitigating radiation exposure and streamlining hotspot localization.Our approach integrates a Compton camera based 3D reconstruction algorithm with a single Timepix3 detector. This eliminates the need for a second detector and significantly reduces system weight and cost. Additionally, combining a 3D camera with the setup enhances hotspot visualization and interpretation, rendering it an ideal solution for practical nuclear decommissioning applications.In a proof-of-concept measurement utilizing a 137Cs source, our system accurately localized and visualized the source in 3D with an angular error of 1° and estimated the activity with a 3% relative error. This promising result underscores the system's potential for deployment in real-world decommissioning settings. Future endeavors will expand the technology's applications in authentic decommissioning scenarios and optimize its integration with robotic platforms.The outcomes of our study contribute to heightened safety and accuracy for nuclear decommissioning works through the advancement of cost-effective and efficient gamma-ray detection systems.

Rupture disk corrosion test (RDCT) method has been recently developed for real-time measurement of initiation of stress corrosion cracking (SCC) in a high-temperature water. This work presents a parametric study on the stress distribution of a thin disk specimen of RDCT to consider the fixture shape and friction using finite element analysis (FEA). The FEA results showed a dome-shaped deformation of the specimen. From the stress analysis as a function of friction coefficient, it was suggested that the maximum stress was applied around the dome-edge, which was invariant with change to the friction coefficient. This indicates that friction between the fixture and the specimen has little effect on stress distribution. On the other hand, the stress analysis as a function of a rounded-corner radius (Rc) revealed the location at which the maximum stress was applied shifted from the dome edge to the dome center as Rc increased. From SCC initiation tests using the RDCT apparatus in a primary water environment, it was found that SCC initiates at the dome edge when Rc is 0.5 mm, while SCC initiates near the disk center when Rc is 2.0 mm. This experimental result is in good agreement with the results of FEA.

International Commission on Radiological Protection (ICRP) recently developed the adult and pediatric mesh-type reference computational phantoms (MRCPs) in high-quality/fidelity mesh format, featuring high deformability into various body sizes and poses. Utilizing this feature, the adult MRCPs-based body-size-dependent phantom library was developed for individualized dosimetry. To complete the full phantom library set, the present study produced the pediatric-MRCPs-based body-size-dependent pediatric phantom library. The library comprises a total of 637 phantoms (356 males and 281 females) with varying standing heights and body weights, covering a wide range of body sizes (i.e., including from 1st to 99th percentile height and weight values) for infants, children, and adolescents, offering a realistic representation of body shapes by reflecting ten secondary anthropometric parameters. The phantoms were automatically constructed utilizing automatic deformation program. The dosimetric impact of the library was investigated by calculating organ doses for external exposures to broad parallel photon beams in anterior-posterior direction. Compared with the values of the pediatric MRCPs, significant differences were observed at energies

While several nuclear energy newcomer (NEN) countries have shown interest in small modular reactors (SMRs) as a potential energy source, this interest can generate new uncertainties regarding future nuclear weapons proliferation risks. Therefore, this research seeks to determine whether future SMR deployment in NEN countries will contribute to nuclear weapons proliferation, and how the risks can be mitigated. This research uses the Bayesian network statistical approach in conjunction with surveys of experts to assess nuclear proliferation risks when NEN countries deploy SMRs or a large commercial nuclear reactor. The results indicate that an NEN with a strong commitment to the nuclear non-proliferation norms and a stable security environment will experience a lower probability of having higher proliferation risks relative to the United Arab Emirates. Specifically, we demonstrate that experts anticipate a minimal escalation in proliferation risks across different SMR types. Instead, the results show that enrichment or reprocessing (E&R) facilities, if associated with an SMR, exert a substantial influence on proliferation risks. Lastly, implementing a spent nuclear fuel (SNF) retrieval system could serve as an option to mitigate proliferation risks in an NEN country. These findings offer insights for leading nuclear supplier countries to alleviate the potential proliferation risks by NEN countries.

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.

Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.

Traditional text-based system engineering, which has been used in the design and application of passive residual heat removal system (PRHRS) for lead-cooled fast reactors, is prone to several problems such as low development efficiency, long iteration cycles, and model ambiguity. This study aims to effectively address the above-mentioned problems by adopting a model-based system engineering (MBSE) method, which has been preliminarily applied to meet the design requirements of a PRHRS. The design process has been implemented based on the preliminary design of the system architecture and consists of three stages: top-level requirement analysis, functional requirements analysis, and design requirements synthesis. The results of the top-level requirements analysis and the corresponding use case diagram can determine the requirements, top-level use cases, and scenario flow of the system. During the functional requirements analysis, the sequence, activity, and state machine diagrams are used to develop the system function model and provide early confirmation. By comparing these sequence diagrams, the requirements for omissions and inconsistencies can be effectively checked. In the design requirements synthesis stage, the Analytic Hierarchy Process is used to conduct preliminary trade-off calculations on the system architecture, after which a white box model is established during the system architecture design. Through these two steps, the analysis and design of the system architecture are ultimately achieved. The resulting system architecture ensures the consistency of the design requirements. Ultimately, a functional hazard analysis was conducted for a specific incident to validate case requirements and further refine the system architecture. Future research can further reduce the design risk, improve the design efficiency, and provide a practical reference for the design and optimization of PRHRS in digital lead-cooled fast reactors.

Amid the ongoing discourse on clean energy solutions, the reopening of decommissioned plants, such as the Bataan Nuclear Power Plant (BNPP) in the Philippines has become a focal point in the country. This study delved into the complex web of factors influencing public acceptance of BNPP, employing the values-beliefs-norms theory. By utilizing partial-least square structural equation modeling, the research unravelled the intricate relationships among biospheric values, altruistic values, egoistic values, ecological worldview, awareness of consequences, personal norm, social norm, and the broader acceptance of BNPP establishment. With 434 respondents participating in a self-administered online survey, the study identified key correlations. Emphasizing the collaborative impact on decision-making processes by social and personal norms, the study also highlighted the role of ecological values in shaping awareness. The foundational impact of values on ecological worldviews was explored, shedding light on public attitudes toward nuclear energy. This research offers actionable insights for policymakers, advocating for targeted communication strategies and public engagement initiatives to navigate barriers and promote informed decision-making in the dynamic landscape of nuclear energy development. The study contributes to the global conversation on sustainable energy strategies, emphasizing the pivotal role of public perception in shaping the trajectory of nuclear power.

Isotopes are an important aspect of modern medical and scientific research and cyclotron-based isotope production is of particular interest. Cooling devices are required to manage the heat generated by high-energy particle beams to ensure that they can be delivered securely. However, there is considerable scope for further advancements in the design of cooling systems. Therefore, this study uses simulations to investigate the design of innovative cooling systems. Various designs and materials are considered using a variable four-sector collimator and various particle-beam conditions. Thus, new design principles and material and structural optimization strategies are proposed, which can be used to improve the stability and efficiency of isotope production. The results provide insight into technical challenges and may contribute toward advances in medical diagnostics, radiotherapy, and fundamental scientific research. In particular, they can guide the development of isotope-based technologies in nuclear medicine.

This paper introduces the creation of a module for Uncertainty Quantification (UQ) specifically designed for VVER analysis through the implementation of the STREAM/RAST-V two-step approach. The aim was to expand the range of use by developing a UQ module tailored for analyzing VVER. This research presents two innovative computational functionalities: (1) development of a library for the pin-based pointwise energy slowing down method (PSM), and (2) extension of the analysis area to study hexagonal-geometry fuel assemblies. The proposed UQ scheme was evaluated through verification using UAM benchmark, and comparative analysis between codes using SCALE 6.2.2 for. STREAM provides an accuracy comparable to that of SCALE 6.2.2. Additionally, a PSM covariance library was utilized in the calculations, achieving 0.7941% and 0.7907% accuracies in the hot full power and hot zero power calculations, respectively. To assess the UQ sequences in the two-step method, the STREAM/RAST-V calculation scheme was verified using the STREAM lattice code. To conclude, this study furnishes comprehensive insights into the development of the UQ module within the two-step method for VVER analysis, and it validates its performance through utilization of the UAM benchmark.

This paper presents a high-fidelity simulation of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) 3D whole-core Watts Bar benchmark using the UNIST in-house STREAM3D (Steady State and Transient Reactor Analysis code with Method of Characteristics) neutronic code. The benchmark encompasses various whole-core exercises, including single physics problems, multi-physics simulations, and depletion problems. When comparing parameters during the zero-power physics tests, including ITC, DBW, CRW, and criticality tests, STREAM3D results indicate a strong agreement with the measured data and KENO-VI. The comparison with the MC21/CTF code in 3D HFP BOC condition demonstrated strong agreement, with only a 0.42% difference in the normalized radial power distribution, a 0.38 K difference in the RMS of the assembly coolant exit temperature, and a mere 4 ppm difference in CBC.

With the increasing usage of nuclear energy, how to properly dispose nuclear waste becomes a critical issue. In this study, a multiscale modeling approach combining the experimental findings is presented to address the illitization process, its impact on transport properties, and system behavior of bentonite buffer in engineered barrier systems (EBS). Through the pore-scale modeling, reactive transport properties such as illite generation rate and effective diffusion coefficient of potassium ion as a function of porosity and temperature are quantified by employing the findings of hydrothermal reaction experiments of Bentonil-WRK. The capability of pore-scale modeling has been developed based on the Darcy-Brinkmann-Stokes equation, involving the processes of smectite illitization and clay swelling. Obtained reactive transport properties are utilized as input parameters for the macroscale modeling to predict the long-term behavior of bentonite buffer in EBS. As such, this study involves the whole workflow of quantifying the reaction parameters of smectite illitization through the hydrothermal reaction experiments, and numerically modeling the reactive transport process of smectite illitization in bentonite buffer of EBS from pore-scale to macroscale. The presented multiscale modeling findings are expected to provide reliable solution for safe nuclear waste disposal with EBS.

For traditional spent fuel shielding materials, due to physical and chemical defects and cost constraints, they have been unable to meet the needs. Therefore, this paper carries out the first discussion on the application and performance of bismuth in neutron shielding by establishing Monte Carlo simulation on the neutron flux model of shielded spent fuel. Firstly, functional fillers such as bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide are added to the matrices to compare the shielding rates of aluminum alloy matrix and silicone rubber matrix. The shielding rate of silicone rubber mixture is higher than aluminum alloy mixture, reaching more than 56%. The optimal addition proportion of bismuth oxide and lead oxide is 30%, and the neutron radiation protection efficiency reaches 60%. Then, the mass attenuation coefficients of bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide in silicone rubber matrix are simulated with the change of functional fillers proportion and neutron energy. This simulation result shows that the mixture with functional fillers has good shielding performance for low energy neutrons, but poor shielding effect for high energy neutrons. Finally, in order to further evaluate the possibility of replacing lead oxide with bismuth oxide as shielding material, the half-value layers and various properties of bismuth oxide and lead oxide are compared. The results show that the shielding properties of bismuth oxide and lead oxide are basically the same, and the mechanical properties, heat resistance, radiation resistance and environmental protection of bismuth oxide are better than that of lead oxide. Therefore, in the case of neutron source strengths in the range of 0.01–6 MeV and secondary gamma rays produced below 2.5 MeV, bismuth can replace lead in neutron shielding applications.

Void drift between subchannels in a rod bundle is a crucial phenomenon affecting the calculation accuracy of thermal-hydraulic parameters in SMRs. It holds significant importance in enhancing the precision of safety analysis for SMRs. Existing research on experiment and model of void drift between rod bundle subchannels is relatively rare, and the accuracy of model calculations requires improvement. In this study, experiments on gas-liquid two-phase non-equilibrium flow were conducted to measure the redistribution of two-phase flow induced by void drift in a 1 × 2 rod bundle. The experiment results indicated that in bubby flow regime with void fraction less than 0.3, the void diffusion coefficient showed little variation with changes in void fraction. However, in slug flow and annular flow regimes with void fraction exceeding 0.3, the void diffusion coefficient significantly increased with an increase in void fraction. Furthermore, a new void drift model was developed and validated based on a subchannel code. The overall predicted uncertainty for the outlet void fraction in the rod bundle benchmark was less than 13%.

Our research seeks to assess the influence of nuclear energy technology on carbon emissions in the top 10 European economies comprising the topmost nuclear energy R&D budget (France, Germany, Russia, the Netherlands, the UK, Finland, Spain, Sweden, Italy, and Switzerland). Unlike prior investigations predominantly relying on panel data methodologies without considering the distinctive characteristics of each economy, our study employs the advanced 'Quantile-on-Quantile' approach. This novel methodology enables us to investigate the interactions between variables within each unique nation, thereby improving the precision of our analysis. As a result, the study provides a thorough global perspective, revealing nuanced findings pertinent to each economy's specific attributes. Our outcomes demonSstrate a positive interconnection between nuclear energy technology and carbon emissions across various quantiles in our analyzed nations. Additionally, the study highlights diverse patterns in these associations within individual economies. These findings emphasize the significance of policymakers performing comprehensive measurements and devising effective strategies to monitor fluctuations in nuclear energy technology and carbon emissions.

Small Modular Reactors (SMRs) adopt the Helically Coiled Once-Through Steam Generators (OTSG) extensively for its compactness and higher heat transfer efficiency. As a heat exchanger between the primary side (reactor coolant system) and the secondary side (feedwater and steam system) of nuclear steam supply system, the inlet/outlet conditions both of shell side and tube side of OTSGs have significant impacts on overall system response. Considering the flexible operation of SMRs and heat application by extracting steam, a simulation tool for accurate prediction of the OTSG dynamic behaviors would be required for optimizing design and control. In this study, the OTSG dynamic simulation model has been developed. Mathematical governing equation has been derived by using moving boundary approach and a simulation module has been developed by using Modelica Language. The developed module has been compared with publicly available experimental results and benchmarked with MARS-KS calculation results. Also, it has been incorporated into the integrated SMR model (i.e., reactor core, primary side, secondary side) and dynamic behaviors with reactivity feedback and heat balancing have been investigated. In both of steady-state and transient conditions, it shows the promising accuracy.

Investigations were conducted on the addition of barium's impact on the radiation shielding and physical attributes of five different glasses, designated S1–S5, with varying BaO contents. Using two point sources namely Co60 and Cs137 along with a scintillation detector [NaI(TL)], experimental measurements were made of the shielding parameters of γ-rays, namely the effective atomic number (Zeff), electron density (Nel), half-value layer (HVL), linear attenuation coefficient (μ), mass attenuation coefficient (μm), mean free path (λ), and radiation protection effectiveness at the energies of 0.664, 1.177, and 1.334 MeV, and comparisons made with recently considered glasses as well as frequently employed materials for γ-ray shielding. The results show that the examined glasses' physical and radiation shielding qualities are improved by the addition of BaO. The μ values increased from 0.245 to 0.275 cm−1 (0.662 MeV), from 0.174 to 0.198 cm−1 (1.173 MeV), and from 0.161 to 0.189 (1.332 MeV). The observed values of HVL decreased from 2.83, 3.98, and 4.3 cm to 2.5, 3.5, and 3.62 cm at 0.662, 1.173, and 1.332 MeV, respectively, for the samples S1 and S5. In addition, the S5 glass sample was determined to have the best protection against photon among all the samples that were evaluated, as well as against recently considered glasses and those materials often utilized for gamma-ray shielding purposes.

Addressing the challenge of identifying an appropriate set of material and irradiation parameters for accurate simulation models using crystal plasticity finite element method (CPFEM), this study proposes a novel two-stage method for nano-indentation modeling of ion-irradiated face-centered cubic (FCC) materials. It includes implementing the strain-gradient crystal plasticity (SGCP) theory with irradiation effects and the calibration of simulation parameters using the particle swarm optimization (PSO) algorithm with experimental data. The proposed method consists of two stages: establishing CPFEM without irradiation effects in stage 1 and modeling irradiation effects based on CPFEM in stage 2. Modeling the nano-indentation test of ion-irradiated stainless steel 304 (SS304) using real experimental data is conducted to evaluate the efficiency of the proposed method. The accuracy of the calibration method using PSO is verified through comparisons between simulation and experimental results for force-indentation depth and hardness-indentation depth relationships under both unirradiated and irradiated conditions. Moreover, effect of ion-irradiation on the mechanical behavior during the nano-indentation of single crystal SS304 is also examined to demonstrate that the proposed method is a powerful approach for nano-indentation modeling of ion-irradiated FCC single crystals using SGCP theory and the PSO algorithm.

Copper canisters are commonly employed to contain HLW for the long-term, making it crucial to understand how corrosion affects the canister. This study conducted a comparative analysis of two widely used calculation methods for modeling canister corrosion within a 3-D DGR domain. The first method, termed transport-limited corrosion, assumes an immediate sulfide-copper reaction and has been traditionally used due to its conservative nature. The second method, known as the potential-limited corrosion, considers coupled redox reactions at the canister surface and computes corrosion rates through anodic current density. From the results, we found that the edge of the canister geometry and the omission of electrochemical kinetics impose critical limitations with the transport-limited corrosion method. These limitations include the singularity problem, excessive sensitivity to the curvature of the canister's edge, and an inability to evaluate the distribution of corrosion rate over the canister surface as a function of the sulfide concentration. On the other hand, the potential-limited corrosion method avoided the limitations found in the other method. Since the factors relating to these limitations are critical to the design and optimization of the copper disposal canister, careful consideration when selecting appropriate calculation methods for corrosion will be required.