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The CFETR multipurpose overload robot (CMOR) is a critical component of the fusion reactor remote handling system. To accurately calculate and visualize the structural deformation and stress characteristics of the CMOR motion process, this paper first establishes a CMOR kinematic model to analyze the unfolding and working process in the vacuum chamber. Then, the dynamic model of CMOR is established using the Lagrangian method, and the rigid-flexible coupling modeling of CMOR links and joints is achieved using the finite element method and the linear spring damping equivalent model. The co-simulation results of the CMOR rigid-flexible coupled model show that when the end load is 2000 kg, the extreme value of the end-effector position error is more than 0.12 m, and the maximum stress value is 1.85 × 108 Pa. To utilize the stress-strain data of CMOR, this paper designs a CMOR morphology prediction control system based on Unity software. Implanting CMOR finite element analysis data into the Unity environment, researchers can monitor the stress strain generated by different motion trajectories of the CMOR robotic arm in the control system. It provides a platform for subsequent research on CMOR error compensation and extreme operation warnings.

This paper details a computational framework to produce automated, graphical workflows, and how this framework can be deployed to support complex modeling problems like those in nuclear engineering. Key benefits of the framework include: automating previously manual workflows; intuitive construction and communication of workflows through a graphical interface; and automated file transfer and handling for workflows deployed across heterogeneous computing resources. This paper demonstrates the framework's application to probabilistic post-closure performance assessment of systems for deep geologic disposal of nuclear waste. However, the framework is a general capability that can help users running a variety of computational studies.

The physical protection systems of Nuclear Power Plant are utilized to safeguard targets against intrude by attacker. As the methods employed by attackers to intrude Nuclear Power Plant become increasingly complex and diverse, there is an urgent need to identify optimal defense strategies to interrupt adversary intrusions. This paper focuses on studying the defense of security personnel against adversary intrusions and utilizes an evolutionary game approach to select the optimal defense decisions for physical protection systems. Under the assumption of bounded rationality for both the attacker and defender, the paper constructs replication dynamic equations for attack and defense strategies, investigating the process of strategy selection and the stability of evolution. Finally, a minimal model is proposed to validate the feasibility of utilizing the evolutionary game model for defense strategy selection.

Reactor core transient calculation is very important for the reactor safety analysis, in which the kernel is neutron kinetics calculation by simulating the variation of neutron density or thermal power over time. Compared with the point kinetics method, the time-space neutron kinetics calculation can provide accurate variation of neutron density in both space and time domain. But it consumes a lot of resources. It is necessary to develop a surrogate model that can quickly obtain the temporal and spatial variation information of neutron density or power with acceptable calculation accuracy. This paper uses the time-varying characteristics of power to construct a time function, parameterizes the time-varying characteristics which contains the information about the spatial change of power. Thereby, the amount of targets to predict in the space domain is compressed. A surrogate method using the machine learning is proposed in this paper. In the construction of a neural network, the input is processed by a convolutional layer, followed by a fully connected layer or a deconvolution layer. For the problem of time sequence disturbance, a structure combining convolutional neural network and recurrent neural network is used. It is verified in the tests of a series of 1D, 2D and 3D reactor models. The predicted values obtained using the constructed neural network models in these tests are in good agreement with the reference values, showing the powerful potential of the surrogate models.

In the paper we validate theoretical models of the pulse against experimental data from the Jozef Stefan Institute TRIGA Mark II research reactor. Data from all pulse experiments since 1991 have been collected, analysed and are publicly available. This paper summarizes the validation study, which is focused on the comparison between experimental values, theoretical predictions (Fuchs-Hansen and Nordheim-Fuchs models) and calculation using computational program Improved Pulse Model. The results show that the theoretical models predicts higher maximum power but lower total released energy, full width at half maximum and the time when the maximum power is reached is shorter, compared to Improved Pulse Model.

In this paper, modal effective mass for asymmetric fluid-structure interaction system is defined and equations for its calculation is derived. To establish consistency, modal effective mass in symmetric structure only system is briefly reviewed, followed by a definition of the modal effective mass in asymmetric system. The equations for calculating modal effective mass in asymmetric system are derived by utilizing the properties of left and right eigenvectors. To simplify the equations, the assumption is made that the mass matrix is only affected by the fluid. The simplified equation is then compared to the equation already used in ANSYS. Finally, the validity of the modal effective mass definition and derivation in this paper is demonstrated through a simple example.

This paper describes the beam dynamic simulation with transfer matrix method for four sector cyclotron. Starting from a description on the equation of motion in the cyclotron, lattice functions were determined from transfer matrix method and the solutions for the 2nd-order nonlinear Hamiltonian were introduced and used in phase space particle tracking. Based on the description of beam dynamics in the cyclotron, simulation code was also developed for cyclotron design.

Due to its low background, alpha spectrometry is capable of determining very low activity concentrations of alpha emitters in all types of samples, including environmental samples. This low background can be lost due to detector contamination caused by recoil nuclei, which can even render the detector useless. In this paper, the thickness of a thin film used to cover the source is calculated analytically. This film, capable of retaining these nuclei and avoiding this contamination, allows all the alpha particles emitted by the source to reach the detector. The energy loss of the alpha particles passing through it was also estimated in order to identify the peaks in the spectrum, including the unexpected ones. The validity of the calculations carried out and the effectiveness of the film in retaining the recoil nuclei were verified using thorium samples. Finally, the effect of applying a negative potential to the source on the retention of these nuclei was studied. The study was carried out using a PIPS detector under vacuum conditions (0.1 mm Hg) with the samples placed at a distance of 5 mm from the detector. The recoil suppression systems of some commonly used commercial instruments were also tested.

This paper presents the development process of a high-dose X-band linear accelerator (LINAC) for radiotherapy machines, specifically focusing on stereotactic body radiation therapy (SBRT). The development process includes a multi-physics design, approach for the radio frequency cavity in the LINAC, with an emphasis on minimizing beam size and maximizing shunt impedance of the LINAC. Commissioning tests were conducted to evaluate the performance of the LINAC, which involved measurements of beam energy, beam size, beam current, and X-ray dose rate.

Storage rings that use the hybrid multi-bend achromat (MBA) lattice concept [1] have been realized. The main characteristics of hybrid MBA as a typical lattice for a 4th generation storage ring are to adopt two large dispersion bumps in cell, and to make a -I-transformation between two dispersion bumps to cancel nonlinear effects. To increase the effectiveness of nonlinear control, analytical investigation of nonlinear phenomenon is essential. In this paper, we describe analytical investigation with a nonlinear driving term and simple model. We obtained similar results to the existing time-consuming MOGA optimization, but use a simplex optimizer with fewer nonlinear control knobs.

Identifying important phenomena and parameters of LBLOCAs is an important step in nuclear power safety evaluation. Traditional identification processes are based on experience and incorporate important phenomena and parameters without quantifying specific aspects of the models. To accurately identify physical effects, this paper presents the use of multiphase field subchannel code for simulation analysis as applied to numerical examples and specific reactor thermal-hydraulic problems. The results indicate that the code can simulate the relevant phenomena, and the calculation band of the refill stage is narrow, while the calculation band of the reflood stage is wide. In addition, the key impact models are captured. The radiation model most significantly impacts the cladding temperature in the core, and the heat transfer model of transition boiling to the gas phase is most strongly influenced by the quenching time. The interfacial heat transfer model of the large bubble regime for the liquid plays a major role in influencing the water inventory of the lower plenum in the downcomer.

This paper elucidates a comprehensive derivation of the variational formulation pertaining to the static and spatial kinetics self-adjoint angular flux (SAAF) neutron transport equation. The methodology employed for discretization of the spatial variable is the finite element method, while the energy group discretization is executed via the group method, and the directional discretization is conducted using the discrete ordinates method. Analytic expressions for the discretization to variable separation under both vacuum and reflective boundary conditions are furnished. Constructed upon the MOOSE framework, a code designated as SAAFCGSN has been developed for the resolution of the SAAF neutron transport equation. The zero-order scattering matrix within this computational framework is managed through an innovative "decoupling" method, thereby enhancing the computational efficiency significantly. The functionality and robustness of the SAAFCGSN code are corroborated through meticulous evaluation involving seven distinct steady-state scenarios as well as two transient states. Empirical outcomes verify the compatibility of the SAAFCGSN code with both structured and unstructured mesh, inclusive of their amalgamations, thus facilitating maintenance and ensuring elevated computational accuracy. In addition, a performance analysis benchmarked against IAEA standards reveals that the adoption of the scattering matrix "decoupling" method propels a computational speed increase exceeding 30 %, signifying a notable advancement in calculation efficiency.

Scientific literature studies irradiated graphite treatment. Research also covers graphite conditioning and its long-term behavior under disposal conditions. The European Commission's CARBOWASTE project, titled “Treatment and disposal of irradiated graphite and other carbonaceous waste”, is a key reference for state-of-the-art studies on alternative solutions. It identified 24 strategic options for managing irradiated graphite throughout its complete life cycle. The methodology proposed in this paper entails the application of the Analytic Hierarchy Process (AHP) method to rank the 24 options, placing particular emphasis on the weighting of seven criteria for selecting management options for the irradiated graphite. The highest weights were assigned by experts to ‘environment and public safety' (28.05 %) and ‘worker safety' (26.16 %). The objective is to develop a standardized approach enabling waste management companies to identify the most appropriate management option, considering structural and legislative constraints in their operating country. Examining the study findings, option 19 “In-situ entombment” stands out as the best choice in both the CARBOWASTE project and the proposed methodology. Thus, this methodology could assist hypothetical entities in examining management options for irradiated graphite, with the aim of identifying the optimal solution for graphite waste disposal.

The SPHERE experiment aimed at studying the behaviour of Minor Actinide-bearing Driver Fuel (U,Pu,Am)O2-xby comparing sphere-packed and pelletized fuels. The irradiation experiment was performed in the High Flux Reactor at Petten from August 2013 to April 2015, and was followed by post-irradiation examinations up to mid-2017. The present work consists in a new analysis of the SPHERE experiment, focusing on the pelletized fuel, by the means of both neutronics and fuel performance codes. This study is performed in the frame of the European Project PATRICIA. The adopted methodology and the main results achieved, assessed in particular against inert gas-related experimental data, are presented in the paper.

Nuclear energy is used more and more widely as a clean energy source, but nuclear energy facilities are risky, and when a nuclear leak occurs, it is necessary to detect equipment in a nuclear radiation environment. In the nuclear radiation environment, due to the impact of high-energy particles on the camera sensor, the collected image contains a lot of radiation noise, which greatly reduces the visual perception of the image. Aiming at the problem that radiation noise reduces image quality, a radiation image compound filtering algorithm combining median filtering and multi-frame average filtering is proposed based on radiation noise characteristics. Compared with several common filtering algorithms, the radiation noise image is filtered under the same radiation dose and achieve the highest Peak Signal Noise Rate (PSNR) and Structural Similarity (SSIM), and compared with the multi-frame average filtering method, the number of image frames required by the algorithm in this paper is greatly reduced. Experimental results show that the algorithm can effectively eliminate radiation noise and is more suitable for image filtering in radiation environment.

Reinforced concrete structures are subject to exposure to chloride ions in the air, leading to chloride penetration, and carbonation attacks resulting from exposure to carbon dioxide. This chemical degradation process induces corrosion of reinforcing bars within concrete, significantly impacting durability. Structures situated in coastal areas, such as nuclear power plants, are particularly susceptible to rapid chloride penetration due to the high chloride concentration in the air. This study utilizes existing experimental data to forecast the chloride diffusion coefficient employing artificial neural network (ANN technology). The total number of experimental data was 535 gathered from 18 papers. Through analysis of the chloride coefficient and predicted degradation depth, the effective cross-sectional area of concrete is examined, and the deterioration of wall performance is forecasted.

Some core designs integrate high-enriched fuel and moderator materials to enhance neutron utilization. This combination results in a broad spectrum within the system, posing challenges in resonance calculation. This paper introduces a general framework to realize resonance self-shielding treatment in broad-spectrum fuel lattice problems. The framework consists of three components. First, a new energy group structure is devised to support resonance calculation in the entire energy range and capture spectral transition and thermalization effects during eigenvalue calculation. Second, the subgroup method based on narrow approximation is selected as a universal method to perform resonance calculation. Finally, transport equations for each fissionable region are solved for neutron flux to collapse the fission spectrum. The proposed method is verified against fast, intermediate, and thermal spectrum pin cell problems and an assembly problem featuring a fast-thermal coupled spectrum. Numerical results affirm the accuracy of the proposed method in handling these scenarios, with eigenvalue errors below 154 pcm for pin cell problems and 106 pcm for the assembly problem. The verification results revealed that the proposed method enables accurate resonance self-shielding treatment for broad-spectrum problems.

During the process of bilateral teleoperation, operators not only need to perform complex maintenance tasks but also have to constantly monitor the safety of the operation, leading to reduced operational efficiency. Therefore, in this paper, we introduce a shared autonomous scheme that intervenes in the operator's command input when necessary, autonomously ensuring the safe operation of the manipulator by employing a rolling horizon planning controller based on Nonlinear Model Predictive Control (NMPC). This controller considers the motion boundaries and collision avoidance constraints of the manipulator, accompanied by the design of corresponding objective functions. To validate the effectiveness of the proposed method, we conduct tests on collision-free trajectory tracking and comprehensive performance with collision constraints, confirming the feasibility and excellent performance of the approach.

This paper presents a seismic analysis method that can evaluate a very large number of cases for the free standing dry storage cask by proposing a methodology that has short analysis time as well as accuracy. This study also performed a seismic analysis of a dry storage facility with multiple casks to show a tip-over phenomenon from earthquake accident conditions. The earthquake accident condition is long-term event that occur during about 20 s long, and lots of seismic analysis cases should be performed to consider various real conditions because the free-standing spent-fuel dry storage cask has many nonlinear responses. The soil–concrete pad–cask interaction was considered in the seismic analysis and finite element model was made using concrete pad, soil and cask models. In the reinforced concrete pad, the rebar was excluded to reduce the analysis time, but the thickness was corrected to maintain the bending rigidity. Additionally, the analysis time reduced by modeling the cask as a rigid body rather than a flexible body. 35-cases of seismic analysis were performed to determine a tip-over phenomenon from each earthquake. The analysis revealed that no tip-over phenomenon of the cask was observed in all analyses from 0.2 g to 0.6 g, however the tip-over of the cask were observed from 0.8 g with friction coefficients of 0.8 and 1.0.

The selection of appropriate transport parameters of radionuclides is of utmost importance in the safety assessment of radioactive waste disposal sites. Both batch and column methods are widely performed to determine the transport parameters of radionuclides through soil layers. In this paper, the distribution coefficients Kdof Co ions in sandy mixtures containing soil samples taken from Anarak near-surface repository (Iran) were determined by batch and 3 different column methods namely, breakthrough curve, thin-section, and gamma spectroscopy methods. The distribution coefficients Kdof the mixture was obtained as 160 L kg−1for the batch method, 23 L kg−1, 19 L kg−1, and 18 L kg−1for the breakthrough curve, the thin-section, and the gamma spectroscopy methods, respectively. These results emphasize proper selection of transport data for using in the safety assessment programs.

The release source term of radioactivity becomes a critical foundation for emergency response and accident consequence assessment after a nuclear accident Rapidly and accurately inverting the source term remains an urgent scientific challenge. Today source term inversion based on meteorological data and gamma dose rate measurements is a common method. But gamma dose rate actually includes all nuclides information, and the composition of radioactive nuclides is generally uncertain. This paper introduces a novel nuclear accident source term inversion model, which is Improve Snow Ablation Optimizer-Sensitivity Analysis Pruning Extreme Learning Machine (ISAO-SAELM) model. The model inverts the release rates of 11 radioactive nuclides (I-131, Xe-133, Cs-137, Kr-88, Sr-91, Te-132, Mo-99, Ba-140, La-140, Ce-144, Sb-129). It does not require the use of the physical field of the reactor to obtain prior information and establish a dispersion model. And the robustness is validated through noise analysis test. The mean absolute errors of the release rates of 11 nuclides are 15.52 %, 15.28 %, 15.70 %, 14.99 %, 14.85 %, 15.61 %, 15.96 %, 15.42 %, 15.84 %, 15.13 %, 17.72 %, which show the significant superiority of ISAO-SAELM. ISAO-SAELM model not only achieves notable advancements in accuracy but also receives validation in terms of practicality and feasibility.

This paper explores the alternating conditional expectation (ACE) algorithm-based surrogate model to advance the state-of-practice in uncertainty and sensitivity analysis methodologies for severe accident codes. For engineering purposes, the ACE algorithm has been used as an alternative means to find the optimal functional forms of the multiple input variables and response variables of interest. Analysis results here demonstrate that compared with the reference cases the proposed surrogate model provides much higher performance in terms of the coefficient of determination (R2) and normalized root mean square error (NRMSE), thus giving more robust insights into the relationship and correlation between the input parameters and figures of merit (FOMs) of interest. Relevant results and insights are summarized in terms of points of interest.

To reduce the losses caused by aging failure of insulation gate bipolar transistor (IGBT), which is the core components of nuclear power plant rod position indicating and rod control (RPC) system. It is necessary to conduct studies on its life prediction. The selection of IGBT failure characteristic parameters in existing research relies heavily on failure principles and expert experience. Moreover, the analysis and learning of time-domain degradation data have not been fully conducted, resulting in low prediction efficiency as the monotonicity, time correlation, and poor anti-interference ability of extracted degradation features. This paper utilizes the advantages of the stacked denoising autoencoder(SDAE) network in adaptive feature extraction and denoising capabilities to perform adaptive feature extraction on IGBT time-domain degradation data; establishes a long-short-term memory (LSTM) prediction model, and optimizes the learning rate, number of nodes in the hidden layer, and number of hidden layers using the Gray Wolf Optimization (GWO) algorithm; conducts verification experiments on the IGBT accelerated aging dataset provided by NASA PCoE Research Center, and selects performance evaluation indicators to compare and analyze the prediction results of the SDAE-LSTM model, PSO-LSTM model, and BP model. The results show that the SDAE-LSTM model can achieve more accurate and stable IGBT life prediction.

The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel pellet can be obtained by coupling of Monte Carlo neutron transport codes with thermal-hydraulics solvers. However, this approach is not practical for standard nuclear design calculations, and computationally cheaper and faster methods must be used. In nuclear core simulators, a concept of a single "effective temperature" that yields the same neutron response as in the case of the actual temperature shape is mainly applied. This paper evaluates various fuel temperature models used in nuclear core simulation calculations, ultimately recommending a new effective temperature model that considers the burnup correction.

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm2O3and Gd2O3demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a 252Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

The high-temperature molten salt pump is the core equipment in the small modular molten salt reactor with media temperatures up to 700 °C. The cantilever ratio of the molten salt pump is usually large. Excessively large cantilever ratios cause increased deformations and rotational amplitudes at the impeller, thus affecting the operational stability of the main pump; small cantilever ratios cause heavy temperature gradients, thus affecting the structural integrity evaluation. This paper used numerical simulation methods to calculate and analyze the temperature field, stress, and structural integrity, optimized the pump shaft cantilever length of the original scheme based on structural integrity using the dichotomy method, and analyzed the rotor dynamics of the optimization results. The results of this study show that the thermal expansion load caused by the temperature difference has a significant mechanical effect on the structure; the first-order critical speed of the rotor system of the optimized schemes has been improved, and the amplitude of the unbalanced response has been significantly reduced, which not only improves the operational stability of the rotor, also contributes to the compact design of the main pump of a small modular molten salt reactor.

It is difficult to detect a small-scale leakage in a nuclear power plant (NPP) quickly and take appropriate action. Delaying these procedures can have adverse effects on NPPs. In this paper, we propose leak flow rate prediction using the bidirectional gated recurrent unit (Bi-GRU) method to detect leakage quickly and accurately in small-scale leakage situations because large-scale leak rates are known to be predicted accurately. The data were acquired by simulating small loss-of-coolant accidents (LOCA) or small-scale leakage situations using the modular accident analysis program (MAAP) code. In addition, to improve prediction performance, data were collected by distinguishing the break sizes in more detail. In addition, the prediction accuracy was improved by performing both LOCA diagnosis and leak flow rate prediction in small LOCA situations. The prediction model developed using the Bi-GRU showed a superior prediction performance compared with other artificial intelligence methods. Accordingly, the accurate and effective prediction model for small-scale leakage situations proposed herein is expected to support operators in decision-making and taking actions.

Enhancing the capabilities of unmanned space exploration, such as satellite monitoring and space science missions, requires efficient and reliable nuclear power systems. A viable solution is found in the 1–10 kWe power level of space nuclear reactor power systems, offering advantages such as a manageable research and development process, and relatively low investment requirements. This paper introduces a conceptual design for a 5 kWe space nuclear reactor power system, outlining its components and characteristics. The study includes a thorough analysis of potential challenges, encompassing heat pipe failure accidents, re-entry scenarios, and weight estimation considerations. The results demonstrate that the proposed space nuclear reactor power system effectively meets the safety requirements. The total mass of the power system is estimated at approximately 1.5 tons, with a specific mass of around 300 kg/kWe. This research contributes valuable insights for the design of space nuclear reactor power systems operating within a similar power range.

Radioisotopes tend to play important roles in many fields, such as industry, healthcare, agriculture, aerospace, etc. Radioisotope production is mainly through accelerators or research reactors, and high flux research reactor is one of the most effective approaches for radioisotope production. The physical basis of preparing radioisotope relies on nuclear reactions occurring in the reactor core, which includes fission, (n,γ), (n,α), and (n,p) reaction, etc. Among them, fission and (n,γ) reaction are most important in the nuclear reactor. For example, the 99Mo could be generated by uranium fission and extracting from the fission products, or through the radiative capture reaction from enriched 98Mo. As for the fission method, the irradiation target is gradually transitioning from high enriched uranium (HEU) target to low enriched uranium (LEU) target due to the requirement of non-proliferation. In this paper, studies on the impacts of different fission targets on radioisotope productions are conducted. Moreover, an optimized study on the radiative capture method is performed to improve the production efficiency. It is concluded that it is advantageous to use radiative capture method to generate radioisotopes in high flux reactor, which helps to improve the specific activity with environmental friendliness.

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

The paper concerns a room-temperature cross-bar H-mode (CH) drift tube linac (DTL) with KONUS (Kombinierte Null Grad Struktur) [1,2] beam dynamics. To make the acceleration in DTL cell more efficient, we studied the correlation between transit time factor (TTF) and structural coefficients, first. Furthermore, we developed a new code with Python to demonstrate the longitudinal dynamics more clearly. The code computationally generates clusters, bunch centers, and emittance growth in a single figure. Thus, the stabilization region and cluster evolution at various negative phases can be studied. Based on the above studies, we designed a 162.5 MHz CH-DTL to accelerate 10 mA D+from 2.11 MeV to 3.25 MeV in continuous-wave (CW) mode. The proposed CH-DTL is a part of the Middle Energy Neutron Source (MENS). The dynamics and RF design were iterated to make the gap voltage error lower than 1 %. The initial beam is assumed to come from a Radio Frequency Quadrupole accelerator (RFQ). The geometries of the CH-DTL are optimized by using CST. Multiparticle tracking from LEBT to RFQ is performed with TraceWin and the transmission efficiency in the CH-DTL is 100 %.

The Korea Atomic Energy Research Institute (KAERI) is developing a novel Two-Region Arc Plasma Ion Source (TRAP) as a negative hydrogen (deuterium) ion source for a Neutral Beam Injection (NBI) system in a fusion tokamak. The TRAP ion source is based on a two-region configuration, comprising a high energy electron region that creates highly vibrationally excited molecules and a low electron temperature region that generates negative ions by attaching electrons to molecules. This configuration can be achieved by optimizing the filament position and magnetic cusp field. In order to optimize the TRAP configuration, the plasma parameters are investigated under various operating conditions, such as filament position, gas pressure, and arc power. Electron density and temperature are determined using Langmuir probe measurements. In this paper, the detailed experimental results are described and discussed.

This paper proposes a combined plastic and creep constitutive model of A533B1 pressure vessel steel to simulate progressive deformation of nuclear pressure vessels under severe accident conditions. To develop the model, recent tensile test data covering a wide range of temperatures (from RT to 1,100 °C) and strain rates (from 0.001%/s to 1.0%/s) was used. Comparison with experimental data confirms that the proposed combined plastic and creep model can well reflect effects of temperature and strain rate on tensile behaviour up to failure. In the companion paper (Part II), the proposed model will be used to simulate OECD lower head failure (OLHF) test data.

This paper defines a 3D full core neutronics benchmark which is based on the NuScale small modular reactor (SMR) concept. The paper provides a detailed description of the NuScale-like core, a list of expected outputs, and a reference solution to the benchmark exercises obtained with the Monte Carlo code Serpent.

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 ×5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.