Search

The binary decision diagram (BDD) method provides a means to calculate the exact probability of a fault tree, but cannot solve large fault trees for nuclear power plant PSAs. However, the BDD method can be used as a supplementary means to increase the accuracy of PSA quantification, especially for a seismic PSA which includes many non-rare events.Since it is difficult to solve a large PSA model using BDD method, several approaches have been developed that partially apply the BDD technique to small-sized branches of a large PSA model. This paper proposes an approach of partially applying BDD technique to large-sized branches. The results from the pilot application to a seismic PSA model shows that the partial BDD approach of this paper provides a relatively accurate way to quantify each sequence as well as the overall core damage in a seismic PSA model.

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.

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.

As the realization of carbon neutrality has been a main assignment for coping with the global climate change, it became necessary to analyze upcoming changes in electricity mix with economic and technical viewpoints. This paper presents a newly-developed simulation model that reflects the daily intermittency of renewable energy by applying daily average power supply-demand patterns for each season. Also, the paper provides an economic analysis in the viewpoint of investment cost, annual cost and power generation cost by utilizing the calculations from the simulation model. Four scenarios are selected for the analyses, one based on the Korean Government's 2050 Carbon Neutrality Scenario and three Nuclear Power Build-up scenarios, which are newly suggested by the authors. The simulation results show that the increase of nuclear energy from 5.7% of Government's Scenario to 37.7% of the proposed Nuclear Power Build-up Scenario leads to the decrease of about 704 billion US$ in investment cost and about 181 billion US$ in annual cost; with reduction in the increase of 2050 expected generation cost from 3.1 to 1.6 times compared with the referenced 2021 average cost. Further, this study has significance in performing the economic analysis with the expected daily power supply-demand patterns in 2050.

In this paper, the similarity criterion and dimensionless conversion method combined with the elasticity condition and Hooke's law are used to derive the functional relationship of the maximum effective value of the vibration velocity between the prototype pump and the model pump. The seawater circulation pump of a nuclear power plant is used as the prototype pump, and the model pump is obtained by performance conversion and choosing the appropriate scale, and the vibration state of the model pump under different flow rates is measured and analyzed. The vibration data of the model pump through the function relationship to find out the vibration parameters of the prototype model pump, and compare with the vibration data of the seawater circulation pump in reality. It can be seen that with the increase of flow rate, the maximum effective value of the vibration velocity of both model and prototype decreases and then increases, and the relative error is small, the maximum value is 7.7757%. Therefore, it can be considered that the functional relationship of model pump converted to prototype pump derived in this paper can be used to analyze the vibration of the actual seawater circulation pump of coastal nuclear power plant.

For the pressurized water reactor two-step calculation, the traditional assembly homogenization and two-group neutron diffusion calculation have been widely used. When it comes to the core pin-by-pin simulation, many models and techniques are different and unsettled. In this paper, the homogenization methods based on the pin discontinuity factors and super homogenization factors are used to get the pin-cell homogenized parameters. The heterogeneous leakage model is applied to modify the infinite flux spectrum of the single assembly with reflective boundary condition and to determine the diffusion coefficients for the SP3 solver which is used in the core simulation. To reduce the environment effect of the single-assembly reflective boundary condition, the on-line method for the SPH factors updating is applied in this paper, and the functionalization of SPH factors based on the least-squares method will be pre-made alone with the table of the group constants. The fitting function will be used to update the thermal-group SPH factors with a whole-core pin-by-pin homogeneous solution on-line. The three-dimensional Watts Bar Nuclear Unit 1 (WBN1) problem was utilized to test the performance of pin-by-pin calculation. And numerical results have demonstrated that PWR pin-by-pin core calculation has more accurate results compared with the traditional assembly-homogenization scheme.

As a typical active equipment, pump machinery is widely used in nuclear power plants. Although the mechanism of pump machinery in nuclear power plants is similar to that of conventional pumps, the safety and reliability requirements of nuclear pumps are higher in complex operating environments. Once there is significant performance degradation or failure, it may cause huge security risks and economic losses. There are many pumps mechanical parameters, and it is very important to explore the correlation between multi-dimensional variables and condition. Therefore, a condition monitoring model based on Deep Denoising Autoencoder (DDAE) is constructed in this paper. This model not only ensures low false positive rate, but also realizes early abnormal monitoring and location. In order to alleviate the influence of parameter time-varying effect on the model in long-term monitoring, this paper combined equidistant sampling strategy and DDAE model to enhance the monitoring efficiency. By using the simulation data of reactor coolant pump and the actual centrifugal pump data, the monitoring and positioning capabilities of the proposed scheme under normal and abnormal conditions were verified. This paper has important reference significance for improving the intelligent operation and maintenance efficiency of nuclear power plants.

Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations.To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes.This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.

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.We evaluate the uncertainties in pulse physical parameters (maximum power, total released energy and full width at half maximum) due to uncertainties in reactor physical parameters (inserted reactivity, delayed neutron fraction, prompt neutron lifetime and effective temperature reactivity coefficient of fuel). It is found that taking into account overestimated correlation of reactor physical parameters does not significantly affect the estimated uncertainties of pulse physical parameters. The relative uncertainties of pulse physical parameters decrease with increasing inserted reactivity. If all reactor physical parameters feature an uncorrelated uncertainty of 10 % the estimated total uncertainty in peak pulse power at 3 $ inserted reactivity is 59 %, where significant contributions come from uncertainties in prompt neutron lifetime and effective temperature reactivity coefficient of fuel. In addition we analyse contribution of two physical mechanisms (Doppler broadening of resonances and neutron spectrum shift) that contribute to the temperature reactivity coefficient of fuel. The Doppler effect contributes around 30 %–15 % while the rest is due to the thermal spectrum hardening for a temperature range between 300 K and 800 K.

Commercial deployment of nuclear microreactors presents an opportunity for the industry to rethink its approach to manufacturing, siting, operation and maintenance, and fuel cycle management as certain principles used in grid-scale nuclear projects are not applicable to a decentralized microreactor economy. The success of this nascent industry is dependent on its ability to reduce infrastructure, logistical, regulatory and lifecycle costs. A utility-like ‘Central Facility’ that consolidates the services required and responsibilities borne by vendors into one or a few centralized locations will be necessary to support the deployment of a fleet of microreactors. This paper discusses the requirements for a Central Facility, its implications on the cost structures of owners and suppliers of microreactors, and the impact of the facility for the broader microreactor industry. In addition, this paper discusses the pre-requisites for eligibility as well as the opportunities for a Central Facility host site. While there are many suitable locations for such a capability across the U.S., this paper considers a facility co-located with the Vogtle Nuclear Power Plant and Savannah River Sites to illustrate how a Central Facility can leverage the existing infrastructure and stimulate a local ecosystem.

Drawing on the deep experience and understanding of the principles of nuclear safety, as well as many years of nuclear power plant design and operation, the EDF led NUWARD SMR Project is developing a design for a Small Modular Reactor (SMR) of 340 MWe composed of two 170 MWe independent units, that will supplement the offering of high-output nuclear reactors, especially in response to specific needs such as replacement of fossil-fuelled power plants. NUWARD SMR is a mix of proven and innovative design features that will make it more commercially competitive, while integrating safety features that comply with the highest international standards. Following the principles of redundancy and diversity and rigorous application of Defence in Depth (DID), with an international view on nuclear safety licensing, the Project also incorporates new safety approaches into its design development.The NUWARD SMR Project has been in development for a number of years, it entered conceptual design formally in mid-2019 and entered Basic Design in 2023. The objective of the concept design phase was to confirm the project technological choices and to define the first design configuration of the NUWARD SMR product, to document it, in order to launch pre-licensing with the French Safety Authority (ASN) and to define its estimated cost and its subsequent development and construction schedules. As a delivery milestone the Safety Options file (called the Dossier d’Options de Sûreté (DOS)) has been submitted to ASN in July 2023 for their opinion.An integral part of the NUWARD SMR Project, is not only to deliver a design suitable for France and to satisfy French regulation, but to develop a product suitable and indeed desirable, for the international market, with a first focus in Europe. In order to achieve its objectives and realise its market potential, the NUWARD SMR Project needs to define and realise its safety approach within an international environment and that is the key subject of this paper. The following paper:• Summarises the foundation principles and technological background which underpin the design;• Contextualises the key design features with regard to the international safety regulatory framework with particular emphasis on innovative passive safety aspects;• Illustrates the Project activities in preparation for first licensing in France, and also a wider international view via the ASN led Joint Early Review of the NUWARD SMR design, including Finnish and Czech Republic regulators, recently joined by the Swedish, Polish and Dutch regulators;• Articulates the collaborative approach to design development from involvement with the Project partners (the Commissariat à l’énergie atomique et aux énergies alternatives (CEA), Naval Group, TechnicAtome, Framatome and Tractebel) to the establishment of the International NUWARD Advisory Board (INAB), to gain greater international insight and advice;• Concludes with the focus on next steps into detailed design development, standardisation of the design and its simplification to enhance its commercial competitiveness in a context of further harmonisation of the nuclear safety and licensing requirements and aspirations.

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.

After a loss of coolant accident (LOCA) in the prestressed concrete containment vessels (PCCVs) of nuclear power plants, the coupling of temperature and pressure can significantly affect the mechanical properties of the PCCVs. However, there is no consensus on how this coupling affects the failure mechanism of PCCVs. In this paper, a simplified finite element modeling method is proposed to study the effect of temperature and pressure coupling on PCCVs. The experiment results of a 1:4 scale PCCV model tested at Sandia National Laboratory (SNL) are compared with the results obtained from the proposed modeling approach. Seven working conditions are set up by varying the internal and external temperatures to investigate the failure mechanism of the PCCV model under the coupling effect of temperature and pressure. The results of this paper demonstrate that the finite element model established by the simplified finite element method proposed in this paper is highly consistent with the experimental results. Furthermore, the stress-displacement curve of the PCCV during loading can be divided into four stages, each of which corresponds to the damage to the concrete, steel liner, steel rebar, and prestressing tendon. Finally, the failure mechanism of the PCCV is significantly affected by temperature.

In this paper, elastic-plastic fracture mechanics analysis is performed to determine the critical crack sizes of the multipurpose canister (MPC) manufactured using austenitic stainless steel under dynamic loading conditions that simulate drop accidents. Firstly, dynamic finite element (FE) analysis is performed using Abaqus v.2018 with the KORAD (Korea Radioactive Waste Agency)-21 model under two drop accident conditions. Through the FE analysis, critical locations and through-thickness stress distributions in the MPC are identified, where the maximum plastic strain occurs during impact loadings. Then, the evaluation using the failure assessment diagram (FAD) is performed by postulating an external surface crack at the critical location to determine the critical crack depth. It is found that, for the drop cases considered in this paper, the principal failure mechanism for the circumferential surface crack is found to be the plastic collapse due to dominant high bending axial stress in the thickness. For axial cracks, the plastic collapse is also the dominant failure mechanism due to high membrane hoop stress, followed by the ductile tearing analysis. When incorporating the strain rate effect on yield strength and fracture toughness, the critical crack depth increases from 10 to 20%.

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.

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.

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.

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 Kd of 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 Kd of the mixture was obtained as 160 L kg−1 for the batch method, 23 L kg−1, 19 L kg−1, and 18 L kg−1 for 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.

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 Sm2O3 and Gd2O3 demonstrate 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.

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.

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.

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.

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.

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.

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 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.

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.

The transportation of spent nuclear fuel is an important process that involves road and sea transport from an interim storage facility to storage and final disposal sites. As spent nuclear fuel poses a significant risk, carefully evaluating its vibration and shock characteristics under normal transport conditions is essential. In this regard, full-scale multi-modal transport tests (MMTT) have been conducted domestically and internationally. In this paper, we discuss the process of developing a multi-body dynamics (MBD) model to analytically simulate conditions that cannot be considered in tests. The MBD model is based on the KORAD-21 transportation system was validated using the Korean MMTT results from 2020 to 2021. This paper summarizes the details of the development and verification of the MBD model for the KORAD-21 transportation system under normal transport test conditions. This approach can be applicable to various transportation scenarios and systems, and the results of this study will help to ensure that nuclear fuel transportation is conducted safely and effectively.

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.

A pressurized vessel-pipe-safety valve (PVPSV) combination is a commonly used configuration in nuclear power plants, and a good numerical model is essential for the system design, sizing and performance optimization. However, owing to the large-scale and cross-scale features, it is still a challenge to build a system level numerical model with both high accuracy and efficiency. To overcome this, a novel system level modeling method which can synthesize the advantages of various models is proposed in this paper. For system modeling, the analytical approach, the method of characteristics (MOC) and the surrogate model approach are respectively adopted to predict the dynamics of the pressure vessel, the connecting pipe and the safety valve, and different models are connected through data interfaces. With this system model, dynamic simulations were carried out and both the stable and the unstable system responses were obtained. For the model verification purpose, the simulation results were compared with those obtained from experiments and full CFD simulations. A good agreement and a better efficiency were obtained, verifying the ability of the model and the feasibility of the modeling method proposed in this paper.

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 °C and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ∼268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m2 for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

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.

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.

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 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.

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.

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.

This paper presents a computational framework for drop time assessment of a control element assembly (CEA) under fuel assembly (FA) deformations. The proposed framework consists of three key components: 1) finite element modeling of CEA, 2) fluid-structure interaction to compute drag force, and 3) modeling of frictional contact between CEA and FA. Specially, to accommodate the large motion of CEA, beam elements based on absolute nodal coordinate formulation (ANCF) are adopted. The continuity equation is utilized to calculate the drag force, considering flow changes in the cross-sectional area during the CEA drop. Lastly, beam-inside-beam frictional contact model is employed to capture practical contact conditions between CEA and FA. The proposed framework is validated through experiments under two scenarios: free falls of CEA within FA, encompassing undeformed and deformed scenarios. The experimental validation of the framework demonstrated that the drop time of CEA can be accurately predicted under the complex coupling effects of fluid and frictional contact. The drop times of the S-shaped deformation case is longer than those of the C-shaped deformation case, affirming the time delay due to frictional force. The validation confirms the potential applicability to access the safety and reliability of nuclear power plants under extreme conditions.