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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%.

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.

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 °C was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

Pressure relief valve (PRV) is one of the important control valves used in nuclear power plants, and its sealing performance is crucial to ensure the safety and function of the entire pressure system. For the sealing performance improving purpose, an explicit function that accounts for all design parameters and can accurately describe the relationship between the multi-design parameters and the seal performance is essential, which is also the challenge of the valve seal design and/or optimization work. On this basis, a surrogate model-based design optimization is carried out in this paper. To obtain the basic data required by the surrogate model, both the Finite Element Model (FEM) and the Computational Fluid Dynamics (CFD) based numerical models were successively established, and thereby both the contact stresses of valve static sealing and dynamic impact (between valve disk and nozzle) could be predicted. With these basic data, the polynomial chaos expansion (PCE) surrogate model which can not only be used for inputs-outputs relationship construction, but also produce the sensitivity of different design parameters were developed. Based on the PCE surrogate model, a new design scheme was obtained after optimization, in which the valve sealing stress is increased by 24.42% while keeping the maximum impact stress lower than 90% of the material allowable stress. The result confirms the ability and feasibility of the method proposed in this paper, and should also be suitable for performance design optimizations of control valves with similar structures.

Nuclear power plants have recognized the importance of nuclear cybersecurity. Based on regulatory guidelines and security-related standards issued by regulatory agencies around the world including IAEA, NRC, and KINAC, nuclear operating organizations and related systems manufacturing organizations, design companies, and regulatory agencies are considering methods to prepare for nuclear cybersecurity. Cryptographic algorithms have to be developed and applied in order to meet nuclear cybersecurity requirements. This paper presents methodologies for validating cryptographic algorithms that should be continuously applied at the critical control system of I&C in NPPs. Through the proposed schemes, validation programs are developed in the PLC, which is a critical system of a NPP’s I&C, and the validation program is verified through simulation results. Since the development of a cryptographic algorithm validation program for critical digital systems of NPPs has not been carried out, the methodologies proposed in this paper could provide guidelines for Cryptographic Module Validation Modeling for Control Systems in NPPs. In particular, among several CMVP, specific testing techniques for ECB mode-based block ciphers are introduced with program codes and validation models.

In this paper, the evaluation method of fluidelastic instability (FEI) of newly designed steam generator tubing in pressurized water reactor (PWR) nuclear power plants is discussed. To obtain the parameters for prediction of the critical velocity of FEI for steam generator tubes, experimental research is carried out, and the design parameters are determined. Using CFD numerical simulation, the tube array scale of the model experiment is determined, and the experimental device is designed. In this paper, 7 groups of experiments with void fractions of 0% (water), 10%, 20%, 50%, 75%, 85% and 95% were carried out. The critical damping ration, fundamental frequency and critical velocity of FEI of tubes in flowing water were measured. Through calculation, the total mass and instability constant of the immersed tube are obtained. The critical damping ration measured in the experiment mainly included two-phase damping and viscous damping, which changed with the change in void fraction from 1.56% to 4.34%. This value can be used in the steam generator design described in this paper and is conservative. By introducing the multiplier of frequency and square root of total mass per unit length, it is found that the difference between the experimental results and the calculated results is less than 1%, which proves the rationality and feasibility of the calculation method of frequency and total mass per unit length in engineering design. Through calculation, the instability constant is greater than 4 when the void fraction is less than 75%, less than 4 when the void fraction exceeds 75% and only 3.04 when the void fraction is 95%.

Kori Unit 1 is planning a system decontamination project to reduce radiation exposure of decommissioning workers, prevent the spread of contamination and down-grade the level of classification of radioactive waste. The system decontamination range for Kori Unit 1 will be the entire primary system, including RCS, CVCS and RHRS. Some system design modifications are required for the system decontamination operation. In this paper, major system design modifications were evaluated based on the conditions that system restoration is needed after completion of system decontamination. The major system design modifications are CIDF connection location to system, system decontamination operating pressure control, RCP seal water injection and formation of letdown flow. It was evaluated that there was no negative effect on the system due to the system design modifications. However, as the RCP seal water is injected into the system in the oxidation process, the concentration of the oxidizing agent is diluted. Therefore, the oxidizing agent injection and system decontamination operation procedures should be developed to address the dilution effect of the oxidizing agent. The system design modifications dealt in this paper will be finally confirmed through on-site investigation in the future, and if necessary, the system design modifications will be re-evaluated.

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on “Neutronics Benchmark of CEFR Start-Up Tests” offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

The variational nodal method for solving the neutron transport equation has evolved over 40 years. Based on a functional form of the Boltzmann neutron transport equation, the method now comprises a complete set of variants that can be employed for different problems. This paper presents an extensive review of the development of the variational nodal method. The emphasis is on summarizing the whole theoretical system rather than validating the methodologies. The paper covers the variational nodal formulation of the Boltzmann neutron transport equation, the Ritz procedure for various application purposes, the derivation of boundary conditions, the extension for adjoint and perturbation calculations, and treatments for anisotropic scattering sources. Acceleration approaches for constructing response matrices and solving the resulting system of algebraic equations are also presented.

The supercritical carbon dioxide (S–CO2) Brayton cycle is an important energy conversion technology for the fourth generation of nuclear energy. Since the printed circuit heat exchanger (PCHE) used in the S–CO2 Brayton cycle has narrow channels, Rayleigh-Bénard (RB) convection is likely to exist in the tiny channels. However, there are very few studies on RB convection in supercritical fluids. Current research on RB convection mainly focuses on conventional fluids such as water and air that meet the Boussinesq assumption. It is necessary to study non-Boussinesq fluids. PRB convection refers to RB convection that is affected by horizontal incoming flow. In this paper, the computational fluid dynamics simulation method is used to study the PRB convection phenomenon of non-Boussinesq fluid-supercritical carbon dioxide. The result shows that the inlet Reynolds number (Re) of the horizontal incoming flow significantly affects the PRB convection. When the inlet Re remains unchanged, with the increase of Rayleigh number (Ra), the steady-state convective pattern of the fluid layer is shown in order: horizontal flow, local traveling wave, traveling wave convection. If Ra remains unchanged, as the inlet Re increases, three convection patterns of traveling wave convection, local traveling wave, and horizontal flow will appear in sequence. To characterize the relationship between traveling wave convection and horizontal incoming flow, this paper proposes the relationship between critical Reynolds number and relative Rayleigh number (r).

Based on the Deep Q-Network(DQN) algorithm of reinforcement learning, an active fault-tolerance method with incremental action is proposed for the control system with sensor faults of the once-through steam generator(OTSG). In this paper, we first establish the OTSG model as the interaction environment for the agent of reinforcement learning. The reinforcement learning agent chooses an action according to the system state obtained by the pressure sensor, the incremental action can gradually approach the optimal strategy for the current fault, and then the agent updates the network by different rewards obtained in the interaction process. In this way, we can transform the active fault tolerant control process of the OTSG to the reinforcement learning agent's decision-making process. The comparison experiments compared with the traditional reinforcement learning algorithm(RL) with fixed strategies show that the active fault-tolerant controller designed in this paper can accurately and rapidly control under sensor faults so that the pressure of the OTSG can be stabilized near the set-point value, and the OTSG can run normally and stably.

There is an increasing interest in passive safety systems to minimize the need for operator intervention or external power sources in nuclear power plants. Because a passive system has a weak driving force, there is greater uncertainty in the performance compared with an active system. In previous studies, several methods have been suggested to evaluate passive system reliability, and many of them estimated the failure probability using thermal–hydraulic analyses and the Monte Carlo method. However, if the functional failure of a passive system is rare, it is difficult to estimate the failure probability using conventional methods owing to their high computational time. In this paper, a procedure for the application of the Chernoff bound to the evaluation of passive system reliability is proposed. A feasibility study of the procedure was conducted on a passive decay heat removal system of a micro modular reactor in its conceptual design phase, and it was demonstrated that the passive system reliability can be evaluated without performing a large number of thermal–hydraulic analyses or Monte Carlo simulations when the system has a small failure probability. Accordingly, the advantages and constraints of applying the Chernoff bound for passive system reliability evaluation are discussed in this paper.

The once-through steam generator used in the small modular reactor needs to consider the stability of the outlet steam pressure and steam superheat of the secondary circuit to achieve better operating efficiency. For this reason, this paper designs a controllable operation scheme for the steam pressure and superheat of the small reactor once-through steam generator. On this basis, designs a variable universe fuzzy controller, first, design the fuzzy control rules to make the controller adjust the PI controller parameters according to the change of the error; secondly, use the domain adjustment factor to further subdivide the input and output domain of the fuzzy controller according to the change of the error, to improve the system control performance. The simulation results show that the operation scheme proposed in this paper have better system performance than the original scheme of the small reactor system, and controller proposed in this paper have better control performance than traditional PI controller and fuzzy PI controller, what's more, the designed control system also showed better anti-disturbance performance in lifting experiment between 100% and 80% working conditions. Finally, the experimental platform formed by connecting the digital small reactor with Matlab/Simulink through OPC(OLE for Process Control) communication technology also verified the feasibility of the proposed scheme.

The present paper proposes some new computational methods and results in the framework of flow computation through congested domains seen as porous media, as it can be found in the core of a Pressurized Water Reactor (PWR). The flow is thus mostly governed by the distribution of pressure losses, both through the porous structures, such as fuel assemblies, and in the thin fluid layers between them. The purpose of the present paper is to consider the question of the interaction of a flow and a rod bundle from an analytical point of view gathering all the contributions through a set of equations as simple and representative as possible. It aims at demonstrating a sound understanding of the relevant phenomena governing the flow establishment in the geometry of interest instead of relying mainly on a posteriori observations obtained both experimentally and numerically. Comparison with two set of experimental results showed good agreement. The model proposed being analytical it appears easily implementable for studies needing an expression of fluid forces in a rod array as for fuel assembly bowing issue. It would be interesting to test the reliability of the model on other geometry with different P/R ratios.

Abstracts: The research on the flow and heat transfer characteristics of lead bismuth(LBE) is significant for the thermal-hydraulic calculation, safety analysis and practical application of lead-based fast reactors(LFR). In this paper, a new CFD model is proposed to solve the thermal-hydraulic analysis of LBE. The model includes two parts: turbulent model and turbulent Prandtl, which are the important factors for LBE. In order to find the best model, the experiment data and design of 19-pin hexagonal rod bundle with spacer grid, undertaken at the Karlsruhe Liquid Metal Laboratory (KALLA) are used for CFD calculation. Furthermore, the turbulent model includes SST k−ω and k−ε; the turbulent Prandtl includes Cheng-Tak and constant (Prt =1.5,2.0,2.5,3.0). Among them, the combination between SST k−ω and Cheng-Tak is more suitable for the experiment. But in the low Pe region, the deviation between the experiment data and CFD result is too much. The reason may be the inlet-effect and when Pe is in a low level, the number of molecular thermal diffusion occupies an absolute advantage, and the buoyancy will enhance. In order to test and verify versatility of the model, the NCCL performed by the Nuclear Thermal-hydraulic Laboratory (Nuthel) of Xi'an Jiao tong University is used for CFD to calculate. This paper provides two verification examples for the new universal model.

Energetic particle strikes the device and induces data corruption in the configuration memory (CRAM), causing errors and even malfunctions in a system on chip (SoC). Software-based fault injection is a convenient way to assess device performance. In this paper, dynamic partial reconfiguration (DPR) is adopted to make fault injection on a Xilinx 16 nm FinFET Ultrascale+ MPSoC. And the reconfiguration module implements the Sobel and Gaussian image filtering, respectively. Fault injections are executed on the static and reconfiguration modules' bitstreams, respectively. Another contribution is that the failure modes and effects analysis (FMEA) method is applied to evaluate the system reliability, according to the obtained injection results. This paper proposes a software-based solution to estimate programmable device vulnerability.

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. Accordingly, HRA community has emphasized the accumulation of HRA data to support HRA practitioners for many decades. To this end, it is critical to resolve practical problems including (but not limited to): (1) how to collect HRA data from available information sources, and (2) how to inform HRA practitioners with the collected HRA data. In this regard, the U.S. Nuclear Regulatory Commission (NRC) and Korea Atomic Energy Research Institute (KAERI) independently initiated two large projects to accumulate HRA data by using full-scale simulators (i.e., simulator data). In terms of resolving the first practical problem, the NRC and KAERI developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. In addition, to inform HRA practitioners, the NRC and KAERI proposed several ideas to extract useful information from simulator data. This paper is the first of two papers to discuss the technical underpinnings of the development of the SACADA and HuREX systems.

This paper proposes an algorithm for signal validation using unsupervised methods in emergency situations at nuclear power plants (NPPs) when signals are rapidly changing. The algorithm aims to determine the stuck failures of signals in real time based on a variational auto-encoder (VAE), which employs unsupervised learning, and long short-term memory (LSTM). The application of unsupervised learning enables the algorithm to detect a wide range of stuck failures, even those that are not trained. First, this paper discusses the potential failure modes of signals in NPPs and reviews previous studies conducted on signal validation. Then, an algorithm for detecting signal failures is proposed by applying LSTM and VAE. To overcome the typical problems of unsupervised learning processes, such as trainability and performance issues, several optimizations are carried out to select the inputs, determine the hyper-parameters of the network, and establish the thresholds to identify signal failures. Finally, the proposed algorithm is validated and demonstrated using a compact nuclear simulator.

This paper aims to present a new method for obtaining an analytical solution for the Kaniadakis Doppler broadening (KDB) function. Also, in this work, we report the computational efficiencies of this solution compared with the numerical one. The solution of the differential equation achieved in this paper is free of approximations and is, consequently, a more robust methodology for obtaining an analytical representation of ψk. Moreover, the results show an improvement in efficiency using the analytical approximation, indicating that it may be helpful in different applications that require the calculation of the deformed Doppler broadening function.

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. In this regard, it is vital to provide credible HEPs based on firm technical underpinnings including (but not limited to): (1) how to collect HRA data from available sources of information, and (2) how to inform HRA practitioners with the collected HRA data. Because of these necessities, the U.S. Nuclear Regulatory Commission and the Korea Atomic Energy Research Institute independently developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. These systems provide unique frameworks that can be used to secure HRA data from full-scope training simulators of NPPs (i.e., simulator data). In order to investigate the applicability of these two systems, two papers have been prepared with distinct purposes. The first paper, entitled “SACADA and HuREX: Part 1. The Use of SACADA and HuREX Systems to Collect Human Reliability Data”, deals with technical issues pertaining to the collection of HRA data. This second paper explains how the two systems are able to inform HRA practitioners. To this end, the process of estimating HEPs is demonstrated based on feed-and-bleed operations using HRA data from the two systems.

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (θ) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified θ projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.

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.