Prescriptive approaches for the cybersecurity of digital nuclear instrumentation and control (I&C) systems can be cumbersome and costly. These considerations are of particular concern for advanced reactors that implement digital technologies for monitoring, diagnostics, and control. A risk-informed performance-based approach is needed to enable the efficient design of secure digital I&C systems for nuclear power plants. This paper presents a tiered cybersecurity analysis (TCA) methodology as a graded approach for cybersecurity design. The TCA is a sequence of analyses that align with the plant, system, and component stages of design. Earlier application of the TCA in the design process provides greater opportunity for an efficient graded approach and defense-indepth.

The TCA consists of three tiers. Tier 1 is design and impact analysis. In Tier 1 it is assumed that the adversary has control over all digital systems, components, and networks in the plant, and that the adversary is only constrained by the physical limitations of the plant design. The plant’s safety design features are examined to determine whether the consequences of an attack by this cyber-enabled adversary are eliminated or mitigated. Accident sequences that are not eliminated or mitigated by security by design features are examined in Tier 2 analysis. In Tier 2, adversary access pathways are identified for the unmitigated accident sequences, and passive measures are implemented to deny system and network access to those pathways wherever feasible. Any systems with remaining susceptible access pathways are then examined in Tier 3. In Tier 3, active defensive cybersecurity architecture features and cybersecurity plan controls are applied to deny the adversary the ability to conduct the tasks needed to cause a severe consequence. Tier 3 is not performed in this analysis because of the design maturity required for this tier of analysis.

Fluid elastic instability (FEI) is the most harmful excitation mechanism in flow-induced vibration. It is one of the important reasons related to the safety of reactor steam generator. In this paper, the FEI characteristics of regular triangular arrangement bundle and rotating regular triangular arrangement bundle with pitch-diameter ratio of 1.5 in water were studied experimentally. The results show that the critical reduced velocity of the regular triangular bundle is 3.0. The critical reduced velocity of the rotating triangular bundle is 1.7. In different stages of flow-induced vibration, the frequency spectrum characteristics of the vibration response of the rod bundle are significantly different. With the increase of reduced velocity, the vibration direction of the bundle changes from resistance direction to lift direction. It can be concluded that when the pitch-diameter ratio is the same, the regular triangle arrangement is more stable than the rotating regular triangle arrangement.

The purpose of the present study is to develop cyberattack early recognition system and to evaluate the effectiveness of the developed system by applying to the chemical plant testbed, which physically simulate the small scale chemical plant with industrial control system(ICS). The developed system is called CAeRS(Cyber Attack early Recognition System). The functional purpose of CAeRS is to help operators in recognizing the possibility of a cyber attack based on data from diverse sources. CAeRS has been developed from a hypothesis-based diagnosis system using Bayesian networks. The role of CAeRS is not to perform automatic recognition of cyber attacks but to provide operators with information required to enhance situational awareness concerning the state of the objective system. Information about the system is obtained in a step-by-step manner until the probability of a specific hypothesis becomes prominent. The performance of CAeRS has been validated using the chemical plant testbed.

The conventional failures have been physically simulated in the testbed to obtain cause-consequence relationship, which is required to build Bayesian Network. The cyber attacks have been simulated by the unauthorized parameter change, which result in the similar symptom with physical failure. Based on the scenario based evaluation, it has been demonstrated that the developed system can provide information to the operator, which helps to distinguish a specific cyber attack from a physical system failure with similar symptoms.

Accidental release of radioactive aerosols is a high possibility under severe nuclear accident. The aerosols may be transported and diffused to downwind of the nuclear site, and simultaneously settle on the ground. The movement of radioactive aerosols in the atmosphere would lead to severe radioactive contamination on the ground, and further polluting the environment and harming human health. Therefore, it is reasonable to make numerical study on the transport and deposition characteristics of radioactive aerosols in the atmosphere. In the present work, estimation method based on Mesoscale Eulerian coupling model was established. Meteorological condition and source term data of Chernobyl accident, aerosol concentration and deposition distribution after Chernobyl accident were simulated. The present study made it possible to apply mesoscale model to severe nuclear accident research, and can be helpful for the aftermath evaluation of severe nuclear accident in the future.

Since the 20th century, the world terrorist attacks, violent terrorist activities, extremist organization activities are increasing, the international community is facing the threat of nuclear terrorism and nuclear proliferation risk increasingly serious, anti-terrorism has become a regular trend. Digitalization and intelligentization have become the theme of today's era. With the continuous escalation of threats and risks, the security of nuclear materials and nuclear facilities is also facing new challenges. In order to better deal with threats and prevent and control risks, it is necessary to improve the technical level of risk analysis and assessment. This paper briefly describes the safety incidents of nuclear materials in various nuclear facilities and the significance of maintaining the safety of nuclear materials in nuclear facilities. Combined with the fault tree analysis method under risk management, this paper simulates the case of nuclear material theft and carries out event-based risk assessment and analysis. The analysis results show the effectiveness of this assessment method, which will provide technical support for the safety prevention of nuclear facilities and nuclear materials.

In pressure vessels, internals, etc., stresses can be identified as membrane, bending, or peak. This characterization is common in many design codes of different countries, such as the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

Finite element analysis is a common tool to provide accurate total stress results used for designing components. The resulting total stress needs to be characterized into membrane, bending, or peak categories to be compared to the design code criteria. Usually, the design codes provide calculation rules to linearize the stresses along a line section through the thickness of the analyzed component, usually called a Stress Classification Line (SCL) which is based on stress distribution in an axisymmetric pressure vessel submitted to internal pressure. On non-axisymmetric components, the SCL method can overestimate or underestimate the linearized stresses.

Even if the SCL is based on stress distribution in axisymmetric pressure boundary, the design code criteria are based on a limit load analysis of a rectangular crosssection bar under tension and bending loads. This article proposes to extend the actual SCL to a Stress Classification Plane (SCP), i.e. defined by a cross-section. The stresses are linearized over the entire cross-section that defines the SCP which makes the method consistent with the design code definitions and criteria. This article proposes to show the benefits, on a simple structure, of using the SCP approach.

Uncertainty Quantification(UQ) is widely used in the design and safety assessment of nuclear reactor system. As for UQ process, including forward UQ and inverse UQ, large quantities of numerical simulations are required for the uncertainty propagation which is very time consuming. To reduce the time costs, surrogate model is normally used to replace the calculation of reactor programs and with the development of technology, high precision surrogate model is constantly optimized by recent researchers. In this paper, three classical surrogate models which are Radial Basis Function (RBF) artificial neural network, Kriging, Support Vector Machine (SVM) regression and two optimized models which are Ensemble Surrogate (ES) model and Particle Swarm Optimization (PSO) Kriging model are introduced. These models are applied to different numerical examples with various numbers and distribution of parameters and with different complexity and degree of nonlinearity. Accuracy, efficiency and robustness are analyzed and compared with each models. Then surrogate models are tested with thermal hydraulic case of reflooding based on FEBA experiment. The results indicate that with small samples, PSO-Kriging model has the highest accuracy and robustness and has broad applicability for different cases. Thus, it is adopted to build the surrogate model of a typical Small Modular Reactor(SMR) for future process of uncertainty propagation, sensitivity or reliability analysis to replace the expensive models and reduce the computational cost.

A three-dimensional seismic isolation system is planed for application to the conceptual design of a sodium-cooled fast reactor in Japan. The main steam crossover piping is laid between the nuclear building with the isolation system and the turbine building without the isolation system. A large displacement of the nuclear building with the isolation system is imposed on the crossover piping, which situation is a particular seismic issue because of the isolation system employment. Furthermore, because the SFR operates at elevated temperatures compared with light water reactors, the crossover steam piping design must comply with the design code for elevated temperature components. In this study, seismic evaluation using an example of a crossover piping layout was performed in accordance with the elevated temperature code of Japan Society of Mechanical Engineers. According to the evaluation results and the up to date technologies such as knowledge obtained from existing dynamic failure tests of piping components, an appropriate seismic evaluation method for the crossover piping was studied to realize the sodium-cooled fast reactor with a three-dimensional isolation system.

This paper describes the contents of “Guidelines on Reliability Target Establishment and Conformity Evaluation for Passive Components” newly developed by the Japan Society of Mechanical Engineers (JSME). JSME developed the guidelines to provide a methodology that allows for the application of risk-informed and performance-based approaches to nuclear passive components. They are technology-inclusive covering new and existing reactors and applicable to all conduits in a lifecycle of a nuclear plant, i.e., design, fabrication, installation, inspection, operation and maintenance. The methodology consists of four steps: 1) task definition, 2) reliability target establishment for a component of interest, 3) structural reliability assessment of the component, and 4) conformity evaluation of the component to the reliability target (integrated decision making). One of the features of the guidelines is that conformity evaluation is performed in an integrated fashion, i.e., not only by comparing an established reliability target and assessed structural reliability but also by ensuring the validity of assessments, the sufficiency of structural reliability considering associated uncertainties, and the consistency with the design philosophy of nuclear power plants, i.e., defense-in-depth and maintaining margins. The first edition of the guidelines will focus on providing a general framework, while more detailed guidance alongside some worked examples is being elaborated to be implemented into the guidelines at the next revision.

Since the NRC issued the PSA policy in 1995, PSA containing risk information has been widely used in supervision and design. Subsequently, the concept of "risk-informed" has gained more and more recognition, and its application has expanded from the initial equipment maintenance to the current design. By analyzing the regulations, standards and other guidance documents of various countries and institutions on safety design, we reviewed the history of risk-informed safety design concepts, methods and their applications, especially the progress of LMP projects in the United States in recent years. We put forward prospects for the development and application of risk-informed concepts and methods in new reactors in the future.

The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes such as inherent safety, sustainable development, nuclear nonproliferation, natural resource protection, and economic efficiency. MSR can passively transfer the residual heat in the core to the environment by establishing the natural circulation after the reactor deactivates. However, the decay heat in the circuit piping will affect the heat removal capacity of the circuit. In this paper, to obtain the thermal characteristics of this passive system, the analysis basis of the natural circulation model is given to clarify the influence of molten salt properties and circuit structure on the natural circulation. By writing the code to design and establish the natural circulation model of the main loop of decay heat discharge from the liquid molten salt reactor, calculate and give the temperature distribution in the loop. The alteration of critical parameters can investigate the capacity of the reactor vessel to take away decay heat through natural circulation under certain temperature rise conditions. The results show that the decay heat in the system loop pipe reduces the heat that can take away from the reactor core. According to the results of the importance analysis, fuel salt density, specific heat capacity, and height difference between hot and cold cores are the critical parameters affecting the natural circulation capacity. When the parameter value changes by 15%, the capacity of decay heat discharge increases by 17.37%, 13.99%, and 11.82%.

The molten-core-coolant interaction is important in assessing the integrity of a reactor pressure vessel (RPV) and containment building (CB). In case of RPV failure during in-vessel retention (IVR), the breakage of the RPV will most likely occur at the level of the upper metallic layer due to the focusing effect. A possible steam explosion may result from the interaction between the metallic melt and water in the CB. If the metallic melt, composed of steel mixed with metallic species such as zirconium and uranium, undergoes oxidation during the premixing, triggering, and explosion phases, the melt oxidation influences the progress of the steam explosion. In this study, a small-scale experiment was conducted by dropping molten droplets composed of stainless steel mixed with zirconium into a water pool. The oxidation characteristics in water, such as drop oxidation, oxide film thickness, and element mapping of the solidified drops, were evaluated by an oxygen analyzer and Scanning Electron Microscope-Energy Dispersive X-ray Spectrometry (SEM-EDX) to clarify the effect of the metal composition of zirconium and stainless steel.

The probability of core disruptive accident (CDA) occurrence in sodium-cooled fast reactors (SFRs) is considered extremely low. However, for further verifying the safety and reliability of SFRs, the CDA sequence is still worth studying. In a case of SFR’s severe accident, such as unprotected loss of flow (ULOF), the CDA may be triggered, and then fuel and some fission products (or called source terms) may be released instantaneously from a CDA bubble through the potential leak paths on the vessel top slab or released with a delay from boiling sodium pool after vessel melt-through, widely known as instantaneous and delayed source terms. Therefore, reasonable prediction of CDA bubble behavior is necessary to investigate instantaneous source terms migration in the vessel pool. In this study, a simplified one-dimensional CDA bubble model that includes the formulation of thermal-hydrodynamic behaviors of the bubble mixture rising through the sodium pool toward the cover-gas region is proposed. The model includes mass transfer processes such as the condensation of gas mixture on liquid fuel/steel/sodium and the bubble interface. In this model, droplet entrainment phenomena at bubble interface are modeled based on Rayleigh-Taylor instability and Kelvin-Helmholtz instability, and the effect of non-condensable gas on condensation process is also considered. To validate the developed model, a past experiment on the expansion of high-pressure bubble in a stagnant liquid pool conducted by Purdue University in the late 1970’s using a 1/7-scale model of Clinch River Breeder Reactor was analyzed. The results showed generally good agreement with measured data and demonstrate that the developed model can reasonably represent the essential characteristics of dynamic behaviors of a high-pressure large-size bubble with heat and mass transfer at the bubble interface. This supports subsequent calculations to carry out the migration of transient source terms in the future.

Despite available correlations that can indicate the conditions for the point of onset of significant void (POSV) with reasonably high accuracy, the mechanism to cause OSV is still not understood sufficiently. The bubble coalescence-based mechanistic model proposed by Okawa (2021) yields a good agreement with the Saha-Zuber empirical correlation on the prediction of subcooling conditions at OSV. Therefore, experiments were carried out to investigate the possible role of bubble coalescence in OSV for subcooled flow boiling. The high-speed camera (HSC) was used to observe the bubble dynamics before and during the OSV occurrence. In contrast, a gamma densitometer and a void probe are used to measure the global and local void fraction at POSV.

The in-pile EAGLE ID1 test was conducted by Japan Atomic Energy Agency to demonstrate the effectiveness of the fuel assembly with an internal duct structure during a core disruptive accident in a sodium-cooled fast reactor. Post-test analysis using the SIMMER-III code are based on a multi-fluid model that uses empirical models in constitutive equations, making it difficult to accurately simulate multi-component, multi-phase flows with complex heat and mass transfer. In this study, a new computational fluid dynamics code based on the fully Lagrangian particle method was developed for the purpose of clarifying the failure mechanism of the inner duct wall of FAIDUS. The three-dimensional simulation of the ID1 test was performed to analyze a series of thermal hydraulic behaviors leading up to duct wall failure for a computational domain that included six fuel pins, i.e., 1/12.5 of the circumference of the test section. The simulations reasonably reproduced the heat transfer characteristics observed in the test, showing that the local contact of liquid steel with high thermal conductivity with the duct wall greatly enhances the heat transfer from the nuclear heating fuel to the duct wall. The present large-scale simulation produced the results that were essentially equivalent to those obtained in a smaller simulation system with three fuel pins in our previous work. The results support the validity of the conclusions of our analytical study regarding the molten pool-to-duct wall heat transfer mechanism that caused the thermal failure of the duct wall.

In the two-phase flow model, the accurate prediction of Critical Heat Flux (CHF) is related to the safety margin of reactor design. In this paper, the CHF calculation model is developed based on the Artificial Neural Network (ANN) and random forest model, the accuracy of the model is verified, and the most suitable model for application is selected. The selected CHF model was coupled with RELAP5, and the coupled program was verified based on the Thermal-Hydraulic Test Facility of the Oak Ridge National Laboratory (ORNL-THTF). The results show that the CHF model based on the machine learning algorithm has good accuracy in calculation, and RELPA5 with the new CHF model is closer to the experimental data in wall temperature calculation. The purpose of optimizing the program is achieved.

Whole process exploration and transient thermal-hydraulic parameters acquisition of Steam Generator Tube Rupture (SGTR) Accident are of significant importance for safety analysis of Lead-cooled Fast Reactors (LFRs). The Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST) for transient experiments of lead-based eutectic (LBE)-water-steam three-phase was built. Three thermal-hydraulic phenomena were analyzed. These include 1) initial pressure peak measured and pressure attenuation equations fitted, 2) LBE pressure rising curves and its equilibrium with water side, 3) steam cavity migration monitored by thermocouples. The latest results showed that the accident pressure in LBE side rose with a higher water initial pressure 7.5MPa. The attenuation curve of pressure peak was represented by simple harmonic vibration. Besides, the injection of supercooled water incurred temperature fluctuations in LBE pool, with an instantaneous 81.6 ℃ temperature drop near the injection nozzle, and then LBE temperature re-entered a new dynamic equilibrium.

The present study is relating the reactor power control using a system for removing fission product gases from a molten salt fast reactor. Because molten salt reactors are equipped with a fission gas removal system, such as Xe gas, controlling the reactor power using this system has been studied. In this system, the Xe gas is removed by injecting helium gas bubbles into the molten fuel salt. When the void fraction of helium gas increases by 1%, the reactivity of approximately -380 pcm δk/k is introduced. Transient behaviors under a single-phase flow are analyzed by RELAP5-3D code, and FLUENT code with a user defined function which incorporates the point kinetics model, assuming reactivity insertion. To verify that the single-phase flow calculations make sense, we are also calculating two-phase flow using FLUENT. As a result, it is conceivable that the reactor could be controlled by increasing or decreasing the void fraction, and the reactor is tripped if a void fraction increases by more than 4%. At reactor start-up, operation starts with a void fraction of about 4%, and by increasing the flow rate of molten fuel salt resulting in reducing the void fraction to about 0.4%, the reactor power can be increased to the rated power.

High flow rate and complex structure of sodium-cooled fast reactor may cause gas entrainment phenomena. A kind of entrainment induced by unstable vortex is relatively more likely to occur in the hot pool, seriously threatening the safe operation. Computational fluid dynamics (CFD) is a significant and suitable tool to investigate this phenomenon and reveal its mechanism. Nevertheless, the vortex dynamics and its formation mechanism as well as interface topological changes caused by entrainment represents still a challenge. In order to build up a reliable numerical method for unstable vortex entrainment, some efforts were devoted to the proper choice of turbulence models, wall function and other numerical settings. The simulation attempts as well as preliminary results were presented and discussed. The experiment case for comparison was also described briefly. The comparison showed that a method of 3D transient VOF simulation with RNG k-ε model is able to reproduce the unstable vortex entrainment well, but there are still some points that need further improvement and optimization, especially concerning the theoretical analysis of swirl intensity under different conditions and the selection of appropriate turbulence models and near-wall treatments.

In this paper, the transport phenomena and heat transfer mechanisms associated with saturated flow boiling of water in a large aspect ratio narrow channel were computationally investigated. A two-dimensional computational model has been developed for analysis of the interfacial behavior of flow boiling with flow patterns transiting from bubbly to full confined slug flow in the narrow rectangular channel based on the multi-phase volume of fluid (VOF) model combined with phase change model. The numerical results are compared with visualized experimental results of detailed information on flow pattern and phase distribution. The computed results show reasonable agreement with experimentally captured interfacial behavior. The experimental and numerical results provide better understanding of the mechanisms of flow boiling heat transfer in narrow channels. Moreover, the established numerical modelling framework in the present paper can provide an alternative and a promising tool for the design and optimization of compact heat exchangers with narrow channels such as in the integrated small modular reactors (SMR) or other industrial applications.

This study examines the drag force on the spherical fuel element in the vertical pipe of the fuel handling and storage system for HTRs. The flowing media in this study are high-pressure helium and atmospheric air. The force on the spherical fuel element in the pipe is examined by the simulation computation of the flow field using SST (Shear-Stress Transport) k-ω and LES (Large Eddy Simulation) models. For various inlet flow velocities, the relationship between the drag force on the spherical element in the cylindrical pipe and the inlet flow velocity is computed. Compared with the SST k-ω turbulence model, the LES model simulation data are closer to the experimental results. Results show that the blockage ratio of the sphere to the pipe has a great influence on the drag force of the spherical element. When the blockage ratio exceeds 0.887, the drag force and drag coefficient significantly rise as the blockage ratio rises. The study of the force characteristics of spherical fuel elements in vertical pipes of fuel handling and storage systems provides technical support for improving the efficiency of spherical element transportation. Future investigation on the flow characteristics of narrow spaces around the sphere inside the pipes would support further engineering improvements for the pneumatic transportation of solids.

Since Three Mile Island accident, the small break losing of coolant accident (SBLOCA) was identified as the important accident during the nuclear reactor operating. Many facilities were built to understand the phenomena during SBLOCA and worked to improve the safety of 2nd generation nuclear power plant (NPP). Now the days, the core makeup tank （CMT）, used to be a coolant source during the high-pressure phase in SBLOCA, had been introduced with the ideal of passive core coolant system (PCCS) into 3rd generation NPP. The earlier study focus on the automatic depressurization system (ADS), accumulator (ACC), and passive residual heat removal system (PRHR). In our simulation on the advanced plant experiment (APEX), the integral test facility of AP-1000, and the advance core-cooling mechanism experiment (ACME), the integral test facility of CAP-1400, by RELAP 5 MOD3.2 shows that the main error on the core makeup tanks (CMTs) injection phase is an important error source in the SBLOCA simulation of PWR with PCCS. In the high-pressure phase of the nature-cycle, the phenomena in the CMT and Pressure Balance Line (PBL) loops lead the CMT level decrease slowly. In this study, we tried to identify the main leading phenomenon.

In the lead-cooled fast reactor, solid-phase oxygen control is recognized as an efficient approach to mitigate the liquid lead-bismuth eutectic (LBE) corrosion on the structural steel materials. The dissolution of solid lead oxide (PbO) particles in the mass exchanger (MX) is the oxygen source to make sure the oxygen concentration in LBE in the reasonable ranges. In this work, a non-equilibrium mass transfer model is developed for the solid-phase oxygen control loop to discuss the oxygen control mechanism. The numerical model was established by the LBE mass, momentum, temperature and oxygen concentration balance equations. The fluid-particle mass transfer coefficient is calculated by the heat-mass analogy theory with the well-known Ranz model of the heat convection in packed bed, which is applied in many related experimental and computational studies. With the simulation results under different conditions, it is found that the oxygen concentration in LBE is enhanced greatly under high operating temperature by higher solubility, higher diffusion coefficient and higher Sherwood number with lower LBE viscosity. Moreover, the effects of inlet velocity and void fraction are much less than the temperature. Present work provides meaningful simulation methods and advices for controlling oxygen concentration in LBE and designing the engineering loop.

Fission products are released from the core as aerosols during severe nuclear reactor accidents. Most aerosols dispersed in containment will be deposited on the floor and could be washed by condensate flow with the migration of radioactivity. Investigating the aerosol wash-down process is interesting for evaluating radioactivity in nuclear power plant containment. THAI-AW3-LAB experiment was carried out by Becquerel to study the wash-down process of insoluble aerosols on walls, based on which parameters such as the wash-down rate and flow coverage were obtained. However, injecting water flow in the investigation simulated the condensation process on a plate, and experimental verification of the natural aerosol wash-down process on walls of a condensation environment needs to be improved. To further explore the aerosol erosion and wash functions on containment floor, an experiment on aerosol wash-down characteristics of a horizontal plate in condensing environment is carried out in this paper. Experimental visualization records the complete process of aerosol erosion and wash-down, focusing on how aerosols move with condensate in environments with different deposition densities of aerosols. The washing rate and share of aerosols under different deposition environments are discussed. Experimental results show that wash-down process may have three different stages with time. This paper describes a complete process in which aerosols on a horizontal plate are washed by the growing or flowing condensate. For example, a considerable part of deposited aerosols can quickly enter the condensate and move with condensate in an early stage of the growth of the droplets. Still, the flowing condensate may hardly erode the remaining aerosol on the plate. The washing rate varies with the washing process and shows a long-term exponential washing process lasting for several hours.

Natural convection heat transfer outside the horizontal cylinder is a basic heat transfer process, which has a wide range of applications in engineering fields, such as pipe transportation, column heat sink, heat dissipation of containment, etc. In response to the current lack of experimental correlations at high Ra numbers, this paper adopts an experimental approach to carry out an experimental study of the natural convection heat transfer characteristics outside the cylinder under high Ra number conditions by constructing a large-size (d=2 m) horizontal cylindrical natural convection heat transfer experimental device. Through the analysis of available data, it is found that the Morgan formula and Churchill-Chu formula, which are commonly used for the calculation of natural convection heat transfer outside the horizontal cylinder, have large deviations from the experimental data under high Ra number (Ra>10⁹) conditions. In this paper, a new experimental correlation suitable for high Ra number (10² <Ra <10¹³) is obtained by collecting previous published experimental data and combining them with the data obtained in this experiment.

The current reactor at the NIST, the National Bureau of Standards Reactor (NBSR) was operational in 1967, serving as a premiere user-facility to the international neutron scattering research community. The NBSR’s age has contributed to difficulties like longer outage times and increased maintenance costs, prompting an investigation into a replacement design. The NIST Neutron Source (NNS) is a proposed replacement design to replace the current NBSR. This paper investigates the thermalhydraulic behavior of the NNS’ compact core preliminary design using computational fluid dynamics (CFD) analysis. While developing the CFD model, any flow irregularities may significantly affect thermal-hydraulic characteristics such as the core’s pressure or velocity profiles. Therefore, a turbulence model must be carefully selected to balance computational costs and model uncertainties. This paper details a sensitivity analysis with multiple turbulence models to evaluate the resulting hydraulic behavior of the coolant flowing from the inlet plenum to the core. Various Reynolds Averaged Navier stokes (RANS) turbulence models in ANSYS Fluent® such as k-ε, k-ω, k-ω SST, realizable k-ε and Spallart-Allmaras are compared. The resulting velocity and pressure profiles for each turbulence model are compared for fit. Discussions of the mesh, assumptions, and boundary conditions are also provided in the text, demonstrating the limitations and methodologies of the study.

The NIST Neutron Source, or NNS, is a proposed new research reactor at the NIST Center for Neutron Research in the United States of America. The NNS will serve as a state-of-the-art source for neutron scattering and irradiation experiments for the domestic and international community, replacing the currently operational but aging National Bureau of Standards Reactor (NBSR) onsite. The reactor will utilize U-10Mo LEU plate fuel assemblies and be moderated and cooled with light water in an upflow forced convection setup. A custom thermal-hydraulics 1D model is developed to perform preliminary assessments of the core’s flow behavior and evaluate the core’s safety margins. This paper is concerned with the critical heat flux ratio (CHFR) and the onset of flow instability ratio (OFIR) as safety margins. Under nominal conditions, it is found that both limits remain above the minimum threshold of 2, which is recommended by the US nuclear regulatory commission’s NUREG-1537 publication. To evaluate the uncertainty in the safety margins, input sensitivity analyses are performed with stochastic and deterministic approaches. The custom model and sensitivity analysis methodologies are detailed in the text, where uncertainties are provided for both safety margins in terms of expected power, mass flow rate, and inlet coolant temperature variations. The results demonstrate how the sensitivity of the safety margins to the inputs varies based on the sensitivity analysis method, where a deterministic approach shows more linearized trends compared to the stochastic trends. Discussions regarding how the results affect the design are included in the text.

The once through steam generator has great heat transfer capacity and compact structure. Some small module reactors (SMR) applied once through steam generatorfor its high thermal efficiency. The thermal hydraulic characteristic of once through steam generator is different from traditional natural cycle steam generator as it doesn't contain water level. Thus the related feed water system is important for the steady operation of the power plant. Base on the thermal hydraulic characteristic of once through steam generator, SMR feed water system that contains process and control model is developed using APROS. The operating parameters are simulated under both different steady states and transient condition. The results show that both the thermal hydraulic and automatic control parameters of SMR feed water system can be simulated accurately in APROS. Reference and guidance can be provided for system optimization and safety analysis.

The vapor-liquid one way and two-way coupling numerical simulation of the subcooled bubbly flow boiling in a vertical heated circular tube with diameter of 15.4 mm at 4.5 MPa is carried out by using discrete phase model (DPM) Combined with the boiling bubble boundary layer model based on the Eulerian-Lagrangian (EL) method in this paper. Initial bubble distribution information in flow boiling in the bubble boundary layer can be obtained by calculating the bubble active nucleation sites density, the lift-off frequency, the lift-off diameter and the lift-off velocity using the bubble boundary layer model. The motion of the bubble in the subcooled main flow can be described by solving the displacement equation and the motion equation for the bubbles. The DPM accounts for bubble–liquid interaction using two-way fluid–bubble coupling which is accomplished through the interphase momentum transfer rate. The void fraction variation results of numerical simulation are compared with the corresponding experimental results and those obtained by one-way coupled E-L numerical simulation. Results show that the two-way coupled simulation method predicts more accurate void fractions than the one-way coupled simulation method, although still underpredicts the void fractions compared with the experimental values.

In a nuclear reactor accident, radioactive aerosols may be released into the gas and retained in the pool through engineering safety design or the deposition of aerosols. The bubbles generated in the pool through gas injection or heating and boiling rise to the free surface and burst to produce film and jet droplets, which can entrain the aerosol in the liquid into the gas. In this process, many parameters influence each other and lead to changes in the generation of droplets. The applicability of the calculation model of key parameters under different working conditions needs to be further verified. This paper uses high-speed photography to visualize the phenomenon of a single bubble burst to produce film droplets. The bubble diameter, bubble lifetime, bubble film opening rate, bubble film thickness, and the size and number of film droplets are obtained by image processing software. It is found that the bubble lifetime distribution obeys the Weibull distribution with a shape parameter of 4/3 in deionized water, but the shape parameters shift to 3/2 in 0.05 g/L TiO2 aerosol suspension. The film thickness and bubble lifetime show an exponential decay law with a parameter of 2/3. The number of film droplets decreases with the increase in film thickness, and the size of film droplets increases with the increase of the film thickness.

Passive safety systems, which are independent of any external input or energy to operate, are considered the most significant candidate to increase the inherent safety of nuclear power plants. Due to the lower driving forces, passive safety systems are more susceptible to interference from external conditions compared to active safety systems, which can affect system performance. Functional failure is an important factor contributing to the operational failure of passive safety systems and needs to be considered in their reliability analysis. However, the assessment of functional failure depends on numerous T-H simulations, which in reality takes a lot of time. Two approaches have been proposed to solve this problem: one is a low-fidelity model that ignores the details of the original model and retains only the main framework of the original model; the other is to use a data-driven model to approximate the model response, which can also be called a metamodel. However, its effectiveness is severely constrained by the quantities of input parameters. The increase in the number of parameters generates a geometric increase in uncertainty. And in order to avoid "dimensional disasters", it is crucial to identify sensitive parameters, which helps to reduce the model complexity and improve the analysis efficiency. In this paper, the T-H model of the Integral-type Pressurized Water Reactor was developed, and based on this model, the elementary effects method and the variance-based method were used to analyze the parameter sensitivity. An efficient and accurate sensitivity analysis framework was proposed in which the number of factors was reduced using the elementary effects method, and then the ANOVA was executed on initially screened parameters to identify key parameters of the PRHRS in IPWR200.

This research focuses on safety criteria for the design of high flux research reactors with downflow systems. Generally, critical heat flux must be considered during the design of the reactor system. The Sudo93 correlation is widely used for prediction of critical heat flux due to its wide applicability for high flux research reactors. In addition, high flux research reactors typically operate under low pressure conditions and they are prone to various flow instabilities especially under accident conditions. Instabilities are likely to cause premature critical heat flux, and thus the onset of flow instability must be given enough attention. The onset of flow instability can be predicted by Whittle-Forgan correlation, Saha-Zuber correlation or Dougherty correlation. To investigate the relationship between critical heat flux predicted by Sudo93 correlation and the onset of flow instability, the critical heat flux databases of Sudo85 and Sudo93 correlations are studied and analyzed. It’s found that the critical heat flux databases contain not only critical heat flux of stable flow but also critical heat flux of unstable flow, which means that critical heat flux predicted by Sudo93 correlation may be affected by flow instabilities. Furthermore, Sudo93 correlation is used to deal with experimental data of the onset of flow instability. It’s found that Sudo93 can predict the onset of flow instability for downward flow. To account for this finding, Sudo93 correlation is compared with Dougherty correlation and Whittle-Forgan correlation. This paper finds that Sudo93 correlation has the similar characteristics with Dougherty correlation and Whittle-Forgan correlation. Besides, the uncertainty of Sudo93 correlation can be accounted for from the perspective of onset of flow instability.

In this paper, the effect of inclination angle on the heat transfer characteristics of water vapor condensation in a tube is studied. The experiments are conducted in a 25 mm inner diameter 1050 mm long inclined tube with an inclination angle of -40 to 40°, a pressure of 0.2 MPa, and a mass flux of 20 to 50 kg/m2s. The results show that the inclination angle and the heat transfer coefficient show non-monotonic variations. At low mass flux and low quality, the inclination angle significantly affects the heat transfer coefficient of condensation and peaks at around ±25° in all cases, and the maximum improvement is 72% compared to the horizontal heat transfer coefficient. The effect of the inclination angle on the heat transfer coefficient diminishes as the mass flux and vapor quality increase. Based on the force balance analysis, the dimensionless parameter F is obtained, and the heat exchange region is divided into the inclination effect dependent zone and the inclination effect independent zone. The experimental values under horizontal conditions are compared with the heat transfer correlations, and it is found that Shah and Cavallini heat transfer correlations can accurately predict the experimental values within ±25% deviation.

Motivated by understanding of severe accident phenomena, extensive studies have been conducted to investigate the thermal-hydraulics of both debris bed and melt pool which may form during a severe accident of light water reactors. However, little work has been done to address the debris remelting phenomenon following a uncoolable debris bed, although it plays an important role in melt pool formation and accident progression. The present study is intended to fill the knowledge gap. For this purpose, a test facility was developed to investigate the remelting process of a particulate bed packed with two-component particles. Electric heating is employed to simulate the decay heat. This paper presents the detailed description of the test facility design, including the rationales of the simulant materials for the oxidic and metallic components of corium, estimation of heating power distribution, design simulation of the heating rod, selection and validation of instrumentation as well as commissioning of the experimental facility.

Steam jet direct contact condensation is widely used in various fields due to its extremely high heat and mass transfer rate. The software FLUENT was employed to numerically study the direct contact condensation process of pure steam jet through square nozzle with different length-width ratios. The thermal equilibrium phase change model was inserted into FLUENT as a user defined function (UDF) to simulate the heat and mass transfer process in DCC. The results show that the shape of the plume under the condition of square nozzle jet is similar to that of round nozzle jet, and the distribution of axial thermal parameters is similar. For square nozzles with the same equivalent diameter, the peak velocity and jet penetration length increase with the increase of nozzle length width ratio. For square nozzles with the same short side length, the peak velocity and jet penetration length increase with the increase of nozzle length of the long side. However, when the length of the long side is greater than a certain value, further increasing the length of the long side has little effect on the jet penetration length.

scheme of the full-cycle of bubble growth size is proposed. Based on several databases of public experimental data, results are calculated according to the proposed calculation scheme. The relative error of the experimental data is within 20%. The influence of different boundary conditions on bubble size and characteristic time points is also explored in this paper. Except the validation of proposed scheme, the sensitivity analysis of boundary conditions is also carried out in this paper. Results shown the important effect of pressure, heat flux density, inlet mass flow rate and inlet supercooling on the of bubble diameter and characteristic time points.

The single-cell thermionic fuel element (TFE) is a promising candidate for the compact nuclear power system. Fuel mass transfer occurs due to high operating temperature. On the other hand, the single-cell TFEs are connected in series in the thermionic nuclear reactor. In view of the above challenge to the safety and efficiency of the single-cell TFE, the thermionic conversion model was modified and the circuit model was developed in FROBA-THERMION, a steady-state performance analysis code for the single-cell TFE. Further, a group of the single-cell TFEs connected in series was chosen and a steady-state performance simulation was carried out. The results indicate that the thermionic conversion efficiency will be decreased due to the emitter temperature change caused by fuel mass transfer. Influenced by the connection in series, the output voltage of the single-cell TFE with a higher emitter temperature is increased to match the output current of the one with a lower emitter temperature.

The SPECTRA code has been developed as an integrated safety analysis tool for comprehensive analyses of in- and ex-vessel phenomena during a severe accident (SA) in a sodium-cooled fast reactor (SFR). The individual modules, such as sodium fire and sodium-concrete reaction, have been implemented into SPECTRA to simulate specific phenomena in an accident of SFRs. However, SPECTRA has no capability for the overlapped event involving sodium fire and sodium-concrete reaction because the sodium pool and the floor concrete are modeled in each module independently. In this study, the capability of SPECTRA is enhanced by integrating the analyses of sodium pool fire and concrete ablation for overlapped events of the ex-vessel phenomena. The sodium pool fire module is connected to the shared module for the sodium pool and the floor concrete developed in our previous study. The developed model is validated through the benchmark analysis of the F7-1 pool fire experiment. The calculation result of the pool and catch pan temperature shows good agreement with the experimental data. A demonstration analysis is also conducted for an overlapped event of the ex-vessel phenomena. SPECTRA can simulate a reasonable heat and mass transfer behavior.

This paper numerically studied natural convection heat transfer augmentation from a local heat source placed in a vertical annular channel, using annular fins. The effects of fin spacing, fin span, and fin arrangement were investigated at constant wall heat flux boundaries. The results of the present analyses indicated that heat transfer rates had a strong dependence on fin spacing (number of fins) and fin span. The fin span was varied from 4.5% to 18% of the annular gap. On increasing the fin span at a particular fin spacing, the heat transfer coefficient increased continuously. When the fin span was largest in the middle section of the heat source, heat transfer was increased by 18%, while it increased by 19% for the uniformly spanned fins of the span l = 0.18b . The results of this study may serve as a reference resource for the thermal design of an intricate air-cooled passive system.

If the water piping system under high temperature and pressure is damaged and water is ejected into the atmosphere, a jet with depressurization boiling (Flashing) occur. Human safety and safety equipment may be damaged by the impingement of the flashing jet. Therefore, it is necessary to evaluate the region affected by the flashing jet impinging on the surrounding equipment and people.

In this study, we investigated experimentally to confirm the validity of the region affected by the flashing jet with the saturated and subcooled water under the conditions of low pressure at the jet inlet up to 2 MPa, and further investigated under the conditions of high pressure up to 7 MPa, which are difficult in the experiment, using Computational Fluid Dynamics (CFD).

From the experimental results, as the subcooling temperature increased, the mass flux and spread angle of the flashing jet increased and the region affected by the flashing jet increased. The mass flux calculated by the Homogeneous Equilibrium Model tended to be in good agreement with the experimental results.

The asymptotic plane distance and the jet width of the flashing jet were well organized by the density ratio at the nozzle outlet. By using the correlations between the density ratio and the asymptotic plane distance and the jet width, it is possible to evaluate the impact region of the flashing jet, which includes conditions from subcooled water to saturated water.

MARS is a thermal-hydraulic system code developed for nuclear power plant design and safety analysis. The code is well known for its robustness and versatility. It has the advantage of realistic calculations of the two-phase flow behaviors in transient and steady states and has been verified using various experimental data. In the previous study, we tried to apply the MARS code to refrigeration cycle analysis. We implemented the thermodynamic properties of refrigerant R410A into the code and modified further. The code showed good prediction performance for experimental data such as boiling, condensation heat transfer, and pressure drop of the refrigerant. In this study, to determine the applicability of the MARS code to more complex refrigeration cycle analysis, we simulated a system air-conditioner performance test using the experimental data of LG Electronics’s multi system air conditioner. For the simulation of the system air conditioner, we inserted the compressor performance map into the code and implemented the loss coefficient according to the opening ratio of the expansion valve. Also, considering the characteristics of the various heat exchangers installed respectively, the fin and shape of each heat exchanger were modeled using the heat transfer area option. From the simulation results, it was shown that the code could predict the steady state and transient state of the multi system air conditioner. The MARS code reasonably predicted the mass flow rate, pressure behavior compared to the experiment. Also, the limitations and further improvements of the code are identified from the simulation results.

The accurate identification of the flow regime is important for two-phase simulation in nuclear reactor design and analysis. In this paper, the flow regime identification model was developed based on the decision tree (DT) algorithm, random forest (RF) algorithm, and artificial neural network (ANN) respectively. Training data was collected from available literature and classified into nine flow regimes. The input parameters of the flow regime identification model were superficial liquid velocity, superficial gas velocity, hydraulic diameter, pressure, and flow direction. Then the three flow regime identification models were evaluated by precision, recall, and F-score which is the composite value of precision and recall. Moreover, the computation time of the three models was also compared. The results show that flow regime identification models based on the RF algorithm and DT algorithm have better accuracy than ANN. The accuracy of the model based on RF and Dt is 94.2% and 97.0% respectively while the accuracy of the model based on ANN is 77.0%. What’s more. the models based on the DT algorithm and ANN have less time-costing than the RF algorithm. The predicting time cost of the models based on the DT algorithm and ANN is 0.47s and 0.74s respectively while the predicting time cost of the model based on RF is 47.0s.

Helical coil once-through tube steam generator (HCOTSG) has the advantages of compact structure and higher heat transfer efficiency due to its special structure. In order to study the flow boiling heat transfer characteristics in HCOTSG, this paper used the FLUENT commercial software, based on the Critical heat flux(CHF) wall boiling model, to conduct numerical simulation in the spiral tube under the given wall temperature boundary conditions, and verified its reliability by comparing the relevant experimental research. On this basis, the flow boiling heat transfer process in the spiral tube was simulated under different flow parameters, and the velocity field, temperature field and the distribution of vapor-liquid two-phases in the spiral tube were studied, It is found that under the condition of constant wall temperature boundary, the influence of different inlet temperature only exists in the single-phase water region near the inlet, and has little influence on the boiling two-phase region and single-phase steam region. The void fraction decreases at the same height in the boiling two-phase region with the increase of inlet flow velocity, which is due to the decrease of the heating time when the fluid reaches the same height.

After a severe PWR accident, the residual heat of the core cannot be discharged in time which results in an increase of the temperature of fuel rods. When the temperature is larger than the melting point, the reactor core would melt and fall to the lower head of the vessel. Severe reaction would occur between the melting reactor core and the residual coolant which make the melt change to debris bed. In the following process of severe accident, the remelting of debris bed may occur if the cooling ability is not sufficient. Finally, a molten pool may form in the lower head of reactor pressure vessel.

In this paper, Models are established for the remelting of debris bed and formation of molten pool separately based on the energy conservation equation. Inside, the convective and radiative heat transfer are considered among the debris bed, the double-layer molten pool structure and the reactor internal components. The model was validated using the IVR analysis results calculated by the ERI, INEEL and DOE models. The AP600 and AP1000 are tested under different conditions. The results show good performance in all the cases. In the azimuthal angle of the lower head, all the models show same results in the bottom of the vessel while the largest error is found in the side aspect. The reason may be the simplified module for the calculation of metal layer of molten pool. However, the largest error of proposed model is lower than 15% for all the parameters when compared with the other models. So, it is concluded that the analytical model can be used to analyze the process characteristics of debris bed remelting and molten pool formation through a simplified method, which provides technical support for in-depth understanding of the serious accident process of pressurized water reactors.

A unique bubble foam flow presents in subcooled flow boiling of 3.5 wt% artificial seawater. Depending on the inlet temperatures of seawater, slug bubbles resulting from foam ruptures might prevail during the boiling period. It is of importance to reveal the Sauter mean diameters of bubbles in foam flow under different experimental conditions for seawater. Such characteristic of bubbles including maximum/minimum diameters and Sauter mean diameters is collected and analyzed from shadowgraph measurements by Image J as current datasets covering various inlet temperatures, heat and mass fluxes for both seawater and de-ionized water. Evidence is presented in this paper to describe the relationship between the maximum/minimum diameters and the Sauter mean diameters. Two existing correlations assuming the Sauter mean diameter is proportional to the maximum diameter are adopted in comparisons with current datasets. One relationship correlating the Sauter mean diameter with the shape characteristic including maximum and minimum diameter is also compared with the dataset. Large mean deviations of 57.10-80.27% and 54.06-76.79% for seawater and de-ionized water are found, respectively. It suggests that the poor applicability of these linear correlations. A new linear relationship with a different constant of 0.376 as its slope for seawater is shown at last with the mean deviation of 8.99% only. It provides the possibility to predict the Sauter mean diameters only with known measurements of geometric characteristics of bubbles from footages by high-speed cameras.

In the framework of the IAEA Coordinated Research Project on In-Vessel Melt Retention, a benchmark of CFD simulations, devoted to thermal-hydraulic behavior of the light metal layer involves several research organizations: KTH of Sweden, SSTC NRS of Ukraine, ÚJV Řež of Czech Republic and CEA of France.

This work aims at better simulating the focusing effect phenomenon leading to a heat flux peak along the height of the light metal layer, which is formed above the oxide layer in a stratified corium pool configuration during a PWR severe accident. This is a known safety issue compromising the reactor vessel integrity. The first benchmark step provides a solid foundation to the CFD schemes (physical models, meshes) by comparing the results of CFD simulations with thermal-hydraulic experimental data obtained using water as simulating fluid in a representative and quite laminar configuration. Then a similar but highly turbulent case, of higher height, is considered for more complex validation of the numerical simulation approach. Results with different turbulent models are compared against experimental data. On the strength of this encouraging work, a simulation of the same height configuration but considering steel fluid under severe accident conditions is foreseen at the final stage of this benchmark.

In Floating Nuclear Power Plant, the equipment with free liquid surface will bring heave, sway and other motions. The fluid dynamic potential energy conversion and additional inertial force will cause the free liquid surface wave, resulting large nonlinear fluctuation of pressure and flow in the system equipment, especially for the downstream pipeline. and aggravating the instability of the system equipment operation. Therefore, it is necessary to simulate the variation of typical free liquid surface equipment of dynamic system in Floating Nuclear Power Plant, and clarify the effect of pressure fluctuation on the accessorial piping system under the sloshing effect. In this study, the sloshing effect on fluid in equipment was simulated by CFD software, the pressure fluctuation data was obtained and transferred to the one-dimensional system analysis software. It is concluded that the sloshing effect of the free liquid surface in vessel only has certain influence on the pressure fluctuation at the inlet of piping system, while the pressure fluctuations of the downstream is dominated by the floating movement.

Supercritical carbon dioxide (SCO₂) has the advantages of high thermal efficiency and compact layout in applications due to its excellent transport properties and lower critical point (p_cr = 7.38 MPa, T_cr = 30.98 ℃). It can be considered as a modelling fluid instead of water for due to significantly lower critical parameters. This paper concentrates on the thermal-hydraulic characteristic of SCO₂ in tube to provide a reliable database for future verification. An experimental loop was designed and constructed for SCO₂ recently. Detailed component description and parameter information are described in the text. The designed maximum pressure and temperature of the system are 15 MPa and 200 °C, respectively, accompanied with circulating mass flux of 0-22 L/min in the loop. Calibrations and uncertainty analysis has been made before experimental tests. Preliminary experiments were carried out to collect heat-transfer data at three pressures above the critical point (7.6, 8.4 and 9.5 MPa), mass fluxes from 200 to 1000 kg/m²s, heat fluxes up to 300 kW/m² and inlet temperatures from 20 to 40°C. The results are used to verify previous data and scaling laws.

When core meltdown accident occurs in PWR nuclear power plant, core melt will gather at the lower head of reactor pressure vessel (RPV) and then form a molten pool, which will threaten the integrity of reactor pressure vessel. The in-vessel retention (IVR) through external reactor vessel cooling (ERVC) technology is a performance method to alleviate core meltdown accident. Therefore, accurately predicting the heat transfer characteristics of external channel of the RPV lower head will significantly mitigate the severe core meltdown accident of nuclear power plant. By analyzing the flow heat transfer characteristics of the internal molten pool and the external flow passage involved in ERVC, the simulation calculation method of the flow heat transfer of the internal molten pool and the external flow passage of the coupled pressure vessel was established. Based on ULPU structural parameters and the input description of Energy Research, Inc. (ERI) benchmark calculations, a homogeneous flow natural circulation flow model and a two-layer melt pool heat transfer model were established. By calculating the natural circulation rate and convective heat transfer coefficient and other parameters, the feasibility of the established ERVC simulation calculation method was verified.

In the event of severe accidents in light-water reactors, the core melt, which falls to the bottom of the containment vessel, is expected to be cooled in the prestored water. The particle debris is generated after the molten material falls from the pressure vessel. The cooling characteristics remains to be further elucidated, for example, in a system where particle debris and structure material coexist, and the gas-liquid two-phase flow inside particulate debris. In this study, air–water two-phase flow inside a particulate bed and near a structure wall was investigated using a high-speed camera, with index matching of the bed with pure water and a wire-mesh sensor. Particles with diameters of ø3, 5, and 10 mm were used and subjected to air–water two-phase flow tests with the superficial gas velocity of 4–2000 mm/s and superficial liquid velocity of 0.5–75.3 mm/s. Bubble behavior near the wall and inside the bed was observed, with bubbles splitting in the Y-shaped channels; however, the frequency of bubble coalescence was relatively low. The spatial connection between the pores in the bed enhanced the advection of bubbles from within the bed to the vicinity of the structure wall.