The Lead-cooled Fast Reactors are one of the most promising Gen. IV nuclear systems, currently under development. They can ensure enhanced performances and minimal waste production, but they also have some issues to be addressed. The long-term degradation process caused by Pb corrosion effects, in structural materials facing liquid lead is one of the most critical. The present state of the art showed that commercial stainless steels as AISI 316L and 15-15Ti uncoated can be adopted as structural materials in lead environments below 500 °C, beyond which they start to experience dissolution of constituting alloying elements (e.g., Ni, Cr, Fe).

Current R&D activities on corrosion-resistant materials are pursuing mainly two treads to push temperature threshold beyond 550 - 600 °C: alumina-based coating to be applied on commercial alloys and advanced alloys able to withstand lead corrosion. Among alumina-based coating technologies, Pack Cementation is an attractive option because of its potential good corrosion performance up to 500 °C and capability to coat complex geometries and the already existing availability at an industrial scale. On the other hand, advanced alloys such as the Kanthal FeCrAl have proven excellent corrosion resistance performances at temperatures beyond 600-650 °C.

In MSR, liquid fuel uses molten salt, usually a eutectic mixture with a low melting point and a high boiling point. At high temperatures, molten salt has heat transfer properties similar to water. Molten salt has a strong tendency to corrode materials, its properties are easily changed because it reacts sensitively to oxygen and moisture. In the reactor, molten salt circulates in an airtight structure with a metallic material and is exposed to nuclear fission reactions, corrosion reactions, and transmutation reactions, producing various by-products. By-products make to change and accumulate the component and composition of molten salt fuel over time. The associated thermal physics and thermal chemicals properties can be changed. This study fabricated the thermal convection loop for the multi-purpose molten salt experiments. The molten salt experiment was performed using a multi-purpose molten salt experimental loop to evaluate the corrosion and thermodynamic properties of the molten salt.

Air ingress into the core of molten salt reactor (MSR) in an accident scenario can cause serious safety problems and especially damages to molten salt infiltrated graphite components. However, the effect of air exposure on the oxidation behaviour of molten salt infiltrated graphite remains elusive. In this study, the effect of alkali metal fluorides, i.e., LiF, NaF, KF, and ternary salt FLiNaK, on the oxidation behaviour of graphite powder was systematically investigated under air atmosphere. Oxidation tests show that the molten-salt-infiltrated graphite samples oxidized at 600–900 °C in air exhibit increased weight loss compared with the virgin sample, and the weight loss follows the order of KF-G > NaF-G > LiF-G. The fluoride salts also reduce the activation energy of the air-exposed graphite at elevated temperatures, and the catalytic oxidation efficiency of the alkali metal fluorides follows the order of KF > NaF > LiF. Scanning electron microscopy reveals that the fluoride salts induce active sites on the surface of graphite plane, which can accelerate the oxidation reactions of graphite in air. It is explained that the alkali metal atoms can be adsorbed on the surface of the graphite planes to activate the carbon atoms to show a high affinity for binding of the oxygen atoms and, thus, to accelerate the graphite oxidation. The results have gained deeper insight into the air oxidation mechanism of fluoride-saltinfiltrated graphite and are meaningful for improving the fabrication technology of nuclear graphite to ensure the operation safety of MSR during air ingress accidents.

Accurate fault diagnosis of nuclear power plants (NPPs) is essential to the prevention of catastrophic accidents and ensures a stable operation condition. Digital twin (DT) has emerged as a data and model-based systematic approach which can generate fault state data similar to the actual system and provide a new paradigm for fault diagnosis. This paper firstly puts forward the concept of DT and its evolution progress. Then a conceptual architecture of DT based on the five components is presented due to lack of studies on the application of DT for fault diagnosis in NPPs. Given the nonlinear dynamics and uncertainty involved during the process of fault diagnosis, challenges and future development prospects to proper design and adaptability of a DT model in fault diagnosis of NPPs are also discussed.

The Radiation Monitoring System (RMS) at the NIST Center for Neutron Research (NCNR) Reactor (National Bureau of Standards Reactor – NBSR) has demonstrated reliability issues due to outdated hardware and lack of vendor support. The existing radiation monitors communicate to a controller in the control room, which then displays an LCD segmented display to operators. If the value is over its pre-defined threshold, the machine actuates an annunciator to indicate an unsafe condition. Over time, multiple RMS devices have gone out of commission, being unable to communicate with the centralized controller, and localized monitors had to be used to monitor area radiation safety. The original manufacturer of the system has since gone out of business and attempts to repair and replace broken components have been fruitless. It is due to these reasons that a new system has been engineered using modern control and communication systems integrated into the NBSR Control Console Upgrade.

Intermediate heat exchangers transfer high-temperature heat from the very-high-temperature gas-cooled reactor (VHTR) to the process heat application facility such as hydrogen production, which is a key component in nuclear heat utilization. To ensure that the IHX provides sufficient heat for process heat applications, different types of corrugated tubes are used to enhance the heat transfer capacity of IHX. Precise geometry parameters are crucial in the study of the thermalhydraulic performance of the corrugated tubes. Using the highprecision industrial cone beam CT can achieve the nondestructive and high-precision measurement, and obtain the precise geometric parameters of the tubes inner surface. However, due to the strong attenuation characteristics of metal materials on X-ray, there are serious scattering, hardening and statistical noise in CT when scanning the corrugated tubes. These factors lead to strong metal artifacts in the reconstructed images, which seriously affect subsequent process such as segmentation and parameter measurements. We propose a method based on generative adversarial networks in order to correct metal artifact while preserving the details of the CT image features well. The proposed method generates CT image samples through simulation methods and obtain the mapping relationship between metal-artifact-containing and metalartifact-free images by training. Validation of simulated CT images and corrugated CT image datasets, the method proposed in this paper preserves the feature information such as CT image size and boundary while correcting metal artifacts well. And it provides accurate parameters for further research on the thermal-hydraulic performance of corrugated tubes, reduces the differences or deviations between experiments and numerical simulations..

Water cooled reactors experience reactivity changes when voids are present in the coolant. The fraction of coolant in the core diminutions in the event of boiling, and the reactivity feedback can either be positive due to reducing neutron absorption in the coolant or can be negative due to reduction in the slowing down of neutrons to thermal energy because of a reduced moderator to fuel ratio. Similarly, mechanical movement of core internals, specifically within the active core volume do cause power fluctuations. Nevertheless, coolant voiding, and mechanical movements are among the main perturbations that induce reactor power oscillations, which can also be defined as the reactor power noise. A real-time noise detection system can track the reactor power measurements, and alarm if the oscillations exceed statistically significant levels. This study presents a method for estimating the noise in reactor power measurements using a real-time signal-to-noise ratio calculation. The effectiveness of the proposed noise calculation methodology is demonstrated using data from the National Bureau of Standards Reactor (NBSR) under both normal operating conditions and fuel failure events. Additionally, this paper describes the detailed design of a noise detection system hardware based on this methodology.

Modular Pebble-bed HTGR (High-Temperature Gascooled Reactor) employs online HPGe (High Purity Germanium) detector systems to measure the nuclide inventory in spherical fuel elements. The determination of the detector’s DE (Detection Efficiency) curve involves Monte Carlo modeling of the detection system. However, the actual dimension parameters of the HPGe detectors often deviate from the nominal values provided by the manufacturer due to manufacturing limitations and accumulated dead layer growth. Therefore, the calculated DE is subject to systematic errors when the detector model is constructed according to nominal values in Monte Carlo simulations.

Focusing on the HTR-10 online burnup measurement system, this work first performs a sensitivity analysis of the DE curve to relevant dimension parameters using Monte Carlo method and analyzes the effect of each variable on the DE profile between 344 and 1408 keV. Then four meta-heuristic algorithms including GA (Genetic Algorithm), DE (Differential Evolution), PSO (Particle Swarm Optimization) and SA (Simulated Annealing) are incorporated to find the optimal modeling parameter combination that minimizes the discrepancy between the measured and calculated DEs. In addition, different algorithms are compared in terms of accuracy and convergence rates to assess the applicability of meta-heuristic algorithms to such detector parameter optimization problems.

Sensitivity analysis reveals that the side dead layer thickness, germanium crystal radius, crystal height, copper radius, and source-detector distance are five key parameters in the determination of DE profiles with different energy selectivity characteristics. After the optimization, the difference between the calculated and measured DEs is reduced to less than or equal to the measurement uncertainty of the measured DEs. Therefore the optimized model can be applied to the efficiency calibration of the above-mentioned detector system to improve its measurement accuracy. Finally, three group intelligence algorithms present satisfactory optimization capability and convergence rate, and the method developed in this paper can be extended to other HPGe systems.

District heating based on fossil energy is an important source of carbon emissions, and district heating of nuclear power plants is one of the ways to alleviate this problem. Since the 1960s to 1970s, international studies on nuclear energy for district heating have been conducted, and these studies have mainly focused on 2nd-generation nuclear power plants. At present, there are fewer units for district heating based on 3rd-generation passive nuclear power technology. In this paper, we analyze the impact of district heating on NSSS (Nuclear Steam Supply System) control system of passive nuclear power plants (AP1000 and CAP1400), including reactor control system, rapid power reduction, steam dump control system, and feedwater control system, taking Haiyang nuclear power plant and Shidaowan nuclear power plant for example. And the corresponding I&C modification schematic is given according to the analysis.

Chinese Modular High-Temperature Gas-Cooled Reactor HTR-PM is a very promising way of energy utilization. Unlike PWR, the primary coolant of HTGR is helium, which follows the ideal gas equation. Due to the helium purification and supporting system, when the pressure of the primary circuit drops, the system will conduct gas supplement and pressure adjustment. Therefore, for the primary circuit small LOCA, it is difficult for the operator to find the problem in a short time, which leads to the leakage of radioactive helium into the containment. If the problem is handled for a long time, it will lead to the final shutdown of the reactor due to negative feedback adjustment, causing unnecessary economic losses. In view of the above problems, the amount of helium in the primary circuit really reflects whether there is an abnormal leakage of helium, but the amount of helium cannot be directly measured. On the premise of not adding additional sensors, the expanded state observer (ESO) can be used to dynamically observe the quantity change of helium in the primary circuit, which can effectively monitor the abnormal state of helium and timely reflect the occurrence of small LOCA.

Offshore small natural circulation lead-cooled fast reactors have harsh working environment and few people on duty. The operation is heavily relied on physical quantities measured by sensors to control the reactor. If the sensor fails and the physical quantity measurement of the reactor is inaccurate, the real operating state of the reactor cannot be reflected, thus affecting the operation safety. Therefore, a fault diagnosis method is necessary to diagnose faults and deal with them in time.

In this paper, the fault diagnosis method based on the combination of Principal Component Analysis (PCA) and Support Vector Machine (SVM) is proposed for sensor fault diagnosis. This data-driven method can effectively reduce the data dimension and does not require the specific state of the system and the precise mathematical model of the processing object. It can play a good role in the diagnosis object such as the reactor with complex internal structure and difficult to detect characteristics. The fault diagnosis model is established using the sensor data obtained from the dynamic model of the small lead-cooled fast reactor constructed in MATLAB/Simulink. The sensor fault data is introduced for testing. The results show that this fault diagnosis method has higher diagnosis accuracy and faster diagnosis time than other methods.

In this paper, the pressurizer system of 150MW pressurized water reactor is taken as the research object, and the reactor model is established based on Relap5. The thermo-hydraulic program Realp5 and the pressurizer control system are coupled through 3Keymaser simulation platform to realize data interaction. According to the characteristics of dynamic process of pressurizer, fuzzy control system and fuzzy PID control system for water level and pressure of pressurizer are proposed respectively. The performance of the pressurizer water level and pressure control system is tested in the process of the pressurizer water level and pressure set point change and power change. The simulation results show that compared with the traditional PID control and fuzzy PI control system, the fuzzy control can better respond to the changes of the system under different working conditions, and the settling time is shorter. The fuzzy controller can effectively improve the control performance of the water level and pressure of the pressurizer under different working conditions.

In the paper, a simulation method is proposed for generation of dynamic detection field of PPS system in 3D modeling environment. Within the method, the 3D models of securitycritical infrastructures are built based on the Autodesk 3ds MAX and Unity real-time development platforms. The detection field is generated based on the scene 3D models. According to the steps of the ASD method, the elements included in the sequence diagram of the PPS system of a sample facility are assigned. The focuses of the paper are put on two parts: 1) construction of dynamic detection field in line with the PPS design with rotating security cameras; 2) calculation of the cumulative detection probability of multiple adversary paths elaborated in 3D models for PPS effectiveness analysis. The 3D visualization of detection field is then created by taking into account the additive effects of both stationary and rotating sensing elements. The cumulative detection probabilities of adversary paths are comparatively estimated in both static and dynamic detection fields for the most vulnerable path identification. The simulation results shows that the 3D dynamic detection fields and probability methods of their construction by considering the agent’s activities and camera rotation can be readily implemented in 3ds MAX and Unity platform. The regional blind spots for adversary intrusion detection can be also discovered by the simulation exercise practice tests on active and intentional avoidance behaviors (attack strategies) of adversaries in dynamically coupled detection filed.

Fault detection and identification are of great significance to the safe and economical operation of nuclear reactors. The fault detection and identification method based on signal and knowledge is usually applied in nuclear reactors. However, operators cannot judge the type and reason of fault from huge amount of data with traditional method. This paper uses one simulation model to deeply analyze various fault characteristics of NUSTER-100, especially the key system parameters and the coupling among them, such as power, core temperature, heat pipe temperature. Using the data obtained from the simulation model, a reasonable and feasible condition monitoring algorithm is designed considering the accuracy and timeliness requirements of the actual system. This paper first uses the principal component analysis (PCA) method to monitor the status of NUSTER-100. When a fault of NUSTER-100 is detected, the back propagation (BP) neural network and the support vector machine (SVM) method are used for fault diagnosis. By comparing the results of the two fault diagnosis methods, the accuracy of fault diagnosis using the SVM method is higher.

In this paper, an automated GO-FLOW modeling paradigm is proposed for facilitating system reliability analysis. Within the proposed algorithmic framework, GO-FLOW models can be directly converted from system Piping and Instrumentation Diagram (P&ID) drawings based on a list of componentized models developed for commonly used components. The library of componentized GO-FLOW models is developed in accordance with the general type classification of system components. Through an example illustration of a water supply system, the GO-FLOW model file generated using the tool is sent to the existing GO-FLOW software package for model verification analysis. The validation process shows that consistent model data formats and analysis results can be obtained using the automated GO-FLOW modeling tool. In addition, the GO-FLOW model file generated by the algorithmic procedure is easy for model modification and update. The concept of algorithm tool is supportive for dynamic reliability and risk mapping.

The emergency operating under accident condition is the most important and complex operation task in nuclear power plant. It is necessary to design a matching, reasonable and highly compatible display specifically for emergency operating. The emergency operating display and the procedures are generally one to one correspondence in design scheme of PWR, which resulting in a large display scale. Sodium-cooled fast reactor has high inherent and passive safety, simple and universal emergency operating measures, high tolerance for operator, and its emergency operation procedures have unique content and structure characteristics. To satisfy the requirements of emergency operating task, take full advantage of the sodium cooled fast reactor safety features and matching the procedures characteristics, in this paper, the research is develop innovatively about the modular design of emergency procedures display. The paper summarizes the emergency procedures task in to several functional requirements, and formed a set of universal functional module display for emergency procedure. To achieve the complete coverage of the core operation content of emergency procedures, without the need to set up a special task display for each procedure. Improve the utilization rate of the emergency procedure display, greatly reduce the number of display, and simplify the scale of display. It has been verified that the modular emergency procedure display design meets the task requirements of emergency procedure operation, improves the execution efficiency, reduces human error, improves the emergency response ability of sodium cooled fast reactor, and improves the safety of power plant.

Many human reliability analysis (HRA) methods have been developed to evaluate the reliability of Main Control Room (MCR) of Nuclear Power Plants (NPP). Performance Shaping Factors (PSFs) were used in most of them to depict the contextual factors with the combination of different dimension, and to qualitatively and quantitatively evaluate the human reliability. The correlations among PSFs were one of the most important challenges for the improvement of HRA analysis quality, while most of existing HRA methods consider PSFs independently. Based on the analysis of some of the existing HRA methods and literature review of published research in this field, this paper presents a report on techniques for identifying the correlations among PSFs and some insights for future research on revealing their actual correlations.

Successful applications of the instrumentation and control (I&C) system are ubiquitous in nuclear power plants (NPPs), contributing to the safe, reliable, and sustainable operation of the current and future fleet of NPPs. However, it is not seen as much in use for the temporary or long-term storage, nonetheless the transportation, of spent nuclear fuel (SNF). With the saturation of wet storage capacity in NPPs and the growing demand for SNF reprocessing, the need for safe maintenance and transportation (M&T) of SNF is bound to increase significantly. The M&T process involves many factors, including the operation and refueling schedule of the NPP, the acceptance capacity and waste reprocessing capabilities of the waste disposal plants, the availability of transportation containers, vehicles, and its crew, etc. Considering the number of factors to be coordinated, this complex process is prone to human error.

This paper describes an integrated management platform for spent nuclear fuel storage and transportation capable of removing human error altogether in the M&T process of SNF. The platform encompasses categorized temporary storage, in-plant transshipment, and out-of-plant long-distance transportation of SNF from the currently operating NPPs. Real-time monitoring of the SNF and dispatch of the shipping crew and emergency response team is made possible based on accurate positioning, tracking data, and instant communication made possible by modern information technology featuring the Internet of Things (IoT). This is of great significance for the safe transportation of spent fuel and for avoiding excessive radioactive contamination and radiation exposure.

This platform also features an embedded decisionmaking module for spent fuel transportation, which can automatically coordinate the generation and reprocessing speeds of SNF at the NPPs and the waste disposal facilities, respectively. The decision-making module also considers resource information, including transportation containers, vehicles, crew, etc., optimizes resource allocation, formulates transportation plans, and outputs resource requirements and capacity building gaps for the next year, ensuring safe, orderly, and efficient transportation of spent fuel. The platform and its related technologies described in this paper can be extended to the storage and transportation of other radioactive waste or critical substances besides spent nuclear fuel.

The tasks of operators in an emergency situation are to identify the event, plan a counter measure to minimize the influence of the event and control the plant condition to become a normal one. If the event is supposed, operators can focus on selecting a suitable counter operation procedure from emergency operation procedures, severe accident management guidelines, etc. and following the selected procedure to settle the event. However, if the event is not included in a list of supposed events, operators should set a goal of counter measure and plan a suitable counter operation procedure by themselves. The authors propose a technique to generate plausible counter operation procedures based on a functional model and its applicability is demonstrated by some case studies. However, there remain some topics to be challenged. The selection of suitable counter operation procedure is one of the topics. This is usually conducted by the consideration of system, material, human, and information resources. System resources include devices and systems used for counter operations. Material resources include water to cool reactor systems and electricity and fuels to work devices and systems. Human resources include staffs and their skills and supplemental personnel dispatched. This paper proposes a systematic selection technique of counter operation procedures by especially considering system and material resources. The applicability of the proposed selection technique is demonstrated by a case study of an accidental situation.

Integrated System Validation is performance-based assessment of integrated human-system interface（HSI） prior to loading in nuclear power plants. Appropriate integrated system validation scenario selection method is to confirm that the activity is sufficient, necessary and orderly. Based on the HSI design of nuclear power plant, this paper proposes a set of systematic and implementable integrated system validation scenario selection method on meeting the review requirements of Human Factors Engineering Program Review Model (NUREG 0711) on sampling of operational conditions, so that the integrated system validation work is more reasonable and efficient.

Since a nuclear power plant design consists of hundreds or thousands of individual HSI devices and plant operations are complex and varied, ISV activities for all operating conditions and human tasks are neither practical nor necessary. Therefore an appropriate context selection approach to ISV is required. ISV scenarios select three elements as tasks, procedures and complex factors, among which tasks and procedures are general elements, and complex factors are considered as additional elements. The three elements (tasks, procedures, and complex factors) ISV scenario selection method introduced in this paper can avoid the disorder of previous scenario selection, and make the scenario selection more systematic and complete. It can not only meet the requirements of regulations and standards, but also make the validation work more reasonable and efficient, and can meet the requirements of all power plants, including new advanced nuclear power units, for the comprehensive validation of integrated HSIs.

In consideration of disturbing factors in the marine nuclear-powered plant, such as permittivity change, density difference and tilting and swaying angles of ship, this paper presents an improved guided wave radar level gauging system. With respect to the physical constraints of nuclear-powered plant, such as spatial limitations, radiation and electromagnetic protections, an overall engineering scheme is proposed, including the physical structure, hardware components and software configurations.

After the serious accident happened in March 2011 at Fukushima Dai-ichi Nuclear Power Plant, Nuclear Power Plants (NPPs) in Japan stopped by sequentially. For the NPPs to be re-started, additional radiation monitors those can be used during a severe accident are required. Therefore, we have developed a small ionization chamber detector and a fiber optic radiation monitor as a new radiation instrumentation system which improved environmental resistance.

A small ionization chamber detector can be improved heat resistance by adopting a MI cable to the signal and power source cable. A fiber optic radiation monitor is a gamma-ray dose rate detector using an optical fiber and a scintillator crystal. When gamma-rays or X-rays are irradiated on to this crystal, scintillator crystal generate photons. The basic performance examination and the environmental resistance examination that simulated the severe accident are carried out. In the environmental examination that simulated the severe accident condition, it succeeded for a measurement under most of conditions. However, there have been several problems with a fiber optic radiation monitor such as BG increases at high temperatures. Some of these problems are expected to be solved by revising the structure of the sensing unit and improving the arithmetic logic.

With the overall improvement of the digital and intelligent level in the nuclear power industry, the failure of the Digital Control System (DCS) has become one of the major factors affecting the stable operation of nuclear power units. DCS safety problems can easily lead to the unit outage or shutdown, power reduction operation, system or equipment degradation events and other serious consequences, which will significantly threaten the safety and reliability of the unit operation. This paper studies the real-time safety problem of DCS based on the monitoring data of DCS itself, focusing on DCS current operating risk state identification, operating risk state transition model construction, residual operating risk prediction, to evaluate DCS operating risk state and predict the risk evolution trend. Furthermore, this study focus on the design and development of the software tool for risk state identification and prediction of DCS, so as to implement the methods and algorithms presented herein.

In this paper, a network switch performance anomaly detection method of the instrumentation and control (I&C) systems in nuclear power plants (NPP) based on network calculus is proposed. In the Ethernet-based packet switching I&C system network, the delay of a network switch is an important indicator of network performance. With the development of small modular reactor (SMR) technology, the scale of the I&C systems in NPPs has increased. The I&C systems of SMRs are often constructed in batches with a modular architecture, which imposes higher requirements for the extensibility of the network architecture and network performance. Traditional network performance analysis uses the queuing theory based on the statistic method and the stochastic process to analyze the average performance of the network, instead of analyzing the deterministic boundary of system performance. In this paper, anomaly detection methods based on network calculus are proposed for network switches in the NPP I&C system by detecting rate-latency performance, delay, and data backlog anomalies. Network calculus is a deterministic queuing theory based on nonlinear algebra. It takes the data flow as the research object, obtaining the performance bound of network nodes through the convolution and deconvolution of the min-plus algebra. The MATLAB simulation experiments are conducted and verify the effectiveness of the anomaly detection methods.

New perspectives of safety called Safety-II emphasize the importance of the resilience potentials of operator personnel for maintaining the safety of socio-technical systems such as nuclear power plants. However, the common safety measures based on traditional safety perspectives called Safety-I strongly require the elimination of failures by stricter compliance to rules and procedures, which can lead to organizational climates that do not tolerate even small failures. The concern is that such organizational climates can reduce learning opportunities through small failures and make it difficult to improve the resilience potentials of operator personnel. Therefore, this study conducted a cognitive experiment requiring the operation of Fire-fighting Command and Control Simulator (FCCS) to analyze the effects of failure tolerance on the operator’s attitude in the training session and on performance in responding to unexpected and novel events. The participants, twenty-eight university students, were divided into two groups depending on training conditions: low-failure-tolerance condition (Group-A) and high-failure-tolerance condition (Group-B). The results showed that, in the training session, more participants of Group-B showed the tendency of trying to learn new procedures in the scenarios that were unsuitable to be dealt with by standard procedures, while the participants of Group-A tended to avoid potential failure. This fact strongly implies that tolerance for failure can contribute to promoting the proactive learning attitude of operators, which can be indispensable to enhancing operators' and the safety of complex systems.

Anomaly detection in nuclear power plants enables us to early execute prevention and mitigation measures against accident progression. In sodium-cooled fast reactors, coolant boiling in reactor cores is one of the important phenomena in the safety assessment. Our final target of the present study is to realize the acoustic anomaly detection of the boiling inception in actual reactors. In the actual environment, various sorts of noises are expectedly superposed on accidental boiling sounds. It is inevitable to distinguish the boiling sounds from the superimposing hostile disturbance with high accuracy. To achieve this, we utilize machine learning techniques and assess the feasibility of boiling sensing based on acoustic recognition and deep learning. In the present study, we employ an autoencoder to denoise boiling sounds, and a convolutional neural network to detect the boiling inception. The boiling acoustics have not been fully understood yet. In the present study, we find that some characteristics of the boiling acoustics are consistent with the resonance vibration of the heating body. This finding contributes to elucidating the physics of boiling acoustics. In addition, it helps us detect boiling occurrences with high accuracy judging from the acoustic characteristics’ patterns.

High temperature gas-cooled reactor (HTGR) is expected to be used for various industrial applications such as hydrogen production with 950°C of high temperature heat. Towards commercialization of HTGR hydrogen production technologies, it is important to develop a system analysis code which can simulate dynamic behavior of a HTGR hydrogen production system to design a plant control system for the effects of circulated helium heat through both facilities.

As a first step of the development, we developed a heat and mass balance evaluation model of a helium-heated steam reformer. The model can simulate distributions of gas composition, pressure and temperature including transient behavior. The accuracy of the model was evaluated by comparing calculation results with experimental data on temperature distributions and the hydrogen production rate in a helium-heated steam reformer in an out-of-pile test facility, which is capable to produce about 100 Nm3 h−1 of hydrogen.

The several test conditions including (1) rated operation and (2) start-up operation in the out-of-pile test facility were selected and simulated by the developed model to validate the thermal-hydraulic model and the reaction characteristics in the steam reformer. Analyzing the simulation results, the performances of the developed model such as accuracy in heat and mass balance evaluation in the steam reformer were quantified.

The results showed that the calculated outlet gas composition from the reformer agreed well with the experimental result with high accuracy. The temperature distributions of helium, process and generated gases in the reformer could be evaluated with an accuracy of about 30°C or less, except for the outlet temperature of the generated gas.

In Japan Atomic Energy Agency (JAEA), an artificial intelligence (AI) aided integrated digital system named “Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)” is being developed to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle. Until 2023, the platform for a common function and the sub-systems: the ARKADIA-Design for design study, the ARKADIA-Safety for safety assessment, and the ARKADIA-KMS for the knowledge base, are separately being developed. This paper describes the development concepts of the platform, the progress of the application study of design optimization in ARKADIA-Design, the progress of optimization model developments and optimization functionality based on AI technology in the ARKADIA-Safety, and the structure of the knowledge base and application of AI technology to the ARKADIA-KMS. In the development of ARKADIA for the next five years until 2028, a strategy of unification of the sub-systems with the AI-aided platform to one system, ARKADIA, and the extension of capabilities of the numerical analyses and evaluation technologies required for plant design are presented in prospects.

The Super LWR is the Japanese representative design concept of the thermal reactor type of the Supercritical Watercooled Reactor (SCWR). The reactor operates at 25 MPa and the coolant undergoes large enthalpy rise as its density decreases to about 1/10 from the core inlet to the outlet. The large water density change indicates possibility of spectral shift operation, in which the neutron spectrum is gradually softened during the operation by increasing the core average water density to gain reactivity and improve utilization of the uranium (U) resources. However, the core is cooled by single-phase flow and the core outlet coolant is directly fed to the turbine in the once-through direct cycle plant system. One of the issues of spectral shifting of such a system is controlling the core average water density without significantly changing the core average coolant outlet temperature. This study proposes two approaches. The first approach is controlling the core inlet coolant temperature and flow rate simultaneously and the second approach is inserting / withdrawing water displacer rods into / out of the core. Hence, this study aims to reveal potential improvements and issues of the Super LWR core with spectral shift operation with the above proposed approaches.

The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishment of the safety approaches and validate the performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WPSO-1 “Methods, Models and Codes” is devoted to the development of tools for the evaluation of safety. WP-SO-2 “Experimental Programs and Operational Experience” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600, EBR-II and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as active and passive safety systems and other specific design features. This paper reports recent activities within the SO project

In terms of commercialization to the future on Sodiumcooled Fast Reactors (SFR), it is important that the feasible study to be with large electrical power as an operational cost advantage. In this study, assuming a demonstration SFR at the next generation being medium power SFR in Japan, the large scaled SFR with 1GWe is designed tentatively. This paper describes following optional studies about each design field.

In the first study, the design case of 6 intermediate heat exchangers installing into the reactor vessel are prepared, differences between 6 loops and 3 loops as for the secondary loops which transfer heat from the reactor to the steam generators are examined on the plant commodity and the plot plan as a twin reactor. As a result, in the case of 3 loops, the building volume is approx. 6% smaller than 6 loops.

In the second study, the core arrangement which has 24 months of operation cycle is prepared as a reference. Parametric study under the constant average burn-up examines variability of the diameter of circumcircle around control rods relating the above core structure.

In the third study, the structural stiffness of the reactor vessel is evaluated, since the stiffness due to form and thickness of parts such as strongback, lower plate, cylindrical shell etc., is a dominant factor for the seismic response.

A sodium-cooled fast reactor (SFR) considers adopting 3dimensional seismic isolation devices for withstanding seismic loads not only horizontal but also vertical direction. A seismic isolation device consists of a laminated rubber bearing and horizontal oil dampers for horizontal direction, coned disc springs and vertical oil dampers for vertical direction, respectively. In order to investigate the performance of each component and the feasibility of integrated system for SFR, the experiments such as load-displacement tests, vibrating tests, etc., to each component of seismic isolation devices and seismic response analysis are carried out. As those experimental results, the mechanical characteristics of each component and the devices are grasped, then it is demonstrated that components and devices have expected performances to reduce the seismic loading within the design range. As the analytical results of seismic response, it is indicated that this 3-dimesional seismic isolation device and system can reduce the seismic response on horizontal and vertical simultaneously.

Based on the analytical studies and experimental results, the feasibility of newly developed 3-dimensional seismic isolation system is obtained and the prospect of practical application of 3D seismic isolation system for fast reactor is implied in this paper.

Tritium produced by various nuclear reactions in light water coolant in LWR and in heavy water coolant and moderator in HWR is one of the radionuclides that limits release of radioactive water into environment in currently operated reactors. On the other hand, fusion reactors would require large amount of tritium to start their operation. Tritium-producing burnable absorber rods made from lithium pellets placed in PWR and tritium removal facility processing heavy water from CANDU reactors are the two options for the start of fusion reactor fleet before its sustainable production from lithium blanket in fusion reactors. In the paper, tritium production in coolant and moderator of four reactors is calculated and compared in order to determine whether the production scale is large enough to be interesting for fusion applications or it would remain a limiting activity of radioactive water release. Two large-scale reactors (VVER-1000 and CANDU 6) and two small-scale reactors (NuScale and Teplator) are analysed by Monte Carlo simulation to compare two LWR and two HWR reactor designs as well as two large and two small designs.

Regenerator and cooler as advanced nuclear power system critical equipment are given more and more attention from scholars recent years. The Printed circuit heat exchanger (PCHE) has an advantage of high pressure-bearing capacity, high efficiency and compact structure is ideally suited as regenerator and cooler in Supercritical Carbon Dioxide (S-CO₂) Brayton Cycle System. However most of published studies focus on one set of channels ignoring the header and pipeline bundle. Moreover, S-CO₂ has the high density and high specific heat compared with general heat transfer medium. But heat exchange complexity also needs for further research. In order to master flow and heat transfer in full size PCHE, 12-channel geometric models are set up. Numerical calculation is applied to analyze flow and heat transfer in full-scale PCHE (including heads) which focus on inlet and outlet headers. Further, full-scale and one set of channels calculation results are compared. Calculation results show that pressure drop in headers section cannot be ignored. The research results can provide references for follow-up research.

The characteristics of the axial evolution of interfacial structure in a narrow rectangular channel are expected to differ from that in other channel geometries because of the significant restriction from the wall. Moreover, the interfacial area concentration (IAC) is one of the essential structure parameters indispensable for predicting the two-phase flow. From this point, a visualization investigation was carried out to study the IAC of flow boiling in a narrow channel. The experiments were conducted in a narrow rectangular channel of 2 mm × 65mm × 700mm, heated by one broad side of the stainless steel for a better visualization from the other side. IAC along the flow direction was abstracted by the image processing method at three different flow rates: 0.08 m3/h, 0.12 m3/h, and 0.2 m3/h. Bubbly flow, churn flow, and annular flow were observed as the void fraction increased during the flow boiling in the narrow rectangular channel. The IAC lines along the flow direction showed an “N” shape during the development. At first, the IAC increased with the increase of void fraction. However, the IAC had a sharp decrease; the drop in the IAC was nearly 50%. And then it increased again until at the outlet. The drop interval was the transition from dense bubbly flow to churn flow from the visualization images. Combined with the void fraction and the bubble number density, IAC was analyzed by the bubble behaviors. The data in this study will be helpful for the development of the two-fluid model in narrow rectangular channels.

The small modular reactor (SMR) is considered to be an enabling technology for providing economical and clean energy in remote areas in Canada. To ensure the SMR technology is developed within a robust framework that addresses environmental and waste management concerns, data are required on radionuclide inventory and characteristics of SMR depleted fuel at the end of reactor service life and at various times thereafter. These data provide essential inputs to assessment of fuel recycle analysis, understanding of environmental impact, and strategy development for waste disposal and management. In this paper, radionuclide inventories of depleted fuel in a small fluoride molten salt reactor (sm-FMSR) and a micro-sized high-temperature gas-cooled reactor (m-HTGR) are calculated using the Monte Carlo neutron transport code Serpent and the point neutron activation and decay code SCALE/ORIGEN. The inventory calculation methods for two selected small modular reactors are described, and radionuclide inventory results from Serpent and ORIGEN are compared. Overall, ORIGEN produces more conservative results than Serpent for both sm-FMSR and m-HTGR. The major characteristics of the radionuclide inventories are discussed for both SMRs.

HTGR has a potential to produce competitive and large amount of carbon-free hydrogen because of its capability to supply high temperature heat of nuclear energy. JAEA started a project of HTTR heat application test producing hydrogen with the HTTR to develop a massive, cost-effective and carbon-free hydrogen production technology utilizing a HTGR.

In present study, the safety design philosophy is developed for the HTTR heat application test facility connecting HTTR and the hydrogen production plant. The postulated events in HTTR heat application facility were identified to develop the safety design philosophy conforming to the requirements in the applicable laws and/or standards.

Considering the characteristics of HTTR heat application test facility, basic philosophy for the safety design was proposed to apply “Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors” for the nuclear reactor plant including the equipment with pressure boundary of secondary helium coolant. For the hydrogen production plant composed of systems for methane supply, product gas combustion and steam supply, “High Pressure Gas Safety Act” was applied to for ensuring high level protection against hazards of combustible gases. A demarcation boundary for applicable law is set on valves connecting the nuclear reactor plant and the hydrogen production plant.

The design philosophy of HTTR heat application test facility was proposed considering each identified postulated event including the internal fire, the tornado, the aircraft crash, the forest fire, the fire and explosion in adjacent facility and poisonous gas leak to conform safety requirements and standards. For example, the facility will be designed to protect the important safety systems against the internal fire by appropriate implementation of fire protection measures and by implementation of the offset distance and/or barrier against postulated explosion of leaked combustible gases from the SR.

Employing high-temperature sodium heat pipes to remove heat from the reactor core and realizing energy conversion through the cascaded thermoelectric generators, a conceptual design of a 100 kWe level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) is presented. Aiming at providing sufficient cooling capacity for NUSTER-100, the development of high-performance sodium heat pipes is faced with challenges for the large length-diameter ratio, high thermal power density, material properties decay at high temperature, etc. In this study, the experimental evaluation of the heat transfer performance of sodium heat pipes is conducted. It is found that the maximum thermal load of sodium heat pipes (φ30×3×2000 mm) is more than 10 kWt, which shows the design of NUSTER-100 is reasonable and available. Besides, the failure of hightemperature sodium heat pipes is observed. With a high thermal load (> 10 kW) and high temperature (> 1000 °C), the upper part of the evaporator section of sodium heat pipes is overheated, which leads to the breach in the container of sodium heat pipes. Due to the difference in the container material, one small rupture with a diameter of 7 mm (HAYNES® 233™ alloy) and two large ruptures with a diameter of more than 10 mm (stainless steel 310S) occur near the outlet of the evaporator section. Furtherly, the attenuation effect of the oxidation of the working fluid on the thermal performance of sodium heat pipes is verified experimentally. Through heating and oxidating the sodium with a purity of 99.7% in the air at 250 °C for 5 hours, the heat transfer performance of sodium heat pipes with the same specification is compared, in which the heat pipe with the oxidated sodium reduces the heat transfer power by 59.4%. This paper summarizes the research progress of sodium heat pipes for NUSTER-100 and provides data support for the research and development of heat pipe cooled reactor.

NuScale VOYGR™ SMR power plant is an innovative design developed by NuScale Power in the United States that utilizes proven light water reactor technology as well as reducing the size and simplifying the design. The simplified design, adoption of passive safety systems, and fewer fission product inventories have significantly improved safety compared to conventional reactors, and the integrated NuScale Power Module™ (NPM) design eliminates the need for large primary coolant piping loops outside the NPM. The design and the technology eliminate a chance of large break loss-of-coolant accident and therefore the risk to the public (e.g., probability of core damage) is reduced to much lower than that of conventional reactors. Currently, the NuScale VOYGR-6 SMR power plant construction project in the U.S. with the Utah Associated Municipal Power Systems is planned to start commercial operation in 2029. JGC Corporation and IHI Corporation, in collaboration with NuScale Power, have been developing innovative nuclear technologies since 2019 through the NEXIP (Nuclear Energy x Innovation Promotion) initiative. In addition, JGC Holdings invested in NuScale in April 2021 and IHI in June of the same year to strengthen this collaboration. This article will describe our participation in the development of the NuScale VOYGR™ SMR power plant.

High temperature gas-cooled Reactor (HTGR) is expected to extend the use of nuclear heat to a wider spectrum of industrial applications such as hydrogen production, high efficiency power generation, etc., due largely to high temperature heat supply capability as well as inherent safe characteristics. Japan Atomic Energy Agency (JAEA) initiated the HTTR (High Temperature Engineering Test Reactor) heat application test plan to develop for coupling technology between HTGR and hydrogen production facility.

The principal objective of this test plan is to establish the high safety coupling technology for coupling a hydrogen production facility to HTGR utilizing the HTTR as a high temperature heat source. The other objective is to develop for coupling equipment required for coupling between HTGR and hydrogen production facility. This paper describes the overview of the HTTR heat application test plan to establish high safety coupling technology mainly for the former objective. The latter one is described separately in the other paper, “(2) Development Plan for Coupling Equipment Between HTTR and Hydrogen Production Facility”.

During the unprotected loss of flow accident of Sodium cooled Fast Reactor (SFR), the coolant sodium in the fuel assemblies may boil, which would induce the sodium void effect and increase the reactor power, leading to serious accident consequences. Under the condition of sodium boiling, the flow and pressure characteristics of sodium will affect the molten core debris distribution and the process of severe accidents.

A sodium boiling model in a single fuel assembly is established with the porous medium approach and two-fluid model, to study the changes of flow and pressure field during sodium boiling in the fuel assembly. The porous medium approach is used to average the parameters in every control volume. The constitutive relations of mass exchange, convective heat transfer and energy exchange are considered in the mass, momentum and energy conservation equations of gas and liquid phases. Based on the semi-implicit differencing scheme and numerical iteration technique, the conservation equations are solved and the thermal-hydraulic parameters of coolant and fuel assembly can be obtained. The KNS-37 loss of flow experiments carried by Kernforschungszentrum Karlsruhe (KfK) is used to be analyzed, and the results are compared with the experimental results and the results of SAS-SFR.

The model is suitable for simulating sodium boiling and the local region drying out during the experimental process. The established model can predict the sodium boiling time and sodium boiling position accurately, and the overall variation trend of temperature and flow rate is in good agreement with the experimental data.

High Temperature Gas-cooled Reactor (HTGR) is expected to extend the use of nuclear heat to a wider spectrum of industrial applications such as hydrogen production, high efficiency power generation, etc., due largely to high temperature heat supply capability as well as inherent safe characteristics. Japan Atomic Energy Agency (JAEA) initiated the HTTR (High Temperature Engineering Test Reactor) heat application test plan to develop for coupling technology between HTGR and hydrogen production facility.

The principal objective of this test plan is to establish the high safety coupling technology for coupling a hydrogen production facility to HTGR utilizing the HTTR as a high temperature heat source. The other objective is to develop for coupling equipment required for coupling between HTGR and hydrogen production facility. This paper describes the overview of developing plan for coupling equipment required for coupling between HTGR and hydrogen production facility. The former one is described separately in the other paper, “(1) Overview of the HTTR Heat Application Test Plan to Establish High Safety Coupling Technology”.

Advanced reactor concepts, such as Small Modular Reactors (SMR) are receiving increasing attentions. There are different concepts which are in a late development stage and are considered as an option for future new builds. Because of this, system codes should be able to simulate phenomena in SMR. To investigate the capability of AC²-ATHLET to simulate SMR and the occurring phenomena in helically coiled steam generators (HCSG), the experiment OSU-002 at the OSU-MASLWR (Oregon State University- Multi Application Small Light Water Reactor) is simulated and analysed. In the experiment, the natural circulation and the HCSG behavior under different conditions were investigated. First results show an overestimation of the temperatures at the outlet of the HCSG and a slight deviation of the qualitative course at higher power levels. Moreover, the pressure loss in the HCSG is underestimated by ATHLET. Due to this, adequate correlations are implemented in an in-house PSS version of AC². The pressure loss is simulated in better agreement with the experiment and consequently shows higher pressure losses with the new implemented correlations. The modifications of the heat transfer calculation show improvements regarding the qualitative temperature course at the outlet of the HCSG, but also slightly higher temperatures.

The transient thermal safety characteristics of the NUSTER-100 were analyzed by applying the commercial CFD software FLUENT. The program module of the point reactor neutron kinetic model was developed through the user-defined function (UDF), coupled with the transient analysis code of heat pipe, and the transient thermal safety analysis of the NUSTER-100 was carried out. According to the power distribution of core obtained by MCNP, the 1/8 core physical model of NUSTER-100 was established. The operating conditions of the NUSTER-100 core and the single/multiple heat pipe failure accidents are analyzed. The results show that NUSTER-100 has good safety; the failure of a single heat pipe has a small impact on the core, and failure of two or more heat pipes may have a greater impact on the core. This article can provide methods and ideas for transient thermal analysis of the heat pipe cooled reactor.

Swimming Pool-type Low-Temperature Heating Reactor (SPLTHR) is a reactor developed to meet the needs of clean heating in northern China. But the core inlet/outlet temperature of SPLTHR is only about 80/100 ℃. In order to ensure the heating capacity of SPLTHR in winter, it is necessary to increase the outlet temperature of the reactor. Increasing the outlet temperature of SPLTHR will certainly lead to the rise of the pool water temperature, which will affect the bearing capacity and service life of the concrete around the reactor. It is necessary to reduce the concrete temperature around the reactor. In this research, the scheme of constructing a water layer in SPLTHR for thermal insulation is proposed, which is expected to reduce the temperature of the concrete outside the reactor. And we carry out a preliminary study of the static water layer scheme through principle experiments. The results show that there is natural convection in the static water layer. And the heat transfer can be effectively reduced by inverting a vertical partition in the static water layer. After analyzing the heat transfer of the SPLTHR external soil, we find that the disadvantages of the static water layer scheme. Therefore, we put forward an improved scheme, which is to use the low temperature reinjection water of the 1^st purification system to cool the thermal insulation layer.

The Helium-Xenon (He-Xe) gas-cooled reactor uses a He-Xe gas mixture to remove the heat from the reactor core. It has the characteristics of a long life-cycle, simple operation, and high power-mass ratio, which has an excellent application prospect in the field of Space Reactors Power Systems. The high operation temperature distinguishes the He-Xe gas-cooled reactor from the traditional reactor in reactivity feedback characteristics. Except for the Doppler effect of the fuel and monolith, the thermal expansion also introduces reactivity feedback during reactor operation. In this work, reactivity feedback characteristics, steady-state characteristics, and transient characteristics of a hundred-kilowatt-level He-Xe gas-cooled reactor (Submersion-Subcritical Safe Space reactor, S⁴) were analyzed. The Doppler feedback characteristics and thermal expansion feedback coefficients of fuel, matrix, and reflector from the cold to the hot state of the S⁴ reactor core were calculated based on the reactor Monte Carlo code (RMC). The results show that the S⁴ reactor core is mainly affected by the fuel Doppler effect and the matrix thermal expansion effect. Furthermore, the temperature distributions of fuel and He-Xe gas of the S⁴ reactor were calculated using the He-Xe reactor transient analysis code (HXRTRAN). The transient characteristics of the S⁴ reactor core when different reactivity was introduced were analyzed. This study may provide a reference for the operation state of the He-Xe gas-cooled reactor.

Self-actuated shutdown system (SASS) is one of the innovative technologies for application to an advanced-loop-type sodium-cooled fast reactor designed in Japan. SASS is a passive reactor-shutdown system that utilizes a Curie-point electromagnet (CPEM), which features the characteristic of loss in magnetism when the magnet temperature reaches the Curie point. A control rod with SASS is inserted into the core by gravity without recourse to any active shutdown system. To allow the SASS to effectively function, efficiently guiding high-temperature fluid from the fuel assembly to CPEM is important. Therefore, CPEM features a complicated shape such as having 45 fins, and a flow collector is installed upstream of CPEM to direct the flow from the fuel subassembly outlet to CPEM. In this study, a full-scale water experiment was performed to understand the flow field around CPEM and obtain data for code validation. The test section that simulated the structure from the fuel subassembly outlet to CPEM was mainly made of acrylic to enable flow visualization and particle image velocimetry (PIV) measurement. The flow velocity conditions were set from 0.26 m/s (Re = 48,000) to 1.55 m/s (Re = 280,000) at the fuel subassembly outlet. We confirmed from the PIV measurements that the flow from the subassembly outlet was directed to CPEM by the flow collector. The velocity at the inlet of the fins was approximately 60% of that at the fuel subassembly outlet. In addition, the velocity distributions around CPEM were almost the same regardless of Re number in the Re range in this study. Moreover, after passing through the flow collector, the dependence on the circumferential direction of the velocity distribution was almost eliminated. These experimental results were useful not only for understanding the flow field around CPEM but also for code validation.

The small module reactor has the characteristics of high energy density and long-term stable operation. It can be used as an energy supply device to supply power for mines, islands, scientific research and national defense equipment located in remote/closed areas. The main deployment scenario of small module reactor is remote areas with weak infrastructure such as transportation, water source and power supply, which puts forward high requirements for the safety, reliability, flexibility, robustness, rapid deployment and "unattended" capability of reactor operation. It is important to study the intelligent operation and maintenance system of mobile small module reactor (SMR) that can realize the ability of "unattended monitoring". However, the research on intelligent operation and maintenance system of SMR belongs to innovative research. From the perspective of demand, it should be designed in a positive iterative way to form an overall conceptual scheme, laying the foundation for the subsequent preliminary design and detailed design.

System engineering method is a forward design method suitable for complex systems, which is widely used in aerospace and other industrial manufacturing fields. At present, the application of system engineering method in the field of nuclear power is in the exploratory stage. This paper studies the applicable system engineering method for the intelligent operation and maintenance system of SMR, and gets the overall conceptual design the intelligent operation and maintenance system of SMR based on this method. The overall conceptual design of SMR intelligent operation and maintenance system based on system engineering is a beneficial attempt to apply system engineering method to the forward design of nuclear power system.

High flux reactor with considerable neutron flux and spectrum plays an important role in medical care, industry, agriculture, energy, public safety, etc. In this paper, the neutron flux and energy spectrum characteristics of high flux reactors in three main application fields are reviewed and analyzed. The neutron flux level that could be achieved is corresponding to the design features of high flux reactor. Technical measures are adopted in order to enlarge neutron flux and optimize energy spectrum. Furthermore, the requirements of nuclear fuel and material irradiation test, radioisotope production and neutron scattering / imaging technique for high flux reactor are summarized. Finally, it is concluded that the construction of ultra-high neutron flux with broad energy spectrum is essential for the development of innovative high flux reactors.

Experiment of a floating seismic isolation technology for application to small modular reactor (SMR) is conducted on a small-scale apparatus. The SMR is assumed to be located on land rather than offshore, and installed on a newly proposed floating structure in pool. The floating structure acts to mitigate the propagation of the horizontal component of seismic motion between the ground and the SMR buildings, and to reduce the excitation force on the buildings caused by the vertical wave propagation through water. The floating structure is designed to have a gaseous space called Air Cavity at the bottom of the structure. Experiment is conducted to verify the performance of the floating seismic isolation structure against various spectrums of seismic motions. The results demonstrate that the performance varies with earthquake input frequency and that the Air Cavity is effective to reduce peak response to horizontal and vertical earthquake accelerations particularly in the range of frequencies that are sensitive to nuclear plant piping.