The enhanced capability to nuclear power plant (NPP) severe accident management and emergency response depends heavily on exercises. Since the exercise scene is usually monotonous and not realistic, and conduct of exercise has a high cost, the effect of enhancing the capability is limited. Thus, the development of a Sever Accident Emergency Simulation System (SAESS) is necessary. SAESS is able to connect NPP simulator, and simulates the process of severe accident management, personnel evacuation, the dispersion of radioactive plume, and emergency response of emergency organizations. The system helps to design several of exercise scenes and optimize the disposal strategy in different severe accidents. In addition, the system reduces the cost of emergency exercise by computer simulation, benefits the research of exercise, increases the efficiency of exercise and enhances the emergency decision-making capability. This paper introduces the design and application of SAESS.

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Subcooled flow boiling in inclined tube occurs widely in many industries and applications. An experimental study on the local interfacial characteristics of subcooled flow boiling was carried out under vertical and inclined conditions. The test section is a circular tube (i.d. 24mm) of which axial heated length is 1m. A double-sensor optical probe was used to investigate the radial distribution of interfacial parameters including local void fraction, bubble frequency, interfacial velocity, Sauter bubble diameter and Interfacial Area Concentration (IAC). The range of heat flux and mass flux are 140-900kW/m^2 and 200-800kg/m^2s, respectively. The inclination angles are 5°, 10°, 20°, 30°, and the polar angle between measurement direction and inclination direction in the probe measurement cross-section are 0°,45°, 90°. From the test, the local interfacial parameters were measured at 15 radial locations at probe elevation. Based on these data obtained in the previous test loop, the influence of flow condition and inclination angle on the profiles of local interfacial parameters was discussed

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Reactor vessel depressurization is an important measure before an accident sequence progresses to the point of vessel bottom head penetration failure under high-pressure condition. The passive depressurization systems can depressurize the reactor coolant system to prevent or to mitigate the effects of direct containment heating instead of the safety/relief valves (SRV) if SRV is inoperable or it is stuck in the closed position. The sensitivity analysis performed with SAMPSON code in this study demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

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Application of partial penetration welding (PPW) to ITER Toroidal Field Coil structure has been proposed because of limited access for weld due to complex shape and low stress components. In order to obtain fatigue crack growth (FCG) behavior of PPW joint in cryogenic environment, Japan Atomic Energy Agency performed FCG test at 4K by using Compact Tension (CT) specimens having as-weld notch of PPW. These CT specimens were sampled from mockups having one of actual joint shape of PPW, double J-groove. As the result of these tests, it was observed that crack propagated in weld metal having inclination from as-weld notch. Moreover it was shown that as-weld CT specimens had high FCG rate region in early stage of crack propagation due to residual stress distribution. In addition, it was proposed and verified application method of this FCG rate to designing of PPW joint.

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The twelve by twelve spacer grid of NHR200-II is developed on the basis of spacer grid of NHR200-I. As the spacer grid plays an important role in the fuel assemblies, it is reasonably important to determine the integral properties of spacer grid. In this work, static compression experiments with specially designed equipment are utilized to investigate the compression properties of the spacer grid at room temperature. The FE model of the spacer grid is also employed to obtain the stiffness of the spacer grid, using the finite element code Abaqus/Explicit. The result of experiments and the finite element (FE) model shows good agreements, which verifies accuracy of the FE method.

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Spacer grid is an important component in PWR fuel assemblies for its significant influence on thermal-hydraulic characteristics of the reactor core. In this study, single-phase CFD technology is used to study the flow and resistance in a 5×5 rod bundle with spacer grid. CFX 12 is chosen as the tool software. The geometries of spacer grid and rod bundle are finished by UG code. To simulate the contact of springs and dimples with fuel rods, small gaskets are induced. The SST model is chosen for turbulent simulation. The results of resistance coefficients including both local loss coefficients and frictional loss coefficients under different Re numbers are conformed well to the experiment. The agreement of calculated liquid velocity distribution downstream the spacer grid with the experiment is also well but less satisfactory. We draw attention to the influence of mixing vanes and find that mixing vanes are the main components that induce the lateral flow, but take up little percents in the local loss coefficient of spacer grid. The lateral flow caused by mixing vanes will increase the frictional loss coefficient downstream the spacer grid. These results are beneficial to the optimization design of spacer grid.

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Based on the results of electron back scattered diffraction (EBSD) analysis of samples hydrogenated at temperatures from 900 to 1200 K (which is a typical temperature range for loss-of-coolant accident (LOCA)) some questions related to hydrogen embrittlement of zirconium and Zircaloy - 4 are being discussed in the present work. These questions affect: hydride origination and development; evolution of microstructure and microtexture; material properties degradation; the application of already known hydrogen embrittlement mechanisms (due to hydride formation, hydrogen-enhanced decohesion, hydrogen-enhanced localised plasticity); a hydrogen-induced crack initiation and propagation phenomena detected inside the furnace at constant temperature of 1000 and 1100 K. The Scanning Electron Microscopy of the ruptured surface was performed to observe the fracture behaviour with respect to hydrogen content. Change in lattice parameters "a" and "c" of hexagonal close-packed α zirconium lattice and the presence of δ - and γ - hydrides were measured after each hydrogenation test by means of the X - ray diffraction (XRD) analysis. This series of experimental tests was performed at Karlsruhe Institute of Technology (KIT) in the framework of the new QUENCH-LOCA program.

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The objective of this paper is a study on thermal-hydraulic and neutronic aspects of a pressure-channel Supercritical Water-cooled Reactor (SCWR) with a focus on determination of fuel and cladding/sheath temperatures as well as a pressure drop across a fuel channel. With these objectives, a thermal-hydraulic code has been developed in MATLAB, which calculates a fuel centerline temperature, sheath temperature, coolant temperature and heat-transfer-coefficient profiles. The developed thermal-hydraulic code is coupled with a lattice code and a diffusion code. The neutronic codes were used in order to determine a power distribution inside the core. This paper presents a fuel centerline temperature of a 73-element fuel bundle with UO_2 as a reference fuel, while results are presented for high thermal-conductivity fuels such as UC and UO_2+SiC. The results show that the maximum fuel centerline temperature is significantly lower for high thermal-conductivity fuels. The total pressure drop varied between 108 to 133 kPa per fuel channel.

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Fuel assemblies with tight lattice may allow us to have fast neutron spectrum in boiling water reactor (BWR), which yield to the breeding ratio nearly equal to unity. Such fuel assemblies need wire spacers rather than grid spacers. The effect of wire spacer on coolability in tight lattice core of BWR was investigated by means of an experiment for critical heat flux (CHF) in single pin and three-pin-bundle with triangular arrangement. The experiments were performed with the parameters of mass flux, steam quality and geometry. The results showed that the heat removability of flow channel with wire was better than that without wire. The result was compared with some correlations by means of analytical consideration.

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SuperCritical Pressures (SCPs) and SuperCritical Fluids (SFCs) are widely used in many industries worldwide. The largest application of SCPs is in the power industry in advanced coal-fired power plants. It is well-known that moving from subcritical-pressure power plants to SCP power plants increases gross thermal efficiency from 38-42% to about 50-55%. Despite all advances in thermal power-plants design and operation worldwide, they are still considered as not "environmentally friendly" due to significant carbon-dioxide emissions and air pollution as a result of the combustion process. In addition, coal-fired power-plants also produce virtual mountains of slag and ash, and other gas emissions that may contribute to acid rains. Therefore, the demand for clean, non-fossil-based electricity is growing. Due to this, nuclear power is considered as a basis for future electricity generation in the world. One of the major problems with current fleet of Nuclear Power Plants (NPPs) is their relatively low thermal efficiencies, especially, of water-cooled-reactor NPPs (the vast majority of NPPs) (30-36%), compared to those of advanced thermal power plants (55-62%). Based on that, next generation or Generation-IV reactors corresponding to those NPPs should definitely be more efficient. Higher level of thermal efficiencies can be reached only due to higher temperatures and, in some cases, higher pressures inside reactors and, especially, in power cycles of Generation-IV NPPs. Analysis of the six concepts of Generation-IV reactors and NPPs shows that three reactor concepts will use SCFs as reactor coolants (helium and water) and all concepts can be linked to SCFs as working fluids in power cycles (SC helium and /or carbon dioxide in the Brayton gas-turbine cycle, and SC water in the Rankine steam-turbine cycle). Therefore, the exact knowledge of specifics of thermophysical properties of SC helium, water and carbon dioxide is very important for any advances in these new nuclear-power technologies. Also, scientists, researchers and engineers should use the same terms describing reactor coolants and power-cycle working fluids at various subcritical and SCPs. Based on that, the objective of the current paper is to define all possible terms to be used at SCPs and to link them with the appropriate regions in the Pressure-Temperature and Temperature-Specific Entropy diagrams.

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Computational Fluid Dynamics (CFD) is a numerical approach to model fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system. A numerical study on heat transfer to supercritical water (SCW) flowing in a vertical tube is carried out using the ANSYS FLUENT code and employing the SST k-ω turbulence model. The 3D mesh consists of a 1/8 section (45 ° radially) of a bare tube. The numerical results on wall temperature distributions under normal and deteriorated heat transfer conditions are compared to experimental results. The same geometry is then simulated with an orifice to study the effect of geometrical perturbation on the flow and heat transfer characteristics of SCW. The orifice is placed areas to test the effect on normal, deteriorated and enhanced heat transfer regimes. The flow effects and heat transfer characteristics will be studied around the appendages to arrive at a fundamental understanding of the phenomena related to supercritical water turbulence.

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Traditionally, the "added hydrodynamic mass" approach was widely used for the flow-induced vibration analysis of nuclear power plant reactors and reactor internals. The added hydrodynamic mass method is simple and effective for structures with regular geometries. Over the years, efforts have been made to account for the fluid-structural interaction among the reactor vessel, core barrel, thermal shield, and core shroud using mass-spring representation of the fluid. These efforts were generally benchmarked against test data to generate conservative results for a specific application. More recently, fluid elements have been developed to model the fluid in a structural finite element analysis. A study was performed herein to evaluate the effectiveness and the accuracy of using different fluid elements to model fluid-structural interaction of reactors and reactor internals. Several different fluid elements were used, including a contained fluid element and an acoustic fluid element. First, the fluid elements were used to model submerged beams and submerged concentric cylinders. The results were compared to empirical equations or available test data. The behaviors of these fluid elements were evaluated. The desirable fluid elements were chosen for further study of their behavior in modeling the actual reactor and reactor internals. The fluid elements were used to develop a reactor equipment system model of a Westinghouse three-loop reactor. The fluid elements transfer the fluid-structural interactions among the reactor vessel, thermal shield, core barrel, and core shroud. The reactor system dynamic analysis results were compared to the plant hot functional test data. Fluid elements, when carefully chosen and properly used, are superior to traditional methods to calculate flow-induced vibration responses because they can be adapted to very complicated geometry and they provide a unique hydrodynamic response for each mode shape requested. Fluid element capacities and how they shall be used in modeling with application to reactors and reactor internals are summarized herein. Using fluid elements and the approach herein, the dynamic response of the reactor equipment system can be calculated more accurately. This method is a powerful tool in the dynamic design of new reactor components. It also yields more realistic acceptance criteria for inspections which will benefits plant aging management programs. The fluid element application can be extended to calculate the dynamic responses of reactor and reactor internals due to operating basis earthquake, safe shutdown earthquake, and loss-of-coolant accident.

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The Province of Saskatchewan and Hitachi Canada have established a collaborative relationship to research design options and conduct feasibility analyses of Small Modular Reactor (SMR) technology with the goal of safely and reliably generating clean energy and helping to achieve a low-carbon society. The focus of this study is to optimize the design of an SMR Balance of Plant (BOP) for the supply of steam and heat to various residential, industrial and commercial applications. The study includes a review to examine if waste heat available from the proposed SMR could be used for producing potable water using nuclear desalination technology. It is in the context that large volumes of groundwater and brackish water are available in Saskatchewan, Canada. The existing literature on desalination processes are being reviewed, including technologies for water production such as Multiple Effect Distillation (MED), Multi Stage Flash (MSF), and Reverse Osmosis (RO). The review includes comparison of various technologies for energy efficiency and overall economics of the process. The Desalination Economic Evaluation program (DEEP) computer model available from International atomic energy agency (IAEA) is also used as a tool for examination. Factors such as salinity of the feedwater, advantages gained by pre-heating the feedwater and efficient utilization of waste heat generated in the SMR showed that technology based on RO will be the most cost effective technology. The conclusions are also supported in previous reviews done by GE and AMEC for the government of Alberta, Canada. The quality of product water produced using RO technology depends on impurities in the feed water. Therefore, trials in a pilot plant and in a semi commercial plant are proposed as next steps.

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A micro-scale crystal plasticity model, recently developed by CEA, is used to simulate a nonlinear mechanical response of type 304L austenitic stainless steel subjected to neutron irradiation. The model is implemented in codes ABAQUS and Cast3M to perform finit element simulations of the polycrystalline aggregates loaded in tension. Grain topology is modeled either by Voronoi tessellations or by a realistic microstructure of the stainless steel wire specimen. The macroscopic stress-strain curves are calculated and compared with the measurements for different irradiation doses. A very good agreement is found for the Voronoi model and a slightly stiffer response is predicted for the realistic wire model. Some preliminary calculations of the local stress distributions at grain boundaries are performed and compared, showing similar evolution trends with applied strain and irradiation levels in both considered spatial models. A brief sensitivity analysis is furthermore performed to estimate the effects of various Voronoi geometries, mesh densities and initial grain orientations.

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The fault injection/insertion testing (FIT) is one of the key techniques applied for independent verification and validation (IVV) of software and FPGA-based safety critical NPP instrumentation and control systems (I&Cs). The technique is based on design fault injection into the software code including VHDL code, physical faults into programmable CPU-based or FPGA chips and modules. The requirements for FIT as a verification technique are described by the standard NUREG/CR-7151 which in addition to the injection of single faults recommends employing a multifault injection technique (MFIT). Requirements of the NUREG/CR-7151 are analyzed and normative profile related to FIT and MFIT are discussed considering FPGA features. The application of MFIT, on the one part, increases the verification time and complicates procedure and tools supporting testing, and, contrariwise, significantly improves the quality of the system and trustworthiness of safety and dependability assessment. This paper offers the approach to development of MFIT for FPGA-based NPP I&Cs taking into account features of such systems. Injection of faults is performed into VHDL code, chip and FPGA-based module. Injection may be fulfilled by use of a few procedures such as (1) single fault by fault injection with (a) and without (b) elimination of injected faults, (2) injection of multi-faults. Besides, two levels of I&Cs are considered in point of view MFIT: first one is FPGA module level, second one is system level. Industrial case of single and multi-fault injection techniques application is described for the FPGA-based platform RadICS. The proposed approach (FIT and MFIT) can be applied to different technologies of NPP I&Cs development.

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There have been relatively few publications detailing correlations for heat-transfer in SuperCritical Water (SCW) flowing inside bundle geometry compared to SCW heat-transfer in bare-tubes. Since flow geometry is impeded by bundle geometry, heat-transfer to the coolant occurs differently in bundles than in bare tubes. In this paper, results of heat-transfer to supercritical water flowing upward in a vertical 1-rod annular channel and 3-rod bundle (rods are equipped with helical ribs) are discussed. The data obtained in this study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles. The Jackson (2002) bare-tube heat-transfer correlation most accurately predicted Heat Transfer Coefficients (HTCs) in annular channel trials at heat fluxes less than 2.5 MW/m2. However, this correlation (along with others) did not predict HTCs accurately in annular channel trials at heat fluxes above 2.5 MW/m2 and in 3-rod bundle geometry at all values of heat flux tested. Changes in thermophysical properties of a coolant, especially, within the pseudocritical region, caused discontinuity in the calculated HTC profiles in most correlations.

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When using CFD for nuclear safety demonstration purposes, EDF applies a methodology based on physical analysis, verification, validation, application to industrial scale and uncertainty quantification (VVUQ), to demonstrate the quality of, and the confidence in results obtained. By following this methodology, each step must be consistent with the others and with the final goal of the calculations. The physical analysis, based on a PIRT (Phenomena Identification and Ranking Table) dedicated to the specific CFD scenario, has a key role to achieve this consistency. The goal of this paper is to describe a way to visually represent the "validation domain" and "application domain" for CFD scenarios, based on data given by the PIRT. The paper first focuses on how the PIRT can be used to describe and analyse any CFD transient. This leads to a practical definition of validation and application domain, based on the figure of merit, physical phenomena, main parameters and dimensionless numbers. Once the notion of domain is precisely defined, a step by step explanation of how these domains can be represented is given in the form of an exhaustive and quantitative chart. A generic overview of the chart is given to understand how the representation deals with validation hierarchy and application range. The representation is tested on an industrial example of a CFD application; the boron dilution transient. The representation of validation and application domain is based on nuclear reactor safety studies and the synthesis validation report of Code_Saturne (EDF in-house CFD code). An example shows how the representation is an efficient tool to evaluate the state of the art of the validation process by focusing on the overlapping of application domain with validation domain. The authors consider this representation to be a clear and simple decision making tool for those involved in safety demonstration. Decisions such as further development of experimental setup or acceptance of safety studies could be supported by this tool if shared with all parties involved. This paper is focused on CFD applications, but the graphical representation of domains as defined could be easily adapted to other scientific fields with few adjustments.

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When using CFD for nuclear safety demonstration purposes, EDF puts industrial applications at the core of an overall methodology including physical analysis, code verification, code validation and uncertainty quantification. This methodology ensures the quality of the results and strengthens the confidence in them. By following this methodology, it must be proved that each step is consistent with the others, and with the final goal of the calculations. The physical analysis, based on a PIRT (Phenomena Identification and Ranking Table) dedicated to the specific CFD scenario, has a key role to achieve this consistency. This paper first focuses on the methodology used to fully understand a specific accidental transient by using a PIRT approach. This approach identifies the different physical phenomena (such as stratification, jets, plumes…) involved in the transient which influence the parameters of concern for the safety analysis. The parameters driving these phenomena are then listed (boundary conditions, modeling parameters...) and their level of influence on the output is quantified, thanks to numerous CFD computation sensitivities. This work enables a hierarchy of the input parameters and phenomena influences to be built. Through the example of a heterogeneous inherent boron dilution accidental transient, this paper will show how useful the quantified PIRT methodology is in increasing confidence in the quality of CFD results, and in the identification of the input parameters leading to the most severe case.

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The SuperCritical Water-cooled Reactor (SCWR) concept is one of six Generation-IV nuclear-reactor systems currently under development worldwide. One of the benefits of such a reactor is the increase in thermal efficiency due to the reactor coolant operating above the critical point of water (22.064 MPa and 373.95℃). Pressure-Tube (PT) SCWRs are being designed to work at pressures of 25 MPa with reactor outlet temperatures up to 625℃. These operating conditions make them a suitable candidate for thermochemical hydrogen cogeneration. This work investigates the use of SCWR process heat for the thermochemical production of hydrogen. A thermochemical cycle currently being studied for this purpose is the 4-stage Copper-Chlorine (Cu-Cl) cycle, due to its relatively low temperature requirements when compared to other existing thermochemical cycles. To achieve this, an intermediate Heat eXchanger (HX) linking an SCWR Nuclear Power Plant (NPP) and a hydrogen production facility is considered. The objective of this work is to complete a heat-transfer analysis on an intermediate counter-flow double-pipe HX to be used for the cogeneration of hydrogen. The HX is located downstream of an SCWR using the reactor coolant, SuperCritical Water (SCW), through the inner tube and a separate supercritical working fluid in the outer tube. In this work the thermal energy requirements for the 4-stage Cu-Cl cycle are identified. A numerical model is developed in MATLAB and a sensitivity analysis is conducted. The sensitivity analysis determines the effect that various parameters, such as pressure, mass flux, percentage of power diverted and pipe thickness will have on the overall system. Based on the results obtained from the numerical model, the design of the counter-flow double-pipe HX can be optimized to improve its efficiency. Ultimately, this will give an indication to a size of the HX depending on input parameters that are selected.

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To meet the safety design criteria (SDC) and contributing to the establishment of the safety design guideline (SDG) for a Generation-IV Sodium-cooled fast reactor (SFR), the heat removal function of JSFR was enhanced to avoid loss of the function even if any internal events exceeding design bases or severe external events happen. As a countermeasure against loss of reactor level (LORL) type event, double failure of reactor vessel (RV) and guard vessel (GV) is prevented by securing a margin to earthquake resistant performance and a reliability of RV and GV. Furthermore, direct reactor auxiliary cooling system (DRACS) is extended functionally to keep heat removal capability even if siphon break occurs by multiple leakages on the primary cooling circuit. Against loss of heat sink (LOHS) type event, effective measures are enhancement of natural circulation function of DRACS and primary reactor auxiliary cooling system (PRACS), application of accident management and introduction of auxiliary core cooling system independent from DRACS and PRACS to prevent loss of all heat removal function from severe external events. It is important that RV melt-through can be practically eliminated by these design measure to achieve in-vessel retention (IVR) without significant core damage in the loss of heat removal type events.

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In nuclear power plants, the capacity of heat transfer is related with flow characteristics in fuel channels. It is significant to choose the suitable turbulence model for analysis and calculating friction pressure drop across the fuel channel for understanding the system features and promoting system design. In this paper, computational fluid dynamics (CFD) method was used to calculate friction factor for CANDU-6 37-element bundle fuel channel. The calculation was carried out with four turbulence models which are standard k-ε model, SST model, SSG Reynolds Stress model and BSL Reynolds Stress model respectively in the range 3.0×10^4<Re<5.8×10^5. The results were compared with empirical correlations. It shows that the results of SST model have the best agreement among that of four models and that of k-ε model and BSL RS model take the second place. The flow friction is large in sub-channels between outer ring and intermediate ring of rods but there is no significant effect on results of friction factor calculated whether to solve out secondary flow on cross section or not. This paper provides the basis for solving complex thermal hydraulic problems using CFD method with appropriate turbulence model in the next study.

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Forced convective subcooled boiling is an effective heat transfer way and used widely in industry. In order to obtain a better understanding of the thermal and hydrodynamic mechanism of the forced convective subcooled nucleate boiling, the characteristics of bubble behaviors of the forced convective nucleate boiling have been studied by many researchers. In this paper, an experimental study was conducted on the bubble behaviors in forced convective subcooled nucleate boiling in an upward annular flow at relatively high subcooling degrees. The images of bubble behaviors during its whole lifetime were captured with high spatial and temporal resolutions by using a high speed video camera and a Cassergrain microscopic lens. The temperature of liquid at the inlet of test section of experimental system was in the range of 50-70℃ at atmosphere pressure, so the subcooling of the forced convective boiling was quite high, above 30K in this study. Some interesting phenomena of the bubble behavior were observed, the bubble shape and behavior were quite different from those of lower subcooling case in forced convective subcooled boiling, the deformation at the bottom of bubble when it is departing and lifting off from the wall was observed, the phenomenon of two consecutive bubbles was also observed, the 2^<nd> bubble grows from the remaining part of the first bubble after it lifts off from the wall.

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For the application to the developing Nuclear Power Plant (NPP) safety analysis code SPACE, Korea Atomic Energy Research Institute (KAERI) proposed a new mechanistic droplet entrainment rate model which is applicable to the horizontal stratified flow condition. In the present study, an experimental work was conducted in the droplet entrainment conditions under the horizontal stratified flow condition. For this, visualization experiments for the interfacial wave were performed in the air-water horizontal rectangular channel of which width, height and length are 40 mm, 50 mm, and 4.2 m, respectively. In the test, a slope, a hypotenuse length and a surface area of an interfacial wave were measured in the droplet entrainment conditions. From this work, we improved original wave height model proposed by Taitel and Dukler [1] by using experimental data. Finally we could formulate three coefficients models in the KAERI droplet entrainment model.

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Reliability is an important issue affecting each stage of the life cycle ranging from birth to death of a product or a system. The reliability engineering includes the equipment failure data processing, quantitative assessment of system reliability and maintenance, etc. Reliability data refers to the variety of data that describe the reliability of system or component during its operation. These data may be in the form of numbers, graphics, symbols, texts and curves. Quantitative reliability assessment is the task of the reliability data analysis. It provides the information related to preventing, detect, and correct the defects of the reliability design. Reliability data analysis under proceed with the various stages of product life cycle and reliability activities. Reliability data of Systems Structures and Components (SSCs) in Nuclear Power Plants is the key factor of probabilistic safety assessment (PSA); reliability centered maintenance and life cycle management. The Weibull distribution is widely used in reliability engineering, failure analysis, industrial engineering to represent manufacturing and delivery times. It is commonly used to model time to fail, time to repair and material strength. In this paper, an improved Weibull distribution is introduced to analyze the reliability data of the SSCs in Nuclear Power Plants. An example is given in the paper to present the result of the new method. The Weibull distribution of mechanical equipment for reliability data fitting ability is very strong in nuclear power plant. It's a widely used mathematical model for reliability analysis. The current commonly used methods are two-parameter and three-parameter Weibull distribution. Through comparison and analysis, the three-parameter Weibull distribution fits the data better. It can reflect the reliability characteristics of the equipment and it is more realistic to the actual situation.

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Thermophoresis theory of solid particles in liquid are selected to research thermophoresis phenomenon in liquid Lead-Bismuth Eutectic (LBE). Thermophoretic velocity of different particles in LBE and stainless steel particles in different fluid are calculated. The results showed that, thermophoretic velocity of particles in LBE increase with the increase of temperature gradient and the decrease of particle radius. And the thermophoretic velocity of stainless steel particles two orders of magnitude lower than the Carbon Nanotubes (CNT) particles, at the same time, it is similar to copper particles in LBE. What's more, the thermophoretic velocity of stainless steel particles in LBE would one order of magnitude lower than that in water and R134a. Of course, it is still faster than that in Engine Oil and Ethyl Glycol two orders of magnitude.

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The high temperature gas cooled reactor (HTR) is characterized by agraphite core structure, helium gas coolant and ceramic fuel elements. With inherentsafety feature, the HTR is considered to be one of the generation IV nuclear reactors.The pebble bed design has been used for HTR-10 and HTR-Pmin China, so thatallows a continuous fuel loading and discharging operation. However, the pebble'smovement through the fuelling system and the core produces graphite dust. It has been estimated that abrasion due to pebble's circulation contributes the major part of the graphite dust production. In the process of circulation the pebbles are liftedpneumatically from the exit at the bottom of the core to the top and re-inserted intothe core. The pebbles make multiple collisions with the stainless steel lifting pipe,thereby causing abrasion of the graphite pebbles. Abrasion due to pebble to steelcontact is more severe than pebble to pebble contact and pebble to graphite core wallcontact, so that a test platform was designed to investigate the abrasion behavior in the lifting pipes in this study. A stainless steel pipe with a height of c.a. 9 m and innerdiameter of 62 mm is used to simulate the lifting pipe; graphite pebble with diameterof 60 mm is lifted pneumatically in the pipe. The control of mean lifting velocity was evaluated and the effect of lifting speed on the abrasion behavior of graphite pebble was investigated. The results obtained in this study will help the understanding of graphite dust generation in HTR-10 and HTR-PM.

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In the present study, the concept of Generalized Failure Mechanism Knowledge (GFMK) has been proposed and applied practically to Rokkasho reprocessing plant. GFMK is a knowledge scheme that describes failure mechanism independent of particular subject. In this scheme, sets of conditions leading to specific failure are generalized, which allows us to apply to different equipment. Once GFMK have been constructed, it is possible to estimate the likelihood of the failure without expert judgments. Following two issues have been studied in this study. 1) Building a GFMK knowledgebase based on previous troubles cases 2) Application of GFMK to specific plant sub-systems to derive failure mechanism For issue 1), the GFMK knowledgebase has been constructed based on the previous failures cases in Rokkasho reprocessing plant with emphasis on the events related to aging. For issue 2), it has been confirmed that the possible failure mechanism can be obtained by applying GFMK to the realistic scale sub-system of Rokkasho reprocessing plant. The derived failure mechanism has been compared with the ones on the design specification and has been confirmed that most of the derived failure mechanism are feasible.

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The helium circulator is the key equipment to drive the helium gas flowing in the primary loop for energy exchange in HTGR. Active magnetic bearings (AMB) have been considered as an alternative to replace traditional mechanical bearings in the helium circulator. Such contactless bearings do not have frictional wear and can be used to suppress vibration in rotor-dynamic applications. It is necessary to study the vibration characteristics of the maglev helium circulator to guarantee the reactor safety. Therefore, a maglev circulator test rig was built. The power of the circulator is 180 kW and the maximum speed is 17000 rpm. For the time being, the test atmosphere is air. In this paper the test rig was introduced. Vibration test work of the maglev circulator was also carried out. The measuring points were arranged at the seat because the seat vibration level is important to evaluate the machine noise. The measuring points were also arranged at the base of the circulator housing to better study the vibration characteristics. The vibrations were measured by the LC-8024 multichannel machinery diagnoses system. At each measuring point the vibrations were detected in three directions(X, Y and Z) with the vibration acceleration sensors. The test speeds varied from 1000 rpm to 17000 rpm with an increase of 1000 rpm each time. The vibration values of the seat are from 89.5 dB at 1000 rpm to 113.3 dB at 17000 rpm. The test results showed that the maglev circulator exhibits good vibration properties. This work will offer important theoretical base and engineering experience to explore the high-speed helium circulator in HTGR.

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Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries.

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In order to investigate ablation failure mode of lower head with molten core during a severe accident of light water reactors, ablation and melting relocation experiments were conducted with a hemisphere vessel with a drain hole. Three different silicone oil were used to investigate an effect of fluid viscosity to simulate a molten core. The hemisphere vessel is molded with lead bismuth eutectic alloy. The vessel wall thinning and melt relocation occurred just below the silicone oil level by ablation due to natural convection. For the lowerviscosity silicone oil, it results in breaking all around the vessel wall. On the contrary for the higher-viscosity silicone oil, the drain hole were ablated as well which enhance drainage flow. The time series of ablated molten wall and silicone oil weights drained from the hole were quantified separately on the basis of measured volume and weight of drained fluids for the code validation.

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The Russian TOPAZ-II reactor is a single-cell thermionic, Zirconium hydride moderated, epithermal design with a highly enriched UO2 fuel. The waste heat from the reactor is removed by the eutectic sodium-potassium coolant, and given off to the space by a radiator. According to the special feature of this space power system, a new transient analysis code (TASTIN) is developed to analysis its thermal-hydraulic characters during steady state, transient events and some typical accidents. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculated the core power considering temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes, coolant and other components in the core. Multiple-channel model is applied to depict the core. The one-dimensional modules for the main components of the coolant loop have been complied and the models are introduced briefly in this paper. A radiator model also be established in details. The steady state calculation results are in good agreement with the design values. During the partial loss of flow accident, in a certain period of time, temperature of each layer material and coolant are lower than the safe limits. It proved that the TASTIN program can be achieved on the system safety analysis of space thermionic reactor.

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In this study, the RF (Radio Frequency) tag with radiation shield is developed and its gamma ray durability is evaluated. RFID (RF Identification) is a radio-wave-based identification technology that can be used for various items. . RF tags find use in many applications, including item tracing, access control, etc. RF tags can be classified as active RF tags, which have inbuilt voltaic cells, and passive RF tags without these cells. Passive RF tags, known for their low price and durability, are used in various fields. For instance, they are used for equipment maintenance in factories and thermal power plants. Several frequencies are used for RF tags. Further, RF tagging on the UHF (Ultra High Frequency) frequencies allows a communication range of approximately 10 m, and thus, remote reading is possible. When used in radiation environments such as in nuclear power plants, remote reading can contribute to the reduction of radiation exposure. However, because semiconductors are the primary elements used in the manufacture of RF tags, they can be damaged by radiation, and operational errors can occur. Therefore, this technology has not been used in environments affected by relatively high radiation levels. Therefore, in nuclear power plants, the use of RF tags is limited in areas of low radiation levels. In our study, we develop and manufacture a new RF tag with a radiation shield cover that provides error correction functionality. It is expected that radiation shielded RF tags will improve the radiation-proof feature, and its application range will be expanded. Using the radiation-proof RF tag, we have conducted radiation durability tests. These tests are of two types: one using low energy gamma ray, and the other using high-energy gamma ray. Experimental results are then analyzed. The number of applications for radiation shielded RF tags is considerably increasing, because it can be used in various radiation environments other than nuclear power plants as well, such as spent nuclear fuel storage facilities, decommissioning sites of nuclear power plants, and for decontamination operations management around Fukushima area.

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The accident was caused by the Great East Japan earthquake and Tsunami at TEPCO's Fukushima Daiichi Nuclear power plants. In order to evaluate the plant status after the accident of Fukushima Daiichi Nuclear Power Plants , the simulation has been investigated by the SAMPSON severe accident code. In this paper, the status of Fukushima Daiichi Nuclear Power Plant unit 1 analyzed by the SAMPSON code is shown. At unit 1, water injection by isolation condenser was terminated at 0.8 hour after scram. After termination, water level inside reactor pressure vessel was decreased by evaporation of water because of the decay heat of fuel rod and etc. The gradual decrease in pressure of reactor pressure vessel by the beak of pipe for source range monitor and the leak of the gasket of safety relief valve was assumed in this simulation. After melting core structure, because the model of the behavior of debris has been improved, a status of the debris was influenced. The debris was retained on the core plate until melting core plate as conventional model, but the model was improved based on results of XR2-1 experiment so that the debris felled to lower plenum by considering a other process of the fall of the debris .After falling debris, by the model of reactor pressure vessel failure improved, the behavior of debris falling from reactor pressure vessel was influenced. The improved model was considered not only the failure of reactor pressure wall but melting of instrument pipe penetrated to the inside of reactor pressure vessel. Applying these improved models, the results of the simulation after operating sea water injection was shown.

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The reliable empirical correlations and models are one of the important ways to predict the interfacial area concentration (IAC) in two-phase flows. However, up to now, no correlation or model is available for the prediction of the IAC in the two-phase flows in large diameter pipes. This study collected an IAC experimental database of two-phase flows taken under various flow conditions in large diameter pipes and presented a systematic way to predict the IAC for two-phase flows from bubbly, cap-bubbly to churn flow in large diameter pipes by categorizing bubbles into two groups (group-1: spherical and distorted bubble, group-2: cap bubble). Correlations were developed to predict the group-1 void fraction from the void fraction of all bubble. The IAC contribution from group-1 bubbles was modeled by using the dominant parameters of group-1 bubble void fraction and Reynolds number based on the parameter-dependent analysis of Hibiki and Ishii (2001, 2002) using one-dimensional bubble number density and interfacial area transport equations. A new drift velocity correlation for two-phase flow with large cap bubbles in large diameter pipes was derived in this study. By comparing the newly-derived drift velocity correlation with the existing drift velocity correlation of Kataoka and Ishii (1987) for large diameter pipes and using the characteristics of the representative bubbles among the group 2 bubbles, we developed the model of IAC and bubble size for group 2 cap bubbles. The developed models for estimating the IAC are compared with the entire collected database. A reasonable agreement was obtained with average relative errors of ±28.1%, ±54.4% and ±29.6% for group 1, group 2 and all bubbles respectively.

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SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV nuclear-reactor concepts. They are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluidand transport-properties packages, thus enabling a transition from subcritical to supercritical fluid states. In previous work, the applicability of ATHLET code to predict supercriticalwater behaviour in various heat-transfer conditions was assessed, and it was concluded that ATHLET can be used to develop preliminary design solutions for SCWRs. In this work, experimental investigations of the local convection heat transfer of supercritical water in a 10 mm diameter bare tube are presented for upward flow for low and moderate mass flux ranges (200 kg/m^2s and 500 kg/m^2s, respectively). The effects of heat flux, the influence of buoyancy and other flow parameters are investigated to further understand the convection heat transfer of supercritical water. A numerical model in ATHLET is created to represent the experimental test section and the results for the inside-wall temperature are compared with the experimental data. Some of the empirical correlations used in the ATHLET simulations are able to capture trends; however, important phenomena like Deteriorated Heat Transfer (DHT) are not being accurately predicted. It is clear that further work is needed to understand the flow parameters and that govern the heat transfer of fluids at supercritical pressures and appropriate heat transfer correlations need to be developed accordingly.

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In the wake of Fukushima Daiichi nuclear disaster, reviews of the safety of nuclear facilities have been conducted in the world beginning with Japan. Countermeasures against severe accidents in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing diffusion of the radioactive materials, it is important to install filtered venting devices to release high pressure pollutant gas to the atmosphere with elimination radioactive materials in the gas. One of the devices for the filtered venting is a Multi venturi scrubber system (MVSS), which is used to realize filtered venting without any power supply in European reactors. The MVSS is composed of a "venturi Scrubbers" part, in which there are hundreds of the venturi scrubbers, and a "bubble column" part. In the MVSS, all of the venturi scrubbers is branched off from a vent line which connect between the containment and the MVSS. In an operation mode of the MVSS, the radioactive materials are eliminated through the gas-liquid interface from the pollutant gas to the liquid phase of a dispersed flow in the venturi scrubber and a bubbly flow in the bubble column part. The dispersed flow is formed from the liquid, which is suctioned from around the venturi scrubber through the hole for suction (called self-priming).In previous studies, an evaluation method for the scrubbing performance of the venturi scrubber was developed. However, actual hydraulic behavior in it is too complicated, the previous evaluation was not validated the hydraulic behavior and studied the effect of differences between the simulated hydraulic behavior and an actual one on the performance of the venturi scrubber. To develop a validated evaluation method for the scrubbing performance, it is important to develop detailed evaluation method for the hydraulic behavior in the venturi scrubber. To simulate the complicated hydraulic behavior, we consider to use analysis code TPFIT. Then, the objective is to validate the hydraulic behavior simulated by TPFIT. As approaches, numerical analysis by TPFIT under air-water conditions was conducted and its hydraulic behaviors were compared with the previous studies qualitatively.

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The cryostat, made of type 304 stainless steel, is required to fulfil the structural integrity and the vacuum tightness at room temperature, and this paper focuses on the fillet welding mechanical properties as a vacuum seal, especially tensile behavior and fatigue strength. Although the lid at the top is a first major part to be removed when the devices inside would be stated in faulted conditions, the closure process is expected to be low cost and simple, and examined with structural clamping and fillet welding as a vacuum seal since the cryostat is not an usual pressure vessel. This standard strength is designed as a 12 mm leg length and reduction of the welding deposition is surveyed with the other comparative specimens of two leg lengths (6 mm, 9 mm). As a result, the region linearly responded to the loading of the 9 mm specimen sufficiently envelops the standard design strength, and then sufficient fatigue strength is confirmed with the linear response limit load as an amplitude until 2000 cycles. Application of the fillet welding to the closure welding is discussed in this paper.

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Slow source convergence of Monte Carlo criticality calculation is a significant problem. Several acceleration techniques have been proposed, whereas Wielandt's method is an efficient method. However a bad estimated eigenvalue (k_e)which is selected often according experience makes the calculation time consuming. In this study, automatic optimized Wielandt's acceleration method has been developed to get optimized estimated eigenvalue based on dominance ratio (DR) for improving the efficiency of source convergence and implemented in Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC). Numerical calculated results of 5x5x5 array of metal spheres and Hoogenboom benchmark showed that automatic optimized Wielandt's acceleration method can enhance the convergence efficiency.

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The flow characteristics of unsteady flow through a symmetrical sudden expansion were investigated experimentally using digital particle image velocimetry (PIV). Expansion ratio (downstream to upstream area ratio) of 4 and 2 were selected comparatively for the study. The instantaneous two dimensional velocity profile and flow resistance were obtained. The results of pulsatile flow were compared with those of steady flow with the Reynolds number ranged from 300 to 7500, based on the upstream hydraulic diameter (De). The time average pulsatile pressure drop between the sudden enlargement drop-sill and downstream 74De is larger than that of steady flow, and the ratio of the former to the latter has positive correlation with pulsatile amplitude and frequency. The velocity profiles in the regions of separation, reattachment and re-development fluctuate periodically, and so as the recirculation length. The recirculation eddy strength, as well as the corner eddy strength, has positive relationship with the acceleration, for the acceleration intensifies the flow mixing in the recirculation zone and corner eddy, thus enhancing the dissipation of energy.

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The throttling orifice is a common steam pressure reducing equipment and steam flow limiting devices in nuclear power plant, which is used in bypass discharge tank and steam generator. To determine the value of the critical mass flow rate of steam through orifices, several groups of throttling orifices was measured. A 6mm-diameter test orifices was installed in pipes with pipe diameters of 12, 25 and 80 mm. An 8mm-depth test orifices, a 10mm-depth test orifices and a 12mm- depth test orifices, whose diameter all are 6mm, were installed in a same pipe. The 4-hole test orifices with varying orifice diameters of 4, 6 and 8 mm were installed in an 80-mm pipe. A one-hole test orifice, a two-hole test orifice and a 4-hole test orifice, which have the same area of hole, were installed in a same pipe. The critical mass flow of steam with a low degree of superheated through the straight-bore orifices was measured. Within a certain range, the bigger rate of orifice opening, the greater critical mass flow factor; the bigger rate of length to diameter, the greater critical mass flow factor of orifice; the smaller area of hole, the greater critical mass flow factor of orifice. The experimental research can provide reference for the design and optimization of the bypass discharge tank and steam generator.

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This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in the past several years, which includes new findings regarding the design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. As a practical matter, tritium production in the reactors depends on the design of the overall configuration of in-vessel components as well as breeding blanket design including materials selection. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of Li_2TiO_3 and Be_<12>Ti pebbles. The coolant was assumed to be pressurized water in the condition of 15.5 MPa and 290-325 ℃. Considering a lot of design requirements such as plasma positional control and maintainability of in-vessel components, a reasonable dimension of breeding blanket was determined. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR > 1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m^2 which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR. The result is favorable in terms of the access of remote handling equipment for maintenance and the installation tolerance of blanket modules. Based on the neutronics analysis, the latest design of breeding blanket and the other in-vessel components is presented.

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The resource-renewable boiling water reactor (RBWR) is an innovative BWR which has a capability to breed and burn transuranium elements (TRUs) using a multi-recycling process. One of RBWR core designs RBWR-TB2 was designed to be a TRU burner similar to sodium fast reactors (SFRs). The RBWR-TB2 is intended to burn TRUs from light water reactor (LWR) spent fuels and is assumed to be applied for reducing the amount of TRUs to be managed in storage facilities. The RBWR-TB2 can fission TRUs at a rate more than twice the rate of TRU production by the advanced BWR (ABWR). It is important to develop reactor internals and fuel assemblies that ensure the feasibility of the RBWR-TB2. RBWR-TB2 reactor internals are developed based on existing proven ABWR technology which is already being used in commercially operating plants. The RBWR-TB2 system is almost the same as that of the ABWR except for the reactor core. The tight pitch fuel lattice and fuel rods are introduced to the reactor core of the RBWR-TB2 to burn TRUs more efficiently. We organized the issues associated with the reactor internals that need to be tackled in order to realize the RBWR-TB2. Consequently, we confirmed that the reactor internals and fuel assemblies could be designed as commercial products. We also proposed the reactor internals and fuel assemblies that were suitable for the RBWR-TB2 which has a closed packed core. We designed a fuel assembly which had lower flow-induced vibration of fuel rods than that in the ABWR. The seismic response estimation of fuel assemblies is one of the most important design tasks for ensuring the seismic safety of nuclear reactors. The resonant frequency of our designed fuel assembly was calculated by a finite element method. The resonant frequency of the RBWR fuel assembly was 11 Hz and more than that of the ABWR assemblies. The designed RBWR-TB2 fuel assembly has higher rigidity. We concluded that the core structure of the RBWR-TB2 was feasible from the viewpoint of vibration characteristics. The RBWR appears to be a promising candidate energy source for energy security, reducing greenhouse-gas emissions, and mitigating the negative environmental impact of TRUs.

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In the large advanced passive PWR nuclear power plant, the long term core cooling (LTCC) following loss of coolant accidents (LOCA) is provided by passive safety system. The purpose of this paper is to study whether the factors that could affect LTCC challenge the safety margin. According to the preliminary PIRT (Phenomena Identification and Ranking Table) of the large advanced passive PWR nuclear power plant, the following factors have been selected: the timing of the recirculation initiation ahead, the swing check valves in the safety injection lines partially opened, core inlet blockage due to debris, the resistances of automatic depressurization system ADS valves, the containment water flooding inventory, the containment pressure and the recirculation temperature. The cases of LTCC following a double-ended direct vessel injection (DEDVI) line break are analyzed. The effects of the increase of ADS valves resistances, the decrease of containment pressure and the increase of recirculation temperature are calculated respectively. It's found that the containment pressure and the containment sump temperature may also play important roles during LTCC. A limiting case combined with all of these factors is performed finally, which the results demonstrates that the passive systems of the large advanced passive PWR nuclear power plant will provide adequate core cooling performance during LTCC.

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Simulation of transition for internal flow is a desirable work for engineering application. We tried to modify SST transition model for transition prediction of internal flow by adjusting the source term of transport equation for transition onset momentum thickness Reynolds number. As the first step towards this goal, a constant Reynolds number of transition onset was employed. The value was selected by a sensitive study. The modified model was applied to calculate 3D steady and pulsating forced convection heat transfer in a mini-rectangular channel. Comparison with the experimental result shows that the model can successfully predict the trend of steady friction factor, Nusselt number, the amplitude of wall and fluid temperature fluctuation. The fluctuation amplitude of wall temperature decreased first and then increased, but finally decreased again as observed both in simulation and experiment. However, as a result of too big Reynolds number of transition onset we chose, the critical Reynolds number predicted by simulation is bigger than that obtained by the experiment. The value of above mentioned parameters also deviated from experimental data. Further improvement is needed.

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Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aim at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO_2 -Zr fuel system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of an extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF.

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The natural circulation operational characteristics of passive residual heat removal system (PRHRS) in secondary loop under ocean conditions were simulated. Mathematical models of PRHRS natural circulation loop under ocean conditions were established and a simulation program was developed. The flow rate, pressure of steam generator, drivinghead and additional pressure drop of PRHRS under ocean and motionless conditions were investigated and compared. The results indicated that natural circulation capacity of PRHRS decreases under ocean conditions. Rolling has the biggest impact on this system among rolling, pitching and yawing. Average flow rate and driving head decrease by increasing rolling amplitude but remains unchanged with the change of rolling period, while average additional pressure drop in one period is zero. The flow rate fluctuates more severely under conditions of short rolling period or large amplitude. Moreover, it is found that the pressure of steam generator decreases more slowly under ocean conditions than that under motionless condition due to flow rate fluctuation and reduction.

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For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, "Nuclear safety and radioactive pollution prevention "Twelfth Five Year Plan" and the 2020 vision" raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs.

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CFETR (Chinese Fusion Engineering Testing Reactor) is a superconducting Tokamak device. The simulation was performed on the cooling circuit inside water cooled breeder blanket which is one of the breeding blanket candidates for CFETR to compare the thermal hydraulic characteristics based on superheated steam and PWR (pressurized water reactor) water conditions. The work is carried out using MELCOR code. As a result, under the optimal flow distribution, it turns out that the thermal hydraulic characteristics meet the design requirements and the nuclear heat inside blanket module as well as heat flux from plasma could be removed completely by the cooling circuits on both conditions. It's also found that the temperature of first wall (FW) is higher on PWR condition than superheated steam. But due to the generation of superheated steam, the wall temperature would grow significantly in the backward zone of cooling circuit. Based on the steady state, LOFA (loss of flow accident) was simulated. It was predicted that FW will melt anyway due to the high nuclear heat without considering thermal radiation. As for different ways of FPSS, a 100s delay is invalid and a 3s delay could buy us time to take measures.

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In the present work, a CFD model is presented to simulate the wall condensation from steam-noncondensable gas mixture. In the proposed CFD model based on the Fluid Film Model built in STAR-CCM+, the behavior of the fluid film on the cold solid surface is predicted by the conventional transport equations and it is coupled with the flow calculation for steam-air mixture region through source terms describing diffusion of steam towards liquid film. To account for enhancement of heat and mass transfer due to suction, an existing suction correction factor is also incorporated into the condensation model. For validation of the proposed CFD model, several simulations are conducted and the results are compared with well-known correlations and experimental data for flat plates and cylindrical condenser tubes. The comparison results overall show good agreements.

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In the sodium-cooled fast reactors (SFRs), the hypothetical energetic core disruptive accident (HCDA) is one of the major concerns for the safety of the SFRs. If the molten fuel is not fragmented and dispersed, there would be a possibility of recriticality of significant energetic reaction. This means that it is necessary to identify whether the molten jet is broken and fragmented well enough in the channel or core. In particular, metal fuel is known to have relatively lower possibility to reach up the HCDA compared to oxide fuel. The current Korean SFR program emphasizes the adoption of the metal fuel design for the reason. In order to investigate the dominant phenomena for the jet breakup and the fragments, the visualization experiments were conducted by injecting the molten jet into the coolant in a pool tank. The wood's metal and water were used to simulate the molten jet and the coolant, respectively. The experimental data showed that the jet breakup and the size of fragments were dependent on the jet velocity, the jet diameter, the initial temperatures of two simulants, and the injection gap length. High speed camera was used to observe the breakup behavior of the molten jet for each experimental condition. Thermal effects on the fragmentation were observed only with the initial temperature conditions. Other experimental variables showing hydraulic effects were closely related to the jet breakup phenomena. After each experiment, the debris was characterized to analyze characteristics of the jet breakup. In addition, the current study shows effects of the horizontal and narrow-down injection into a subchannel of a fuel assembly.

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