The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group cross sections and constants from the assembly-level homogenization process. The traditional deterministic lattice codes are incompetent in the more and more complex geometries and complicated physics. In contrast, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. A code, named "Monte Carlo Multi-group Constants Generation Code", or MCMC, as part of RMC Program, is being developed in Tsinghua University. The methods to estimate multi-group cross sections are introduced in this paper. The code is applied in four types of assembly configurations to validate the accuracy and the applicability. At core-level, a PWR prototype core is examined. The effective multiplication factors and flux distributions are comparable with the references.

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With the rapid development of nuclear power, there is a great lot of spent fuel removed from reactor every year. Because of its high radioactivity and decay heat, how to safely dispose of it has became an important problem. There are two methods for interim storage of spent fuel, one is the dry storage and the other is wet storage. But before interim storage the spent fuel unloaded from reactor should be put into the spent fuel storage pool in plant for several years. Whether a spent fuel storage system is intact or the outer main cooling system is failure, there will be a natural circulation in the whole spent fuel storage pool. The understanding of its characteristics is very significant to system security. In this paper, RELAP5 code is used as the research tool to study a spent fuel storage pool. And then corresponding nodalizations are established to discuss various thermal hydraulic characteristics of the entire pool with natural circulation, including the flow and temperature distribution when coolant pump is in normal operation and the flow and temperature history when the pump is failure. In addition, the paper also discusses the effect of different location of cooling system drain pipe on natural circulation in the pool. The results show that: the decay heat of spent fuel can be perfectly removed when cooling system is on normal condition; furthermore, when coolant pump is failure, the decay heat can be absorbed by pool water utilizing the natural circulation in pool, which is enough to remove the decay heat for a long time before boiling, and the time need by pool water to reach boiling from steady state is about 17 hours. Finally, the effect of the location of cooling system drain pipe on natural circulation is significant and some suggestions are given. The paper can be a reference of design and optimization for spent fuel storage pool and has important engineering significance.

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The concept of inert matrix fuel (IMF) has been proposed to reduce the worldwide growing stockpiles of plutonium oxide (PuO_2). Small disks of a simulated IMF were synthesized by a low-temperature fabrication route not exceeding 1050℃. The samples on a silicon carbide (SiC) base possessed theoretical densities (TD) ranging from 68% to 80%. Therefore coarse and fine SiC particles were mixed with a liquid polymer precursor that yielded a one-to-one Si-to-C ratio. Five weight percent of ceria (CeO_2) were added to act as a surrogate for PuO_2. The specimen received thermal shocks up to 900 K by applying the water-quenching method. The mechanical properties: Vickers hardness, fracture strength and fracture toughness were determined and compared to the published data of other IMF, UO_2 and MOX fuels. Although the fuel possessed a significantly lower density, the mechanical properties proved to be adequate when compared to conventional UO2 fuel. The temperature shock resistance of the fabricated fuel was found to be exceptionally good. XRD analyses indicated a chemical reaction between CeO_2 and the applied polymer precursor during the sintering process. A crystalline compound, namely a cerium oxysilicate was detected.

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The wall boundary layer is an inalienable part of the contours with heavy liquid metal coolants (HLMC) that are used in the fourth generation nuclear reactors. The properties of the wall boundary layer determine a reactor's efficiency and influence hydraulic characterictics and heat exchange. Characteristics of the wall boundary layer "HLMC-constructional material" have been studied by various techniques and methods at the Nizhny Novgorod State Technical University (NNSTU). The study included: - ultrasonic analysis; - determination of the contact thermal resistance; study of the influence of the wall boundary region characteristics on the MHD resistance of the HLMC flow. Due to the results of this research the modern model of the wall boundary layer in the medium of heavy metal coolants was built. The following characteristics were experimentally found in the wide range of parameters: - the magnitude of the contact thermal resistance of the wall boundary layer in the Peclet number range from Pe=260 to Pe=1430 with the oxygen concentration varied in the range from 10^<-7> to 10^0; - the dependences of the hydraulic loss coefficients on the Stuart criterion in the magnetic field.

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The most effective methods for nondestructive control of a cable condition, which are used for a periodic assessment of expected life of power and control cables at nuclear power plants of Rosenergoatom Concern OJSC, are presented. In spite of the fact that many methods are based on physical principles that are well-known for a long time, they have been improved due to the experimental data received lately.

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There are several methods for enhancement of heat transfer; for example, there are attaching various fins on the heat transfer surface, processing the surface roughly, and so on. When cooling high temperature circular or rectangular channels by forced convection of gas, there are several methods for enhancement of heat transfer such as attaching radial or spiral fins on the channel surface or inserting twisted tape in the channel. In the case of the gas heating type steam reformer, disk type fins are attached on the outside surface of the reformer tube, and the tube is inserted into the guide tube to increase an amount of heat transferred from the high temperature gas. However, it has to take into consideration the deterioration of the structure strength by attaching the fins on the tube surface with the design of the steam reformer. The objective of this study is to clarify performances of a method for heat transfer enhancement using porous material with high porosity. The experiment has been performed using an apparatus which simulated the passage structure of the steam reformer to obtain characteristics of heat transfer and pressure drop. From the results obtained in this experiment, the heat transfer rate by this method showed a good performance in the laminar flow region. It was also found that the method for heat transfer enhancement using porous material with high porosity is further improved under the high temperature condition as compared with the other methods for heat transfer enhancement.

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Based on the Defense-in-depth Concept in nuclear and radiation safety, Defense-in-depth Concept for design management of Nuclear Power Plant (NPP) is developed in this paper to analyze the feasibility and importance of the application of the basic principle: Defense-in-depth concept in NPP systems performed during the design control of NPP. This paper focuses on the NPP engineering management process, and according to the analysis of such process, 5 principles of Defense-in-depth Concept applied in NPP design management are raised: (1) preventing the non-conformities of design via effective design quality management system; (2) discovering and correcting non-conformities of design quality in time via design checkup and design review meeting; (3) carrying out timely analysis and treatment against design non-conformities which have been transferred to construction phase; (4) Assessing and judging the severe non-conformities in construction phase, putting forward treatment opinions and remedies accordingly so as to avoid the existence of such non-conformities in physical construction of NPP; (5) Paying "return-visit" and performing "post-assessment" for NPP design to assess the designed functions and safety of NPP comprehensively.

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Tube support plate clogging of steam generators affects their operating and requires frequent maintenance operations. A diagnosis method based on dynamic behaviour analysis is under development at EDF to provide means of optimisation of maintenance strategies. Previous work showed that the dynamic response to a power transient of the wide range level measurement contains informations about the clogging state of steam generators. The diagnosis method consists of comparisons of the measured dynamic response with simulations on a mono-dimensional dynamic steam generator model for various input clogging configurations. In order to assess the potential of this method, a sensitivity analysis has been conducted through a quasi-Monte Carlo scheme to compute sensitivity indices for each half tube support plate's clogging ratio. Sensitivity indices are usually defined for scalar model outputs. Principal component analysis has been used to determine a small subset of variables that condense the information about the shape of the response curves. Finally, estimation variability was assessed by construction of bootstrap confidence intervals. The results showed that half of the preselected input variables have negligible influence and allowed to rank the most important ones. Interactions of input variables have been estimated to exert only a small influence on the output. The effects of clogging on the steam generator dynamics has been characterised qualitatively and quantitatively.

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An improper cleanliness in the secondary circuit of nuclear plant steam generators may result in tube support plate blockage by deposits, and tube fouling. In order to improve our understanding of those two phenomena, a model for the growth of solid deposits on the secondary side of steam generators has been developed by the R&D Division of EDF. This model has been implemented in the frame of THYC, which is the EDF reference code for the modelling of two-phase thermal-hydraulic phenomena at the steam generator scale. It aims to evaluate the localization and the growth rate of deposits, as well as the resulting additional pressure losses. The model considers that the deposits are corrosion products that are introduced in the steam generator either as suspended particles or as dissolved species. The transport of corrosion products in the secondary circuit is modelled, as well as the mass transfers between dissolved species, suspended particles and wall deposits. Several deposit mechanisms have been investigated, such as boiling, diffusion, turbulence, and deposit removal. Mass transfers being complex processes which depend on several physical and chemical local parameters, they cannot be represented by simple laws. It is therefore necessary to introduce some empirical correlations between thermalhydraulic parameters and deposit growth rates. Those correlations depend on parameters whose values can only be determined by specific experiments performed on devoted test-facilities. Authors of the domain provide different values for those parameters, depending on experimental conditions. At the steam generator scale and in Pressurized Water Reactors (PWR) conditions, their values are not determined. We hereafter present an inverse method, that could enable one to evaluate those parameters in PWR conditions, by fitting the results of simulations to the actual levels of deposits observed in some French nuclear plants. In this paper, we only discuss the case of tube support plate blockage but the method is in principle also applicable to deal with tube fouling. The limits of the model and the strategy to improve it are discussed.

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Forced convection subcooled water boiling experiments were conducted in inclined rectangular channels. The inclination angle was 45° and the heating surface was downward facing upflow. Bubble diameter distributions in three different axial position of the heating surface had been determined from high-speed digital video camera and image processing. The bubble size had been statistically analyzed under each condition. The result showed that the Probability Distribution Function (PDF) for the bubble diameter often proved to be well represented by a log-normal distribution law. Because of the slipping bubble growth on the heating surface and the bubble coalescence, the bubble size distribution in downstream position was larger than that in upstream position. The results of the bubble size distribution were also presented as cumulative distribution functions, which exhibited in reality a very wide spread of bubble sizes. Compared with vertical upflow in the outlet position, a large number of big slipping bubbles were observed in inclined down facing upflow and the bubble size distribution was larger than that in vertical upflow.

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A tightly-coupled numerical approach is desired for safety analysis of Very High Temperature Reactors (VHTRs) since its high-level of passive safety design introduces longer transient scenarios, as compared to current nuclear power plants. The commonly employed operator-splitting approach for multiphysics time integration is susceptible to additional truncation errors, and accumulation of the truncation errors can alter the numerical solution of the physics equation system. Multiphysics core simulations for a prismatic-type VHTR are performed in this study. Our solution scheme is based on the Jacobian Free Newton Krylov (JFNK) method, which enables second-order convergence by solving all the constitutive relations consistently at a new time level without forming a complex Jacobian matrix. Under a Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, the code is written in modular fashion, which allows one to restrict or extend the governing physics at the input stage. As a preliminary example, a thermal-fluid calculation is performed with an idealized two-dimensional symmetric representation of the GT-MHR and compared with the RELAP5-3D simulation results. Also, a neutronics calculation is conducted using the same geometry as the thermal-fluid calculation, and using cross section data obtained from an HTGR benchmark problem. In addition, a coupled steady-state thermal-fluids neutronics calculation is performed. The calculation results showed that the developed prismatic VHTR core simulator can perform tightly-coupled multiphysics simulations efficiently, taking full advantage of the MOOSE framework. It is expected that due to the flexibilities of MOOSE the accuracy enhancement of nuclear reactor simulations via higher-fidelity physics models such as Navier-Stokes and transport corrected neutron diffusion equations can be added and utilized while retaining the speed of simulations.

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Public offering system of research and development ideas in The Japan Atomic Power Company (JAPC) is introduced. This system is established in 1999 to solve various subjects existing in JAPC's nuclear power stations by using highly advanced technique possessed by the enterprises or the organizations limited in Fukui Prefecture, and also to contribute them to become more advanced in their skill of technology. We have improved this system to make the results more applicable. We think it is important to coexist with the local community for the company engaged in nuclear energy as to cooperate together for the improvement of the local society.

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The Reactor Cavity Cooling System (RCCS) is one of the new safety systems designed for the next generation of nuclear power plants which will be incorporated into proposed reactor designs for the Very High Temperature Reactor (VHTR). The new RCCS experimental facility at Texas A&M University was conceived to observe, study and investigate the complex thermohydraulic phenomena taking place in the Reactor Cooling Cavity System during an accident scenario when forced convection is lost. A set of simulations using RELAP5-3D was carried out to predict the main thermohydraulic parameters, such as water and wall temperatures, cavity air temperature, flow regimes of water in the standpipes and in the water tank and the overall time response of the system. These predictions were used to provide a clearer picture of the phenomena that will be observed during the experimental session, to confirm the hand calculations made during the scaling of the facility from the full scale plant, and to refine and optimize the facility's configuration and instrumentation layout. The purpose of this paper is threefold: to show the concepts and the techniques adopted to create the RELAP5-3D hydrodynamic model, the assumptions and methods used to optimize the heat transfer mechanism model and the sensitivity analysis carried out in order to improve the code prediction of particular expected phenomena, such as water flashing and liquid recirculation. The main simulation results will be presented with an emphasis on the flow instability, typical of flows through parallel heated channels.

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This paper describes the achievements of a program in which technology education is provided to cultivate practical core engineers for low-level radiation. It was made possible by means of (1) an introductory education program starting at an early age and a continuous agenda throughout college days and (2) regional collaboration. First, with regard to the early-age introductory education program and the continuous education agenda, the subjects of study related to atomic energy or nuclear engineering were reorganized as "Subjects related to Atomic Power Education" for all grades in all departments. These subjects were included in the syllabus and the student guide book, emphasizing a continuous and consistent policy throughout seven-year college study, including the five-year system and additional two-year advanced course. Second, to promote practical education, the contents of lectures, experiments, and internships were enriched and realigned in collaboration with the Japan Atomic Energy Agency, Okayama University and The Cyugoku Electric Power Co., Inc. In addition to the expansion and rearrangement of atomic power education, research on atomic power conducted for graduation thesis projects were undertaken to enhance the educational and research activities. In consequence, it has been estimated that there is now a total of fourteen subject areas in atomic energy technology, more than eight-hundred registered students in the department, and thirteen members of the teaching staff related to atomic energy technology. Furthermore, the "Tsuyama model" is still being developed. This program was funded by the Ministry of Education, Culture, Sports, Science and Technology.

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In order to meet fast development of nuclear power, China should establish its own nuclear power standards system. This paper gives some opinions on establishment of Chinese nuclear island systems and components design and construction standards. It is suggested to draft "Chinese Nuclear Power Utility Requirements Document".

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Critical Heat Flux (CHF) is one of the important parameters for the thermal design of a rod bundle. Subchannel analysis is recommended in the absence of experimental data to determine the thermal margin of a new fuel design using empirical correlations. However, there are a large number of empirical CHF correlations available even for a simple geometry which indicates that a generic correlation with wide range of validity is not available in the literature and hence mechanistic approach could be more appropriate. A mechanistic model based on the Liquid Film Dryout approach with the constitutive models suitable for the BWR conditions have been employed in conjunction with the subchannel code COBRA-IIIC. This paper deals with the film dryout analysis of a typical BWR fuel bundle considering a simplified mechanistic approach in conjunction with the subchannel code.

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An advanced integral pressurized water reactor, SMART (System-Integrated Modular Advanced Reactor) has been developed by KAERI (Korea Atomic Energy Research and Institute). The purposes of the SMART are sea water desalination and an electricity generation. For the safety evaluation and performance analysis of the SMART, TASS/SMR-S (Transient And Setpoint Simulation/Systemintegrated Modular Reactor) code, has been developed. In this paper, the gap conductance model for the calculation of gap conductance has been validated by using another system code, MARS code, and experimental results. In the validation, the behaviors of fuel temperature and gap width are selected as the major parameters. According to the evaluation results, the TASS/SMR-S code predicts well the behaviors of fuel temperatures and gap width variation, compared to the MARS calculation results and experimental data.

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The tendency has always been to build ever larger single-modular reactor plants with the objective of benefiting from economies of scale. These plants have compiled admirable safety records. Nevertheless, there is concern that conventional large single reactors have become too complex by reason of placing too much reliance on engineered safeguards. The multi-modular approach offers a solution in that its use of many small reactors in conjunction with several shared turbines permits a simpler core design while, at the same time, at least partially retaining economies of scale by increasing the number of modules. Specific advantages to the multi-modular approach are as follows. First, the small-sized of the reactor core may allow the incorporation of passive safety features such as natural circulation cooling on loss of off-site electricity. Second, the individual modules are to be sized so that components related to nuclear safety can be factory-fabricated. Moreover, once the major components are made, they are to be transported to the site for rapid installation. This construction method is expected to reduce the licensing effort because the modules will be pre-licensed, and only site-specific issues will have to be considered in the final licensing process. At present, related studies show that the multi-modular approach for Generation IV can retain both the inherent safety and good economies of scale. However, the unbalanced load operation of the multi-modular power plant in which each module operates at a different power level and strong coupling between multi modules creates a complex control challenge to safe operation and control. Firstly, this paper summarizes the unbalanced load operation characteristics and challenges faced by operation and control of multi-modular power plant in the dynamic operational characteristics and requirements of coordinated control between multi modules. Secondly, detailed analysis and comparison are given in the integral supervisory hierarchical control structure, operation and control strategy of existing multi-modular reactor in the world. Finally, this paper indicates that the high-performance autonomous control method is one of the possible research directions to solve the problems of operation and control of multi-modular power plant.

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A code for IVR (In-vessal Retention) analysis has been programmed based on a well-known IVR stable bounding computation methodology. The code has been verified by comparing the calculation results with ERI, DOE, and INEEL results for AP600 and AP1000. The motivation of this research work is to investigate the feasibility of adopting IVR strategy to mitigate the severe accident of a typical large-scale advanced PWR (power level is higher than 1400MWe). According to molten pool parameter analysis and uncertainty analysis, the probability of exceeding-CHF accident in large-scale advanced PWR is smaller than 7% if the flow path of ULPU-2000 configuration for vessel external flooding is used, but the probability that the largest melt thickness exceeds 15cm is very large. Heat flux distribution of the lower head of the pressure vessel plays the main role to influence the molten pool behavior. It is suggested to take engineering action to adjust the heat flux distribution for mitigating heat transfer crisis at the region near oxide zone and metal zone. The results obtained are encouraging for further study on IVR strategy to mitigate severe accident of large-scale PWRs.

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Hydrogen is expected to serve as a clean secondary energy, because it can be manufactured from water, used in a variety of energy end-use sectors as fuel, and returned to water after burning. For the realization of hydrogen energy system, development of efficient and economical hydrogen production methods is required to meet the future huge demand of hydrogen. The Iodine-Sulfur (IS) process is a promising candidate of such hydrogen production methods, in which water reacts with iodine and sulfur dioxide to produce hydrogen iodide and sulfuric acid (Bunsen reaction) and the produced acids are then decomposed to produce hydrogen and oxygen, respectively. This study is concerned with the development of IS process equipment named direct contact sulfuric acid concentrator, in which gaseous mixture produced by thermal decomposition of sulfuric acid contacts directly with sulfuric acid solution. In the concentrator, the high temperature heat of the decomposed gas is recovered and used to concentrate sulfuric acid solution and, at the same time, the undecomposed sulfuric acid is condensed and separated from the decomposition products of sulfur dioxide and oxygen. Although the concept is very attractive from the viewpoint of the development of compact and efficient sulfuric acid concentrator, little is known on the heat and mass transfer relevant to the concentrator. Therefore, experimental methods were discussed to acquire the gas-phase mass transfer coefficient required for the optimal design of the concentrator. Assuming the use of wetted-wall column and also of the sulfuric acid of azeotropic composition as the test solution which could eliminate the liquid-phase mass transfer resistance, the column specification and the measurement conditions were determined by which flooding could be avoided and surface wetting could be assured, as well.

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Three low Reynolds number (LRN) k-ε models, one LRN k-ω model and the k-ω SST model are tested with OpenFOAM for the computation of high-Schmidt- number mass transfer in the flow-accelerated corrosion (FAC), especially for the separated and reattaching flow. Three types of flow are selected for the test of models: 1) the fully developed pipe flow, 2) the axisymmetric flow with an abrupt expansion, 3) the flow through an orifice. The model developed with the aid of direct numerical simulation (DNS) data, the Hwang-Lin model, shows a good performance in the fully developed pipe flow, but its prediction in the latter two flows is far from reliable. The LRN k-ω model and the k-ω SST model predict a low mass transfer rate for all three types of flow. The Lam- Bremhorst model shows abnormal behavior at the reattaching point. Synthetically evaluating all the models in all the computed case, the Abe-Kondoh-Nagano model is the best one; however, the prediction is still not satisfactory.

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This paper presents the coupled RELAP5-3D and MCNP5 baseline PB-AHTR core model. The motivation of the study is to couple the thermal-hydraulics (T/H) with neutronics to investigate the feedback mechanism, coupled calculation viability and disparities of results between coupled and traditional approaches during steady state and transient scenarios in the PB-AHTR innovative core design, and also provide preliminary modeling approaches and results for further study. In order to analyze the steady state T/H behavior of the core and communicate with the neutronics calculation, a PB-AHTR MCNP5 core model, which achieved reliable steady-state, was developed. The power distribution of the PB-AHTR can be calculated by MCNP5 in fine enough resolution that it can be mapped to the RELAP5-3D mesh. Eventually, the preliminary results from the MCNP5 calculations during the transient are compared with those obtained from RELAP5-3D point kinetics model, indicating that the approach of comparing the reactivity evolution calculated by MCNP5 and RELAP5-3D during the transient is feasible. Further work will be focused on the systematic configuration of the RELAP5-3D model for more transient simulations.

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Steam injector (SI) is a passive jet pump which operates without rotating machineries and a high efficiency heat exchanger driven by the direct-contact-condensation phenomena between a supersonic steam flow and a water jet. The objective of this study is to elucidate the mechanisms of the heat and momentum transfer phenomena, as well as to investigate the effect of interfacial behavior on heat and momentum transfer characteristics induced by direct-contact condensation phenomena. In the present study, a transparent test section of SI was adopted to observe the flow structure with a high speed video camera. In addition, special measurement instrumentations of temperature and total pressure were applied to obtain radial distribution of temperature and flow velocity in the mixing nozzle. Several indicators of heat and momentum transfer characteristics were estimated from those measurement results. Additionally, complex wavy behavior on the water jet surface was observed and quantified with the image processing technique and correlation method, and the velocity were correlated against inlet conditions such as inlet water and steam flow rate. Heat transfer characteristics between the flows were well described with Nusselt number and Jacob number, furthermore, it was confirmed that there was correlation between momentum transfer characteristics and interfacial behavior, which was reasonably described with the correlation of slip ratio between the flows and interfacial wave velocity. Mechanisms of heat and momentum transfer induced with the direct-contact-condensation mediated with complex interfacial behavior are discussed from those experimental results.

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Since the beginning of this century, the so-called 3Ss (Nuclear Safety, Nuclear Security and Safeguards) have become major regulatory areas for peaceful uses of nuclear energy. The importance of the 3Ss is now emphasized to countries which are newly introducing nuclear power generation. However, as role models for those newcomers, existing nuclear power countries are also required to strengthen their regulatory infrastructure for the 3Ss. In order to rationalize the allocation of regulatory resources, interrelationship of the 3Ss should be investigated. From the viewpoint of the number of the parties concerned in regulation, nuclear security is peculiar with having "aggressors" as the third party. From the viewpoint of final goal of regulation, nuclear security in general and safeguards share the goal of preventing non-peaceful uses of nuclear energy, though the goal of anti-sabotage within nuclear security is rather similar to nuclear safety. As often recognized, safeguards are representative of various policy tools for nuclear non-proliferation. Strictly speaking, it is not safeguards as a policy tool but nuclear non-proliferation as a policy purpose that should be parallel to other policy purposes (nuclear safety and nuclear security). That suggests "SSN" which stands for Safety, Security and Non-proliferation is a better abbreviation rather than 3Ss. Safeguards as a policy tool should be enumerated along with nuclear safety regulation, nuclear security measures and trade controls on nuclear-related items. Trade controls have been playing an important role for nuclear non-proliferation. These policy tools can be called "SSST" in which Trade controls are also emphasized along with Safety regulation, Security measures and Safeguards. Recently, it becomes quite difficult to clearly demarcate these policy tools. As nuclear security concept is expanding, the denotation of nuclear security measures is also expanding. Nuclear security measures are more and more overlapping with measures for non-proliferation including safeguards and trade controls. Such situation should be clarified in order to design better regulatory systems, especially for nuclear security regulation.

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Most of recent subchannel analysis codes are based on a multi-fluid model, and accurate evaluation of the constitutive equations in the model is essential. In order to get accurate ones of the wall and the interfacial friction forces and validate them, experimental data on these forces have been obtained from the pressure drop and the void fraction data for air-water annular flows in 2×1 rods channel simplifying a BWR fuel rod bundle geometry. In order to know the effects of liquid properties on the data, water temperature was changed from 18 to 50℃. The wall and interfacial friction forces data determined are compared with the existing correlations reported in literatures. As a result, correlations of NASCA code and RELAP5/MOD2 code, and homogeneous flow model with McAdams' viscosity model for the wall friction force, and Fukano and Furukawa's correlation for the interfacial friction force show the best prediction against the present data.

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As the economy grows, demand of electricity is growing in China. However, building more thermal power plant is obviously improper choice, due to the concern about running out of fossil fuel. According to National Program for Medium-to-Long-Term Scientific and Technological development published by Chinese government, 40GWe nuclear power capacity will be achieved by 2020. Regarding to the speed of nuclear power plant construction in China now, recent report said that the nuclear capacity might rise to 60 GWe even 70GWe by 2020 and the further substantial increase to 200 GWe by 2030. However, to guarantee sustainable supply of electricity by nuclear power, a large amount of uranium is needed. While, there is a limitation of uranium resources, too. A light water reactor (LWR) and fast breeder reactor (FBR) matched scenario should be considered to prevent the crisis of running out of nuclear fuel. The purpose of this study is to find out the best LWR-FBR matched scenario which can reduce uranium requirement efficiently. Four scenarios which consist of 10 cases are selected in this study. After simulating each case by two computer codes, it is clear that scenario 3-1 is the most efficient case in the aspect of saving natural uranium. Scenario 3-1 is a scenario that LWRs loaded with MOX fuel partially (MOX fuel is a kind of nuclear fuel which is a mixture of PuO_2 and UO_2) are introduced in 2020 and then they were replaced by Fast Breeder Reactors in 2050.

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High temperature gas-cooled reactor coupled with helium turbine cycle (HTGR-GT), based on the closed Brayton cycle, can take advantage of high temperature and achieve the high efficiency in power generation. The typical power conversion unit (PCU) of HTGR-GT is the vertical, single-shaft and compact configuration. All components of HTGR-GT are housed in pressure vessel, connected each other by pipes and irregular flow channels. Therefore the flow losses, including pressure drops and leakages inside pressure vessel, are unavoidable in actual projects and had significant influence on cycle efficiency. This paper presented a dimensionless model to analyze the effects of flow losses on cycle's features. (1) The effects of pressure drops: It showed that the pressure drop ratios of different components were inherently connected by Reynolds number due to the closed cycle. And the effectiveness of the heat exchangers can also be linked with the pressure drop ratios by Reynolds number. So the cycle's efficiency just depends on the effectiveness of recuperator and compression ratio, and there exists optimal recuperator's effectiveness and maximum cycle's efficiency. (2) The effects of pressure drops and leakages: leakages inside the pressure vessel could be divided into two groups by the seal's structure. It is found that leakage ratio could be connected to pressure drop ratios, seal's structure and other thermodynamic parameters. Therefore, the cycle's efficiency can also be expressed as the complex function of the effectiveness of recuperator and compression ratio, and leakages significantly reduce the cycle's efficiency. In brief, flow losses pose the limitation of cycle's efficiency for engineering design of HTGR-GT projects.

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Some damages were found at the forks of low pressure turbine blades due to high-cycle fatigue caused by random vibration and steam-flashback vibration. In this study, laser peening technology was developed to improve high-cycle fatigue properties of 12Cr stainless steel of blades material, and the effect on the material properties was examined. Laser peening is a process to induce compressive residual stress to the material surface. Fatigue specimens which simulate the stress concentration zone of the forks were fabricated and laser peening was applied to the surface of the specimens. Residual stress was measured by X-ray diffraction method and it was confirmed that compressive residual stress was formed on the peened surface. As the results of the fatigue test, fatigue strength of the laser peened specimens was shown to improve by about 40 percent compared to that of the unpeened specimens.

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The Japan Atomic Power Company (JAPC) is the only electric power company in Japan solely engaged in nuclear energy. In order to fulfill our role as a pioneer in nuclear power generation, various projects have been undertaken, including the construction of the first commercial nuclear power plant in Japan and the construction, operation and decommissioning of power plants in Tokaimura in Ibaraki prefecture and in Tsuruga city in Fukui prefecture. JAPC is an electric nuclear power generation company which sells electricity directly to the electric companies. Since JAPC doesn't have retail customers, it has limited opportunities to come in contact with local residents. It is essential to gain the confidence and understanding of nuclear power by local residents in order to promote our projects and to manage our nuclear power plants. Under these circumstances, JAPC has steadily developed public relations in local areas and surrounding neighborhoods through an action policy of two-way communication. In this presentation, the two-way communication public relations policy will be explained. I would like to describe the achievements of the two-way communication policy by referring to the results of public opinion surveys conducted in Ibaraki and Fukui prefectures

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This study is about the acceptability of the society of utilizing nuclear power in China. There are different interactions and opinions has been raised that leads conflict between the society and the nuclear technology people. Many emotions and views have mixed up: people have image of environment friendly, also have image of explosion, fright and danger. The purpose of this study is to make clear the social consciousness for utilizing nuclear power in China and to improve social acceptability. A survey by the questionnaire was conducted for investigation of the features of the social consciousness for necessity and safety for utilizing nuclear power in China. The questionnaire has been sent to 10000 citizens of Zhejiang province in south China from June to August 2010 (Zhejiang province is one of the earliest city for utilizing nuclear power plants in China).There were 4255 available respondents (42.6%) with 1851 men and 2404 women of adult who answered the questionnaire. The survey was including 37 items about energy problem, social consciousness for utilizing nuclear power, safety, the reliability of government and information sources. As a result, it was found that some 57.5% of adults are in favor of nuclear power plants. There are differences of sense between men and women. The reliability for safety of men for nuclear power is higher than that of women. Some 59.6% of men respondents are in favor of nuclear power plants, and about 10% respondents of women are in favor of nuclear power plants. Social acceptance is still one of the major barriers for further development of nuclear power, although recent technological and institutional innovation is clearly reduced its risk and enhanced its relative and absolute benefit compared to other energy resources. Consequently, this result indicates that women with maternal instinct may not be favor of nuclear power plants compared with men in China.

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Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA.

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The ASME pressurization transient analyses are performed by the TPC/INER Transient Analysis Methodology (TITRAM). Three typical ASME pressurization transients including main steam isolation valves closure (MSIVC), load rejection (LR), and turbine trip (TT) are analyzed to evaluate the maximum vessel pressure. The conservative assumptions including failure of direct scram signals, relief valve function, and 7 of 18 safety valves, and larger initial core power and pressure are utilized for conservatism. The analysis results indicate that MSIVC event leads to largest maximum vessel pressure of 9.17 MPaG, which is below the acceptance criterion of 9.48 MPaG. The sensitivity analysis is conducted to identify the sensitive parameters associated with ASME pressurization by MSIVC. The sensitivity analysis results indicate that SV setpoint is the most sensitive parameters among the input parameters for ASME pressurization with MSIVC. It is concluded that the performance of SVs is capable of preventing the vessel pressure from reaching safety limit (9.48MPaG).

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Thermal fatigue is a safety relevant damage mechanism in pipework of nuclear power plants. A well-known simplified method for the assessment of thermal fatigue due to turbulent mixing is the so-called sinusoidal method. Temperature fluctuations in the fluid are described by a sinusoidally varying signal at the inner wall of the pipe. Because of limited information on the thermal loading conditions, this approach generally leads to overconservative results. In this paper, a new assessment method is presented, which has the potential of reducing the overconservatism of existing procedures. Artificial fluid temperature signals are generated by superposition of harmonic components with different amplitudes and frequencies. The amplitude-frequency spectrum of the components is modelled by a formula obtained from turbulence theory, whereas the phase differences are assumed to be randomly distributed. Lifetime predictions generated with the new simplified method are compared with lifetime predictions based on real fluid temperature signals, measured in an experimental setup of a mixing tee. Also, preliminary steady-state Computational Fluid Dynamics (CFD) calculations of the total power of the fluctuations are presented. The total power is needed as an input parameter for the spectrum formula in a real-life application. Solution of the transport equation for the total power was included in a CFD code and comparisons with experiments were made. The newly developed simplified method for generating the temperature signal is shown to be adequate for the investigated geometry and flow conditions, and demonstrates possibilities of reducing the conservatism of the sinusoidal method. CFD calculations of the total power show promising results, but further work is needed to develop the approach.

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Uranium crystallization system has been developed as innovative technology for advanced aqueous reprocessing of FBR spent fuel in Japan Atomic Energy Agency (JAEA). In the crystallization system, a great part of uranium is separated from dissolver liquor with high heavy metal concentration by cooling as uranyl nitrate hexahydrate (UNH) crystal. However, the purity of UNH crystal is not high, so we have discussed the application of crystal purification technology to improve the purity of UNH crystal and selected Kureha crystal purifier (KCP) as one of the desirable purifiers for UNH crystal. In the previous study, we have found that the purity of UNH crystal depends on the behavior of solid impurities in the crystallization system. In this study, the effects of grain size and density of solid impurities on the purification performance were evaluated in the UNH crystal purification tests to discuss the purification mechanism of KCP. From the results, a decontamination factor (DF) of UNH crystal with the smaller grain size of solid impurities was larger than that of the bigger one, on the other hand, the density effect was not significant. It is explained that the smaller size of solid impurities is easy to pass through between UNH crystal grains and the decontamination of the solid impurities in the purifier is promoted by pushing down them along the flow of melt.

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A numerical study on fuel rod surface temperature distributions in liquid metal cooled sub-channel flow is presented in this paper. Special interests are focused on the effects of misaligned central fuel rod in typical fuel rod cluster geometries. Numerical models based on FLUENT, a Computational Fluid Dynamics (CFD) commercial software, are developed to model steady state fuel thermal conductions, fully-developed turbulent convective flows, and local flow mixing. Numerical results show that fuel rod surface temperatures are higher for surfaces facing the narrow gap of the sub-channel. The temperature differences are mainly dependent on the fuel pitch-to-diameter ratio, flow patterns, and other flow physical properties. Situations are worse for cases where misaligned fuel rods exist in tightly packed fuel rod clusters. Even for a small amount of rod center deviation, such temperature differences can be as high as 54 K around a fuel rod whose diameter is only 4.7 mm. The situations could be very serious and might generate unnecessary thermal stresses and cause fuel failure.

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The steady-state turbulent heat transfer coefficients in a short vertical Platinum (Pt) test tube for the flow velocities (u=4.01 to 13.62 m/s), the inlet liquid temperatures (T_<in>=294.00 to 304.29 K), the inlet pressures (P_<in>=802.12 to 859.08 kPa) and the increasing heat inputs (Q_0exp(t/τ), exponential periods, τ, of 6.04 to 25.67 s) were systematically measured by an experimental water loop comprised of a multistage canned-type circulation pump with high pump head. Measurements were made on a 59.2 mm effective length and its three sections (upper, mid and lower positions), which was spot-welded four potential taps on the outer surface of a 6 mm inner diameter, a 69.6 mm heated length and a 0.4 mm thickness of the Pt test tube. The outer surface temperature distribution of the Pt test tube was also simultaneously observed by an infrared thermal imaging camera at intervals of 3 seconds. Theoretical equations for turbulent heat transfer in a circular tube of a 6 mm in diameter and a 636 mm long were numerically solved for heating of water with heated section of a 6 mm in diameter and a 70 mm long by using PHOENICS code under the same condition as the experimental one considering the temperature dependence of thermo-physical properties concerned. The surface heat flux, q, and the surface temperature, T_s, on the circular tube solved theoretically were compared with the corresponding experimental values on heat flux, q, versus the temperature difference between heater inner surface temperature and liquid bulk mean temperature, ΔT_L [=T_s-T_L, T_L=(T_<in> + T_<out>)/2], graph. The theoretical solutions of q and ΔT_L are almost in good agreement with the corresponding experimental values of q and ΔT_L with the deviations less than ±10 % for the range of ΔT_L tested here. The theoretical solutions of local surface temperature, (T_s)_z, local average liquid temperature, (T_<f,av>)_z, and local liquid pressure drop, ΔP_z, were also compared with the corresponding experimental data on (T_s)_z, (T_<f,av>)z and ΔP_z versus heated length, L, or distance from inlet of the test section, Z, graph, respectively. The theoretical solutions of local surface temperature, (T_s)z and local average liquid temperature, (T_<f,av>)_z are within ±10 % of the corresponding experimental data on (T_s)_z and (T_<f,av>)_z, although those of local liquid pressure drop, ΔP_z, become 37.6 % lower than the experimental ones. It was confirmed in this study that authors' steady-state turbulent heat transfer correlation, Eq. (1), based on the experimental data (Hata and Noda, 2008) Nu_d=0.02Re^<0.85>_dPr^<0.4>(L/d)^<-0.08>(μ_1/μ_w)^<0.14> (1) can not only describe the experimental data of steady-state turbulent heat transfer but also the theoretical solutions within ±10 % difference for the wide ranges of temperature differences between heater inner surface temperature and liquid bulk mean temperature (ΔT_L=5 to 200 K) and flow velocity (u=4.01 to 13.62 m/s).

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Internal events as well as external events as initiators that may entail potential consequence have to be considered when we discuss the comprehensive risk of Nuclear Power Plant (NPP). Two or more NPPs that are located at a multi-unit site or close to each other could fail simultaneously at seismic event as one of external events, depending on the degree of seismic ground motion and its impact to units of the site. The seismometers at the Kashiwazaki-Kariwa NPPs recorded different time-histories of ground motion due mainly to the folding structure beneath the free field rock surface and to the installation condition of the embedded buildings, when the Niigata-ken Chuetsu-oki earthquake in July, 2007, impacted on the Site significantly. This suggested the different input ground motions on all plants at the site. Then the method is being enhanced to be applicable even if different input ground motions to respective plants at the site may happen to be generated. The interface of seismic hazard evaluation method, structure, system and component (SSC) fragility evaluation method, and accident sequence evaluation method were carefully and systematically considered. The enhanced method was preliminary applied to a hypothetical site where a highly-aseismatic and a lesser-aseismatic designed plant were located, and the applicability of the developed method to LWRs was discussed, applying typical correlation factor to the analysis model. Some insights on accident sequences for both plants failure and for either plant failure ("plant failure" is defined as "serious damage to the reactor core" in this paper.) were obtained from the analysis results. The developed method is robust to delineate the risk profile of multiple NPPs close to each other, being impacted by a strong seismic motion.

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Understanding the complex dynamic behavior of Underwater Fuel Storage Racks (UFSR) is of prime importance for the safety of nuclear plant facilities. This equipment is submitted to a very strong fluid-structure coupling because of the small layers of fluid surrounding each of its modules. Various studies have been made on this subject. Moreira & Antunes developed a simplified linearized model for the fluid-structure interactions between the modules composing the rack, they estimate the hydrodynamic masses and include a dissipative effect. However, the pressure dissipation in the fluid bifurcation (at the corner of each module) is neglected. This can lead to an overestimation of the coupling between adjacent modules. In this article, we carry on with the average flow velocity method developed by Moreira & Antunes. The focus is on the determination of accurate hydrodynamic masses by including the pressure drops at the corners of each module. This pressure drop is directly linked to the flow distribution at each intersection of the fluid. We have developed an analytical method to determine this flow distribution and then obtain the hydrodynamic added masses.

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In the nuclear energy industry, most of the major components are anchored to the civil works using numerous types of supports devices. These anchorages are big issues of the nuclear plant design: the implantation of the components has to be fixed definitely, stress concentration in the surroundings of the anchorage, and for immersed structure, possible loss of the impermeability. Thereby, under certain safety regulations, some structures lay directly on the ground. This is the case for in air or underwater structure, such as fuel storage racks. This solution gives more flexibility in the use of the components and a decrease of the stress. However, one has to evaluate precisely the behavior of this sliding structure, and in particular, the cumulated sliding displacement during a seismic event in order to prevent any impact with other components. During a seismic event, the unanchored structure can slide, rotate and tilt. The aim of this paper is to present analytical solutions to estimate the sliding amplitudes of different simplified systems which represent a given dynamic behavior. These simplified models are: a sliding mass and a complex sliding structure defined by its eigenmodes. Each simplified system corresponds to a different set of assumptions made on the flexibility of the structure. Two analytical solutions are presented in this article: single sliding mass and a vertical sliding beam. In each model, the fluid-structure interaction between the immersed body and the pool is modeled as hydrodynamic masses. The sliding is represented by Coulomb friction. The seismic loading can be any 3D seismic accelerogram. The analytical solutions are obtained considering the different phases of the movement and the continuity between each phase. The results are then compared to the values computed with the commercial Finite Element package AN SY S^<TM>. The analytical curves show a good fit of the computational results.

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The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700℃. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750℃ and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into a initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223℃. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 ℃. It was a little higher than 1230℃--the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620℃, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

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This paper describes a simple method of a flow rate monitoring using Ultrasonic Velocity Profile method (UVP), in the case of the disturbed flow just after the double elbows and the large pipe at high Reynolds number. This method utilizes the linearity between the flow rate and the velocity at the pipe center. And the experimental results show that this method can monitor the flow rate changes accurately at the high Reynolds number, around 4 million.

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In order to reveal the deactivation mechanism of the hydrogen recombination catalyst of off-gas treatmen system, we investigate by using multi-level computational chemistry simulation methods. The recombiner apparatus is modeled by the numerical mesh system in the axial coordinates, and unsteady, advection and reaction rate equations are solved by using a finite difference method. The chemical reactions are formulated to represent adsorption-desorption of hydrogen and oxygen on Pt catalyst, and time developments of the coverage factors of Pt are solved numerically. The computational simulations successfully reproduce the very similar behaviors observed by experiments, such as increasing of the inversion rates of H_2 to H_2O, the temperatures distributions along the flow direction, dependencies of experimental condition, and so on. Thus Pt poisoning is considered to cause the deactivation of the hydrogen recombination catalyst. To clarify the poisoning mechanism, the molecular level simulation is applied to the system of Pt on boehmite attacked by a cyclic siloxane which has been detected by experiments and considered as one of poisoning spices. The simulation shows ring-opening reaction of the cyclic siloxane on Pt, then attachment of two ends of the chain-like siloxane to Pt and boehmite, respectively, and that finally the recombination reaction is prevented. This may be the first study to find out the detailed dynamical mechanism of hydrogen recombination catalyst poisoning with cyclic siloxane.

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Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO_2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (〜500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified.

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In order to clarify the effect of bottom configuration on pool film boiling heat transfer from a vertical finite-length cylinder with a length comparable to the diameter, three kinds of silver cylinder with flat, hemispherical and conical bottoms were tested by quenching in saturated and subcooled water. For saturated water, the wall heat flux averaged over the entire surface of a finite-length cylinder takes a higher value for the cylinder with a conical bottom than those for the other two types as the wall superheat decreases. For highly subcooled bulk water, the average wall heat flux for the cylinder with a flat bottom becomes larger than those for the cylinders with hemispherical and conical bottoms. This is attributed that the cylinder with a flat bottom has quite a thin vapor film at the lower end of the vertical cylinder due to the edge effect. The wall superheat (ΔT_<min>) corresponding to the vapor-film-collapse is constant at about 133K for three kinds of test cylinder in saturated bulk water. However, for highly subcooled bulk water(ΔT_<sub> =20K), ΔT_<min> with the cylinder with a flat bottom surface is about 60K larger than those with hemispherical and conical bottom surfaces.

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A study to develop a parameter database for Probabilistic Safety Assessment (PSA) for the application of risk information on plant operation and maintenance activity is important because the transparency, consistency, and traceability of parameters are needed to explanation adequacy of the evaluation to third parties. Application of risk information for the plant operation and maintenance activity, equipment reliability data, human error rate, and 5 factors of "five-factor formula" for estimation of the amount of radioactive material discharge (source term) are key inputs. As a part of the infrastructure development for the risk information application, we developed the integrated parameter database, "R-POD" (Rokkasho reprocessing Plant Omnibus parameter Database) on the trial basis for the PSA of the Rokkasho Reprocessing Plant. This database consists primarily of the following 3 parts, 1) an equipment reliability database, 2) a five-factor formula database, and 3) a human reliability database. The underpinning for explaining the validity of the risk assessment can be improved by developing this database. Furthermore, this database is an important tool for the application of risk information, because it provides updated data by incorporating the accumulated operation experiences of the Rokkasho reprocessing plant.

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This paper presents the mechanisms of steam generator fouling, various methodologies for mitigation and the AREVA C^3 (Customised Chemical Cleaning) Concept as a solution to this problem in operating power plants. It also covers the principle of preventative versus curative chemical cleaning and innovative waste treatments options to manage the chemical cleaning liquid waste. Finally it presents the AREVA field experience covering these aspects.

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Primary separators in pressurized water reactors (PWRs), cyclone separators, separate most water from vapor-water two-phase mixture, which is important to safety and economics of nuclear power plants. To improve the performance of cyclone separators, we propose new structures including inlet impeller, porosity and conical angle of the chamber wall. Aiming at the new structures, this paper analyzes separation efficiency and pressure drop experimentally in air-water two-phase flows. Results show that separation efficiency and pressure drop have always opposite trends but with similar transition boundaries under all the new structures. Each of the structural parameters has effect on separator performance, but none of the trends is monotonical as phase velocity increasing. The comprehensive performance is determined by separator structure as well as flow patterns. From the investigations in this paper, it is promising to deeply understand the separation mechanism and further to provide enough data to design large-scaled separators of CAP1400.

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The spent fuel is kept and cooled in the Spent Fuel Pool (SFP). Therefore, there is some risk of the spent fuel damage in the SFP as well as that of the reactor core damage when abnormal events such as loss of cooling accident occur. It is thought that the risk of spent fuel damage in the SFP is lower than that of reactor core, because the decay heat level generated by the spent fuel in the SFP is lower than that of reactor core. However, there is a possibility that the fuel damage risk of the SFP cannot be ignored in comparison with the core damage during the refueling outage, because the fuel that has high decay heat level is transferred to the SFP during the exchange of fuel and the mitigation system for the SFP is limited in comparison with the reactor core. Therefore, in this study, risk monitor to evaluate the SFP fuel damage risk during the plant shutdown conditions was constructed for a BWR plant, and the SFP fuel damage frequency was evaluated. By utilizing the risk monitor, the useful risk information for the safety management activity can be supplied during the refueling outage.

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