In this study, wireless electrical resistance detector is developed as first step in order to develop electrical resistance tomography (ERT) that are attached wireless communication, and miniaturized. And the particle volume fraction measurement results appropriateness is qualitatively examined. The real-time particle volume fraction measurement is essential for centrifuges, because rotational velocity and supply should be controlled based on the results in order to obtain the effective separation, shorten process time and save energy. However, a technique for the particle volume fraction measurement in centrifuges has not existed yet. In other words, the real-time particle volume fraction measurement in centrifuges becomes innovative technologies. The experiment device reproduces centrifugation in two-phase using particle and salt solution as measuring object. The particle concentration is measured changing rotational velocity, supply and measurement section position. The measured concentration changes coincide with anticipated tendency of concentration changes. Therefore the particle volume fraction measurement results appropriateness are qualitatively indicated.

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This study aims to investigate the performances of a near surface repository subject to fuel burning occurring simultaneously or subsequently to a large commercial aircraft impact. Specifically the therm al effects cause d by a Boeing-747 crushing (considered like "beyond design basis accident") are studied. An important part of this study is t he analysis of the possible (thermo-mechanical) degradation effects, as dehydration, degasification, pressurization, etc. that the concrete may undergo, particularly in the case of prolonged fire, and of the resistance of structure itself in this condition. Conservative assumptions and restrictions have been made with regard to the fire scenario, the maximum temperature of which is calculated on the basis of the fuel airplane amount, the normal impact, the variation of the material properties along with the temperature as well the damaging phenomena of concrete. The airplane impact load, calculated with the Riera approach, and the maximum temperature, reached during the fuel combustion, are u sed as in put (boundary condition) in the numerical simulations performed by MARC[○!C] code. The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the local penetration of wall. To reduce the computational cost, the analyses have been made only on a half part of the structure, highlighting the dominance of thermal effects. Despite the ongoing concrete degradation phenomena, the overall integrity of the repository seemed to be guaranteed as well as the containment and th e confinement of radioactive waste.

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EDF R&D is seeking to access the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this paper, we use balance (between primary and secondary systems) to improve the simplified semi-empirical 0D Model for RCS, using a "fitting" method for the bypass coefficient related to the part of the flow which is not in contact with the fuel assemblies in each quarter of a four-looped core. Thus, we get a model that can be used to generate state vectors containing most of primary parameters values. Then we describe how to use this model to define a Data Assimilation Approach, by generating random parameters and thus constructing a sample of random state vectors, from which the background vector and related error covariance matrix can be deduced. Finally, we apply our method with a focus on Normalized Integrated Neutron Powers, using twin experiments to evaluate its performances. Overall, calibrating the random parameter generator for Neutron Mapping on real data does improve the algorithm performance, though only moderately.

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The High Temperature gas cooled Reactor Pebble bed Module (HTR-PM) is in design and construction process in China, which is considered as one of the candidates for the Gen-IV nuclear power plant, and has advantage of inherent safety, avoiding nuclear proliferation, high temperature industry heat production and so on. The sphere fuel element is used in HTR-PM. The fuel particles are spread in the fuel element, and the sphere element's diameter is 60mm after oppression. After discharged from the HTR reactor core, the spent fuel element would be transferred into the spent fuel canister. The spent fuel canister would be stored in the spent fuel storage well after fully loaded. In the process of the spent fuel storage and operation, it is required to ensure the operation safety, subcritical, radiation shielding safety and residual heat removal safety. In order to decrease the price of the spent fuel canister, the canister was designed as a thin shell vessel, which has weak radiation shielding function, and cannot fulfill safety requirement of radiation shielding, so it is required to research and design a set of devices, which could provide enough radiation shielding for the spent fuel canister, and the device could also transfer the spent fuel canister safely and reliably, and then the safety of the operation staff and the facility could be ensured. The concrete shielding well lid is set on top of the storage well, when the spent fuel canister is needed to put into the storage well, the well lid would be taken out from its mounting position. In the operation process of the spent fuel canister and the concrete shielding well lid, the ground crane with accurate positioning function is required to position the spent fuel canister and the concrete shielding well lid to the operating position. The main components of the ground crane system includes: Crane bridge, shielding cask, neutron shielding boron barrel, canister hoisting mechanism, well lid hoisting mechanism, bottom plate opening mechanism, shielding strip mechanism, residual removal blowers, butterfly valves, ground rails, cable sliding bridge, encoder positioning scale and so on. The ground crane could satisfy the accurate positioning and safe operation of the spent fuel canister, and could ensure the operating reliability of the spent fuel canister and the concrete shielding well lid in HTR-PM operational period.

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Natural circulation has been adopted as the way of passive heat removal by some evolutionary and innovative reactors. Sub-cooled boiling is of great importance to the stability of a natural circulation system. An analytical model of two-phase natural circulation system is developed and sub-cooled boiling is taken into consideration. Numerical simulation of both steady state and transient state natural circulation are carried out. Besides, flow oscillation in Nuclear Power Simulation System (NPSS) is studied. One dimension homogeneous flow model is used to solve the governing equations. And lumped parameter model is adopted to establish the wall energy conservation equation. The test section is divided into four parts: single phase liquid, partial sub-cooled boiling, fully developed sub-cooled boiling and saturated boiling. Different heat transfer correlations are employed to calculate the wall temperature and heat transfer coefficient. Fluid temperature, void fraction and wall temperature are obtained by a MATLAB code. The effects of heating power, inlet temperature and system pressure on flow rate are investigated by the steady state code. The transient analysis indicates that flow oscillation occurs when sub-cooled boiling takes place at the exit of the test section. The flow oscillation is enhanced when heating power or inlet temperature is increased before a certain value, the flow oscillation tends to disappear when heating power or inlet temperature is further increased. Increasing system pressure has a stabilizing effect on the flow oscillation.

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Among Gen IV projects for future nuclear power plants, Lead Fast Reactors (LFR) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant-safety and waste management. The novelty of the matter causes some open issues about coolant chemical aspect, structural aspects, monitoring instrumentation, etc. Particularly hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e. source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal-hydraulics and chemical conditions for LFR core environment will be assumed, as the neutron flux will be studied extensively by means of the Monte Carlo transport code MCNPX. The core coolant's high temperature drastically reduces the candidate instrumentation, because only some kind of fission chambers and Self Powered Neutron Detectors can be operated in such an environment. This work aims to evaluate the capabilities of the available instrumentation (usually designed for Sodium Fast Reactors, SFRs) when exposed to the neutron spectrum derived from ALFRED, a pool-type small-power LFR project to demonstrate the feasibility of this technology into the European framework.. This paper shows that such instruments do follow the power evolution, but they are not completely suitable to detect the whole range of reactor power. Some improvements are then possible in order to increase the signal-to-noise ratio, by optimizing each instrument in the range of reactor power, such to get the best solution. Some new detector designs are here proposed, and the possibilities for prototyping and testing by means of a fast reactor investigated.

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Helium circulator is the largest rotatable component of the 10 MW high temperature gas-cooled reactor (HTR-10). There are some radioactive contamination risk on the environment and personnel during the process of maintaining the helium circulator after long term power operation. Radiation dose monitoring, contamination source control, atmosphere isolation, process control, maintenance personnel training, experience feedback and etc. are employed during the maintaining process. Radioactive pollution risk on the environment and personnel are well controlled.

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Separation process is an integral portion of the missions of three offices within the U.S. Department of Energy (DOE), Office of Environmental Management (EM), Office of Nuclear Energy (NE), and the National Nuclear Security Administration (NNSA). The specific needs of these three offices vary widely, as do their operating timelines. Multiple responsibilities are required to recover nuclear materials for their direct use, future reuse or disposition: cleanup of legacy defense-related nuclear facilities, research on advanced commercial fuel cycles, and treatment and disposal of radioactive wastes including spent nuclear fuel. The scales at which separations are applied also vary significantly, from analytical separations at the microliter scale required for some NNSA applications, to hundreds of metric tons per year recycling of commercial fuel ultimately required by DOE-NE, to the 3.4 × 10^5 m^3 of radioactive wastes stored in tanks at the Hanford site and Savannah River Site (SRS) that DOE-EM must process and immobilize. Similarly, the concentrations involved range from ultra-trace level detection of radionuclides to hundreds of grams per liter, and the complexity of the solution matrices can vary from simple dilute solutions (groundwater) to highly concentrated electrolyte mixtures (salts, acids, bases, complexants, and elements across the periodic table). Despite this disparity, common themes exist that tie together the missions of DOE-EM, DOE-NE, and NNSA. Over the past three years, efforts have been ongoing in DOE to coordinate planning for research, development and deployment (RD&D) of nuclear separations technologies. It has been recognized that nuclear material recovery is a topic in which DOE has an important leadership role. Based on that recognition, DOE organized three workshops and brought together subject matter experts from national laboratories, academia, industry, the international community, and government organizations to discuss the role of nuclear separations, technical subjects (chemistry and speciation, design of molecules, scale-up of processes, and interface issues), opportunities for partnerships among the attending organizations, and the advantages that might obtain to the development of a nuclear separations center of knowledge. The extensive record of the workshops has been summarized in a consolidated report which will be issued in November 2014. Each workshop established a current state of knowledge or technology and evaluated the needs to reach near-term (0 to 5 years), mid-term (6 to 15 years), and long-term (> 15 years) objectives. Details and examples of these findings will be presented in our contribution.

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In July 2002, molten salt reactor was selected as one of the six type reactors of the fourth-generation nuclear energy systems at the Generation IV International Forum (GIF). ORNL completed the conceptional design of Molten Salt Breeder Reactor (MSBR) in 1970s. In present study, a new bubble generator was designed for an experimental TMSR with an electricity power about 10MW and was tested in an experimental loop using water as working fluid. A high speed camera was employed to record the distribution and detailed breakup process of bubbles in the generator. Meanwhile the mechanism of bubble breakup was tried to be illustrated. The experimental results showed that the bubbles had average diameters of 0.5〜0.6mm and were uniformly dispersed in the main stream when the void fraction was in the range of 0.2%〜0.3%, satisfying the requirement of a molten salt reactor. On the other hand, sudden deceleration of the bubbles in the diverging portion of the bubble generator led to a sharp velocity gradient in the vicinity of the bubble, resulting in greater shear stress on the surfaces of the bubbles.

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Microbubble emission boiling (MEB) occurs in some cases of subcooled boiling, accompanied by the emission of microbubbles from coalescing bubbles on heating surface and with much higher heat flux than that of conventional nucleate boiling. MEB was studied firstly by Inada et al. in 1981 and has attracted a great deal attention in recent years due to its extremely high heat dissipation capacity and potential application in thermal engineering. Investigations were conducted concerning MEB on a small heating surface as well as the condensation process of a single vapor bubble in a subcooled pool. Characteristics of MEB and the principles dominating the breakup of vapor bubbles on the heating surface was illustrated. It is speculated that the Marangoni convection induced surface wave might result in that the vapor film on the heating surface cannot sustain under a certain condition, leading to the occurrence of MEB.

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Japan Atomic Energy Agency initiated a High Temperature Engineering Test Reactor (HTTR) demonstration program in accordance with recommendations of a task force established by Ministry of Education, Culture, Sports, Science and Technology according to the Strategic Energy Plan as of April 2014. The demonstration program is designed to complete helium gas turbine and hydrogen production system technologies aiming at commercial plant deployment in 2030s. The program begins with coupling a helium gas turbine in the secondary loop of the HTTR and expands by adding the H_2 plant to a tertiary loop to enable hydrogen cogeneration. Safety standards for coupling the helium gas turbine and H_2 plant to the nuclear reactor will be established through safety review in licensing. A system design and its control method are planned to be validated with a series of test operations using the HTTR-GT/H_2 plant. This paper explains the outline of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950°C reactor outlet temperature for production of power and hydrogen as recommended by the task force. Commercial deployment strategy including a development plan for the helium gas turbine is also presented.

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The vitrified high-level radioactive waste from reprocessing spent fuels is planned to be disposed of within a stable geological environment at least 300m below ground surface. The evaluation of the individual dose rate caused by the waste to the biosphere is the primary issue for evaluating the safety of geological disposal. The radionuclide dissolution behavior from the vitrified waste is an especially critical processes for calculating the dose rate. The dissolution model proposed by Japan Nuclear Fuel Cycle (JNC, now JAEA) in the second progress report (H12 report)(JNC, 1999) has been used in Japan. However, various dissolution models have been proposed by researchers and implementers since the H12 report. This paper presents the recent progress of the glass matrix dissolution research and identifies four issues of the dissolution behavior of radionuclides. The four issues are as follows: A) "Identification of alteration rate acceleration mechanism and conditions", B) "Investigation into the protective nature of glass alteration layers", C) "Model development for potential disposal environments", and D) "Investigation into the effective surface area".

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Best-Estimate (BE) calculation is more broadly used in nuclear industries and licensing process to reduce the significant conservatism for evaluating Loss-of-Coolant-Accident (LOCA). A key feature of BE evaluation requires to quantify the uncertainty of the calculations. However, the quantification of input uncertainties, especially related to physical models in thermal-hydraulic code, is difficult on the determination and the justification of the uncertainty range associated with each uncertainty parameter. So, the quantification is often performed mainly by subjective expert judgment or from reference documents. More mathematical methods are needed to determine the uncertainty ranges, objectively. CIRCE (Calcul des Incertitudes Relatives aux Correlations Elementaires) is an inverse quantification method of the uncertainties of the physical models using the Expectation-Maximization algorithm, which has been developed by CEA, FRANCE. In this study, CIRCE method was used to quantify the ranges of most influential uncertainty parameters in MARS-KS thermal-hydraulic code. Six FEBA reflooding tests with the different initial and boundary conditions during reflood phase of L OCA, were chosen to quantify uncertainty ranges in the CIRCE method. Only two uncertainty parameters such as dry/wet wall criteria and interfacial heat transfer of drop-steam are considered due to the limitation of this me thod. For FEBA experiment, measured cladding temperatures, the calculated values at the corresponding measured locations and their derivatives of code response with respect to each uncertainty parameter are used as inputs of CIRCE. The CIRCE method gives uncertainty distribution with the mean and standard deviation of each uncertainty parameter. With the determined uncertainty ranges, 124 calculations for each six FEBA test were performed to check whether the experimental responses such as the cladding temperature or pressure drop are inside the ranges of calculated uncertainty bounds. Also, the extrapolability of the CIRCE results was confirmed by 200 case uncertainty analysis for six reflooding PERICLES tests carried out to investigate 2-dimensional (2D) effects. However, because the uncertainty ranges in CIRCE method may be highly dependent on the selection of the responses and their derivatives, the different set of responses are chosen and evaluated to study this user's effect in the CIRCE applications.

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A sodium-cooled fast breeder reactor (FBR) is now in a developing stage in Japan. A shell-and-tube type once-through heat exchanger is used to generate steam in the design. Low pressure hot sodium flows in the shell side and high pressure water flows in the tube side. Water is heated to generate steam in the tubes. It has been anticipated that a pin hole is formed on the tube wall and high pressure water or water-steam two-phase mixture blows out from the hole. When the high pressure water flows out from the hole of tube, flashing occurs and a high speed steam jet is formed in the sodium coolant. Small droplets of sodium are torn off from the sodium surface and entrained into the high speed steam jet. Water-sodium chemical reaction results an increase the entrained liquid droplet temperature. The hot and high speed sodium entrained liquid droplets attack the wall of a neighboring tube to damage and create a hole on the tube wall, which may result in the chain tube wall damage. The steam jet surface behavior, the sodium surface behavior, and the entrainment behavior are key issue to consider this occurrence. In the present paper, a two dimensional air jet was blown out vertically upward from a rectangular hole at the bottom of a thin rectangular test vessel into stagnant water of 80℃ in the vessel. This obtained results were compared with the former experiment results which obtained by using same vessel, on the other hand, water was used in normal temperature. The test vessel was 5 mm deep, 270 mm wide and 300 mm high. The stagnant depth of water in the vessel was 100 mm. The rectangular hole of air inlet was 1 mm × 5 mm. The air pressure at the upstream of the hole was at 115 〜 250 kPa. The air velocity at the outlet of the hole was 130 〜 300 m/s. The entrainment process and the air jet/water surface behavior were recorded with a high speed video camera. The trajectory of entrained liquid droplets and the velocity variation after entrained were examined from recorded pictures.

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In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code α-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.

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The pressure and temperature of water in the steam generator in high-temperature gas-cooled reactor power plant can be increased to supercritical as the HTR-PM reactor, which was built by the Institute of Nuclear and New Energy Technology of Tsinghua University in China, provides helium up to 750 ℃ and the design and fabrication of supercritical unit in China are sophisticated. Ten 252.9 MW(th) supercritical steam generators, which are exactly the same, including a main heating section and a reheating section, are designed to work with ten HTR-PM reactors and one supercritical turbine unit, and the thermal and hydraulic characteristics are analyzed. The supercritical steam generator is composed of 19 5-layer helically coiled tube components in both these two sections. The tube size, the component diameter and the tube sheet are quite similar with HTR-PM subcritical steam generator, while the total tube height is 6 m higher, so the damping devices might be necessary to be installed at the bottom of the supercritical steam generator to avoid swaying and vibration. The heat transfer coefficient of supercritical pressure water has a peak in the main heating section without any abrupt variation of the inside and outside wall temperatures.

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A speed-up method of a neutron kinetic model was studied. The neutron kinetic model is a on e-energy group three-dimensional (3D) neutron kinetic model that uses a finite-differential method. External neutron sources can be considered by the 3D neutron kinetic model to simulate the reactor start-up of initially subcritical condition. To speed up the neutron calculation, we have applied the odd-even method, also known as the checkerboard ordering method, to the time-dependent 3D neutron diffusion calculation. The number of iterations needed for convergence was decreased using the method. The odd-even method was validated by a simulation of the advanced boiling water reactor (ABWR) transient e vent of the main steam isolation valve closure. The num ber of neutron diffusion calculation iterations was decreased to 47% of the number needed by a general method. The calculation time of a 100s simulated period was 67s using an ordinary PC.

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In the field of severe accident, the description of corium progression events is mainly carried out by using integral calculation codes. However, these tools are usually based on bounding assumptions because of high complexity of phenomena. The limitations associated with bounding situations ([Seiler et al., 2014] e.g. steady state situations and instantaneous whole core relocation in the lower head) led CEA to develop an alternative approach in order to improve the phenomenological description of melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core melt-down to the molten core concrete interaction. This phenomenological approach is based on available data (including learnings from TMI2), on physical models and knowledge about the corium behavior. It provides emerging trends and best estimate intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three dimensional, all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the unit 1 of the Fukushima Dai-ichi Nuclear Power Plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on available results derived from codes like MELCOR calculation code [Gauntt et al, 2012]. The core degradation could then follow different ways: - Axial progression of the debris and the molten fuel through the lower support plate; - Lateral progression of the molten fuel through the shroud.

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In the late phase of a severe loss-of-coolant accident (LOCA) in a light water reactor (LWR), carbon monoxide (CO) may be generated inside the containment due to molten corium concrete interaction (MCCI). As a component of the accident atmosphere, CO will interact with passive autocatalytic recombiners (PARs) which are installed inside LWR containments for hydrogen (H_2) removal. Depending on the boundary conditions, CO may either react with oxygen to carbon dioxide (CO_2) or act as catalyst poison, reducing the catalyst activity and hence the hydrogen conversion efficiency. An experimental test program performed in co-operation between Forschungszentrum Julich and RWTH Aachen University investigates these aspects aiming at providing data for model development for advanced severe accident analyses. In the first test series to be presented, the parallel catalytic reaction of H_2 and CO on the catalyst surface has been studied, i.e. the hydrogen recombination reaction was started before CO was injected. The test results show that under the given conditions the conversion of CO into CO_2 has no negative impact on the parallel hydrogen conversion. Additionally, the efficiency of the CO reaction in terms of molar rates is significantly smaller than the corresponding H_2 conversion efficiency. Due to the exothermal reaction, the parallel CO conversion may also have an impact on the possible ignition of the flammable gases at hot PAR surfaces. A second test series aimed at investigating the influence of low oxygen (O_2) concentrations on the parallel H_2/CO reaction. Basically the same experiments as in the first test series were performed, only with stepwise decreasing O_2 content until the break-up of the catalytic reaction. Main result of this test series was the observation of different break-up mechanisms dependent on the initial gas concentrations. A spontaneous break-up of the catalytic reaction took place at tests with low H_2/CO concentrations (i.e. 2 vol.%). In these tests, no loss of efficiency during O_2 reduction was observed. All tests with higher H_2/CO concentrations revealed a stepwise break-up of the catalytic reaction accompanied by significant loss of conversion efficiency for both species during O_2 reduction. This paper provides an overview of the results of both test series including first modelling approaches.

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In the high temperature gas-cooled reactors, the dust safety issues have been received wide attentions, especially in the pebble bed reactors. Dust complicated the prediction of fission and activation products transport since it presented a parallel transport path and was thought to be an important source term during the depressurization accident scenario. A modification in the helium purification system of HTR-10 has been finished which built a new experimental system to sample the radioactive graphite dust in the primary loop after the restart of HTR-10. During the modification, some pipes, a valve, and a dust filter of the primary loop were cut off and used for further measurement. Besides, an inspection on the helium circulator of HTR-10 has been implemented. Some cotton balls for sampling the radioactive graphite dust during the inspection have been obtained. The measurement of γ dose rates of all samples indicates the existence of fission and activation products. The inspection on site with endoscope inside the old dust filter shows the deposition morphology of radioactive graphite dust. Co-60 and Cs-137 have been determined from a preliminary γ spectrometry measurement. It can supply important information for determination of the source term, decontamination and decommissioning work, and radiation protection during maintenance, and promote the study on the radioactive graphite dust in the pebble bed reactors.

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Research on radionuclide accumulation is significant for analyzing the source term in a severe accident. The previous method utilizing the point depletion computer code to analyze radionuclide accumulation assumes that thermal power density and burnup distribution are uniform. And the different fuel enrichment in different region is also ignored. However, the thermal power density and the burnup distribution are actually not uniform during AP1000 reactor operation. Therefore, the results obtained by the previous method needs further verification. In this paper, a fine structure model was established utilizing CASMO-SIMULATE for AP1000 reactor to calculate the radial and axial thermal power density distribution. Based on this distribution and the different fuel enrichment in different region, the reactor core was divided into ten levels in axial orientation and four rings in radial orientation. A particular level and a particular ring define a core cell. BURNUP code was used to analyze the radionuclide accumulation in each cell. The radionuclides are grouped into three segments as follows: activation products, actinides and fission products. And, each group in each cell was analyzed. Then, radionuclide accumulation characters in entire core were obtained. The result was compared with the result obtained by the previous method. And the influence to the results of each effect factor, fuel concentration, power distribution and burnup, was analyzed. The radionuclide decay characters after reactor emergency shutdown were also obtained.

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As a continuation of a previous theoretical study, we propose here preliminary experimental results about the application of the TDR technique to a coaxial reflectometric sensor used as level monitoring device. This kind of sensor may offer some interesting features for the design of nuclear facilities control systems and oil and gas plants. In order to experimentally verify the theoretical results, the sensor is used as a level control for a hydraulic loop system, firstly in a static condition, then its usage planned under a dynamic variation of the liquid level with different transient times. As a matter of fact this hydraulic loop system, yet simple, can offer short/medium-term transient level dynamics requiring fast sensor response, on-line measurement and some non-linearity features that are crucial for its control. In this paper, a second-order nonlinear ODE governing the system dynamics is theoretically analyzed, and the hydraulic loop prepared for its experimental observation through the use of a coaxial reflectometric sensor and a novel kind of pump with very low-pulsatility flux suitable for this study, managed by an advanced level detection algorithm.

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In the Fugen Nuclear Power Plant, we are going to conduct appropriate classification of the waste according to the contamination level of the material of the plant, to reduce the amount of radioactive waste and to promote dismantling work rationally and efficiently. For this reason, we are going to apply the clearance system to the dismantled material generated from dismantling work of the turbine system, and to reduce the radioactive waste amount as much as possible. In order to operate the clearance system properly, the target nuclides need to be selected accurately, and the evaluation method of radioactive concentration should be established. The assessment was conducted as follows. (1)Selection of the target nuclides for evaluation Ten target nuclides for the clearance evaluation were selected such as H-3, Mn-54, Co-60, etc. based on the investigation on both activation contamination and secondary contamination with consideration for the plant characteristics. (2)Establishment of the evaluation method of radioactive concentration The evaluation method of the radioactive concentration of the target nuclides was verified properly based on the analysis data obtained by sampling from the contaminated components in Fugen NPP.

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Silicon carbide (SiC) is a possible structural material for application with accident-tolerant fuel for light water reactors, which is expected to decrease heat and hydrogen generation resulting from oxidation reaction with high-temperature steam. In addition to the accident tolerance, SiC is also advantageous for fuel economy because of its small neutron absorption cross-section. We had already reported critical experiments to validate nuclear data libraries. In the present work, we performed sensitivity analyses about the critical experiments for cross-sections related to SiC with a continuous-energy Monte Carlo transportation code. Reactivity worth is defined as the difference of reactivity between a test case with SiC sample rods or aluminum sample rods and a reference case with air rods. Three core configuration experiments, each with a different neutron energy spectrum in sample rod position, were conducted. In the soft spectrum configuration, calculation value of reactivity worth shows good agreement with experimental result. In hard spectrum configurations, however, there are some discrepancies between experimental and calculation results for the reactivity worth of SiC when using air as reference. It was suggested that the discrepancy is mainly affected by thermal neutron absorption cross-section 28Si(n,γ). In the hard spectrum configuration, fast neutron elastic scattering cross-section of SiC is also sensitive to the reactivity worth to the same extent as absorption cross-section.

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Thermal fragmentation process plays a key role in Fuel Coolant Interaction (FCI) during NPP's severe accidents, which significantly affects the heat transfer and determines the ratio of heat transferred to mechanical energy. Although various thermal fragmentation models have been raised, the phenomenon is not well understood due to its complicated process. Unstable film boiling is one of the mechanisms that lead to thermal fragmentation of the melt. In this paper, thermal fragmentation process induced by unstable film boiling condition is discussed based on theoretical analysis, including a momentum equation for vapor film dynamics, an energy equation for each phase involved and some appropriate boundary conditions. The effects of the initial melt temperature, coolant temperature and ambient pressure on thermal fragmentation process are also investigated. In order to evaluate this fragmentation model, a set of experiments on typical simulant materials are introduced, which give the fragmentation time and the evolution of mixture region. The evaluation shows that the main results calculated from the model are consistent with the experimental data and reflect the fragmentation process well.

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An experiment on accident management (AM) measures during a PWR station blackout (SBO) transient with the TMLB' scenario and leakage from primary coolant pump seals was conducted using the ROSA/large scale test facility (LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The AM measures considered in this study are steam generator (SG) secondary-side depressurization by fully opening safety valves (SVs) in both SGs and primary-side depressurization by fully opening SV in pressurizer with the start of core uncovery as well as coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed that the decrease was accelerated in the primary pressure when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code well predicted the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of the primary pressure and SG U-tube collapsed liquid level. Influences of the pump seal leakage onto major phenomena until the start of the AM measures were investigated through comparison with results of the authors' previous test on the SBO (TMLB') transient without the pump seal leakage.

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In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R&D risks.

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Analyses of severe accidents for sodium-cooled fast reactors have shown that by assuming pessimistic conditions the accident might proceed into a transition phase where a large whole-core-scale pool containing sufficient fuel to exceed prompt criticality by fuel compaction might be formed. Local fuel-coolant interaction (FCI) in the pool is regarded as one of the probable initiators that could lead to such compactive fluid motions. To enhance the evaluation of severe accidents, a three-step research plan has been determined. In Phase 1, a series of simulated experiments, which covers a variety of conditions including much difference in water volume, melt temperature, water subcooling and water release site, was conducted by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy, while in Phase 2, the interaction characteristics (esp. the pressure buildup) was investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It was recognized that the SIMMER-III code, with its existing heat-transfer models, can reasonably reproduce the experimental evidence observed, thereby greatly stimulating us to initiate the Phase 3, in which further numerical analyses using reactor materials are planned. In this work, based on the latest calculations, it is confirmable that similar to the observations in Phases 1 and 2, for a given melt and sodium temperature within the non-film boiling condition, as the volume of sodium entrapped within the pool increases, a limited pressure-buildup is achievable. In addition, the performed analyses also suggest that despite of a comparatively larger temperature range of molten-fuel and sodium possibly varied during reactor accident progression, the isolation effect of vapor bubbles generated at the melt-sodium interface seems to be the unique dominant mechanism that leads to such limited pressurization. Knowledge and fundamental data from this work might be utilized for future empirical-approach studies (e.g. a lookup table for estimating the critical coolant volume required for achieving the saturated pressurization at varied melt and coolant temperatures).

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In this paper, design study and evaluation related with safety design criteria (SDC) and safety design guideline (SDG) on the balance of plant (BOP) of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system, requirements and relation with safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.

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When pressurized water or vapor leaks from a failed heat transfer tube in a steam generator of sodium-cooled fast reactors, a high-velocity, high-temperature jet with sodium-water chemical reaction may cause wastage on the adjacent tubes. For assessment of the wastage environment, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code was investigated through the analysis of the experiment on water vapor discharging in liquid sodium under actual condition of the steam generator. The computational domain consists of the cylindrical vessel and the simulated two horizontal heat transfer tubes. The cylindrical vessel was initially filled with liquid sodium. Pressurized water vapor goes into the sodium pool vertically upward from the discharging nozzle located at the center of one of the two tubes. The numerical result showed that the underexpanded jet appeared and impinged on the target tube located above the discharging tube. The calculated temperature distribution agreed with the measurement result well. The liquid droplet entrainment and its transport were considered in this analysis. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment.

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Atmospheric dispersion (AD) modeling is the imperative part of accident consequence assessment in nuclear leakage accident. The emergency decision-making depends on the AD model forecast result. The combination of the observation data and forecast result will improve the AD prediction. This paper practices a data assimilation (DA) method in AD based on image registration. The observation data is used to adjust the result of Gaussian plume model, and image affine transformation is used in the DA. The DA method is simple and rapid, and the discrepancy between AD model prediction and true situation is significantly reduced.

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In July 2013, Nuclear Regulation Authority (NRA) in Japan published new regulatory requirements for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFRs) of "Monju" in consideration of severe accidents (SAs) based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as "1F accident") occurred in March 2011. Although the regulatory requirements for Monju will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency set up "Special Committee on Monju Safety Requirements" consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected for the prototype FBR "Monju" (hereinafter written as "SRs for Monju") considering SAs with knowledge from JAEA as well as scientific and technical insights from the experts. Some of the safety features of SFR results from the characteristics of coolant sodium. It does not require pressurization for power generation owing to the high heat conductivity and boiling point of coolant sodium (883 deg C at atmospheric pressure). Reactor coolant level is maintained above the reactor core by guard vessels (GVs) even if a leakage occurred. Therefore, depressurization of primary cooling system and operation of Emergency Core Cooling Systems (ECCSs) in LWRs are not required in SFRs. Liquid sodium can be used in the wide temperature range and natural circulation could be rather easily formed. In addition, multiple accident management (AM) strategies by manual operation can be applied with use of sufficient grace period (several to several ten hours) before significant core damage occurs, because temperature increase is generally gradual even under accident conditions due to large heat capacity of sodium and structures in systems for the loss of heat removal system type accidents. On the other hand, it should take into consideration that not only the possibilities of positive void reactivity and of recriticality due to fuel compaction but also high chemical activity of sodium. For establishing the safety requirements for SFRs, safety characteristics of SFRs, which are different from LWR, shall be taken into account appropriately. This paper summarizes the above mentioned SRs for Monju discussed by the committee, in order to secure the safety of people and to protect environment from accidents.

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The Very High Temperature Reactor (VHTR) is a next generation nuclear reactor system. It has the characteristic that the thermal capacity of the core is so large. Therefore, even if the depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. The VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). The objectives of this study is to investigate the heat transfer characteristics of natural convection of an one-side heated vertical rectangular channel inserting porous materials with high porosity and to develop the passive cooling system for the VHTR. We carried out the experiment and numerical analysis using the one-side heated vertical rectangular channel. The heat transfer and fluid flow characteristics of the natural convection cooling were analyzed using the commercial CFD code. The result shows that the amount of removed heat inserting the copper wire (porosity = 0.9964) was about 8% higher than that without the copper wire.

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We evaluate the heat capacity of actinide dioxides using first-principles density functional theory (DFT). The heat capacity of actinide dioxides mainly originates from lattice vibrations, and the second largest contribution is the Schottky heat capacity, which is caused by the electronic excitation on actinide atoms. The lattice heat capacity is evaluated thorough the first-principles phonon calculation using DFT, and the Schottky heat capacity is calculated with crystal field potentials based on DFT. We obtain the calculated heat capacity of UO_2 and PuO_2 in good agreement with measurements.

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The purpose of this study is to investigate the influence of calculation conditions, such as damping and the boundary conditions between a reactor building and soil, on reactor building vibration due to aircraft impact load. In this study, we created a 3D finite element building model and applied a load-time function for military aircraft. Transient vibrations were analyzed under several calculation conditions. We compared the maximum floor acceleration distributions calculated with and without damping and examined the influence of damping on the acceleration distributions. The calculation results without damping seemed to show an unrealistic distribution; therefore, damping is indispensable for assessing the building vibration. The influence of the boundary condition was examined by varying the support condition of the building. A boundary condition in which the displacement perpendicular to the building walls under the ground was fixed led to underestimation of the acceleration of the first floor; therefore, the fixed condition of the wall should not be applied to the model. Comparison between three spring support conditions (soft, medium, and hard) showed that the maximum accelerations were in almost complete agreement. This indicates that the influence of the spring constants of the soil-structure interaction on a reactor building is small.

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The Nuclear Risk Research Center (NRRC) has been established since October 1st 2014 in Central Research Institute of Electric Power Industry (CRIEPI) to strongly support the Japanese utilities' continuous safety improvement initiative. One of the NRRC's responsibilities is to develop a near-and mid-term roadmap for nuclear risk related research and development (R&D) including modern probabilistic risk assessment (PRA) methods and physical investigations in beyond-design-basis events at a nuclear power plant (NPP). NRRC deals with this nuclear-risk related R&D dividing it into two fields: risk assessment field and natural external event field. This paper presents the activities with respect to the latter; frequencies and magnitudes of natural external events (earthquake ground motion, tsunami, fault displacement, tornado and volcanic eruption) and effects of these events on structures and components of NPP sites. The authors also introduce some large-scale experimental facilities built at the Abiko Research Area in CRIEPI to study nuclear risk related R&Ds.

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The accident occurred at the Fukushima Daiichi Nuclear Power Plant Unit 2 has been investigated by the severe accident analysis code, SAMPSON with more realistic boundary conditions and newly introduced models. In Unit 2, the Reactor Core Isolation Cooling system (RCIC) is thought to have worked for unexpectedly long time (about 70 hours) without batteries. It is thought to be due to balance between injected water from the RCIC pump and supplied mixture of steam and water to the RCIC turbine. To confirm the RCIC working condition and reproduce the measured plant properties, such as pressure and water level in the reactor pressure vessel (RPV), we introduced two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by RCIC was calculated through mass flow rate of steam included in extracted two-phase flow, steam generated from flashing of water included in extracted two-phase flow, and turbine efficiency degradation originated by the mixture of steam and water flowing to the RCIC turbine. To reproduce the dry well (DW) pressure, we assumed that torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Simulation results by SAMPSON basically agree with the measured values such as pressure in the RPV and in the DW until several days after the scram. However, some contradictions between the simulation results and the measured values, such as that inversion of the RPV pressure at 10 hours after scram in the measurement happened at 14 hours in the simulation and that the DW pressure showed different behavior between simulation and measurement when SRV started periodic operation at 71 hours, are still remain and are under consideration. In the current calculation, model for falling core to the lower plenum was modified so that debris is not retained at the core plate based on observation of the XR2-1 experiment. Additionally, model of the RPV failure by melting of the penetrating pipe was newly introduced. Though actual mass flow rate of water injection by fire pump, which affects the core and the RPV damage, is still unclear, simulation result indicates that a part of the core fell to the lower plenum but the RPV did not failed.

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We have produced new nuclear data group constant sets from JENDL-4.0 and FENDL-3.0 for fusion reactor nuclear analyses; FUSION-J40-175, FUSION-F30-175 (40 materials, neutron 175 groups, gamma 42 groups), FUSION-J40-42 and FUSION-F30-42 (40 materials, neutron 42 groups, gamma 21 groups). MATXS files of JENDL-4.0 and FENDL-3.0 were newly produced with the NJOY2012 code. FUSION-J40-175, FUSION-J40-42, FUSION-F30-175 and FUSION-F30-42 were produced with the TRANSX code. KERMA factors, DPA and gas production cross-section data were also prepared from the MATXS files with TRANSX. Test calculations were carried out in order to validate these nuclear group constant sets. They suggested that these group constant sets had no problem.

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The upgrade of the JT-60U to the superconducting tokamak "JT-60SA" has been carried out to contribute the early realization of fusion energy by addressing key physics issues relevant for ITER and DEMO. Disassembly of the JT-60U tokamak was required so as to newly install the JT-60SA torus at the same position in the torus hall. The JT-60U tokamak was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium-deuterium (D-D) reactions of 1.5x10^<20> (n) in total. Since this work was the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. About 13,000 components with a total weight of more than 5,400 tonnes were removed from the torus hall and stored safely in storage facilities. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. It was confirmed that the main radioactive material of the disassembly components was the stainless steel and that its dose rate was almost background level (〜0.1μSv/h) at 〜10m far from the vacuum vessel. It seems that the disassembly components with background dose level are in the clearance level. The assembly of JT-60SA tokamak has started in January 2013 after this disassembly of the JT-60U tokamak.

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When the depressurization accident occurs in the Very-High-Temperature Reactor (VHTR), it is expected that air enter into the reactor core. Therefore, it is important to know a mixing process of different kind of gases in the stable or unstable stratified fluid layer. Especially, it is also important to examine an influence of localized natural convection and molecular diffusion on mixing process from a viewpoint of safety. In order to research the mixing process of two component gases and flow characteristics of the localized natural convection, we have carried out numerical analysis using three dimensional CFD code. The numerical model was consisted of a storage tank and a reverse U-shaped vertical slot. They were separated by a partition plate. One side of the left vertical fluid layer was heated and the other side was cooled. The right vertical fluid layer was also cooled. The procedure of numerical analysis is as follows. Firstly, the storage tank was filled with heavy gas and the reverse U-shaped vertical slot was filled with light gas. In the left vertical fluid layer, the localized natural convection was generated by the temperature difference between the vertical walls. The flow characteristics were obtained by a steady state analysis. The unsteady state analysis was started when the partition plate was opened. The gases were mixed by molecular diffusion and natural convection. After the time elapsed, natural circulation occurred. The result obtained in this numerical analysis is as follows. The temperature difference of the left vertical fluid layer was set to 100K. The combination of the mixed gas was nitrogen and argon. After 76 minutes elapsed, natural circulation occurred.

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The redox behavior of chromium was investigated under three conditions for the nitric acid solutions considering the operating conditions of the reprocessing plant -atmospheric-pressure boiling, reduced-pressure boiling, and atmospheric-pressure non-boiling-to determine the relationship between the corrosion of austenitic stainless steel (R-SUS304ULC) and this behavior. The oxidation of Cr(III) to Cr(VI) in 8 M boiling nitric acid was increased over that in non-boiling nitric acid, even at the same temperature by measuring the time dependence of the chromium concentration and thermodynamic calculations. These results show that Cr(III) is more likely to oxidize under boiling than non-boiling conditions. On the other hand, the apparent rate of Cr(VI) reduction, as consumed by corrosion of the stainless steel, was more than 10 times greater than the rate of Cr(III) oxidation in 8 M nitric acid solution. This is evidence that the rate of Cr(VI) reduction by corrosion is kinetically faster than the rate of Cr(III) oxidation. In addition, the apparent rate of Cr(VI) reduction and the corrosion rate of R-SUS304ULC were positively correlated. Therefore, these indicate that the corrosion of R-SUS304ULC apparently follows Arrhenius' law in 8 M nitric acid solution containing Cr(VI), regardless of the boiling or non-boiling conditions.

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In this paper, a failure criterion applicable to large-strain finite element analysis (FEA) results was studied to predict the limit bending load M_c of the groove shaped wall-thinned pipes, under combined internal pressure and bending load, that experienced cracking. In our previous studies, Meshii and Ito (2012) considered cracking of pipes with groove shaped flaw (small axial length δ_z in Fig. 1) was due to the plastic instability at the wall-thinned section and proposed the Domain Collapse Criterion (DCC). The DCC could predict M_c of cracking for small δ_z by comparing the von Mises stress σ_<Mises> with the true tensile strength σ_B. Because the discrepancy in prediction of the M_c in the case of cracking was within 15%, it was considered that the predictability was could be improved further. Thus, in this work, attempt was made to improve the accuracy of M_c prediction with a perspective that multi-axial stress state might affect this plastic instability at the wall-thinned section. As a result of examination of the various failure criteria based on multi-axial stress, it was confirmed that the limit bending load of the groove flawed pipe that experienced cracking in experiment (Hereafter, it was expressed "flawed pipe that experienced cracking") could be predicted within 5 % accuracy by applying Hill's plastic instability onset criterion (Hill, 1952) to the outer surface of the crack penetration section. The accuracy of the predicted limit bending load was improved from DCC's within 15% to within 5%.

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After the Fukushima Daiichi nuclear power plant accident due to the pacific coast of Tohoku earthquake and Tsunami, we have been developing an analysis method considering creep damage mechanisms based on three-dimensional analysis for the early completion of the decommissioning of nuclear power plants. We have also been obtaining material properties that are not provided in existing databases or literature for the analysis. In this study, we measure the tensile and creep properties of low alloy steel, Ni-based alloy, and stainless steel at high temperatures near the melting points. Using experimental data, some parameters for the creep constitutive law and creep failure evaluation method are determined. In addition, by comparing the creep rupture time between the finite element analysis and the experiment, the validity of the failure evaluation method is confirmed.

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In existing severe accident codes such as MELCOR and THALES2, rupture of reactor pressure vessel (RPV) by relocated molten core is judged using simple models such as temperature and/or stress criteria. However, it is difficult to assess rupture behavior of the lower head of RPV in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi. One reason is that boiling water reactors (BWRs) have geometrically complicated structure with a lot of penetrations. Another one is that BWR lower head is composed of various types of materials of RPV, weld-overlay cladding, control rod guide tubes, stub tubes, welds, etc. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. The detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes, and welds are constructed. TH analysis is performed to obtain three-dimensional temperature distribution in relocated debris. Using TH analysis results, structural analysis is carried out to evaluate creep damage distributions using four types of damage criterions of "considere", strain, Kachanov, and Larson-Miller-parameter (LMP) criteria. Creep damage evaluation based on Kachanov and LMP models is made by using experimentally determined parameters. From comparison of damage criterions, it is shown that failure regions of BWR lower head are only penetrations under simulated conditions, although there is a large difference in failure time.

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The equatorial launcher (EL) has been designed with the poloidal steering with the 20MW beam injection to plasma, expecting the enhancement of driven current at the peripheral region of plasma, ρ = 0.4 〜 0.6. The millimeter (mm)-wave design to attain the poloidal beam steering with transmission efficiency of 99% has been successfully obtained and the structural design of the launcher components such as blanket shield modules (BSMs), port plug and so on are also modified based on the mm-wave design modification. This modification ensures that mm wave beams from both middle and bottom beam row pass through the same BSM opening and will leads to the neutron shielding potential. The electromagnetic analysis of the modified EL shows that induced force and momentum torque on the BSMs and entire EL structure are below the mechanical criteria of the support structure of each BSMs and EL port plug flange.

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AP1000 nuclear power plant adopts several severe accident management strategies aiming at terminating the core melting process and maintaining the containment integrity, thus limiting the risk of significant radioactive releases to the environment. Analyses are carried out to evaluate the effectiveness of these management strategies using integral severe accident computer code MIDAC. Simulations and analyses of severe accidents initiated by 1-in, 2-in and 5-in LOCAs are performed first, then 1-in LOCA is selected to assess the severe accident management strategies. Different assumptions are defined in different cases for a specific management strategy. Results show that ADS can significantly reduce the primary system pressure, thus resulting in the injection of cold water from CMTs and ACCs into the core rapidly. In addition, high-pressure melt ejection is prevented due to low in-vessel pressure and the hydrogen generation in-vessel is significantly decreased. After cavity injection system operates at core exit temperature 1367 K, water from IRWST will flood the cavity up to the hot legs level. Consequently, heat in-vessel is removed by external reactor vessel cooling (ERVC) and the goal of in-vessel melt retention is achieved. As the final heat sink, PCCS can significantly decrease the containment pressure and gas temperature to ensure the containment integrity. Igniters arranged at different locations of the containment will lead to local hydrogen combustion, therefore, the fraction of hydrogen in each compartment is maintained less than 6% and hydrogen deflagration or hydrogen detonation is consequently avoided.

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Intermediate heat exchangers (IHXs) are structural components of the very high temperature gas-cooled reactor (VHTR) which transfer the heat generated in high-temperature core to the secondary circulation loop. In this paper, the properties of Alloy 617, one of the candidate alloys as structural material for IHXs, were evaluated by finite element analysis (FEA) based on ANSYS platform under creep-fatigue interaction conditions. Life prediction approach based on damage mechanics was combined with ANSYS structural analysis. Results of the life prediction and damage calculation of structural component by this method will provide guidance for structural and operational design of IHX. Preliminary comparison between calculation and FEA results indicates that this method is feasible, but many experiments have to be conducted in future because the creep-fatigue experimental results are scarce.

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After a severe accident in nuclear reactor like Fukushima power plant for the on-site scenario, some staffs could not evacuate with personal vehicles especially in China. Faced the simulation requirement of evacuating by foot in microscopic field, here we will introduce the lattice-gas model of pedestrians evacuation with two dimensional grid dynamic analysis, combined with Gaussian puff model to calculate the scale of personal radiation risk for their trajectories discretely and respectively. The purpose of this paper is to give informatively route selection designed for the plan of nuclear emergency preparedness work in responding large amount leakage of radionuclide, and cope with the problems from evaluation of nuclear accident emergency management and regulation. Our results also illustrate comparison data and targeted countermeasure in the corresponding suggestion, and giving advice for individual and crowd. A simplified application is presenting in demonstrations in a nuclear power plant under constructed at south China.

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Due to Fukushima nuclear accident, the attitude of the public to the development of nuclear power changes so much and public have put forward new questions against building nuclear power plants. More and more people in different fields are paying attention to public communication of nuclear power. How to deal with public communication in a more scientific and efficient way to enhance the public's cognition and eliminate their misunderstanding about nuclear power has become a difficulty in the development of nuclear power. Based on the process of public communication accomplished in China before and after Fukushima accident and other countries' experience of public communication, this paper concludes the important questions existing in public communication and researches into the development trend of public communication in order to build a foundation for the scientific development of nuclear power.

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The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a PWR reactor building. Some mitigation against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and combustion were necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on the CFD simulation of hydrogen distribution in penalized reactor conditions. Numerical developments of CFD code and code validation lead to consistent simulations. Results show that three mechanisms rapidly reduce the combustion risk. This paper focuses on a violent scenario occurring on a Pressurized Water Reactor 900 CPY.

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