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In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), the eutectic reaction between boron carbide (B4C) as control rod element and stainless steel (SS) as control rod cladding or related structure materials may take place. Thus, the kinetic behavior of the B4C-SS eutectic reaction is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of the B4C-SS eutectic reaction and compare the pervious findings, the thermal analysis using the pellet samples of B4C and Type 316L SS as a different experimental approach was performed up to 1773 K at different heating rates of 2.5-10 K min-1. The differential thermal analysis (DTA) endothermic peaks for the B4C-SS eutectic reaction appeared from 1483K to 1534K and systematically shifted to higher temperatures with increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B4C-SS eutectic reaction were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values of the thinning experiment at high temperatures. In addition, the microstructure and element distribution formed in the interdiffusion layer composed of the B4C-SS system were analyzed by the electron probe microanalyzer (EPMA), which can provide key validation data on elemental interdiffusion behavior in the early stage of the eutectic reaction for reflecting the reaction kinetic modeling.

International audience; France (CEA and FRAMATOME) and Japan (JAEA, MFBR, and MHI) teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor (SFR) concept. This paper mainly describes the capabilities of ASTRID 600 to demonstrate SFR technologies of both countries. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600, considering Japan’s target for core performance. The neutronics design of the core has satisfied most required design targets and conditions. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device (CSD), called a self-actuated shutdown system (SASS), one of the safest approaches in Japan. Japan’s team used the SASS in place of the hydraulically suspended absorber rod, called RBH, one of the safest approaches of France, to investigate the potential of the SASS as a design measure common to the countries. The preliminary analysis has shown that the SASS can satisfy the countries’ main requirements. This study has also revealed that the mitigation measures of ASTRID 600 against a severe accident are promising to achieve in-vessel retention for both countries.

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

This paper describes the development of a probabilistic risk assessment (PRA) methodology against a combination hazard of strong wind and rainfall. In this combination hazard PRA, a hazard curve is evaluated in terms of maximum instantaneous wind speed, hourly rainfall, and rainfall duration. A scenario analysis has provided event sequences resulting from the combination hazard of strong wind and rainfall. The typical event sequence was characterized by the function loss of auxiliary cooling system, of which heat transfer tubes could crack due to cycle fatigue caused by cyclic contacts with rain droplets. This cycle fatigue crack could occur if rain droplets enter into the air cooler of the system following the cooler's roof failure due to strong-wind-generated missile impact. This event sequence has been incorporated into an event tree which addresses component failure caused by the combination hazard. As a result, a core damage frequency has been estimated to be about 10-6/year in total by multiplying discrete hazard frequencies by conditional decay heat removal failure probabilities. The dominant sequence is the manual operation failure of an air cooler damper following the failure of external fuel tank due to the missile impact. The dominant hazard is the maximum instantaneous wind speed of 20−40 m/s, the hourly rainfall of 20−40 mm/h, and the rainfall duration of 0−10 h.

The effects of B4C addition on the solidus and liquidus temperatures of type 316 austenitic stainless steel (SS), and on the density and surface tension of molten SS, were experimentally studied. The solidus temperature of SS−x mass% B4C (0 ≤ x ≤ 10) monotonically decreased from 1666 to 1307 K with B4C addition. The liquidus temperature had a minimum at around 2.5 mass% B4C, and increased with further B4C addition up to 10 mass%. The density and surface tension of molten SS−x mass%B4C were successfully measured over a wide temperature range (including an undercooling region) via an electromagnetic-levitation technique. The density of each sample decreased linearly with temperature. The density also monotonically decreased with B4C content. Although the addition of B4C had no clear effect on the surface tension of SS−x mass%B4C, sulfur dissolved in SS316L caused a significant decrease in the surface tension.

International audience; In the frame of the France-Japan agreement on nuclear collaboration, a bilateral collaboration agreement on nuclear energy was signed on March 21st, 2017, including a topic dedicated to Sodium-cooled Fast Reactor (SFR). This agreement has set the framework to start a bilateral discussion on a joint view of an SFR concept. France (CEA and FRAMATOME) and Japan (JAEA, MHI and MFBR) have carried out these studies from 2017 to 2019. Based on the beginning of the basic design phase of ASTRID project − ASTRID − 600 MWe (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration), the two countries performed a common work to examine ways to develop a feasible common design concept, which could be realized both in France and in Japan. The subject was then extended and extrapolated with the ASTRID − 150 MWe data (reduced power reactor and enhanced experimental capabilities) in a second phase of this study. France and Japan first focused on design requirements. Common requirements were identified, as well as differences in the safety approach and the structural design requirements, according to national standards and respective site conditions, in particular considering seismic hazards. The teams developed common Top-Level design Requirements (TLRs) to allow common specification data, then joint design. This collaborative work was carried out through the implementation of twelve France-Japan Working Groups, working jointly. This paper is providing a review of this joint synthesis on Sodium Fast Reactor design concept. It is summarizing the context and objectives, then the definition and approaches of the Top Level Requirements. This paper is then dealing with the major design features: the core design and their related safety aspects, and the nuclear island design. Thus, this paper is providing a comprehensive review of this joint work gathering French and Japan nuclear design teams during two full years.

A new method has been developed to assess potential challenges by forest fire smoke on a cooling function of a decay heat removal system (DHRS) of a sodium-cooled fast reactor. Combinational numerical simulations of a forest fire propagation and a smoke transport were performed to evaluate a cumulative amount of smoke captured on air filters of the DHRS. The forest fire propagation simulations were performed using FARSITE code to evaluate a temporal increase of a forest fire spread area, a frontal fireline location, reaction intensity, and fireline intensity. Peripheral boundary of the forest fire spread area is shaped like an ellipse on the terrain, and the active forest fire area from which smoke is produced as a forest fire product is increased with forest fire spread. The smoke transport simulations were performed using ALOFT-FT code where a spatial distribution of smoke density, especially of particle matter (PM), is evaluated. The snapshot (i.e. at a certain time step) outputs by FARSITE on the reaction intensity and the fireline intensity were utilized as the input data for ALOFT-FT, while it was conservatively assumed that the smoke generated from the active forest fire area along the periphery boundary rises up from the frontal fireline location nearest to a nuclear power plant (NPP) and that prevailing wind transports all smoke to an NPP in the leeward side. The evaluated time-dependent changes of spatial PM density were utilized to calculate a cumulative amount of PM captured on the air filters of the DHRS. Sensitivity analysis was performed on prevailing wind speed to which both the fireline intensity and the smoke transport behavior are sensitive. The total amount of PM on the air filters was conservatively estimated around several hundred grams per m2 which is well below the utilization limit.

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior., 炉心損傷事故(CDA)の炉内格納(IVR)はナトリウム冷却高速炉(SFR)の安全特性向上において極めて重要である。SFRのCDAにおいては、溶融炉心物質が炉容器の下部プレナムへ再配置し、構造物へ重大な熱的影響を及ぼし、炉容器の溶融貫通に至る可能性がある。この再配置過程の評価を可能とし、SFRのCDAではIVRで終息することが最も確からしいことを示すため、SFRのCDAにおける物質再配置挙動の評価手法を開発する研究計画が実施された。この計画では、炉心領域からの溶融物質流出挙動の解析手法、溶融炉心物質のナトリウムプール中への侵入挙動、デブリベッド挙動のシミュレーション手法を開発した。

After the Fukushima Daiichi nuclear power plant [NPP] accident, there has been an increased concern with the safety of NPPs in terms of external hazards, one of which is a forest fire which can create potential challenges to safety functions and the structural integrity of an NPP. As a part of the development of a risk assessment methodology for forest fires as an external hazard, forest fire propagation simulations have been performed by using the FARSITE simulator. These simulations have been used to evaluate two intensity parameters (i.e. fireline intensity and reaction intensity) and three other key parameters (i.e. flame length, rate-of-spread, and forest fire arrival time) which are related to “heat” and “flame” effects on an NPP. Sensitivity analyses for a wide range of weather conditions were performed in order to identify the variable ranges of the intensity and other key parameters. The location studied was selected from among areas with typical topographical and vegetation surrounding NPPs in Japan. The NPP is facing the sea and surrounded by hills, distanced from an urban area, with mostly broad leaf forests, several paddy fields and a few pasture areas. Low-to-high frequency weather conditions have been utilized in this analysis; forest fire propagation simulations were performed “with/without prevailing wind” (i.e. 0–24 m/s wind speed) and “high/low values for ambient temperature and relative humidity” (−4.3 to 37 °C and 5–99%, respectively) according to the recorded data ranges for the typical NPP site. The maximum values of fireline intensity and rate-of-spread are 4.7 × 102 kW/m and 2.4 m/min and they depend very much on prevailing wind speed and relative humidity (around 2.3 and 1.8 times respectively) but less on ambient temperature (around 1.1 times). Reaction intensity and flame length change within relatively narrow ranges (around 1.7 and 1.5 times respectively) even for all the variation in weather parameters. The forest fire arrival time at the site is reduced by a factor of 5 with changing prevailing wind speed from the recorded-highest to zero. The arrival time increases some 3.4 times with the highest humidity compared to the recorded-lowest conditions, although it is changed little even by varying ambient temperature. Given that this study shows that the maximum height of a flame on a canopy top is close to the range of power line height, a loss of offsite power is recognized as a possible subsequent event during a forest fire.

Studies on the self-leveling behavior of debris beds are crucial in the assessment of core-disruptive accidents (CDAs) that could occur in sodium-cooled fast reactors (SFRs). To further clarify this behavior, a series of experiments has been performed in which nitrogen gas has been percolated uniformly through particle beds. Although in the past, several experiments have been conducted to investigate this behavior, most of these were under comparatively lower gas velocities, the findings of which might be not directly applicable to actual reactor accident conditions. Current experiments were conducted in a cylindrical tank with the dimensions of 310 mm in inner diameter and 1000 mm in height, in which nitrogen gas, water and different kinds of solid particles, simulate the fission gas, coolant and fuel debris, respectively. During experiments, to accomplish the bubble-based leveling as expected in actual reactor conditions, two experimental approaches, termed respectively as the gas pre-charge method and the pressure-adjustment method, have been attempted. Through elaborate comparisons and evaluations, it is found that compared to the gas pre-charge way the pressure-adjustment method can alleviate the liquid disturbance from the bottom inlet pipelines more effectively throughout the whole experimental process. Further, based on the experimental observations and quantitative data obtained, influence of various experimental parameters on the self-leveling, including gas flow rate, water depth, particle size, particle density as well as boiling mode is checked and compared. From the analyses, it is also observed that, the convection in the water pool, though is not so evident for previous studies at lower gas velocities, is newly identified to play a significant role in this work, especially for the experimental runs using larger-size lower-density particles at comparatively higher gas flow rates and water depths. Current work provides critical evidence and a large palette of favorable data for a better understanding of CDAs and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

In this study, hazard evaluation methodologies were developed for the decay heat removal of a typical sodium-cooled fast reactor in Japan against snow, tornado, wind, volcanic eruption, and forest fire. In addition, probabilistic risk assessment and margin assessment methodologies against snow were developed as well. Snow hazard curves were developed based on the Gumbel and Weibull distributions using historical records of the annual maximum values of snow depth and daily snowfall depth. Wind hazard curves were also evaluated using the maximum wind speed and instantaneous speed. The tornado hazard was evaluated by an excess probability for the wind speed based on the Weibull distribution multiplied by an annual probability of the tornado strike at a target plant. The volcanic eruption hazard was evaluated using geological data and tephra diffusion simulation which indicated tephra layer thickness and tephra diameter. The forest fire hazard was evaluated based on numerical simulation which contributed to creating a response surface of frontal fire intensity and Monte Carlo simulation for excess probability calculation. After developing an event tree and failure probabilities, the snow PRA showed the order of 10−7/year of core damage frequency. Event sequence assessment methodology was also developed based on plant dynamics analysis coupled with continuous Markov chain Monte Carlo method in order to apply to the event sequence against snow. Furthermore, this study developed the snow margin assessment methodology that the margin was regarded as the snowfall duration to the decay heat removal failure which was defined as when the snow removal speed was smaller than the snowfall speed.

As part of the development of a flow-induced vibration evaluation methodology for the primary cooling piping in Japan sodium-cooled fast reactor, important factors were discussed in evaluating the flow-induced vibration for the hot-leg piping. To investigate a complex flow near the inlet of the hot-leg piping, a steady-state numerical analysis was carried out for the reactor upper plenum flow, which was simulated in a 1/10-scale reactor upper plenum experiment. Based on this analysis, experimental conditions on swirl inflow and deflected inflow that were identified as important factors were determined for flow-induced vibration experiments simulating only the hot-leg piping. In this study, the effect of the swirl inflow on flow pattern and pressure fluctuation onto the pipe wall was investigated in a 1/3-scale hot-leg pipe experiment. The experiment has indicated less significant for the pressure fluctuations, while the flow separation region was slightly influenced by the swirl inflow. Computational fluid dynamics simulation with a U-RANS approach results also appear in this paper. Through the simulations under the swirl inflow conditions of 0% and 5%, the validity of the U-RANS simulation was confirmed by comparison to the 1/3-scale hot-leg piping experiments.

Laser Doppler Velocimetry (LDV) measurements in a 90 degree elbow of which the curvature radius coincides with its inner diameter were examined for the cases of inflow from a long pipe, short pipe and swirl generator. Ensemble averaged flow distribution at the Reynolds number of 320000 based on the inner pipe diameter and bulk velocity shows that shortening the upstream pipe length to 4.9D from 10D induces the flow separation downstream of the elbow. Detailed observation suggests that shortening upstream pipe weakens the Prandtl's secondary flow of the first kind. Our swirl generator induced a swirling inflow with the non-dimensional angular momentum of 0.12 based on the inner pipe diameter and bulk velocity. The circumferential velocity distribution formed a shape like a Rankine combined vortex at the elbow inlet, and the accelerated axial velocity was observed at the vortex center. The axial velocity distribution however was found to be almost the same as that of the non-swirl inflow case in the latter half of the elbow. Frequency analyses showed that the Strouhal number by vortex shedding from the boundary layer occurring at the inner side of the elbow become 0.5, except for 0.6 in the case of the long pipe. The change of the Strouhal number is probably related with the boundary layer width and the local flow velocity.

LDV measurements in a 90 degree elbow, of which curvature radius coincides with the diameter have been conducted for cases of developed inflow, undeveloped inflow, swirling inflow, and deflected inflow that became faster at the inside. This paper especially focuses on a result of the deflected inflow, comparing with a result of the undeveloped inflow. The result shows that the deflected inflow reinforces a convex velocity distribution occurring near the curvature inside in the downstream region, concluding that the deflected inflow promotes the secondary flow of Prandtl's first kind in the elbow. Frequency analyses of axial flow fluctuations show that the dominant frequency can be seen in the downstream of the elbow even in the deflected inflow case. However its Strouhal number increases to 0.6 from 0.5, that is for the short pipe case. Results of frequency analyses are also shown for other cases that we have been examined. Dominant Strouhal numbers in most of the cases become 0.5, except for 0.6 in cases of the developed and deflected inflow. This frequency shift might be related with the boundary layer size and the local flow velocity, since the corresponding fluctuation is caused by vortex shedding from the boundary layer at the elbow inside.

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal–fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt–solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and fracture criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.

Studies on the self-leveling behavior in debris beds are crucial in the assessment of core-disruptive accidents (CDAs) that could occur in sodium-cooled fast reactors (SFRs). To clarify this behavior, a series of experiments have been performed in which nitrogen gas has been percolated uniformly through a particle bed. In these experiments, solid particles and water contained in a rectangular tank simulate respectively fuel debris and coolant. Based on the data obtained, an empirical model was developed to describe the transient variation in the bed inclination angle during the self-leveling process. Good agreement has been obtained between calculated and experimental values. Verification of the model has been confirmed through detailed analysis of the effects of experimental parameters such as particle size, particle density, and gas flow rate. Its applicability to extended conditions was further discussed by performing modeling simulations and comparing results against experimental data obtained from a ...

This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in the Japan sodium-cooled fast reactor (JSFR), with particular emphasis on recent research and development activities that investigate unsteady elbow pipe flow. Experimental efforts have been made using 1/3-scale and 1/10-scale single-elbow test sections for the hot-leg pipe. The 1/10- scale experiment simulating the hot-leg pipe indicated no effect of pipe scale in comparison with the 1/3- scale experiment under inlet-rectified-flow conditions. The next experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was deflected at the downstream from the elbow, the experiment clarified a less significant effect of swirl flow on pressure fluctuation onto the pipe wall. An additional experiment was intended to study the effect of elbow curvature. The experiments with water revealed no clear flow ...

A series of experiments on bubbling behavior in particle beds was performed to clarify three-phase flow dynamics in debris beds formed after core-disruptive accident (CDA) in sodium-cooled fast breeder reactors (FBRs). Although in the past, several experiments have been performed in packed beds to investigate flow patterns, most of these were under comparatively higher gas flow rate, which may be not expected during an early sodium boiling period in debris beds. The current experiments were conducted under two dimensional (2D) and three dimensional (3D) conditions separately, in which water was used as liquid phase, and bubbles were generated by injecting nitrogen gas from the bottom of the viewing tank. Various particle-bed parameters were varied, including particle-bed height (from 30 mm to 200 mm), particle diameter (from 0.4 mm to 6 mm) and particle type (beads made of acrylic, glass, alumina and zirconia). Under these experimental conditions, three kinds of bubbling behavior were observed for the first time using digital image analysis methods that were further verified by quantitative detailed analysis of bubbling properties including surface bubbling frequency and surface bubble size under both 2D and 3D conditions. This investigation, which hopefully provides fundamental data for a better understanding and an improved estimation of CDAs in FBRs, is expected to benefit future analysis and verification of computer models developed in advanced fast reactor safety analysis codes.

In a core-disruptive accident of a sodium-cooled fast breeder reactor, core debris may settle on the coresupport structure and/or in the lower inlet plenum of the reactor vessel because of rapid quenching and fragmentation of molten core materials in the subcooled sodium plenum. Coolant boiling is the mechanism driving the self-leveling of a debris bed that causes significant changes in the heat-removal capability of the beds. In the present study, we develop criteria establishing the onset of this self-leveling behavior that we base on a force balance model assuming a debris bed with a single-sized spherical particle. The model considers drag, buoyancy, and gravity acting on each particle. A series of experiments with simulant materials verified the applicability of this description of self-leveling. Particle size (between 0.5–6 mm), shape (spherical and nonspherical), density (namely of alumina, zirconia, lead, and stainless steel), along with boiling intensity, bed volume, and even experimental methods...

This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: fuel-pin disruption, and low energy disrupted core motion. Related experimental database were reviewed to select appropriate experiments. To analyze the fuel-pin disruption behavior, the CABRI-EFM1 and the CABRI-E7 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall fuel-pin disruption and dispersion behavior, which was characterized by a thermal pin-failure mode, observed in the CABRI-EFM1 experiment. Since the code framework of SIMMER-III cannot treat a mechanical deformation and breakup behavior, the SAS4A code was applied to the CABRI-E7 experiment, where a mechanical pin-failure mode was realized. In this study, such a failure mode was also reasonably simulated. The low-energy disrupted core consists principally of fuel particles and liquid steel (or fuel). Under such a mixture condition, significantly reduced melt penetration length was obtained in the THEFIS out-of-pile experiments. SIMMER-III well simulated the melt freezing and blockage behavior observed in the experiment.

For the Japan Sodium-cooled Fast Reactor, an experimental study on the fluctuating pressure of the hot legs was carried out with tests using a 1/3-scale model. The total resistance coefficient is consistent with the published data, and our research has given some additional data up to the Reynolds number of 8.0x106. The flow pattern in the postcritical regime is independent of a Reynolds number. The statistical examination revealed that fluctuating pressures on the pipe wall depend on the mean velocity but not on the viscosity of the fluid. Negative spikes of pressure appeared for high velocity. Based on these experimental data it is concluded that, there are similarity laws for the scale of a model and the property of fluid, but not for the velocity in the pipe. Theoretical considerations also gave a discussion how to extrapolate the fluctuating pressure to the actual hot-leg conditions.