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To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (B4C) and stainless-steel (SS). The influence of the existence of carbon on the B4C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B4C-SS samples, a new layer was formed between B4C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe2B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the reaction. These domains might be formed during the solidifying process.

To enhance resilience of next-generation nuclear structures, it is necessary to develop design methodology that mitigates impacts of failure caused by extremely high temperature conditions which might lead to a severe accident. The purpose of this study is to understand its deformation behavior under extremely high temperature conditions and to identify the areas that should be focused on to mitigate impacts of failure. For this purpose, this study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a commercial finite element analysis code, FINAS/STAR. In a postulated scenario of loss of heat removal system, the RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum for 20 hours, assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of the RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 160 mm, but it resulted in no failure. This analysis implies maintaining the RV sodium level in a long term, thereby enhancing the RV resilience.

For nuclear power plants, probabilistic risk assessment (PRA) should be performed not only against earthquake and tsunami, which are critical events especially in Japan, but also other external hazards such as strong wind. The aim of the present study is to develop a practical PRA methodology for sodium-cooled fast reactors (SFRs) against strong wind, paying attention to the final heat sink, ambient air, that removes decay heat under accident conditions. First, this study used Gumbel distributions to estimate hazard curves of the strong wind based on weather data recorded in Japan. Second, it identified important structures, systems and components (SSCs) for decay heat removal, and developed an event tree that results in core damage, focusing on the impacts of missiles (e.g., steel pipes) caused by strong wind. It also identified missiles that can reach SSCs at elevated places, and calculated the fragility of the SSCs against the missiles as a product of two probabilities. One is a probability of the missiles that would enter an inlet or outlet of the decay heat removal system, and another is a probability of failure caused by missile impacts. Finally, it quantified conditional decay heat removal failure probabilities by introducing the fragilities into the event tree. The core damage frequency (CDF) was estimated at about 5 × 10-10/y. The dominant sequence is that strong wind causes offsite power loss and missiles, the missiles penetrate the diesel fuel tank, cause a fire, and the fire increases air temperature around the reactor building where air cooler inlets of decay heat removal systems are installed, leads to loss of power for the diesel generator for forced circulation cooling, resulting in loss of decay heat removal. Through the above, this study has developed the practical PRA methodology for SFRs against strong wind.

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.

In a core disruptive accident scenario, boron carbide, which is used as a control rod material, may melt below the melting temperature of stainless steel owing to the eutectic reaction with them. The eutectic mixture produced is assumed to extensively relocate in the degraded core, and this behavior plays an important role in significantly reducing the neutronic reactivity. However, these behaviors have never been simulated in previous severe accident analysis. To contribute to the improvement of the core disruptive accident analysis code to simulate these eutectic melting and relocation behavior, the thermophysical property database of the eutectic mixture implemented into the analysis code should be developed. As part of this database development, this study measured the thermophysical properties of the eutectic mixture in the solid state. In this study, pre-alloyed austenitic stainless steel containing boron carbide, which has the prescribed concentration of boron carbide and is homogeneous, is used as eutectic samples to reduce the uncertainty of samples and achieve accurate measurement. Based on these evaluation results, the effect of adding boron carbide on the thermophysical properties of austenitic stainless steel are discussed. Furthermore, regression equations that show the temperature (and boron carbide concentration) dependence are created for each thermophysical property.

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the B4C-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-B4C eutectic melting reaction was carried out by contacting SS melts of several kg with a B4C pellet heated up to about 1500 °C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the B4C-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the B4C-SS contact interface and (Fe,Cr)2B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the B4C pellet.

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 aimed at developing a PRA methodology for the combination of low temperature and snow for a sodium-cooled fast reactor which uses the ambient air as its ultimate heat sink to remove decay heat under accident conditions. The annual exceedance probabilities of low temperature and of snow can be statistically estimated based on the meteorological records of 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., a clogged intake and/or outtake for a DHRS and for an emergency diesel generator, an unopenable door on necessary access routes due to accumulated snow, failure of intake filters due to accumulated snow, and possibility of water freezing 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 intakes and outtakes caused by loss of the access routes, this study has investigated effects of electric heaters installed around the intakes and outtakes as an additional countermeasure. By using the annual exceedance probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence caused by a snow and low temperature combination hazard is the failure of the electric heaters and the loss of the access routes for snow removal due to low temperature and snowfall which last for a day, and daily snowfall depth of 2 m/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.

As a part of external hazard probabilistic risk assessment methodology development, a sensitivity study on forest fire hazard curves was performed on condition parameters where frequency/probability variables were varied within respective fluctuation ranges. Important variables related to forest fire breakout time, forest fire breakout points, and forest firefighting operation were identified. The probability fluctuation on forest fire breakout time only affects the intensities of the hazard curves, but not the frequency, because the intensity increases in daytime due to sunshine and the breakout probability in daytime is statistically higher than the daily average. The hazard curves of the reaction intensity and the fireline intensity increased around 4% and 14% respectively in comparison with those without the forest fire breakout time fluctuation. The probability fluctuation on forest fire breakout points affects only the frequency of the hazard curves, but not the intensities. The hazard curves vary around +70% to -40% in frequency, because each forest fire breakout point has different distance to the plant and the forest fire arrival probability varies with propagation duration. The probability fluctuation due to forest firefighting operation only affects the frequency of the hazard curves, but not the intensity. The effect of the forest firefighting operation was conservatively assumed where there is no forest firefighting operation outside a plant, hence all potential forest fires arrive at the plant. The hazard curves remarkably increase around 40 to 80 times in frequency in comparison with those with considering the forest firefighting operation effect outside the plant. This study indicated that the most significant factor in the forest fire hazard risk is whether the forest firefighting operation outside the plant is expected before the forest fire arrival at the plant.

The objective of this study is to develop probabilistic risk assessment (PRA) methodology for combination event of low temperature and snow by focusing on decay heat removal system (DHRS) of sodium-cooled fast reactor. For this combination event, annual excess probability depending on the hazard intensity was statistically estimated based on the meteorological data. Event tree was developed by considering the impact of low temperature and snow on DHRS: e.g., plug at the air intake of ultimate heat sink and of emergency diesel generator and unopenable door for access route due to accumulated snow, failure of air intake filter due to deposited snow, possibility of freezing of cooling circuits. Recovery actions (i.e., snow removal and filter replacement) were considered in the event tree. By using annual excess probability and failure probabilities, the event tree was quantified and the quantified event tree showed that dominant core damage sequence is loss of access route for snow removal against the combination event at daily snowfall depth of 3m/day continued during 24h. When the access route is secured for all hazard intensity assumed this paper, it was showed that core damage frequency decreases about 1/10 times and securement of the access route has significant sensitivity to the core damage frequency.

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

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.

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.

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.

Wall pressure measurements were conducted for a 90 degree elbow of which the axis curvature coincided with its inner diameter (125 mm). Reynolds numbers examined were 1.0 × 105 (available only for steady components), 3.2 × 105 and 5.0 × 105. Results showed that distributions of fluctuating normalized pressures obtained here and those made by Shiraishi et al. (2006) for the Reynolds number of 3.25 × 10 6 coincided within 0.04 of the dynamic pressure. These distributions had the same tendency: The strong fluctuating region existed in the curvature inside and had concave/convex shapes at the upstream/downstream ends, respectively. Power spectral density functions of fluctuating pressures mostly exhibited the slope of the minus seven-third law, which is seen in the inertial range of turbulence, at large frequencies. The peak spectrum with the Strouhal number of 0.5 could be found in the curvature inside downstream of the elbow. They corresponded to the vortex shedding from the boundary layer developed in the inner and aft part of the elbow. The slope at large Strouhal numbers became negatively steep near the region where the peak spectrum was seen. The peak intensity having the Strouhal number of 0.5 was quantitatively in accordance with that of the data obtained in the experimental setup that Shiraishi et al. used (Yamano et al., 2011), suggesting that the law of dynamical similarity could be applied with regard to this oscillation. Cross correlations of pressure fluctuations showed that the pressure fluctuation having the Strouhal number of 0.5 propagated as a planar wave with the bulk velocity.

This paper describes experimental analyses using SIMMER-III and -IV, which are respectively two- and three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of integral code validation. Two topics of key phenomena in core disruptive accidents of sodium-cooled fast reactors are presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III and -IV computer code for the duct wall failure and fuel discharge/relocation behavior.

Studies on the self-leveling behavior of debris bed are crucial in the assessment of core disruptive accident (CDA) that could occur in sodium-cooled fast reactors (SFR). To clarify the mechanisms underlying this behavior, several series of experiments were elaborately designed and conducted within a variety of conditions in recent years, under the close collaboration between Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The current contribution, including knowledge from both experimental analyses and empirical model development, is focused on a recently developed comparatively larger-scale experimental facility using gas-injection to simulate the coolant boiling. Compared to the previous investigations, this facility can achieve a much wider range of gas velocities (presently up to a flow rate of around 300 L/min). Based on the experimental observation and quantitative data obtained, influence of various experimental parameters, including gas flow rate (～ 300 L/min), water depth (180 mm and 400 mm), bed volume (3 ～ 7 L), particle size (1 ～ 6 mm), particle density (beads of alumina, zirconia and stainless steel) along with particle shape (spherical and non-spherical) on the leveling is checked and compared. As for the empirical model development, aside from a base model which is restricted to predictions of spherical particles, the status of potential considerations on how to cover more realistic conditions (esp. debris beds formed with non-spherical particles), is also presented and discussed. This work, which gives a large palette of favorable data for a better understanding and an improved estimation of CDAs in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.