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Numerical analyses were conducted to replicate several tests for simulating a double-ended cold-leg large break loss-of-coolant accident (LBLOCA) in the Loss-of-Fluid Test (LOFT) using the TRACE (version 5/patch level 4) code. Analytical results by the original TRACE code were so conservative that especially a first peak of cladding temperature was estimated higher than the experimental data at the blowdown phase and subsequent temperature drop corresponding to the temporal quench was not seen. We were interested in minimum film boiling temperature (Tmin) as a heat transfer model factor estimating the quench at the moment, investigated correlation equations for Tmin in previous studies and especially focused on ones given as a function of coolant mass flow because the complicated flow transient and decompression in the core region at the blowdown phase was interpreted as having an influence on the cladding temperature behavior. There are several correlations meeting the above condition but it was revealed that they are insufficient to apply for high pressure especially. Therefore, a new term including an effect of mass flow flux and time derivative of pressure was defined and added with a proportional coefficient hypothetically to the current correlation in the TRACE code for modification. The LOFT analyses were conducted again using the modified TRACE code, and it was shown by applying roughly the same proportional coefficient to all the cases of LOFT analyses that estimation of the cladding temperature behavior was improved more precisely at the blowdown phase. Also, the transition during the phase was explained phenomenologically with the wall heat transfer mode and boiling curve.

As an accident management procedure of light water (nuclear) reactors which are situated along sea shore, sea water will be injected into the reactor pressure vessel to flood the nuclear fuel which is heated by residual heat. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Precipitation behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with condensed (two and half times denser) sea water and a mixture solution of sea water and borated water. Three-dimensional salt-precipitation distributions in the rod bundles were quantified with X-ray CT system. For both solutions, salt precipitated downstream and close to the top of active fuel (TAF) height where the void fraction is the highest. The condensed sea water yields wider precipitation region in height direction than mixture solution does. Mixture solution may give localized precipitates at the same height, which is just below TAF and uniformly spread on the horizontal plane. For both solutions, dryout concentration is larger as collapsed solution level is higher. This is because that lower collapsed solution level gives longer boiling-length and higher void-fraction, which results in larger amount of salt precipitations. The proposed salt concentration is useful to evaluate dryout concentration, which is the almost constant salt concentration for heat flux levels within the experimental ranges.

Rapid thermal elevation in boiling water reactor (BWR) is an important factor for nuclear safety and there is a need to develop an analysis code for the transient phenomenon and its validation process. To evaluate the thermal property of transient boiling and its uncertainty, corroborative experimental information is crucial. In particular, the lateral propagation behavior of a vapor bubble (void) in the cross-sectional direction of fuel assembly has yet to be determined. This study evaluates the void propagation behavior in a 5 × 5 rod bundle with cross-sectional heat distribution that causes only the 3 × 3 rod bundle to generate heat; assuming rapid heating under atmospheric pressure. In this paper, using the maximum heat output applied to the nine heated rods as a parameter, from the visualization of the void behavior and the measurement of the local void fraction, the heat output conditions under circumstances where lateral propagation of voids occurs and where voids are only localized in the heated region are summarized. We quantified the time difference initially detected and the time-averaged void fraction according to the lateral propagation level.

The bubble pairing scheme was devised to quantify three-dimensional velocity of each bubble. We used two sets of Wire-Mesh Sensors to identify locations of each bubble according to bubble identification algorithm, which was developed by HZDR. The devised scheme was applied to the vertical upward air-water flow at 0.64 m/s for both air and water superficial velocities in a large diameter pipe (i.d. 224 mm). The bubble pairing scheme visualized the developing process of two-phase flow: large bubbles coalesced with each other to move toward the center, while the rest of bubbles broke up into smaller bubbles and decelerated.

In order to obtain high-resolution data for modelling of boiling two-phase flow and its validation, we designed and constructed a test loop with a vertical annulus flow path and conducted subcooled boiling experiments to investigate subcooled bubble incipience and its development process under atmospheric condition. Three kinds of the state-of-the art measurement techniques were applied to quantify key parameters such as radial and vertical distributions of void fraction, bubble velocity, interfacial area concentration (IAC), Sauter mean diameters, high-resolution temperature distribution on rod surface, bubble transport behavior, and turbulent velocity components as well as onset of nucleate boiling (ONB), and onset of significant void (OSV).

Natural circulation is a key technology for developing the molten core cooling system without an external power source from the lessons of the severe accident at Fukushima-Daiichi Nuclear Power Station. This study is devoted to quantify the void fraction which is an important parameter for the driving force of natural circulation flow, and to evaluate the effect of the void fraction correlation on the prediction accuracy of the natural circulation flow rate. Test was conducted at atmospheric pressure and room temperature, using the upward air–water two phase flow. Vertical tubes with an inner diameter of 36 and 25 mm were used as the test section. The void fraction was measured by three different methods: quick-closing valve method, pressure drop method, and conductive void-probe method. The following conclusions are obtained from this study: (1) The data of the natural circulation flow rate, void fraction and pressure drop for the upward air–water two phase flow at atmospheric pressure and room temperature were obtained to develop and verify the new model. (2) By improving the void correlation, it was found that the prediction accuracy of the natural circulation flow rate could be improved by about 10% to 5%, that is, the prediction error can be halved in the range of this study. (3) The natural circulation flow rate for 25 mm test section was saturated with increasing the air flow rate at higher air flow condition. The model cannot predict this tendency. From the point of design of the actual molten core cooling system, the model improvements in this region are necessary in the future.

We have established an experimental system to visualize a droplet flow in a simulated BWR fuel sub-channel optically and measure the diameter and velocity of droplet after passing through a spacer. For representative spacers of ferrule and grid type, an effect of them on downstream droplets was evaluated with the experimental system. When a ferrule type spacer was simulated and implemented in both the center and side sub-channel, the vertical velocity of droplets got faster especially in the range of small diameter compared to the case of no spacer. When a grid type spacer was simulated and implemented in the center sub-channel especially, a large dispersion of vertical velocity of droplets occurred especially in the range of small diameter compared to the case of no spacer. By a computational fluid dynamics analysis for gas phase flow to drive the droplets in the sub-channel, it was confirmed qualitatively that the characteristics of droplet behavior observed in this experiment were dependent on the structure and geometry of spacer and sub-channel. Furthermore, it was revealed that a relation between a droplet diameter and velocity can be organized with a non-dimensional function derived from a momentum equation of particle in driving fluid and its drag coefficient has linear correlation with a gas Froude number.

Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

During a severe nuclear power plant accident, the release of fission products into containment and an increase in containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released. Pressure suppression pools, containment spray systems and a filtered containment venting system (FCVS) reduce containment pressure and reduce the radioactive release into the environment. These devices remove radioactive materials via various mechanisms. Pressure suppression pools remove radioactive materials by pool scrubbing. Spray systems remove radioactive materials by droplet−aerosol interaction. FCVS, which is installed in the exhaust system, comprises multi-scrubbers (venturi-scrubber, pool scrubbing, static mixer, metal−fiber filter and molecular sieve). For the particulate radioactive materials, its size affects the removal performance and a number of studies have been performed on the removal effect of radioactive materials. This study has developed a new means of evaluating aerosol removal efficiency. The aerosol number density of each effective diameter (light scattering equivalent diameter) is measured using an optical method, while the decontamination factor (DF) of each effective diameter is evaluated by the inlet outlet number density ratio. While the applicable scope is limited to several conditions (geometry of test section: inner diameter 500 mm × height 8.0 m, nozzle shape and air-water ambient pressure conditions), this study has developed a numerical model which defines aerosol DF as a function of aerosol diameter ( d ) and submergences ( x ).

Steam explosion experiments with a melt layer spreading at the bottom of a shallow water pool, namely the PULiMS-E6 and SES-S1 by KTH, Sweden, were simulated by the steam explosion simulation code, JASMINE. The observed impulses in the experiments were successfully reproduced by simulations with assumed premixing conditions. With those simulation results, the adequacy of the kinetic energy evaluation method used for the experiments were examined by comparison of the kinetic energy directly obtained in the simulation, E k , and the one evaluated based on the impulse and the water mass limited to the center area above the premixing zone, E kic . It showed that the impulse based kinetic energy evaluation gives about five times overestimation. The impact of the water pool geometry on the validity of the impulse based kinetic energy evaluation method was further examined by a parametric study with variations of the pool geometry in the simulations of PULiMS-E6 and SES-S1 as well as high pressure bubble expansion simulations. The results for the relation of E kic / E k and the geometric factors were consistent between the cases for the experiments and the bubble expansion. The results showed that: (1) for the shallow water pool regime, E kic / E k shows a trend of convergence to 4–5, (2) for deep water pool regime, the impulse based kinetic energy evaluation with the whole water mass, E ki , rather than E kic , gives a good estimation. A set of empirical formulas was obtained for E kic / E k .

In order to evaluate residual the stress distribution in oxides formed on zirconium alloys, synchrotron X-ray diffraction (XRD) was performed on the oxides formed on Zircaloy-2 after autoclave treatment at a temperature of 360° C in pure water. The use of a micro-beam XRD and a micro-sized cross-sectional sample achieved the detailed local characterization of the oxides. The oxide microstructure was observed by TEM following the micro-beam XRD measurements. The residual compressive stress increased in the vicinity of the oxide/metal interface of the pre-transition oxide. Highly oriented columnar grains of a monoclinic phase were observed in that region. Furthermore, at the interface of the post-first transition oxide, there was only a small increase in the residual compressive stress and the columnar grains had a more random orientation. The volume fraction of the tetragonal phase increased with the residual compressive stress. The results are discussed in terms of the formation and transition of the protective oxide.

When there is no power for cooling the spent fuel pool and conditioning the air in a boiling water reactor (BWR) reactor building, water vapor is generated from the pool and it affects the atmosphere in the building. To consider the impact of the steam in preparing emergency operation procedures, the building atmosphere under various conditions is to be evaluated with reasonably low computational cost. A lumped parameter model to predict the transient behavior of the building atmosphere was developed, in which the evaporation from the spent fuel pool and the condensation to the wall were taken into consideration. A transient behavior of temperature and vapor concentration in a BWR operating floor was predicted with the model. The results and the prediction speed were compared to those of a three-dimensional computational fluid dynamic calculation, and it was confirmed that the model could obtain almost the same results about 280,000 times faster. Parameter studies are conducted with the model, and dominan...

Given the severe accident of a light water reactor (LWR), stratification and solidification/melting are important phenomena in melt corium behavior within the reactor lower head, influencing the decay heat distribution and ablation of penetration tube and vessel wall. Numerical calculation is a necessary and effective approach for mechanistic study of local melt corium behavior. In this study, the improved moving particle semi-implicit (MPS) method was applied for investigating the stratification and solidification/melting phenomena. The implicit viscous term calculation technique and stability improvement technique were adopted to enable MPS to simulate the stratification process of materials with high viscosity in phase transition stage. The solid–liquid phase transition model was also coupled with MPS method. The validation experiment was carried out with low-melting-point metal tin and NeoSK-SALT. The layer configurations and temperature profiles obtained from MPS calculation showed good agreement with the experimental results. Meanwhile, the calculation results indicated that the material freezing behavior could affect the layer formation, and the layer configurations also significantly influenced the temperature profiles and heat flux distributions. The present results demonstrated that MPS method has the capacity to understand the local melt behavior in detail that is relevant to stratification and phase transition.

An analytical model was developed to describe thermal stratification in a primary containment vessel (PCV) and transient thermal-hydraulics coupled with a reactor pressure vessel (RPV) using TRACE code version 5.0 patch level 4. Geometries of a dry well (D/W) and a suppression chamber (S/C) were represented by a nodalization of TRACE code to simulate multi-dimensional flow in the PCV. An additive loss coefficient (so called ‘K-factor’) was focused as a sensitivity parameter to limit flow rate in a pool. For the first step, a validation analysis was conducted against a steam discharge experiment of S/C. The TRACE result was in good agreement with the measurement and showed a thermally-stratified temperature distribution in the S/C pool. For the second step, an analysis to simulate the accident at Fukushima Daiichi Unit 3 power plant (1F3) was conducted. It was proved to be able to explain the pressure increase in the PCV at the beginning of accident by demonstrating thermal stratification in the S/C pool. Sensitivity study revealed an optimal K-factor value for a macroscopic viscous drag in a liquid phase fluid to demonstrate thermal stratification in a pool.

Stratification behavior is of great significance in the late in-vessel stage of core melt severe accident of a nuclear reactor. Conventional numerical methods have difficulties in analyzing stratification process accompanying with free surface without depending on empirical correlations. The Moving Particle Semi-implicit (MPS) method, which calculates free surface and multiphase flow without empirical equations, is applicable for analyzing the stratification behavior of fluids. In the present study, the original MPS method was improved to simulate the stratification behavior of two immiscible fluids. The improved MPS method was validated through simulating classical dam break problem. Then, the stratification processes of two fluid columns and injected fluid were investigated through experiments and simulations, using silicone oil and salt water as the simulant materials. The effects of fluid viscosity and density difference on stratification behavior were also sensitively investigated by simulations. Typical fluid configurations at various parametric and geometrical conditions were observed and well predicted by improved MPS method.

Cladding materials with superior corrosion resistance and anti-hydrogen pickup have been developed for high burnup nuclear fuel. We have suggested a surface modification of the cladding materials for this purpose and invented a new surface modification method “Fresh-Green”. The Fresh-Green treatment oxidizes and carbonizes a material surface in the same process. Zircaloy-2 with the Fresh-Green treatment showed the improvement of corrosion resistance in autoclave tests. In order to investigate the effect of surface modifications on the corrosion resistance, a synchrotron radiation experiment and a TEM observation were performed on different oxide layers formed on Zircaloy-2. The oxide layers were formed by air-oxidation, an autoclave test and the Fresh-Green treatment. Crystal structures of all the samples were transformed as Zr > Zr3O > tetragonal ZrO2 > monoclinic ZrO2 from the matrix to the surface. Columnar grains of monoclinic zirconia were arranged unidirectionally in the Fresh-Green oxide layer treated at a low temperature. Diffusing capacity for oxygen influenced the crystal structure of the oxide layers.

Experiments were conducted to investigate two-phase flow instabilities due to flashing in a boiling natural circulation loop with a chimney at low pressure. The SIRIUS-N facility was designed to have non-dimensional values nearly equal to those of typical natural circulation boiling water reactor (BWR). The observed instability is suggested to be flashing-induced density wave oscillations, since the oscillation period correlated well with the passing time of single-phase liquid in the chimney section regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before heating according to the stability map.

Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-I category, since they occur at low flow qualities, according to the Fukuda—Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and th...

In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liqui...

An analytical study is presented on the thermo-hydraulic stability of a boiling natural circulation loop with a chimney at low pressure start-up. The effect of flashing induced by the pressure drop in the channel and the chimney due to gravity head on the instability is considered. A method to analyze linear stability is developed, in which a drift-flux model is used. The analytical result of a stability map agrees very well with the experimental one obtained in a previous report. Instability does not occur when the heater power is too low to generate voids in the chimney and only natural circulation of single phase can be induced. Instability tends to occur when boiling occurs only near the chimney exit due to flashing. This instability phenomenon has some similarities with density wave oscillation, such as the phase difference of temperature between the boiling region and non-boiling region, and the oscillation period which is near to the time required for fluid to pass through the chimney. However, there are also some differences from density wave oscillation, such as the boiling region is very short, and pressure fluctuation can affect void fraction fluctuation.

The compact reversed shear tokamak CREST is a cost competitive reactor concept based on a reversed shear high β plasma and water cooled ferritic steel components. The moderate aspect ratio A = 3.4 and the elongation κ = 2.0 of CREST are very similar to the case of the ITER advanced mode plasma. Presentation of such a concept based on the ITER project should be worth while for formulating a fusion development strategy. The achievement of a competitive cost of electricity (COE) is the first priority for electric power industries. High β and high thermal efficiency are the most effective parameters for achieving a competitive COE. In order to achieve a high efficiency power plant, a superheated steam cycle has been adopted which permits a high thermal efficiency (η = 41%). Current profile control and high speed plasma rotation by neutral beam current drive stabilize the ideal MHD activity up to the Troyon coefficient βN = 5.5. A cost assessment has shown that CREST could generate about 1.16 GW(e) electric power at a competitive cost.

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for heat conduction in a fuel-rod and modal point kinetics of reactor neutronics using measured void fractions in reactor core sections of the thermal-hydraulic loop. In order to estimate decay ratios, an auto-regressive method has been successfully applied for time series data of the core inlet flow rate. Experiments were conducted with the SIRIUS-N facility for the rated operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of design parameters such as the power profile, void reactivity coefficients, core inlet sub-cooling, and the fuel rod time constant.

To investigate the stability of a boiling water reactor (BWR), the SIRIUS-F facility was designed and built for highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermal hydraulics-neutronics instabilities of the BWR. By using two sets of measured void-fraction distributions in a reactor core section of the SIRIUS-F facility, a real-time void-reactivity feedback simulation was performed on the basis of the modal point kinetics of reactor neutronics and fuel rod thermal conduction. A noise analysis method was performed to calculate decay ratios and resonance frequencies from dominant poles of transfer function based on the AR method using time-series measurement data of a core inlet flow of the facility. Channel and regional stability experiments were conducted for a wide range of operating conditions, including maximum power points along the minimum pump speed line and the natural circulation line of advanced BWR plants. The experimentally obtained decay ratios and resonance frequencies are in good agreement with those calculated by the linear stability analysis code ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics as a function of power and revealed a sufficiently large stability margin even under hypothetical power level conditions.

A detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-physical property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decrease of the condensation heat transfer coefficient, or increase of the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small when the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers within approximately 10 K. A simple correlation for the stability boundary is proposed by simplifying the above...