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A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should be less than 100 MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be less than 1.0. Unfortunately some criteria could not be satisfied, such as the maximum fuel centerline temperature and maximum cladding stress. Therefore the influence of average linear heat generation rate, Pu enrichment and compressive stress to yield strength ratio are analyzed. Finally the improved fuel rod design of the new core is proposed. [ABSTRACT FROM AUTHOR]

Abstract: The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. [Copyright &y& Elsevier]

Abstract: This paper focuses on the unsteady numerical simulations of the turbulent flow in tight lattice geometries with the explicit algebraic Reynolds stress model. The calculation results showed that the shape of the geometry, gap width and rod diameter as well as the Reynolds number were important parameters that would influence the characteristics of the flow (such as the dominant frequency and wavelength of the pulsation) and the turbulent mixing between the subchannels. And a modified correlation of turbulent mixing was proposed, which showed better agreement with the past experimental results compared to the other existing correlations, especially for the low Prandtl number fluids such as liquid sodium. [Copyright &y& Elsevier]

Abstract: This paper focuses on the unsteady numerical simulation of the turbulent flow in two types of geometry containing a narrow gap with the explicit algebraic Reynolds stress model. The model was first validated through the comparison of simulation results inside a rectangular channel containing a cylinder and the corresponding experimental data. The structures of the oscillation were correctly reproduced. Simulation of turbulent mixing between circular channels connected by a narrow gap was carried out with the validated model. Because of the influence of the strong anisotropic turbulent flow in the gap region, the mixing rate was dramatically enhanced by the cyclic and almost periodic flow pulsation. The calculation results of the turbulent mixing rate showed good agreement with the experiment and the maximum error was less than 15%. [Copyright &y& Elsevier]

Abstract: The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include 99Tc, 129I and 135Cs discharged from LWR or fast reactor. Their isotopes, such as 127I, 133Cs, 134Cs and 137Cs were also considered to avoid the separation. By loading the isotopes (99Tc or 127I and 129I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWeyear and 2.79%/GWeyear can be obtained for 99Tc and 129I, respectively. The transmuted amounts of 99Tc and 129I are equal to the yields from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135Cs and the effect of other cesium isotopes, 135Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99Tc and 129I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of 135Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the transmutation of 135Cs is a challenge not only for the Super FR but also for other commercial fast reactors. [Copyright &y& Elsevier]

Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum supercritical water-cooled reactor (SCWR) that is presently researched in a Japanese project. A preliminary core has been designed with 1.59E+06W/m3 of power density [1]. In order to ensure the fuel rod integrity, the fuel rod behaviors under the normal operating conditions are analyzed using FEMAXI-6 code. Three types of the limiting fuel rods, with the maximum cladding surface temperature (MCST), maximum power peak (MPP) and maximum discharge burnup (MDB), are chosen to cover all the fuel rods in the core. The power histories of these fuel rods are taken from the neutronics calculation results in the core design. The available design range of the fuel rod design parameters, such as the initial gas plenum pressure, gas plenum length, grain size and pellet-cladding gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900°C. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Compressive stress to yield strength ratio should be less than 0.2. (5) Cumulative damage fraction (CDF) on the cladding should be less than 1.0. Finally the improved fuel rod design is proposed. [Copyright &y& Elsevier]

Abstract: The supercritical-pressure water-cooled fast reactor (SWFR) is a fast spectrum supercritical water-cooled reactor (SCWR) studied by the University of Tokyo. The SWFR is designed as a two-pass core with an outlet temperature 500°C. The SWFR has fuel channels cooled by downward flow, higher power density, and smaller coolant density reactivity feedback compared with Super LWR. This paper describes the safety analyses of abnormal events for the SWFR. SPRAT-F code is used for the safety analysis at supercritical pressure considering the downward flow cooled seed fuel channel. This code is based on a 1-D node junction model with point kinetics and decay heat calculations. Flow redistribution among parallel paths is calculated by pressure-loss balance and momentum conservation. The initiating events are selected from those of LWRs. For the safety analysis, nine abnormal transients and four accidents are selected with considering types of abnormality. By the numerical analyses, it was found that the loss of flow events can be mitigated by the “water source” effect of the downward flow blanket channels in the abnormal transients and accidents. All the abnormal events satisfy the criteria with margin. [Copyright &y& Elsevier]

Abstract: A particle-grid hybrid method for simulating the onset of droplet entrainment is described. The air and liquid phases are expressed by grid and particle separately. The effects of surface tension and wall adhesion are calculated by the continuum surface force model. The fluid film variation and the droplet disengaging process are analyzed. The onset of entrainment criterion is obtained which has a good agreement with the correlation of Ishii and Grolmes (AIChE J., 21, 308–318). The influences of surface tension and gas phase density on the criterion are also studied. Several simulation flow fields are given for illustrating the efficiency of the method. [Copyright &y& Elsevier]

Abstract: In the present study, single steam bubble condensation behaviors in subcooled water have been simulated using Moving Particle Semi-implicit (MPS) method. The liquid phase was modeled using moving particles and the two phase interface was set to be a movable boundary which can be tracked by the topological position of the interfacial particles. The interfacial heat transfer was determined according to the heat conduction through the interfacial liquid layer and the coupling between momentum and energy was specially treated. Computational results showed that the bubble experiences various deformations at lower degrees of liquid subcooling while it remains nearly spherical at higher degrees of liquid subcooling. The bubble lifetime is nearly proportional to bubble size and is prolonged at higher system pressures. Bubble lifetime obtained from the MPS method agrees well with the experiments of , however it is lower than the predictions of . The underestimation is caused by severe bubble deformation at lower degrees of subcooling. The present study exhibits some fundamental characteristics of single steam bubble condensation and is expected to be instructive for further applications of the MPS method to evaluate more complicated bubble dynamics problems. [Copyright &y& Elsevier]

Abstract: In present study, the collapse of void bubble in liquid has been simulated using moving particle semi-implicit (MPS) code. The liquid is described using moving particles and the bubble–liquid interface was set to be vacuum pressure boundary without interfacial heat mass transfer. The topological shape of bubble can be traced according to the motion and location of interfacial particles. The time dependent bubble diameter, interfacial velocity and bubble collapse time were obtained under wide parametric range. The comparison with Rayleigh and Zababakhin''s prediction showed a good agreement which validates the applicability and accuracy on MPS method in solving present momentum problems. The potential void induced water hammer pressure pulse was also evaluated which is instructive for further material erosion study. The bubble collapse with non-condensable gas has been further simulated and the rebound phenomenon was successfully captured which is similar with vapor-filled cavitation phenomenon. The present study exhibits some fundamental characteristics of void bubble hydrodynamics and it is also expected to be instructive for further applications of MPS method to complicated bubble dynamics problems. [Copyright &y& Elsevier]

Abstract: This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal–hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal–hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied. [Copyright &y& Elsevier]

Abstract: Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650°C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156W/cm3 and its equivalent diameter is only 2.7m for 1000MWe class reactor core. High average core outlet temperature of 500°C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost. [Copyright &y& Elsevier]

According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident. [ABSTRACT FROM AUTHOR]

Highlights: [•] Safety of Super FR with all-upward two-pass flow scheme has been analyzed. [•] Current core design exhibits unique flow behavior with regard to two-pass. [•] Depressurization strategy is effective to induce flow for once-through SCWR. [ABSTRACT FROM AUTHOR]

Abstract: Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than −30pcm. [Copyright &y& Elsevier]

Abstract: Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5°C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3°C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31°C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681°C (650°C of nominal+31°C) within 95/95 limit for SWFR. [Copyright &y& Elsevier]

Abstract: Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k–ɛ high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface. [Copyright &y& Elsevier]

Abstract: Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25MPa) and high temperature (500°C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853°C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application. [Copyright &y& Elsevier]

Droplet entrainment and deposition are a couple of significant mechanisms for the heat transfer in annular two-phase flows existing in some heat exchange systems. The basic physics include the peeling of droplets from the liquid film due to high friction with the gas phase and the collision of droplets with the liquid film or deposition into the liquid film. Droplet deposition particularly plays a crucial role in the course of film dryout events which might have a vital importance for particular systems. In this study, a new numerical method (named as VOF-MPS hybrid method) based on the moving particle semi-implicit (MPS) method was developed to analyze the droplet deposition onto a stagnant thin liquid film. That proposed method combines the volume of fluid (VOF) solver of the open-source CFD code OpenFOAM with the MPS method. VOF-MPS technique introduces the surface tension force calculation of VOF model into the MPS method. MPS method formerly employed the continuum surface force (CSF) approach based on the particles. Three-dimensional (3D) VOF-MPS simulation with the novel surface tension modeling addition to the MPS-based modeling provides a smoother liquid-gas interface on the crown formed after the impact. Droplet deposition experiments were also carried out for the validation and comparison of two models. VOF-MPS method could predict the crown parameters such as crown thickness relatively better than the MPS method employing the CSF. However, the instabilities formed at the tip of the crown, observed in the experiments, could not be resolved with both methods. Graphical Abstract: [Figure not available: see fulltext.] [ABSTRACT FROM AUTHOR]

Abstract: In the present study, we preformed a two-dimensional numerical simulation of the motion and coalescence of bubble pairs rising in the stationary liquid pool, using the moving particle semi-implicit (MPS) method. Moving particles were used to describe the liquid phase and the vapor phase was evaluated using real vapor sate equation. The bubble–liquid interface was set to be a free surface boundary which could be captured according to the motion and location of interfacial particles. The behaviors of coalescence between two identical bubbles predicted by the MPS method were in good agreement with the experimental results reported in the literature. Numerical results indicated that the rising velocity of the trailing bubble was larger than that of the leading bubble. Both of the leading bubble and the trailing bubble rose faster than the isolated bubble. After coalescence, the coalesced bubble showed velocity and volume oscillations. The time of the volume oscillations increased with increasing initial bubble diameter. The wake flow and vortex would form behind the coalesced bubble. [Copyright &y& Elsevier]

Abstract: In the present study, two-dimensional numerical simulation of single bubble dynamics during nucleate flow boiling has been performed using moving particle semi-implicit (MPS) method. A set of moving particles was used to represent the liquid phase. The bubble–liquid interface was set to be a free surface boundary which can be captured according to the motion and location of interfacial particles. The interfacial heat transfer rate was determined by the energy variety of interfacial particles. The bulk liquid velocities investigated ranged from 0.07 to 0.3m/s. The surface orientations varied from vertical to horizontal through 60°, 45° and 30°. Bulk liquid subcooling varied from 0 to 6.5°C and wall superheat from 2.0 to 20.0°C. The computational results show that the bulk liquid velocity and surface orientation influenced the bubble diameter and liftoff time. Bubble would slide along the heater surface before lifting off and the sliding velocity at liftoff increased with an increase in bulk liquid velocity. Bubble dynamic was related to bulk liquid subcooling as well as wall superheat. The numerical results have been compared with the experimental data. [ABSTRACT FROM AUTHOR]