To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant (NPP) Unit-3 (from hereinafter, Unit-3), the meshless, Lagrangian Moving Particle Semi-implicit (MPS) method has been developed for the evaluation of debris-structure interactions in the pedestal region of Unit-3. The developed MPS method incorporated approaches to enhance numerical accuracy and stability, such as second-order corrective matrix and particle shifting technique, and the calculation cost reduction and efficiency improvement strategies to enable plant-scale simulations. A 1/10 scale numerical model of the pedestal region and structures within of Unit-3 and sensitivity analysis conditions were established for simulation cases. The simulation results showed that the convective vapor cooling from the debris surface in the pedestal region and the debris relocation amount/the time intervals between the relocations play important roles in the structure damage due to thermal attacks from the debris.

The use of different nuclear reactor systems as power source of deep-space explorers has been studied over the past decades as solar power cannot be expected in deep-space beyond Jupiter. In the preceding study, a cylindrical solid moderator reactor concept of about 500 kg in weight, which consists of 20% enriched UN fuel, YH1.5 moderator and Be reflector without using working fluid, was developed. In this study, we propose to further extend use of the developed core concept for the in-depth ice layer's investigation of Europa (one of the moons of Jupiter). At the end of the journey, the bare reactor is to be landed on the ice layer of Europa and sink down through the ice-layer as it melts the ice layer with its thermal power. For this purpose, this study aims to flatten the core power distribution to increase the core surface temperature for efficient ice melting while keeping the peak temperature of the core below a design limit. The core neutronics characteristics and power distributions are evaluated with neutron diffusion approximation and ablation and sinking behavior of the reactor through the ice layer is analyzed with Moving Particle Semi-implicit (MPS) method. Among different designs, the Hollow core could reduce the radial power peaking relatively well. The trial analysis by MPS method showed that modeling convective heat loss of the core surface and / or modifications to the core design may be necessary to prevent excess heat-up of the core before it sufficiently melts the ice and sinks down into the ice layer.

A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. A super fast reactor (Super FR) is a fast reactor type concept of SCWR. Unlike other SCWR core concepts, it adopts the single coolant pass flow scheme, in which the coolant passes the core only once from the bottom to the top without any reverse flows or preheating stages. In the meantime, reducing the core size tends to increase the core power peaking and reduce criticality. Therefore, the key issues with the small Super FR core design is reducing the core power peaking and achieving high average core outlet temperature with the single coolant pass scheme. This study aims to highlight the design issues through conceptual core designs of SMR class Super FR. To evaluate the core characteristics, three-dimensional coupled core calculations are carried out. The proposed design with small fuel assemblies, which are equivalent to those of boiling water reactors, attains a high core average outlet temperature of about 500 °C, which is compatible to that of typical large SCWR core design.

Part of this work was conducted under “Understanding Mechanisms of Severe Accidents and Improving Safety of Nuclear Reactors by Computer Science” of Institute for Advanced Theoretical and Experimental Physics and Waseda Research Institute for Science and Engineering.

Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water Reactor Fuel Performance Conference.All rights reserved. This paper introduces some highlights of the current progress of development of accident tolerant FeCrAl-ODS (oxide dispersion strengthened) fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. Both the experimental and the analytical studies in JFY2018 resulted in a successful step to develop the technical basis to introduce the FeCrAl-ODS fuel claddings to the current BWRs.

The potential reasons for the corium spreading difference over inert ceramic and reactive concrete channels of VULCANO VE-U7 experiment are investigated using Moving Particle Semi-implicit (MPS) method. A new thermal contact resistance model has been developed for MPS so that influence of the subscale heat transfer between the melt/crust and the substrate on spreading can be considered. The results indicate that the spreading difference is not much influenced by heat loss of the melt to different substrates, but more likely due to gas bubbles in the concrete channel. The most likely responsible gas bubble effect could not be well identified with the single channel analysis, because it could not consider the inflow mass interactions at the stabilization pool. The double-channel analysis with such consideration indicated enhancement of the effective thermal conductivity of the melt as the key influence of the gas bubbles that led to the difference.

In a severe accident of a light water reactor, the reactor pressure vessel (RPV) lower head may fail due to ablation at the vessel wall boundary involving natural convection of molten core materials. Accurate prediction of RPV lower head failure is essential for assessing severe accident progression and improving accident management because it greatly influences the subsequent ex-vessel accident progressions. However, there have been still large uncertainties about RPV lower head failure mode in the Fukushima Daiichi Nuclear Accident in 2011. The Lagrangian based MPS (moving particle semi-implicit) method has advantage of analyzing such phenomena involving complex interfaces and liquid-solid phase changes over other Eulerian mesh-based method. In the preceding study, small-scale Pb–Bi hemisphere vessel ablation experiment, with silicone oil as simulated molten core, was reproduced qualitatively by original MPS method. However, ablation mechanism associated with natural convection of the high temperature liquid could not be discussed because of significant influence of numerical discretizing error. In this study, the improved MPS method coupling corrective matrix in the particle interaction model which largely suppress the numerical fluctuation was adopted to analyze the experiment. The results show that the ablated metal relocation may enhance convective heat transfer in the downstream. As a result, ablation of the vessel wall extends from the level, close to the silicone oil surface down to the bottom of the vessel rather than previously simulated localized ablation near the silicone oil surface.

Super FR is the fast reactor version of SuperCritical Light Water Reactor (SCWR), which operates under supercritical condition of light water, so that designing the core with high coolant outlet temperature above the pseudo-critical point is possible. To raise the core outlet temperature of Super FR, axially heterogeneous core concept was proposed by the preceding study. It consisted of alternatively arranged two seed layers and two blanket layers (i.e., four-layers core). Furthermore, the core was axially divided to two sections and fuel shuffling was considered in each of the two axial sections independently to improve breeding performance. However, the study assumed the same fuel depletion history along the axial direction of the core, despite the large coolant density change. The core outlet temperature was only 387 C and the Compound System Doubling Time (CSDT) was about 98 years, which was not as short as desired. Hence, this study aims to reveal more comprehensive understanding of the concept of improving the performance of Super FR with fuel shuffling in multi-axial layers through design analyses. A five-layers core is considered, which consists of the lower blanket, lower seed, inner blanket, upper seed, and the upper blanket layers. Influences of the lower blanket/upper blanket discharge burnup on the core characteristics, such as the CSDT, void reactivity, and core outlet temperature have been studied with different coolant inlet temperatures for the first time. Different fuel depletion histories are considered for each of the axial layers with consideration of the axial coolant density change (which had not been considered in the preceding study). The present results show that it is preferable to keep the lower blanket discharge burnup relatively low, compared with that of the upper blanket to improve breeding performance with respect to achieving short CSDT. However, it leads to slight reduction in the core outlet temperature. Based on the sensitivity analyses, a representative design is proposed, which achieves CSDT of 74 years and average core outlet temperature of 492 C, which show significant improvement from the preceding design.

The Lagrangian nature of the moving particle semi-implicit (MPS) method brings two challenges: disordered particle distribution and particle clumping. The former can cause large random discretization error for the original MPS models while corrective matrix can effectively reduce such large error to the high-order truncation error. The latter can trigger instability easily and thus some adjustment strategies for stability are indispensable, thereby causing non-negligible stabilization error. The purpose of this paper is to compare the relative magnitude of the truncation and stabilization error, which is of great significance for future improvements. An indirect approach is developed because of the difficulty of separating different error from total error in dynamic simulations. The basic idea is to check whether the total error decreases significantly after the truncation error is further reduced. First, a second order corrective matrix (SCM) is proposed for MPS to reduce the truncation error further, as demonstrated by theoretical error analysis. Second, error analysis reveals that the first order gradient model produces less numerical diffusion than the second order gradient model in interpolation after particle shifting. Then, several numerical examples, including Taylor-Green vortex, elliptical drop deformation, excited pressure oscillation flow and continuous oil spill flow, are simulated to test the variance of total error after SCM is applied. It is found that the SCM schemes basically did not remarkably decrease the total error for incompressible free surface flow, implying that truncation error is not dominant compared to the stabilization error. Therefore, reducing the stabilization error is of more significance in future.

Liquid ZrO2 is one of the most important materials involved in severe accident analysis of a light-water reactor. Despite its importance, the physical properties of liquid ZrO2 are scarcely reported. In particular, there are no experimental reports on the viscosity of liquid ZrO2. This is mainly due to the technical difficulties involved in the measurement of thermo-physical properties of liquid ZrO2, which has an extremely high melting point. To address this problem, an aerodynamic levitation technique was used in this study. The density of liquid ZrO2 was calculated from its mass and volume, estimated based on the recorded image of the sample. The viscosity was measured by a droplet oscillation technique. The density and viscosity of liquid ZrO2 at temperatures ranging from 2753 K to 3273 K, and 3170 K–3471 K, respectively, were successfully evaluated. The density of liquid ZrO2 was found to be 4.7 g/cm 3 at its melting point of 2988 K and decreased linearly with increasing temperature, and the viscosity of liquid ZrO2 was 13 mPa at its melting point.

In a postulated sever accident of a light water reactor (LWR), molten core debris (corium) may breach the reactor pressure vessel and be released to the ex-vessel containment floor. A core catcher manages ex-vessel corium cooling by uniformly spreading the corium into a large space, but it requires a dedicated plant design. In contrast, corium shields have been back-fitted to some boiling water reactors to prevent excessive amount of corium to flow into sump pits, where effective corium cooling may be difficult. However, corium shields can only block the corium flow and cannot contribute to uniform spreading of the ex-vessel corium. This study proposes a preliminary concept of “Debris Spreading Floor”, which can be applied to any types of reactor plants including back-fitting to the existing plants. More specifically, the existing containment floor is overlaid with sacrificial material and refractory material is placed around sump pits. It is intended to allow the original function of sump pits to collect leaking water under normal, abnormal transient and design basis accident conditions. However, under postulated severe accident condition, spreading of ex-vessel corium is promoted by ablating itself with the hot corium and guiding corium spreading away from sump pits. To develop the concept, mechanistic analysis of corium spreading, which can consider influence of substrate ablation, is needed. The Moving Particle Semi-implicit (MPS) method is a Lagrangian particle method and thus suitable for mechanistic simulation of free-surface spreading flow involving solid / liquid phase change and interactions. In this research, effect of the proposed concept is firstly presented with the MPS simulations. Preliminary simulations in 2D show that, amount of corium flowing into sump pits is reduced by the concept. Secondly, validity of the MPS simulations is quantitatively discussed by simulating the experiments with simulant. The experiment was carried out by Central Research Institute of Electric Power Industry (CRIEPI) by pouring liquid Pb-Bi onto a Pb-Bi block so that the inflow liquid spreads on the block surface while it also ablates the block. Sensitivity analyses have been carried out with different initial conditions, calculation resolutions, subscale models and parameters of the MPS simulations to identify the key models and parameters for quantitative prediction of the melt / substrate interactions.

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th 2011 put
the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions
resulting from loss of offsite power and AC power. It is believed that the Station Blackout (SBO) and loss
of heat sink led to core meltdown in all Units 1–3. Despite past research efforts on the severe accident
progression in Fukushima NPP Units 1–3, there are still knowledge gaps and uncertainties existing in
understanding of the severe accident scenarios and consequences.
Hence, this study aims at identifying the modeling uncertainties and addressing the sensitivity parameters
in Fukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core
region has been developed to better simulate the thermal-hydraulic behavior of liquid water and steam,
which is considered to be crucial in simulating the core uncovery and degradation process. The boundary
conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High
Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were
determined based on the available reactor water level and pressure measurement data. The current study
suggested that the local vapor heatup behavior could influence the core melting and relocation behavior,
which can lead to different core degradation scenarios. With the current modeling assumptions in
MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 SRVs could have
remained open when the major core slumping took place at ca. 45:20 h (ca. 12:00, March 13) with 50 to
80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of
the AWI to the reactor core could have been only 15% as of reported by TEPCO with the current modeling
conditions if debris dryout was assumed to have occurred at around ca. 54.0 h (20:40 h, March 13th). As
for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-
Miller creep rupture model in the current MELCOR modeling.

Corium (lava-like mixture of fissile material) spreading prediction is of great significance in the severe accidents
of nuclear power plants. Crust formation due to solidification distinguishes corium spreading from common
isothermal spreading. The Lagrangian moving particle semi-implicit (MPS) method is potential for such
spreading flow with both free surface and crust-melt interface. Crust formation is usually represented by viscosity
escalation, but crust creeping is an associated problem. In the original MPS algorithm, creeping velocity
cannot be reduced steadily by the continuous increase of viscosity, owing to the numerical creeping. A new
solution algorithm is proposed for particle methods to eliminate such numerical creeping, so that creeping
velocity decreases proportionally with viscosity rise. In this situation, high enough viscosity can effectively
represent crust behaviors. Three numerical examples, leakage flow with high viscosity, dam break flow with low
viscosity and the VULCANO VE-U7 corium spreading experiment with both high and low viscosities simultaneously,
are investigated to contrast the performance difference between the original and new algorithms. It is
demonstrated that the current algorithm is suitable for crust formation in corium spreading.

Pressurized water reactor (PWR) is the reactor type with the most abundant operation experience in the world. However, studies on designing PWR-type fast reactors have been limited and there have not been any PWR-type fast breeder reactor design concepts. In this study, the concept of seed-blanket PWR-type breeder reactor with tightly packed fuel assembly (TPFA) has been developed by coupled three dimensional neutronics and thermal-hydraulics core calculations. For the seed-blanket heterogeneous core using mixed oxide (MOX) fuel and depleted uranium, it has been shown that the core height is limited to about 1.0 m or less, in order to satisfy the design criterion of negative void reactivity. Moreover, it has been shown that increasing the core power density is difficult, as it leads to substantial increase in the core pressure drop. Consequently, the core characteristics are featured by low breeding performance with compound system doubling time (CSDT) of about 150 years or more (fissile plutonium surviving ratio of below 1.01) and low thermal power of about 700 MW or less. For the core using 4.95 wt% enriched uranium for the blanket assembly, it is possible to improve the void reactivity characteristics, breeding performance and thermal power by reducing reactivity difference between the seed and the blanket fuel assemblies. By utilizing enriched uranium, the concept of breeding PWR core with CSDT of 60 years and thermal power of 1000 MW has been shown. In addition, PWR-type breeder reactor concept using enriched uranium that the fissile surviving ratio including uranium and plutonium exceeds 1 was shown for the first time.

The Lagrangian moving particle semi-implicit (MPS) method has potential to simulate free-surface and multiphase flows. However, the chaotic distribution of particles can decrease accuracy and reliability in the conventional MPS method. In this study, a new Laplacian model is proposed by removing the errors associated with first-order partial derivatives based on a corrected matrix. Therefore, a corrective matrix is applied to all the MPS discretization models to enhance computational accuracy. Then, the developed corrected models are coupled into our previous multiphase MPS methods. Separate stabilizing strategies are developed for internal and free-surface particles. Specifically, particle shifting is applied to internal particles. Meanwhile, a conservative pressure gradient model and a modified optimized particle shifting scheme are applied to free-surface particles to produce the required adjustments in surface normal and tangent directions, respectively. The simulations of a multifluid pressure oscillation flow and a bubble rising flow demonstrate the accuracy improvements of the corrective matrix. The elliptical drop deformation demonstrates the stability/accuracy improvement of the present stabilizing strategies at free surface. Finally, a turbulent multiphase flow with complicated interface fragmentation and coalescence is simulated to demonstrate the capability of the developed method.

The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquid-like high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. When the molten corium, which is generally comprised of UO2, ZrO2 and metals such as zircalloy and stainless steel, is discharged into a dry reactor cavity, a stratified molten pool configuration with two immiscible oxidic and metallic phases can be expected to form and lead to MCCI. Compared to a homogenous oxidic molten pool configuration, the metallic phase in the stratified molten pool might influence the crust formation on the corium-concrete interface and consequently cause different concrete ablation behavior to evaluate MCCI progression concerning containment failure. In terms of this issue, past experimental studies, such as COMET-L, VULCANO VBS and MOCKA test series, have been carried out to investigate the influence of such oxidic and metallic stratified pool configuration on MCCI. The experimental results have shown that the metallic phase can have a significant impact on the axial and radial ablation kinetics that could influence the ablation patterns of reactor pit. As regards numerical studies, past numerical modeling of MCCI was generally based on Eulerian methods and simplified empirical approach to simulate solid/liquid phase change and evolving of corium/crust/concrete interface. Such modeling might be efficient but have shown deficiencies and inadequacies due to its Eulerian and empirical nature, which has suggested a necessity to seek for a more mechanistic approach for modeling of MCCI. In this sense, Moving Particle Semi-implicit (MPS) method is considered suitable for MCCI analysis for its advantages of tracking interfaces and modeling phase change accurately as a Lagrangian particle method. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. Solid/liquid phase change was simulated with types of solid and liquid particles with thermal and physical properties including temperature and solid fraction, which enabled tracking of the solid/liquid status of each particle to achieve accurate free surface and corium/crust/concrete interface capturing. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

<p>A pedestal-floor structure of a nuclear-reactor containment-vessel was additive manufactured with two sump-pits and a doorway at a reduced scale of 1:100 and 1:50. A silicone-oil jet impinged and spread on the pedestal floor. Such three-dimensional flow was visualized from three cameras. A volume of fluid (VOF) simulation was successfully reproduced this three-dimensional spreading behavior, silicone-oil flow-thickness evolution and drainage weights.</p>

The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking.

To utilize the merit of supercritical water cooling, the Super FBR core concept, which is compatible with both the high breeding and the high enthalpy rise needs to be developed. One possible solution to meet such requirements may be to compose an axially heterogeneous core with MOX and blanket layers, with consideration of the large density change and specific heat of supercritical water at vicinity of the pseudocritical point. A new design concept of Super FBR has been proposed with "with multi-axial fuel shuffling", which has flexibility in designing fuel shuffling schemes in the lower part and upper part of the core independently. Fully coupled neutronics and thermal- hydraulics core calculations were carried out to investigate impact of designing independent number of fuel batches and shuffling patterns in the upper and the lower parts of the core. Promising results were obtained, showing possibility of improving the core breeding performance with respect to the compound system doubling time (CSDT) by reducing the reactor doubling time (RDT) with designing of the independent fuel shuffling. Moreover, reduction in the ex-core factor (EF) was shown to be possible with such independent fuel shuffling. The combined effects of reductions in RDT and EF showed significant reduction in CSDT. It is the first design concept of Super FBR with coolant enthalpy rise, which covers from the liquid like state (below the pseudocritical point) to the gas-like state (above the pseudo-critical point) of supercritical water. Further design investigations may be necessary to reduce CSDT and increase the average core outlet temperature.

In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the "immobilization solid fraction" of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination. (C) 2017 Elsevier Ltd. All rights reserved.

In two-dimensional (2-D) molten corium-concrete interaction (MCCI) experiments with prototypic corium and siliceous concrete, the more pronounced lateral concrete erosion behavior than that in the axial direction, namely anisotropic ablation, has been a research interest. However, the knowledge of the mechanism on this anisotropic ablation behavior, which is important for severe accident analysis and management, is still limited. In this paper, 3-5 simulation of 2-D MCCI experiment VULCANO VB-U7 has been carried out with improved Moving Particle Semi-implicit (MPS) method. Heat conduction, phase change, and corium viscosity models have been developed and incorporated into MPS code MPS-SW-MAIN-Ver.2.0 for current study. The influence of thermally stable silica aggregates has been investigated by setting up different simulation cases for analysis. The simulation results suggested reasonable models and assumptions to be considered in order to achieve best estimation of MCCI with prototypic oxidic corium and siliceous concrete. The simulation results also indicated that silica aggregates can contribute to anisotropic ablation. The mechanisms for anisotropic ablation pattern in siliceous concrete as well as isotropic ablation pattern in limestone-rich concrete have been clarified from a mechanistic perspective. (C) 2017 Elsevier B.V. All rights reserved.

The fuel support piece in a boiling water reactor (BWR) is used to brace fuel assemblies. The channel within the fuel support piece is determined to be a potential corium relocation path from the core region to the lower head during the severe accident of BWR. In the present study, the improved ***Moving Particle Semi-implicit (MPS) method was adopted to simulate the flow and solidification behavior of the melt in a fuel support piece. The MPS method was first validated against the Pb-Bi plate ablation test that was perfornied by CRIEPI. The predicted ablation mass of the plate agreed well with the experimental results. Then the flowing and freezing behaviors of molten stainless steel (SS) and zircaloy in the fuel support piece were simulated by MPS method with a three dimensional particle configuration, respectively. In this study, the flow and solidification behavior of SS was simulated first. After all the SS passed through the channel, the flowing behavior of Zr in the fuel support piece was simulated. The simulation results indicated that the crust layer formed on the inner surface of the fuel support piece during the melt discharging process. The fuel support piece was plugged by the solidified zircaloy particles in the lower initial temperature case. The fuel support piece kept intact in all the calculation that were performed under the assumed order of melt injection. The present results could help to reveal the progression of a BWR severe accident. (C) 2017 Elsevier Ltd. All rights reserved.

Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

Melting penetration information of Fe-U system is necessary for simulating the molten core behavior during severe accident in nuclear power plants. For Fe-U system, the information is mainly obtained from experiment, i.e. TREAT experiment. However, there is no reported data on SS304 at temperature above 1350°C. The MPS-LER has been developed and validated to simulate melting penetration on Fe-U system. The MPS-LER modelled the eutectic phenomenon by solving the diffusion process and by applying the binary phase diagram criteria. This study simulates the melting penetration of the system at higher temperature using MPS-LER. Simulations were conducted on SS304 at 1400, 1450 and 1500°C. The simulation results show rapid increase of melting penetration rate.

Anisotropic and isotropic ablation in molten corium-concrete interaction (MCCI) phenomenon was studied with the Moving Particle Semi-implicit (MPS) method by carrying out numerical simulations of CCI-2 and 3 experiments. The interaction of the fully oxidized PWR core melts with specially-designed two-dimensional limestone and siliceous concrete test sections was analyzed, focusing on investigating the two-dimensional ablation behavior with both limestone and siliceous concrete. The phase transition of molten corium and concrete was modeled based on a phase transition model for mixture. Slag film model and crust dissolution models were incorporated in the current MPS code to simulate the effect of gas generation and crust dissolution phenomena in limestone concrete. The effects of gas generation and aggregates on the concrete ablation behavior were investigated by simulating different specially designed cases. The simulation results by MPS method reproduced the isotropic and anisotropic cavity ablation profile and the overall axial and lateral ablation rates agreed well with the experimental measures. The experimental and MPS results both indicate that the crust on the corium-concrete interface can play an important part in concrete ablation process. The simulation results by MPS method also provide evidence to support the theory that aggregates are part of the cause of anisotropic ablation profile in cavity with siliceous concrete because aggregates could delay the axial basemat ablation more significantly than the lateral one and influence the power split in the melt pool. (C) 2016 Elsevier Ltd. All rights reserved.

A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries. (C) 2015 Elsevier B.V. All rights reserved.

A high breeding core of supercritical water cooled fast reactor (Super FBR) is designed with the tightly packed fuel assembly for obtaining a high breeding ratio with negative void reactivity. The coolant volume fraction is substantially smaller than that of the tight lattice fuel assembly in Super FRs. The present study conducted the safety analysis of this reactor for the abnormal transients and accidents at supercritical pressure. The safety system and safety criteria are similar to those of Super FRs. The accident "control rod ejection" gives the highest fuel cladding temperature and the highest peak pressure, although which are still within the limit of safety criteria. The overall results show that all the safety criteria are satisfied at the selected events. (C) 2015 Elsevier B.V. All rights reserved.

In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

The molten corium-concrete interaction (MCCI) phenomenon was analyzed with the Moving Particle Semi-implicit (MPS) method by carrying out a numerical simulation of CCI-3 experiment. The interaction of the fully oxidized PWR core melts with a specially-designed two-dimensional siliceous concrete test section was analyzed, focusing on investigating the anisotropy in cavity ablation with siliceous concrete. The phase transition of the melt and the concrete was modeled based on a phase transition model for mixture. The effects of aggregates on the lateral and axial ablation were investigated by simulating two specially designed cases: one with aggregates in concrete and the other without aggregates in concrete. The simulation results by MPS method reproduced the anisotropic cavity ablation profile and the overall axial and lateral ablation rates agreed well with the experimental measures. The experimental and MPS results both indicated that the crust on the interface between the basemat and melt pool played an important part in axial concrete ablation process by hindering the axial ablation. The simulation results by MPS method also provided an evidence to support the theory that aggregates were the cause of anisotropic ablation profile in cavity with siliceous concrete because aggregates could delay the axial basemat ablation more significantly than the lateral one and influence the power split in the melt pool.

In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

The innovative water reactor for flexible fuel cycle (FLWR) is an advanced reactor concept based on the well-developed light water reactor (LWR) technology. It is to be introduced in two stages to achieve effective and flexible utilization of the uranium and plutonium resources. In the first stage, the high-conversion-type reactor concept (HC-FLWR) is to be introduced, with a core that achieves a fissile Pu conversion ratio of 0.84. Then, in the second stage, the reduced-moderation water reactor (RMWR) concept can be introduced, with a breeder-type core that achieves a fissile Pu conversion ratio of 1.05. From the viewpoint of effective introduction of the high-conversion-type reactor, such as the introduction capacity of the reactor, HC-FLWR is required to further raise the fissile Pu conversion ratio to similar to 0.95.
This study aims to develop a new core design concept for the high-conversion-type core, HC-FLWR+, to achieve the higher fissile Pu conversion ratio of similar to 0.95 under the framework of UO2 and U-Pit mixed-oxide (MOX) fuel technologies for LWRs. For raising the fissile Pu conversion ratio and controlling the void reactivity characteristics of the core, the concept of FLWR/MIX fuel assembly, which uses MOX and enriched UO2 fuel rods, is utilized.
The relationships between the main design parameters and the core performance index parameters are clarified in this study. When the fuel rod diameter and the clearance range from 1.23 to 1.28 cm and 0.25 to 0.20 cm, respectively, under the same pitch of 1.48 cm, the fissile Pu conversion ratio and the core average discharge burnup range from 0.89 to 0.94 and 53 to 49 GWd/tonne, respectively (the fissile Pu conversion ratio and the burnup are subject to a trade-off). Furthermore, when U-235 enrichment in the UO2 fuel rods is increased from 4.9 to 6 wt%, the fissile Pu conversion ratio improves to 0.97.
From these relationships, two representative core designs with fissile Pu conversion ratios of 0.91 and 0.94 and one optional design with a ratio of 0.97 were obtained. Hence, the flexibility of HC-FLWR+ concept to achieve a higher fissile Pu conversion ratio of similar to 0.95 has been revealed. Together with the standard HC-FLWR design, the concept covers a wide range of needs on fissile Pit conversion ratio from 0.84 up to 0.97, with design variations that are expected to be within the scope of current boiling water reactor and MOX fuel technologies.

The idea of recycling minor actinides (MAs) with fast breeder reactors (FBRs) is an effective way to potentially reduce environmental burdens associated with nuclear energy production. For such FBR cores, it is necessary to find one or more promising MA loading methods that can effectively transmute MAs while minimizing deterioration of the core performance and reducing the overall fuel fabrication cost. In this study, the homogeneous MA loading core with 3 wt% MAs is used as a reference design to evaluate the impact of the americium (Am) target in-core loading on reactivity characteristics and unprotected loss-of-flow (ULOF) response of sodium-cooled mixed-oxide FBR.
The Am target loading core of this study is designed by roughly preserving the MA inventory of the homogeneous MA loading core while placing Am and curium (Cm) to the ring-shaped target region between the inner and the outer core regions with 20 wt% content.
This design can flatten core radial reactivity worth distributions and effectively reduce reactivity insertion into the core during ULOF compared with the homogeneous MA loading core. It also has relatively flat and stable radial power distributions, which allow a relatively large coolant flow rate to be distributed to the target region.
During ULOF, the power increase of the Am target loading core of this study is slower than that of the homogeneous MA loading core. The maximum fuel temperature of the target region does not become particularly high compared with that of the inner core, and it is much lower than the melting point. Hence, the proposed Am target in-core loading method does not have a significant influence on ULOF response of the core. It is promising from the viewpoints of the reactivity characteristics and ULOF response.

Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain an understanding of the conceptual design elements and specific analysis methods for supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference for engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology. © Springer Science+Business Media, LLC 2010. All rights reserved.

The advanced reactor concept innovative water reactor for flexible fuel cycle (FLWR) is being studied to achieve effective and flexible utilization of uranium and plutonium resources based on well-developed light water reactor (LWR) technology. In order to design and evaluate FLWR fuel rod behavior, uncertainties in FEMAXI-6 calculations and key models and parameters for predicting LWR MOX fuel rod behavior need to be evaluated. In this study, the test fuel data bases (TFDBs) obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX (IFA-514 rod-1). Based on the evaluation results, fission gas release (FGR), pellet densification, swelling, and relocation models were found to be particularly important. The FGR model has a relatively large uncertainty for predicting MOX fuel rod behavior. However, the uncertainties in the other models are within the range expected by the property variations of MOX fuels. Hence, the densification. swelling, and relocation models of FEMAXI-6 can be applied to MOX fuel analyses.

A detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting Of UO2 pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel. analysis code is used to evaluate the fuel rod integrities in abnormal transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during abnormal transients. (c) 2006 Elsevier Ltd. All rights reserved.

An equilibrium core for the High Temperature Supercritical-pressure Light Water Reactor, now called the Super Light Water Reactor (Super LWR), has been designed. The fuel assemblies loaded in the peripheral region of the core are cooled with descending flow to achieve a high average coolant Core outlet temperature. This flow scheme is compatible with a low leakage fuel loading pattern (LLLP) in which 3(rd) cycle fuel assemblies are loaded in the core peripheral region. Stainless steel is used for fuel rod claddings and for structural materials. Watts correlations are used for predicting heat transfer ill the core. They take into account the improved heat transfer for downward flow. It is found that the water rods with their downward flow need to be thermally insulated with thin ZrO2 layer to keep the moderator temperature below the pseudo Critical temperature and to reduce the thermal stress in water rod walls. Ail average coolant core outlet temperature of 500 degrees C is achieved. The effects of various heat transfer correlations oil the cladding surface temperature are evaluated.

The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 &DEG; C.
In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 &DEG; C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.
In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies. &COPY; 2005 Elsevier Ltd. All rights reserved.

The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500°C is accurately evaluated for the first time in the development of the SCLWR-H. The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in order to establish the concept of this unique reactor design. However, it can only be determined by three-dimensional core design method, taking into account the control rod patterns, fuel loading patterns, coupling of the neutronic and thermal-hydraulic calculations, and burnup distribution of each fuel assembly. The R—Z two-dimensional core design method used in previous studies could not model or evaluate such parameters with sufficient accuracy. In this study, a three-dimensional equilibrium core design method for the SCLWR-H is established. This method can accurately evaluate the average coolant core outlet temperature and has permitted a comprehensive equilibrium core to be developed, which satisfies all design criteria. The design criteria are maximum fuel rod cladding surface temperature of 650°C, maximum linear heat generation rate of 39 kW/m, and a positive water density reactivity coefficient. © 2005 Taylor &amp
Francis Group, Ltd.

This paper describes safety analysis of the high-temperature supercritical water-cooled thermal reactor with downward-flow water rods (called Super LWR) at supercritical pressure. Eleven transients and four accidents are chosen for the safety analysis considering types of abnormalities. The cladding temperature is taken as the important transient criterion instead of the heat flux ratio. The once-through cooling system and the downward-flow water rod system characterize safety of the Super LWR. “Loss of feedwater” is important because it is the same as “loss of reactor coolant flow” unlike BWR and PWR. However, the downward-flow water rods mitigate core heat-up before startup of the auxiliary feedwater system because they remove heat from the fuel channels by heat conduction and supply their water inventory to the fuel channels by volume expansion. During pressurization transients, the reactor power does not increase significantly unlike BWR due to no void collapse in single-phase flow and decrease in coolant density by flow stagnation in the once-through cooling system. All the transients and the accidents satisfy the criteria. Increases in the hottest cladding temperatures are about 50°C at transients and 250°C at accidents at maximum. The period of the high cladding temperature is very short at transients. © 2002 Taylor &amp
Francis Group, Ltd.

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis. © 2002 Taylor &amp
Francis Group, Ltd.

Overview of the conceptual design of the high temperature reactors cooled by supercritical water (SCR) is presented. It is being studied for 13 years at the University of Tokyo. The design was taken as the reference of HPLWR (high performance LWR) study of EU. It was selected as a Generation4 reactor. The research and development started worldwide. We studied major elements of reactor conceptual design and safety evaluation by developing computer codes. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LW R and supercritical FPP, in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The potential cost reduction in both reactor and turbines are the characteristics of the SCR. It should be noted that the SCR is the smallest reactor not in power, but in size. Supercritical water contains higher enthalpy than sub-critical water per volume. Once-through coolant cycle is simpler than the re-circulating one. The containment is smaller than that of BWR due to the small coolant enthalpy inside. These are the reasons why the SCR is the smallest in sizes. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. It is based on the experiences of LWR design and safety and supercritical fossil-fired power plant technology. It is the reactor version of the once-through boiler. Its development follows the history of boilers.

An SCLWR-H core is designed with 3-D coupled thermo-neutronic core calculations for the first time. The change in power distribution is reflected to the evaluation of water density distribution in the core. The thermal-hydraulic and neutronic calculations are alternatively carried out until convergence is obtained. In the 3-D core calculation, each fuel assembly is modeled and the burnup, power and temperature distributions are evaluated. Fuel enrichment, burnable poisons, reload pattern, control rod pattern and the coolant flow rate distribution are designed to give average core outlet temperature 500C.

The elements of design of high temperature LWR operating at supercritical pressure are described. It includes conceptual design of fuel, core, plant, safety, control and start-up systems. The concept is developed at the University of Tokyo by computer analysis. The feature of the reactor such as economic improvement and hydrogen production potential are described as well as the view from the theory of boiler innovation. The technical background of the concept is LWR and supercritical fossil-fired power technologies. The concept was selected as one of the Generation 4 reactor. The research and development in Japan and in the world are underway. Potential of further design improvement exists.