Authorship：Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

researchmap

Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

researchmap

Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

researchmap

Language：English   Publishing type：Research paper (scientific journal)   Publisher：Annals of Nuclear Energy  

Sodium-cooled fast reactors (SFRs) have high safety with an extremely low probability of a core disruptive accident (CDA). However, from a defense-in-depth perspective, the CDA sequence is still worth studying. Severe accidents in SFRs, such as unprotected loss of flow, might lead to a CDA during which fuel and some fission products could be instantly released from a large CDA bubble through potential leak paths in the top shield structure. Therefore, a reasonable prediction of dynamic behavior of a large-scale bubble within the sodium pool is vital for accurately evaluating the migration of source terms. In this study, we propose a simplified one-dimensional CDA bubble model that can handle heat and mass transfer in two-phase multicomponent materials in different computational domains during the rising of bubbles through the sodium pool toward the cover-gas region. In this model, an entrainment model based on the Rayleigh–Taylor instability and Kelvin–Helmholtz instability was used to explain the coolant entrainment through the gas/liquid boundary and jet fragmentation. The model also addresses the mitigation effect of non-condensable gases on the condensation of fuel, steel and sodium vapor at bubble interface. We evaluated the model using a past experiment on the expansion of a two-phase, large bubble with high pressure in a stagnant liquid pool conducted by Purdue University in the late 1970 s using a 1/7-scale model of Clinch River Breeder Reactor. Good agreement with the experimental data demonstrates that the developed model can reasonably represent the essential characteristics of dynamic behavior of a large, high-pressure bubble with heat and mass transfer in two-phase multicomponent materials. This is valuable for evaluating the migration of the source terms, which will be carried out in future studies.

DOI： 10.1016/j.anucene.2024.110567

Web of Science

Scopus

Language：English   Publishing type：Research paper (international conference proceedings)  

researchmap

Language：English   Publishing type：Research paper (scientific journal)   Publisher：ASME International  

Abstract

Following the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident (CDA) prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.

DOI： 10.1115/1.4056834

Web of Science

Scopus

researchmap

Language：English   Publishing type：Research paper (scientific journal)   Publisher：ASME International  

Abstract

A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.

DOI： 10.1115/1.4056854

Web of Science

Scopus

researchmap

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

Feasibility of a concept of innovative subassembly-type passive-reactor-shutdown device has been studied, targeting to strengthen safety-"diversity" and -"robustness" of measures to prevent core damage accidents in sodium-cooled fast reactors. We have investigated target measures to achieve inherent safety capability under unscrammed (Anticipated Transient without Scram; ATWS) events in a 750MWe class mixed oxide-fuel fast reactors. As the countermeasure to prevent occurrence of core disruptive accidents (CDAs), we have built a basic proposal of this passive device designs, taking into accounts for engineering restrictions to be required in some design phase. Two types of the devise subassembly are discussed in this work; one device utilizes metal-fuel-alloys and another device salt compound to meet required passive capability. In this study we have determined the basic specifications of device fuel materials for alloy-type Pu-U-Fe alloy and salt-type (U-Pu) Cl3, respectively. Ternary Pu-U-Zr alloy is selected for the candidate fuel materials used in the pre-heating pins placed within this device subassembly. Through the studies, it has been suggested that the effectiveness and applicability of U-Pu-Fe alloys and low-enriched U (LEU) -Fe alloys as device fuels span a wide range of fast reactors to enhance safety tolerances against CDAs.

DOI： 10.1299/jsmeicone.2023.30.1811

Scopus

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

A new subassembly-type passive reactor shutdown device has been proposed to expand the versatility and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). This device can passively provide a large negative reactivity to the core by rapidly transferring the device fuel, which liquefies as the core temperature rises during an accident, to the lower plenum region of the device pins using only simple physical phenomena such as gravity falls. The fuel used in this device is assumed to be a metal alloy or chloride with the characteristics of fast reactor fuel and a relatively low melting point. In this study, the transient response analysis of the initiating phase during a typical unprotected loss of flow (ULOF) event was performed for a device loaded core of 750 MWe-class MOX fuel SFR, and the effect of different device fuel materials on the event termination was investigated. The results indicate that, no matter what device fuel material is used, it is expected to be possible to terminate the ULOF event without coolant sodium boiling in the core during the initiating phase of the event by replacing about 30 of the 286 fuel subassemblies in the core with device fuel subassemblies.

DOI： 10.1299/jsmeicone.2023.30.1582

Scopus

Language：English   Publishing type：Research paper (international conference proceedings)   Publisher：The Japan Society of Mechanical Engineers  

<p>The probability of core disruptive accident (CDA) occurrence in sodium-cooled fast reactors (SFRs) is considered extremely low. However, for further verifying the safety and reliability of SFRs, the CDA sequence is still worth studying. In a case of SFR’s severe accident, such as unprotected loss of flow (ULOF), the CDA may be triggered, and then fuel and some fission products (or called source terms) may be released instantaneously from a CDA bubble through the potential leak paths on the vessel top slab or released with a delay from boiling sodium pool after vessel melt-through, widely known as instantaneous and delayed source terms. Therefore, reasonable prediction of CDA bubble behavior is necessary to investigate instantaneous source terms migration in the vessel pool. In this study, a simplified one-dimensional CDA bubble model that includes the formulation of thermal-hydrodynamic behaviors of the bubble mixture rising through the sodium pool toward the cover-gas region is proposed. The model includes mass transfer processes such as the condensation of gas mixture on liquid fuel/steel/sodium and the bubble interface. In this model, droplet entrainment phenomena at bubble interface are modeled based on Rayleigh-Taylor instability and Kelvin-Helmholtz instability, and the effect of non-condensable gas on condensation process is also considered. To validate the developed model, a past experiment on the expansion of high-pressure bubble in a stagnant liquid pool conducted by Purdue University in the late 1970’s using a 1/7-scale model of Clinch River Breeder Reactor was analyzed. The results showed generally good agreement with measured data and demonstrate that the developed model can reasonably represent the essential characteristics of dynamic behaviors of a high-pressure large-size bubble with heat and mass transfer at the bubble interface. This supports subsequent calculations to carry out the migration of transient source terms in the future.</p>

DOI： 10.1299/jsmeicone.2023.30.1048

CiNii Research

Language：English   Publishing type：Research paper (international conference proceedings)   Publisher：The Japan Society of Mechanical Engineers  

<p>The in-pile EAGLE ID1 test was conducted by Japan Atomic Energy Agency to demonstrate the effectiveness of the fuel assembly with an internal duct structure during a core disruptive accident in a sodium-cooled fast reactor. Post-test analysis using the SIMMER-III code are based on a multi-fluid model that uses empirical models in constitutive equations, making it difficult to accurately simulate multi-component, multi-phase flows with complex heat and mass transfer. In this study, a new computational fluid dynamics code based on the fully Lagrangian particle method was developed for the purpose of clarifying the failure mechanism of the inner duct wall of FAIDUS. The three-dimensional simulation of the ID1 test was performed to analyze a series of thermal hydraulic behaviors leading up to duct wall failure for a computational domain that included six fuel pins, i.e., 1/12.5 of the circumference of the test section. The simulations reasonably reproduced the heat transfer characteristics observed in the test, showing that the local contact of liquid steel with high thermal conductivity with the duct wall greatly enhances the heat transfer from the nuclear heating fuel to the duct wall. The present large-scale simulation produced the results that were essentially equivalent to those obtained in a smaller simulation system with three fuel pins in our previous work. The results support the validity of the conclusions of our analytical study regarding the molten pool-to-duct wall heat transfer mechanism that caused the thermal failure of the duct wall.</p>

DOI： 10.1299/jsmeicone.2023.30.1062

CiNii Research

Language：English   Publishing type：Research paper (scientific journal)   Publisher：Elsevier BV  

The ID1 test was the final target test of the EAGLE experimental framework program. It was used to verify that during a core disruptive accident, the molten fuel could be discharged via wall failure of an inner duct in FAIDUS, a design concept for the sodium-cooled fast reactor. The ID1 results revealed that the wall failure behavior owed to the large heat flow from the surrounding fuel/steel mixture. The present study numerically investigated the heat transfer mechanisms in the test using the finite volume particle method in the three-dimensional domain. The thermal hydraulic behaviors during wall failure were reproduced reasonably. The present three-dimensional simulation mitigated inherent defects of our previous two-dimensional calculation and clarified that the solid fuel and liquid steel close to the outer surface of the duct can expose the duct to high thermal loads, resulting in the wall failure.

DOI： 10.1016/j.anucene.2022.109389

Web of Science

Scopus

researchmap

Language：English   Publishing type：Research paper (international conference proceedings)  

researchmap

Authorship：Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (other academic)  

Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)   Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

A new subassembly-type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device can passively provide a large negative reactivity to the core. In this study, the nuclear and thermal properties of the device subassembly with metallic fuel were evaluated to determine the basic device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWe-class MOX-fueled SFR core showed that a conventional homogeneous core can maintain stable cooling of the core prior to coolant boiling in the driver fuel subassemblies. On the other hand, the negative reactivity required to terminate the event by device operation was found to be slightly larger in the low sodium void reactivity core than in the conventional homogeneous core.

DOI： doi.org/10.1115/ICONE29-91812

Scopus

Authorship：Lead author,　Corresponding author   Language：Others   Publishing type：Research paper (scientific journal)   Publisher：MDPI AG  

The ability to predict critical heat flux (CHF) is of considerable interest for high-heat equipment, including nuclear reactors. CHF prediction from a mechanistic model for subcooled flow boiling in rod bundles still remains unsolved. In this paper, we try to predict the CHF in an annulus, which is the most basic flow geometry simplified from a fuel bundle, using a liquid sublayer dryout model. The prediction is validated with both water and R113 data, showing an accuracy within ±30%. After the CHF in an annulus is calculated successfully, a near-wall vapor–liquid structure is proposed on the basis of the liquid sublayer dryout model. Modeling of heat transfer modes over the heating surface at CHF is performed, and predictions of the changes in liquid sublayer thickness and heater surface temperature at the CHF occurrence point are carried out by solving the heat conduction equation in cylindrical coordinates with a convective boundary condition, which changes with the change in flow pattern over the heating surface. Transient changes in the liquid sublayer thickness and surface temperature at the CHF occurrence point are reported.

DOI： 10.3390/fluids7070230

Web of Science

Scopus

researchmap

Authorship：Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

Language：Others   Publishing type：Research paper (international conference proceedings)  

Authorship：Corresponding author   Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.nucengdes.2021.111494

Language：Japanese   Publishing type：Research paper (scientific journal)  

Toward mechanistic evaluation of Critical Heat Flux in nuclear reactors (2)

DOI： 10.3327/jaesjb.63.12_820

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.anucene.2021.108512

Authorship：Corresponding author   Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1080/00223131.2021.1931518

Language：English   Publishing type：Research paper (international conference proceedings)  

DOI： doi.org/10.1115/ICONE28-64099

Authorship：Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

DOI： doi.org/10.1115/ICONE28-64022

Language：English   Publishing type：Research paper (other academic)  

DOI： doi.org/10.1115/ICONE28-61469

Authorship：Corresponding author   Language：Others   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.anucene.2020.107867

Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.anucene.2020.107856

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English  

Language：English  

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.anucene.2018.08.047

Language：English  

Language：English  

Language：English  

Language：English   Publishing type：Research paper (international conference proceedings)  

Authorship：Lead author,　Corresponding author   Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English  

Language：Japanese   Publishing type：Research paper (scientific journal)  

Effect of Seawater on Heat Transfer without Boiling in Internally Heated Annulus
<p> Seawater was injected into the reactors during the accident at TEPCO's Fukushima Daiichi NPS. However, the effects of the seawater on the cooling performance of the fuel rods and fuel debris are not clear. As possible effects, the change in the physical properties of the coolant and the sea salt deposition on a heat transfer surface and in the coolant are considered. We conducted thermal-hydraulic experiments using an internally heated annulus to determine the effects of seawater under conditions without boiling. The same experiments for water and sodium chloride (NaCl) solution were also conducted for the purpose of comparison with the artificial seawater. In these experiments, considering the physical properties of the artificial seawater, the thermal-hydraulic behaviors of the artificial seawater under forced convection (Re>2300 [－]) was estimated from the Dittus-Boelter correlation although sea salt was deposited in the fluid. According to the results of particle image velocimetry (PIV), the velocity distribution in the artificial seawater was NOT different from that in the water and the NaCl solution. For a mixed convection regime, the Nusselt number of the artificial seawater was obtained from the correlation of the Grasholf number, Reynolds number and Prandtl number, as well as those for the water and the NaCl solution. Therefore, considering the physical properties of the artificial seawater, the thermal-hydraulic behavior of the seawater in single-phase flow can be estimated from the conventional thermal-hydraulic correlations for a single-phase flow.</p>

DOI： 10.3327/taesj.j15.024

Language：Others   Publishing type：Research paper (scientific journal)  

For a steam generator (SG) in a commercialized sodium-cooled fast breeder reactor (FBR), flow instability in the water side is one of the most important items needing research. As the first step of this research, thermal-hydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop, and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steam-water pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and two-phase flow multipliers under high pressure. As a result, the drift flux model implemented in the TRAC-BF1 code was confirmed to suitably predict the void fraction well under the present high pressure conditions. For the two-phase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present high-pressure conditions.

DOI： 10.1115/1.4023678

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.nucengdes.2011.06.036

Language：English   Publishing type：Research paper (scientific journal)  

Steam Water Pressure Drop under 15 MPa
For a steam generator with straight double-walled heat transfer tubes that will be used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multipliers were checked and then, it was found that both Chisholm two-phase flow multiplier and homogeneous model can predict the present experimental data in high accuracy. A sudden decrease of the pressure drop was observed when flow pattern shifts from bubbly and churn flows to annular flow. The reason for this decrease is tried to be interpreted.

DOI： 10.1299/jpes.5.229

Language：English   Publishing type：Research paper (scientific journal)  

Experimental Research on the Effect of Axial Power Distribution on Critical Power
This paper concerns experimental research to ascertain the effect of axial power distribution on critical power in the positive quality region. Experiments took place at atmospheric pressure in a circular tube. Axial uniform heating and two other axial non - uniform heating cases were selected for detailed evaluation. The effects of relative power ratio on critical power, critical quality and critical boiling length were ascertained in detailed evaluations. Using the experimental data, we evaluated existing correlating concepts with critical power. Result showed a combination of the overall power concept (χ_BT - L_B) and the local conditions concept (χ_BT - q_BT) appearing to be promising in correlating present critical power data in axial non - uniform heating conditions.

DOI： 10.1299/jpes.3.301

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1080/01457630801981614

Language：Others  

Language：English   Publishing type：Research paper (scientific journal)  

Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions
A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

DOI： 10.1299/jpes.2.240

Language：English   Publishing type：Research paper (scientific journal)  

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles, 1; Master plan and executive summary
R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

DOI： 10.1299/jpes.2.229

Language：English   Publishing type：Research paper (scientific journal)  

Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions

DOI： 10.1080/18811248.2007.9711360

Language：English   Publishing type：Research paper (scientific journal)  

An Improved Critical Power Correlation for Tight-Lattice Rod Bundles

DOI： 10.1080/18811248.2007.9711845

Language：English   Publishing type：Research paper (scientific journal)  

Gap Width Effect on Critical Power based on Tight-Lattice 37-Rod Bundle Experiments

DOI： 10.1080/18811248.2007.9711256

Language：English   Publishing type：Research paper (scientific journal)  

Critical Power Experiment with a Tight-Lattice 37-Rod Bundle

DOI： 10.1080/18811248.2006.9711082

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1002/htj.20097

Language：Others   Publishing type：Research paper (scientific journal)  

Characteristics of Boiling Curve in Transition Region between Nucleate Boiling and Film Boiling

DOI： 10.1299/kikaib.71.1390

Language：Others   Publishing type：Research paper (scientific journal)  

Homogeneous nucleation, although being discounted as a mechanism for vapor formation for water in most conditions, is found to possibly occur under some extreme conditions in subcooled flow boiling. Under the conditions, vapor bubbles of molecular dimensions generated in the superheated liquid adjacent to channel wall from homogeneous nucleation due to the local temperature exceeds homogeneous nucleation temperature. The condition is called in this paper as homogeneous nucleation governed condition. Under the condition, conventional flow pattern for subcooled flow boiling, which is characterized by the existence of Net Vapor Generation (NVG) point and the followed bubble detachment, movement and coalescence processes, cannot be established. Critical heat flux (CHF) triggering mechanism so far proposed, which employs a premise assumption that the conventional flow pattern has been established, such as liquid sublayer dryout model, is no more appropriate for the homogeneous nucleation governed condition. In this paper, first, the existence of the homogeneous nucleation governed condition is indicated. In the following, a criterion is developed to judge a given working condition as the conventional one or the homogeneous nucleation governed one. With the criterion, subcooled flow boiling data are categorized and typical homogeneous nucleation governed datasets are listed. The homogeneous nucleation governed data are characterized by extreme working parameters, such as ultrahigh mass flux, ultralow ratio of heated length to channel diameter L/D or ultrahigh pressure. CHF triggering mechanism for the homogeneous nucleation governed condition is proposed and verified. Parametric trends of the CHF, in terms of mass flux, pressure, inlet subcooling, channel diameter, and the ratio of heated length to diameter are also studied.

DOI： 10.1115/1.1844536

Language：English   Publishing type：Research paper (scientific journal)  

Critical Power Correlation for Tight-Lattice Rod Bundles

DOI： 10.1080/18811248.2005.9726362

Language：Japanese  

Transient Critical Power Experiments and Analyses in Tight Lattice Bundles
A major concern in the design of RMWR is that sufficient cooling capability be provided to keep fuel cladding temperature below specified values, even for a postulated abnormal transient process. In this research, power increase and flow decrease transient tests are performed in 7-rod and 37-rod double-humped tight lattice bundles, under RMWR nominal operating condition (P_<ex> = 7.2 MPa, T_<in> =283℃) for mass velocity G = 300, 450, 600 kg/m^2s. Experiments are analyzed with TRAC code, in which new JAERI critical power correlation is implemented for BT judgment. For the postulated nominal power increase and flow decrease transients, when CPR is 1.3, no Boiling Transitions (BTs) are observed in experiments and TRAC code predicts the same trends. For severer conditions that BT occurs, wall temperature jumping points (BT points) can be predicted quite well within the accuracy of the implemented critical power correlation. The traditional quasi-steady-state prediction of BT in transient process is confirmed being applicable for axially double-humped-heated tight lattice bundles.

DOI： 10.1299/jsmemecjo.2004.3.0_231

Language：English   Publishing type：Research paper (scientific journal)  

Development of predictable technology for thermal/hydraulic performance of reduced-moderation water reactors, 1; Master Plan
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR. This series presentation focuses on the feasibility study and shows the R&D plan using large-scale test facility and advanced numerical simulation technology.

DOI： 10.1299/jsmemecjo.2003.3.0_247

Language：English   Publishing type：Research paper (scientific journal)  

Critical Power in 7-Rod Tight Lattice Bundle
The Reduced-Moderation Water Reactor (RMWR) has recently becomes of great concern. The RMWR is expected to promote the effective utilization of uranium recourse. The RMWR is based on water-cooled reactor technology, with achieved under lower core water volume and water flow rate. In comparison with the current light water reactors whose water-to-fuel volume ratio is about 2-3, in the RMWR, this value is reduced to less than 0.5. Thereby, there is a need to research its cooling characteristics. Experimental research on critical power in tight lattice bundle that simulates the RMWR has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3mm equilateral triangular pitch. Each rod is 13mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under G=100-1400kg/m²s, P_ex=2-8.5MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

DOI： 10.1299/jsmeb.47.299

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1002/htj.10116

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1002/htj.10117

Language：Japanese  

Development of Predictable Technology for Thermal/Hydraulic Performance of Reduced-Moderation Water Reactors (2) : Large-scale Thermal/Hydraulic Test and Model Experiments
In this manuscript, the key point of large-scale thermal-hydraulic tests and model experiments which simulate the Reduced-Moderation Water Reactor (RMWR) core are reported. The aim of the large-scale thermal-hydraulic tests using 37-rod bundle test facility is to make clear the rod number effect, rod gap effect etc. for investigating the feasibility of RMWR. Thermal-hydraulic model experiments will be performed in order to verify an advanced 3-D two-phase flow simulation method. Void fraction distribution and its fluctuation and basic thermal characteristics in tight lattice rod bundles will be measured by high-frame-rate neutron radiography technique etc. We will develop the database by 2007 that can resolve the fundamental feasibility subjects for the RMWR.

DOI： 10.1299/jsmemecjo.2003.3.0_249

Language：Others   Publishing type：Research paper (scientific journal)  

Prediction of Position of Moving Heat Source using Inverse Solution

DOI： 10.1299/kikaib.69.1651

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/s0017-9310(02)00510-0

Language：Japanese  

Critical Heat Flux in Tight Lattice Bundles [1] : Effects of Power Distributions and Flow Parameters
CHF experiments have been conducted using the rod bundles of a narrow triangular arrangement, which simulate the Reduced-Moderation Water Reactor core. The purposes of these experiments are (a) to investigate the parameter effects on critical power, (b) to evaluate the existing correlations, (c) to propose the correlation and (e) to make use of the data for the verification of the numerical analysis code. In this paper, parameter effects on critical power or critical quality were focused as a fundamental understanding. It was found from the comparison between the axially uniform and the double humped power distribution that the critical quality of the double humped one decreases significantly when mass velocity>about 150 kg/m^2s.

DOI： 10.1299/jsmeptec.2002.0_325

Language：Japanese  

Critical Heat Flux in Tight Lattice Bundles (2) : Verification of Correlation for Tight Lattice Bundle
Reduced-Moderation Water Reactor (RMWR) will be operated under low core water volume and low water flow rate. In RMWR, Water-to-fuel volume ratio (V_m/V_f) is reduced less than 0.5,in comparison with the values of 2&acd;3 in the current light water reactors. Thereby, there is a need to research the cooling limit. Experimental research on CHF in tight lattice bundles that simulates the actual RMWR have been carried out in JAERI. This paper will focus on the verification of existing CHF correlation for the tight lattice bundle. The Arai correlation is verified with BAPL data and JAERI data and is found that it can not give satisfied CHF prediction to the RMWR working condition.

DOI： 10.1299/jsmeptec.2002.0_327

Language：Others  

Analytical Method of Two Dimensional Inverse Heat Conduction Problem using Laplace Transformation : Effect of Measuring Point Number

Language：Others   Publishing type：Research paper (scientific journal)  

Improvement of Inverse Heat Conduction Solution using Laplace Transformation. (Method of Partial Division of Time).

DOI： 10.1299/kikaib.68.2093

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1002/1521-4125(200204)25:4<447::aid-ceat447>3.0.co;2-p

Language：Others   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/s0017-9310(99)00373-7

Event date： 2023.9

Language：Others   Presentation type：Oral presentation (general)  

Country：Other  

Development of a passive safety shutdown device to prevent core damage accidents in fast reactors (11) Overview of project progress

Event date： 2023.9

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2023.9

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2023.9

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2022.9

Language：Others  

Country：Other  

Development of a passive safety shutdown device to prevent core damage accidents in fast reactors (7) Overview of overall project progress through FY2021

Event date： 2022.9

Language：Others  

Country：Other  

Development of a passive safety shutdown device to prevent core damage accidents in fast reactors (10) Selection of the device assembly design effective in the reactivity control by its material relocation

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2022.3

Language：Others  

Country：Other  

Study on FPs removal effect on the containment vessel penetrations during a severe accident (1) Development of an experimental method for FPs removal effect on the containment vessel penetrations reflecting actual plant conditions

Event date： 2022.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2022.3

Language：Japanese   Presentation type：Oral presentation (general)  

researchmap

Event date： 2021.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2021.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2021.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2021.3

Language：English   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2021.3

Language：English   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2020.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：ウェブ   Country：Japan  

Event date： 2020.6

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2020.6

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2020.3

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2020.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：みなし発表   Country：Japan  

Event date： 2019.12

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2019.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡   Country：Japan  

Event date： 2019.9

Language：Japanese  

Venue：岡山   Country：Japan  

Event date： 2019.8

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：東京   Country：Japan  

Event date： 2019.5

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：徳島県徳島市   Country：Japan  

Event date： 2019.3

Language：Japanese  

Venue：水戸   Country：Japan  

Event date： 2019.3

Language：Japanese  

Venue：水戸   Country：Japan  

Event date： 2018.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：北海道大学   Country：Japan  

Event date： 2018.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：大阪   Country：Japan  

Event date： 2018.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：大阪   Country：Japan  

Event date： 2018.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：大阪   Country：Japan  

Event date： 2017.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：北海道大学   Country：Japan  

Language：Japanese  

Tight-Lattice 37-rod Bundle Thermal-Hydraulic Tests and Model Experiments
Main purposes of the tight-lattice 37-rod bundle thermal-hydraulic tests (37-rod tests) and model experiments are to investigate the scale effect (rod-number effect) on critical power and to obtain the detailed thermal-hydraulic data for understanding the phenomena and estimating the advanced numerical analysis codes, respectively. The 37-rod bundle test section and some test sections for the model experiments simulate the Reduced-Moderation Water. Reactor core. It was found from the comparison of 37-rod test data with existing 7-rod test data that critical quality increase with increasing the rod number. Using the spacer-effect fundamental neutron radiography experiment, void fraction distribution around the object, which simulates the spacer, in a heated tube was discussed. From the 14-rod bundle neutron 3D tomography experiments, it was found that vapor tends to move the center region of the flow channel.

Type：Competition, symposium, etc. 

Number of participants：1,000

Type：Competition, symposium, etc. 

researchmap

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：4

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：16

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Competition, symposium, etc. 

Number of participants：300

Type：Competition, symposium, etc. 

researchmap

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：3

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：4

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Scientific advice/Review 

researchmap

Type：Competition, symposium, etc. 

Type：Competition, symposium, etc. 

researchmap

Type：Competition, symposium, etc. 

researchmap

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：2

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：5

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Competition, symposium, etc. 

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：3

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：7

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Competition, symposium, etc. 

Number of participants：900

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：2

Number of peer-reviewed articles in Japanese journals：1

Proceedings of International Conference Number of peer-reviewed papers：0

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Competition, symposium, etc. 

Number of participants：500

Type：Competition, symposium, etc. 

Number of participants：150

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：3

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：4

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Peer review 

Number of peer-reviewed articles in foreign language journals：2

Number of peer-reviewed articles in Japanese journals：0

Proceedings of International Conference Number of peer-reviewed papers：5

Proceedings of domestic conference Number of peer-reviewed papers：0

Type：Academic society, research group, etc. 

上海交通大学―韓国先端科学技術大学－台湾清華大学の三校の間で発足された原子力工学合同サマースクールに九州大学(KU)も参入し、2019四校合同サマースクールを九大に誘致した。本サマースクールは、日本原子力学会九州支部の共催で、2019年8月5日～9日までに九大伊都キャパスで開催した。上記４校の他、ドイツKITからの参加も加わり、計59人の参加があった。学生から大変好評を受けた。

Type：Other

researchmap

Audience：Infants,　Schoolchildren,　Junior students,　High school students

Type：Other

Type：Other

researchmap

Audience：Infants,　Schoolchildren,　Junior students,　High school students

Type：Other

Type：Other

researchmap

Audience：Infants,　Schoolchildren,　Junior students,　High school students

Type：Seminar, workshop

Type：Seminar, workshop

researchmap

Audience：Infants,　Schoolchildren,　Junior students,　High school students

Type：Seminar, workshop

Type：Seminar, workshop

researchmap

Audience：Infants,　Schoolchildren,　Junior students,　High school students

Type：Other

Type：Other

researchmap