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Journal Articles

Challenge next-generation nuclear system; Development of oxide dispersion strengthened ferritic steel

Otsuka, Satoshi; Kaito, Takeji

Enerugi Rebyu, 39(1), p.44 - 46, 2019/01

For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Nuclear Technology, 205(1-2), p.119 - 127, 2019/01

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.

Journal Articles

Performance evaluation of eddy current flowmeter in Monju

Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohashi, Yuko

Journal of Nuclear Science and Technology, 55(12), p.1393 - 1401, 2018/12

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor's thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter's signal intensity and the primary sodium's flow rate under 10-100% flow rate condition. The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.

Journal Articles

Development of safety design guideline of structures, systems and components of Generation-IV Sodium-cooled Fast Reactor

Okano, Yasushi

Nippon Genshiryoku Gakkai-Shi, 60(12), p.764 - 769, 2018/12

JAEA has developed the second safety design guidelines report, "Safety Design Guidelines on Structures, Systems and Components" for Generation-IV SFR system, following the previously published SFR Safety Design Criteria and the first SFR Safety Design Guidelines report and with the reviews by a technical committee under Atomic Energy Society of Japan. This article explains about 14 key points on reactor core system, coolant system, and containment system and also current consistency of international SFR designs to the safety design guidelines.

Journal Articles

Droplet generation during liquid jet impingement onto a horizontal plate

Zhan, Y.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi

Experimental Thermal and Fluid Science, 98, p.86 - 94, 2018/11

In sodium-cooled fast reactors, liquid sodium leakage may lead to fire accident. In the case that the liquid sodium is discharged as a liquid jet, a number of droplets are produced during liquid jet impingement on the structures. In the present work, a liquid jet was emanated vertically downward from a circular nozzle hole onto a horizontal disk to measure the total amount and the maximum size of splashed droplets. It was found that a significant amount of liquid was splashed when the liquid jet impinged as a broken jet. Thus, a prediction method was first developed for the impact frequency of the primary droplets produced due to the jet breakup. It was then shown that a phenomenological model using the impact frequency and the impact Weber number as the important variables can predict the splashing rate well. It was also indicated that the size of the maximum splashed droplets was fairly proportional to the size of primary droplets.

Journal Articles

Numerical analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

Based on the event tree analysis, the present numerical analyses investigated the capability of fuel discharge through the one-dimensional single fuel assembly geometry and the two-dimensional geometry of a CRGT channel with neighboring fuel assemblies. The single fuel assembly analyses showed that the fuel blockage formed in the lower shielding region because fuel solidified by contacting with cold sodium in case of no fission gas release. On the assumption that fission gas was released, the molten fuel successfully relocated below the core. The next analyses using the CRGT channel indicated a significant fuel discharge through the CRGT channel. This is because the fuel temperature was still high just after the CRGT wall failure and sodium in the CRGT channel was quickly voided just after the ingress of a small amount of molten fuel.

Journal Articles

Effect of porosity distribution on two-phase pressure drop in a packed bed

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.

Journal Articles

Results of an out-of-pile experiment for fragmentation of a simulated molten core material discharged into a shallow sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11

In Core Disruptive Accidents of Sodium-cooled Fast Reactors, molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores. The inlet coolant plenums have quite limited heights and sodium inventories. Therefore, in the inlet plenums, molten core material with a jet-like shape would impinge on the bottom of the plenum before it breaks up into fragments. In this study, to clarify fragmentation behavior in a shallow sodium pool whose height and volume are so limited that jet impingement on the bottom is expected, an out-of-pile experiment discharging molten alumina into a sodium pool was conducted. Although a small amount of alumina agglomeration was found on the bottom plate (steel disk) installed in the sodium pool, most of the molten alumina was fragmented into debris particles. Results obtained in the present experiment suggest that molten core material is fragmented and quenched even in a shallow sodium pool.

Journal Articles

Demonstration of under sodium viewer in Monju

Aizawa, Kosuke; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*

Nuclear Technology, 204(1), p.74 - 82, 2018/10

Development of inspection technique in opaque liquid metal coolant is one of the important issues to ensure the safety of Liquid Metal Fast Breeder Reactor (LMFBR). Performance tests of an Under Sodium Viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of LMFBR Monju, have been carried out. The ultrasonic sensors and reflectors are located across the core inside of the Monju's RV. The USV can detect an obstacle existing in between the core top and the Upper Core Structure (UCS) bottom by identifying differences of echo signals. This report describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated and the signal to noise ratio successfully exceeded the target value. Measured signals clearly differed from with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in the specified place.

Journal Articles

Level 1 PRA for external vessel storage tank of Japan sodium-cooled fast reactor in whole core refueling

Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi; Naruto, Kenichi*

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 15 Pages, 2018/10

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure, which was achieved through probabilistic risk assessment for the EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. The safety strategy for the EVST involves whole core refueling (early transfer of all core fuel assemblies into the EVST) assuming a severe situation that results in sodium level reduction leading finally to the top of the reactor core fuel assemblies in a long time. This study introduces the success criteria mitigation along the decay heat decrease over time. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, a probability analysis for human error, and quantification of accident sequences. The fuel damage frequency of the EVST was evaluated to be approx. 10$$^{-5}$$/year. The dominant accident sequence resulted from the static failure and human error for the switching from the stand-by to operation mode in the three stand-by cooling circuits after loss of one circuit for refueling heat removal operation as an initiating phase.

Journal Articles

Particle-based simulation of heat transfer behavior in EAGLE ID1 in-pile test

Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10

The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m$$^{2}$$ on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.

Journal Articles

Numerical simulation of thermal striping phenomena for fundamental validation and uncertainty quantification; Application of least square version GCI and area validation method to impinging jet in a T-Junction piping system

Tanaka, Masaaki

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 14 Pages, 2018/10

A numerical simulation code MUGTHES has been developed to estimate high cycle thermal fatigue in SFRs. In development of numerical simulation code, verification, validation, and uncertainty quantification (VVUQ) are indispensable. In this study, numerical simulation at impinging jet condition in the WATLON experiment which was the water experiment of a T-junction piping system was performed for the fundamental validation. Based on the previous studies, the simplified least square version GCI method and the area validation metrics were employed as reference methods to quantify uncertainty and to measure the degree of difference between the numerical and the experimental results, respectively. Through the examinations, the potential applicability of the MUGTHES to the thermal striping phenomena was indicated and requirements of modification in the simulation was suggested in accordance with the uncertainty values.

Journal Articles

Numerical simulation on self-leveling behavior of mixed particle beds using multi-fluid model coupled with DEM

Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10

Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.

Journal Articles

Numerical modeling of radiation heat transfer under sodium spray combustion in sodium-cooled fast reactors

Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10

Heat radiation is one of dominant heat transfer process during a sodium fire event which is a concern in sodium-cooled fast reactor plants. This study aims to model radiation heat transfer from combusting droplets. Radiation energy transport on the combustion flame surface around a sodium droplet is formulated considering emission, absorption and scattering through a similar approach to the formulation of the wall boundary condition. The improved model is tested trough a simple verification analysis and a benchmark analysis on an upward sodium spray combustion experiment. As the result, overestimation of atmospheric temperature and pressure is mitigated by the improved model due to increase in heat transfer to structure.

Journal Articles

A Study on splashing during liquid jet impingement onto a horizontal plate

Kuwata, Yusuke*; Zhan, Y.*; Enoki, Koji*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10

This study aims improvement of safety analysis for sodium fire accidents in sodium-cooled fast reactors. In the experiment, effect of viscosity on liquid jet impact on solid surface was studied.

Journal Articles

Modeling of eutectic reaction between molten stainless steel and B$$_{4}$$C for severe accident simulations

Liu, X.*; Morita, Koji*; Yamano, Hidemasa

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10

On the basis of experimental results, growth of the eutectic material is modeled by the parabolic rate law. Heat and mass transfer processes are also modeled considering both the equilibrium and non-equilibrium phase changes of eutectic material. Thermophysical properties of eutectic material obtained from the experimental measurements are also included in the analytic thermophysical property model and analytic equation-of-state model.

Journal Articles

Development of probabilistic risk assessment methodology against volcanic eruption for sodium-cooled fast reactors

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamamoto, Takahiro*

ASCE-ASME Journal of Risk and Uncertainty in Engineering Systems, Part B; Mechanical Engineering, 4(3), p.030902_1 - 030902_9, 2018/09

This paper describes volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3$$times$$10$$^{-6}$$/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. In addition, sensitivity analyses have investigated the effects of a tephra arrival reduction factor and pre-filter covering.

Journal Articles

Selective Sc recovery from rare earths in nitric acid medium by extraction chromatography

Watanabe, So; Suzuki, Hideya; Goto, Ichiro*; Kofuji, Hirohide; Matsumura, Tatsuro

Nippon Ion Kokan Gakkai-Shi, 29(3), p.71 - 75, 2018/09

Chemical compounds containing scandium (Sc) are widely applied to various fields such as catalysts, alloys, lamps and etc. Sc is found in mineral ore, and it is necessary to develop efficient Sc separation and recovery technology for the industrial applications. So far, solvent extraction based processes were proposed as promising Sc recovery procedures. Our group has found that hexaoctyl nitrilotriacetic acid triamide (HONTA) has strong affinity to Sc. In this study, applicability of the extraction chromatography technology for the selective Sc recovery was experimentally evaluated through batch-wise adsorption/elution studies and column separation experiments.

Journal Articles

Improvement of penetrate crack length evaluation method for LBB assessment of sodium-cooled fast reactor components

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*

Nippon Kikai Gakkai 2018-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2018/09

According to the fitness for service code of Sodium-Cooled fast Reactor (SFR), the volumetric tests as in-service inspection can be replaced with continuous leak monitoring, where the Leak Before Break (LBB) is demonstrated, because the primary stress caused by internal pressure is not significant in SFR components. Basically, if the detectable crack length and the penetrated crack length are sufficiently smaller than the unstable critical crack length, it can be concluded that LBB is successfully demonstrated. The authors had already proposed a simplified method to calculate the penetrated crack length both of the circumferential and axial cracks in the pipe as a function of pipe geometry, fatigue crack growth characteristics and loading conditions. However, some problems in the method have been pointed out in the process of the reviewing by the JSME code committee. This study describes an improved method to calculate the penetrated crack length.

Journal Articles

Study on heterogeneous minor actinide loading fast reactor core concepts with improved safety

Ohgama, Kazuya; Oki, Shigeo; Kitada, Takanori*; Takeda, Toshikazu*

Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09

A core concept of minor actinides (MAs) transmutation with improved safety was designed by applying sodium plenum and axially heterogeneous configuration. In this study, heterogeneous MA loading methods were developed for the core concept to explore the potential of further improvement of MA transmutation amount and "effective void reactivity" which was introduced by assuming the axial coolant sodium density change distribution for the unprotected loss of flow accident. By investigating characteristics of heterogeneous cores loading MA in different radial or axial positions, preferable MA loading positions were identified. The core loading MA in the radial position between inner and outer core region attained the largest MA transmutation amount and lowest maximum linear heat rate (MLHR) among heterogeneous cases. The lower region of the core was beneficial to improve the effective void reactivity and MLHR maintaining the nearly same MA transmutation amount as that of the homogeneous core. The radial blanket region was also useful to increased MA transmutation amount without deterioration of the effective void reactivity.

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