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Journal Articles

A Study on sodium-concrete reaction in presence of internal heating

Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800$$^{circ}$$C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800$$^{circ}$$C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

Journal Articles

Computer code analysis of irradiation performance of axially heterogeneous mixed oxide fuel elements attaining high burnup in a fast reactor

Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*

Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements with high burnup in a fast reactor were conducted. Post-irradiation experiments revealed local concentration of Cs near the interfaces between MOX fuel and blanket columns including the internal blanket of the fuel elements as well as an increase in their cladding diameters. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cs-U-O compound was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.

JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements (Translated document)

Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru

JAEA-Technology 2019-020, 167 Pages, 2020/03

JAEA-Technology-2019-020.pdf:21.06MB
JAEA-Technology-2019-020-high-resolution1.pdf:47.3MB
JAEA-Technology-2019-020-high-resolution2.pdf:34.99MB
JAEA-Technology-2019-020-high-resolution3.pdf:48.74MB
JAEA-Technology-2019-020-high-resolution4.pdf:47.83MB
JAEA-Technology-2019-020-high-resolution5.pdf:18.35MB
JAEA-Technology-2019-020-high-resolution6.pdf:49.4MB
JAEA-Technology-2019-020-high-resolution7.pdf:39.78MB

The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.

JAEA Reports

Phase 1 code assessment of SIMMER-III; A Computer program for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu

JAEA-Research 2019-009, 382 Pages, 2020/03

JAEA-Research-2019-009.pdf:22.82MB

The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.

Journal Articles

Post-test material analysis of eutectic melting reaction of boron carbide and stainless steel

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00360_1 - 19-00360_13, 2020/03

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B$$_{4}$$C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the B$$_{4}$$C-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-B$$_{4}$$C eutectic melting reaction was carried out by contacting SS melts of several kg with a B$$_{4}$$C pellet heated up to about 1500 $$^{circ}$$C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the B$$_{4}$$C-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the B$$_{4}$$C-SS contact interface and (Fe,Cr)$$_{2}$$B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the B$$_{4}$$C pellet.

Journal Articles

Development of numerical analysis code LEAP-III for tube failure propagation

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.

Journal Articles

Evaluation of important phenomena through the PIRT process for a sodium fire event

Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00366_1 - 19-00366_8, 2020/03

Sodium fire is one of key issues in sodium-cooled fast reactor plant. JAEA has developed sodium fire analysis codes, such as AQUA-SF and SPHINCS, to evaluate the consequence of sodium fire events. This paper describes the PIRT (Phenomena Identification and Ranking Table) process for sodium fire events. Ranking table for important phenomena and an assessment matrix are completed. As a part of comprehensive validation based on the assessment matrix, experimental analyses using the AQUA-SF and SPHINCS codes for a sodium spray fire experiment Run-E1 show good agreement with the experimental result.

Journal Articles

Prototype fast breeder reactor "Monju" start of unloading operation of the fuel assembly from the core

Koga, Kazuhiro*; Suzuki, Kazunori*; Takagi, Tsuyohiko; Hamano, Tomoharu

FAPIG, (196), p.8 - 15, 2020/02

The prototype fast breeder reactor Monju has already started (from June 2017) the unloading operation period (about 5.5 years: until the end of 2022) of the fuel assembly, which is the first stage of decommission. Among them, the first "Processing of fuel assembly" operation (86 in total) was conducted from August 2018 to January 2019 as the first handling of the fuel assembly. Fuji Electric provided technical support, such as dispatching technicians throughout the period, in cooperation with Japan Atomic Energy Agency for the "Processing of fuel assembly" operation, and contributed to the completion of the operation while experiencing various troubles. This manuscript introduces the contents of the first "Processing of fuel assembly" operation and the overview of the trouble status. This manuscript is a sequel to FAPIG No.194 "Prototype Fast Breeder Reactor Monju Decommissioning and Unloading Operation of the Fuel Assembly from the Core", please refer to it.

Journal Articles

Improvement in flow-sheet of extraction chromatography for trivalent minor actinides recovery

Watanabe, So; Senzaki, Tatsuya; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki; Nakatani, Kiyoharu*; Matsuura, Haruaki*; Horiuchi, Yusuke*; Arai, Tsuyoshi*

Journal of Radioanalytical and Nuclear Chemistry, 322(3), p.1273 - 1277, 2019/12

 Times Cited Count:0 Percentile:100(Chemistry, Analytical)

Journal Articles

Extraction mechanism of lanthanide ions into silica-based microparticles studied by single microparticle manipulation and microspectroscopy

Otaka, Toshiki*; Sato, Tatsumi*; Ono, Shimpei; Nagoshi, Kohei; Abe, Ryoji*; Arai, Tsuyoshi*; Watanabe, So; Sano, Yuichi; Takeuchi, Masayuki; Nakatani, Kiyoharu*

Analytical Sciences, 35(10), p.1129 - 1133, 2019/10

 Times Cited Count:0 Percentile:100(Chemistry, Analytical)

Journal Articles

Thermophysical properties of molten stainless steel containing 5mass%B$$_{4}$$C

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Nuclear Technology, 205(9), p.1154 - 1163, 2019/09

 Times Cited Count:3 Percentile:17.59(Nuclear Science & Technology)

An electromagnetic-levitation technique performed in a static magnetic field was used to measure the density, surface tension, normal spectral emissivity, heat capacity, and thermal conductivity of molten 316L stainless steel (SS316L) and SS316L that contained 5mass%B$$_{4}$$C. The addition of 5mass%B$$_{4}$$C to SS316L yielded reductions of 111 K, 6%, 19%, and 6% in the liquidus temperature, density, normal spectral emissivity, and thermal conductivity at the liquidus temperature of SS316L, respectively. The heat capacity increased by 5% with this addition. Although the 5mass%B$$_{4}$$C addition had no clear effect on the surface tension, sulfur dissolved in the SS316L resulted in a significant decrease in the surface tension.

Journal Articles

Thermophysical properties of stainless steel containing 5 mass%B$$_{4}$$C in the solid phase

Takai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa

Nuclear Technology, 205(9), p.1164 - 1174, 2019/09

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Technological development of the particle size adjustment of dry recovered powder

Segawa, Tomoomi; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Kawaguchi, Koichi; Ishii, Katsunori; Sato, Hisato; Fukasawa, Tomonori*; Fukui, Kunihiro*

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.738 - 745, 2019/09

In the MOX fuel fabrication process, the dry grinding technology of mixed oxide pellets have been developed for the effective use of nuclear fuel materials. To develop a technology to control the particle size of dry recovered powder, the performance of the buhrstone mill and the collision plate type jet mill were studied using a simulated powder of particle size distribution about 500 $$mu$$m. We found that the particle size can be controlled at the range of about 250 $$mu$$m or less by both by adjusting the clearance between the grinding wheels of the buhrstone mill, and the clearance and elevation angle of the clarification zone of the the collision plate type jet mill. And furthermore, the collision plate type jet mill is considered to be suitable for particle size control because the operating parameters of the classifier can be finely adjusted.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09

Eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on B$$_{4}$$C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified B$$_{4}$$C-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.

Journal Articles

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 4; Effect of B$$_{4}$$C addition on viscosity of austenitic stainless steel in liquid state

Ota, Hiromichi*; Kokubo, Hiroki*; Nishi, Tsuyoshi*; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.858 - 860, 2019/09

A viscosity measurement apparatus has been developed. It is known that the measurement of the viscosity of molten alloy at elevated temperatures is difficult due to the difficulty of handling for low viscosity fluids such as the stainless steel (SS)+B$$_{4}$$C alloy. In this study, the viscosities of the molten nickel (Ni) and stainless steel (SS) were measured by the oscillating crucible method to confirm the performance of the viscosity measurement apparatus as a first step. This method is suitable for high temperature molten alloys. A crucible containing molten metal is suspended, and a rotational oscillation is given to the crucible electromagnetically. The oscillation was damped by the friction of molten metal. The viscosity is determined from the period of oscillation and the logarithmic decrement. The crucible was connected to a mirror block and an inertia disk made of aluminum, and whole of them was suspended by a wire made of platinum-13% rhodium alloy. A laser light is irradiated to the mirror. The reflection light is detected by the photo-detectors, and then, the logarithmic decrement of molten metal is determined. The viscosities of molten nickel and SS melts were measured up to 1823 K. In these results, the measured viscosity values of molten Ni and SS were close to those of the literature values of molten Ni and SS. By the equipment, the viscosity of molten SS+B$$_{4}$$C alloys are measured. The B$$_{4}$$C concentration dependence of the viscosity of molten SS+B$$_{4}$$C alloys is to be clarified.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 3; Effect of B$$_{4}$$C addition on thermophysical properties of austenitic stainless steel in a liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.853 - 857, 2019/09

Thermophysical properties of molten mixture of 316L stainless steel (SS316L) and control-rod material (B$$_{4}$$C) are necessary for the development of computer simulation codes that describe core degradation mechanisms during severe accidents in nuclear power plants involving sodium-cooled fast reactors. The effect of B$$_{4}$$C addition to SS316L on the solidus and liquidus temperatures were first measured by differential scanning calorimetry. An electromagnetic levitation technique performed in a static magnetic field was used to measure the density, surface tension, normal spectral emissivity, specific heat capacity, and thermal conductivity of molten SS316L and SS316L containing B$$_{4}$$C. The effects of B$$_{4}$$C addition to SS316L on the thermophysical properties were studied up to 10 mass%.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 5; Validation of a multi-phase model for eutectic reaction between molten stainless steel and B$$_{4}$$C

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

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09

Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a B$$_{4}$$C pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid B$$_{4}$$C and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of B$$_{4}$$C pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.

Journal Articles

Cross-section-induced uncertainty evaluation of MA sample irradiation test calculations with consideration of dosimeter data

Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*

Annals of Nuclear Energy, 130, p.118 - 123, 2019/08

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.

Journal Articles

Comparative study on the thermal behavior of structural concretes of sodium-cooled fast reactor

Kikuchi, Shin; Koga, Nobuyoshi*; Yamazaki, Atsushi*

Journal of Thermal Analysis and Calorimetry, 137(4), p.1211 - 1224, 2019/08

 Times Cited Count:2 Percentile:37.05(Thermodynamics)

In this study, two siliceous concretes with similar specification as structural concretes of SFR were selected for the comparative study of the thermal behavior. The thermal behavior of the structural concretes was investigated in a temperature range from room temperature to 1900 K using TG-differential thermal analysis (DTA) and other supplementary techniques. The softening and melting of the concretes initiated in the thermal decomposition product of the cement portion in the temperature range 1400-1600 K. Because the compositional difference between the cement portion of two different siliceous concretes was characterized by different Ca(OH)$$_{2}$$/CaCO$$_{3}$$ ratios, the melting temperature ranges of those thermal decomposition products are not so significantly different. On the other hand, the melting of the aggregate is directly influenced by the initial composition of SiO$$_{2}$$ compounds.

11063 (Records 1-20 displayed on this page)