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Journal Articles

Kinetic study on eutectic reaction between boron carbide and stainless steel by differential thermal analysis

Kikuchi, Shin; Nakamura, Kinya*; Yamano, Hidemasa

Mechanical Engineering Journal (Internet), 8(4), p.20-00542_1 - 20-00542_13, 2021/08

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) and stainless steel (SS) may take place. Thus, kinetic behavior of B$$_{4}$$C-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of B$$_{4}$$C-SS eutectic melting, the thermal analysis using the pellet type samples of B$$_{4}$$C and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the B$$_{4}$$C-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B$$_{4}$$C-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2019

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

Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 11 Pages, 2021/08

One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses 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 a range from solid to liquid state. The physical model is developed for a CDA 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 conducted until 2019. Specific results in this paper are the validation of physical model describing B$$_{4}$$C-SS eutectic reaction in the CDA analysis code, SIMMER-III, through the numerical analysis of the B$$_{4}$$C-SS eutectic melting experiments in which a B$$_{4}$$C block was placed in a SS pool.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2018

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

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08

One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses 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 a 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 conducted until 2018. Specific results in this paper are 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

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 3; Kinetic study of boron carbide-stainless steel eutectic melting by differential thermal analysis

Kikuchi, Shin; Yamano, Hidemasa; Nakamura, Kinya*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) and stainless steel (SS) may take place. Thus, kinetic behavior of B$$_{4}$$C-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of B$$_{4}$$C-SS eutectic melting, the thermal analysis using the pellet type samples of B$$_{4}$$C and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the B$$_{4}$$C-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B$$_{4}$$C-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.

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

Development of accident tolerant control rod for light water reactors

Ota, Hirokazu*; Nakamura, Kinya*; Ogata, Takanari*; Nagase, Fumihisa

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.159 - 168, 2016/09

Control rods can be disintegrated and neutron absorber would be removed from the core region before most of the fuel pins are still not damaged seriously in severe accidents of LWRs. The present study investigates a concept of accident tolerant control rod (ATCR) with the following characteristics; (1) sufficiently-high melting and eutectic temperatures, (2) high miscibility with molten and solidified fuel materials, and (3) enough control rod worth. It has been shown that rare-earth sesqui-oxides are expected to be compatible with iron up to higher temperatures than the melting points of structure materials of control rods, and that Sm$$_{2}$$O$$_{3}$$, Eu$$_{2}$$O$$_{3}$$, Gd$$_{2}$$O$$_{3}$$, Dy$$_{2}$$O$$_{3}$$ or their mixtures with HfO$$_{2}$$ are available as alternative neutron absorbers to conventional Ag-In-Cd alloy.

JAEA Reports

Fuel and core design studies on metal fuel sodium-cooled fast reactor (4), (5) and (6); Joint research report for JFY2009 - 2012

Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.

JAEA-Research 2012-041, 126 Pages, 2013/02

JAEA-Research-2012-041.pdf:16.49MB

The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.

Journal Articles

U-Pu-Zr metal fuel fabrication for irradiation test at JOYO

Nakamura, Kinya*; Kato, Tetsuya*; Ogata, Takanari*; Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

The first irradiation campaign of U-Pu-Zr metal fuel in Japan is planned in the experimental fast reactor JOYO. In the fabrication of U-Pu-Zr fuel, two methods were adopted for preparing U-Pu alloy from the oxide; one is the electrochemical reduction and the other is the electrorefining followed by reductive extraction. Injection casting for U-Pu-Zr slug was carried out after adding U and Zr metals to meet the target specifications of the irradiated fuel. Several conditions of Na-bonding process were determined from the results of tests using simulated metal fuel pins. Based on these results, six U-Pu-Zr fuel pins for the irradiation tests are now being fabricated.

Journal Articles

Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

Kikuchi, Hironobu; Nakamura, Kinya*; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Ogata, Takanari*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.323 - 331, 2011/12

A high-purity Ar gas atmosphere glovebox accommodating injection casting and sodium-bonding apparatuses was newly installed in Plutonium Fuel Research Facility (PFRF) of Oarai Research and Development Center, Japan Atomic Energy Agency. Past experiences in PFRF led to the establishment of technological basis of fabrication of U-Pu-Zr alloy fuel pin for the first time in Japan. After the injection casting of U-Pu-Zr alloy, the metallic fuel pins are fabricated by welding upper- and lower end plugs with cladding tube of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and cladding tube with the melted sodium, the fuel pins are subjected to the inspection for irradiation tests. This paper summarizes the equipment of the apparatuses and the technological basis for fabrication of U-Pu-Zr alloy fuel pins for the coming irradiation test in the experimental fast test reactor JOYO.

Journal Articles

Fabrication of U-Pu-Zr metallic fuel elements for the irradiation test at experimental fast test reactor Joyo

Nakamura, Kinya*; Ogata, Takanari*; Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Kato, Tetsuya*; Arai, Yasuo; Uozumi, Koichi*; Hijikata, Takatoshi*; Koyama, Tadafumi*; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.245 - 256, 2011/12

Sodium-bonded metallic fuel elements were fabricated for the first time in Japan for the irradiation test in the experimental fast test reactor JOYO. U-20Pu-10Zr fuel slugs of 200 mm in length and approximately 5 mm in diameter were fabricated in a small-scale injection casting furnace. Each fuel slug was loaded into the ferritic martenstic stainless steel (PNC-FMS) cladding tube with the sodium thermal bond, thermal insulator and reflector in a helium gas atmosphere glove box. After top-end plug welding to the cladding tube and heat treatment of the welding area, each fuel element was subjected to the sodium bonding process. After the inspection such as element length, gas plenum length and helium-leak tightness, six metallic fuel elements are transported to the JOYO site for the coming irradiation test.

Journal Articles

Fabrication of U-Pu-Zr metallic fuel elements for irradiation test at Joyo

Nakamura, Kinya*; Ogata, Takanari*; Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Kato, Tetsuya*; Arai, Yasuo; Koyama, Tadafumi*; Itagaki, Wataru; Soga, Tomonori; et al.

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12

CRIEPI and JAEA have fabricated sodium-bonded metallic fuel elements for the first time in Japan as a collaborative research, for use in the irradiation test at the experimental fast test reactor Joyo. The irradiation test aims to assess the irradiation behavior of the fuel and the internal wastage of the stainless-steel cladding by rare-earth fission products at a maximum cladding temperature above 873 K. U-20 wt% Pu-10 wt% Zr alloy fuel slugs of 200 mm length were fabricated in an injection-casting furnace using U metal, U-Pu alloy and Zr metal. Two types of fuel slug were fabricated, i.e., 5.05 mm and 4.95 mm in diameter, and loaded into a ferritic-martensitic stainless-steel cladding tubes, respectively. After top-end-plug welding to the cladding tube, each fuel element was subjected to sodium bonding to fill the annular gap between the fuel slug and the cladding with melted sodium. The fabrication results indicated that the characteristics of the fuel elements were within the required specifications.

Journal Articles

Preparation of TRU alloys by electrochemical or lithium reduction method

Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo; Kurata, Masaki*; Nakamura, Kinya*; Arita, Yuji*

Proceedings of 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (CD-ROM), 9 Pages, 2010/00

Information on phase diagram and thermodynamic data of transuranium (TRU) alloys are required for the development of metallic fuel fabrication technology and fuel design for TRU transmutation. In the present study, TRU alloys were prepared from their mixed oxides to establish these informations. A few grams of mother alloys containing TRU elements were prepared from their mixed oxides in the electrochemical reduction method. However, it was found in the electrochemical reduction of PuO$$_{2}$$ that only the sample surfaces were reduced to metal. It is anticipated that the electrochemical conversion of mixed oxides with high Np and Pu contents into metals are difficult owing to production of low-melting metals. Thus, the Li reduction method was applied to the reduction of mixed oxides with high TRU contents and the alloys with high Np and Pu contents were prepared.

JAEA Reports

Welding of metallic fuel elements for the irradiation test in JOYO; Preliminary tests and welding execution tests (Joint research)

Kikuchi, Hironobu; Nakamura, Kinya*; Iwai, Takashi; Arai, Yasuo

JAEA-Technology 2009-049, 22 Pages, 2009/10

JAEA-Technology-2009-049.pdf:13.99MB

Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests.

Journal Articles

Fabrication of metal fuel slugs for an irradiation test in JOYO

Nakamura, Kinya*; Ogata, Takanari*; Kato, Tetsuya*; Nakajima, Kunihisa; Arai, Yasuo

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1487 - 1495, 2009/09

The U-Pu-Zr fuel slugs for the irradiation test in JOYO were manufactured in a small-scale injection casting furnace. The U-Pu alloy ingots as starting materials were prepared by means of electrochemical reduction of the dioxides. The U-Pu-Zr fuel slugs manufactured met all the specifications determined based on not only the results of preliminary tests but also the specification of EBR-II driver fuels. The americium to plutonium ratio in the fuel slugs slightly decreased after the injection casting process.

Journal Articles

Development of metal fuel fabrication technology for irradiation test in JOYO; Production of uranium-plutonium alloy by electrochemical reduction

Kato, Tetsuya*; Nakamura, Kinya*; Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo

Denryoku Chuo Kenkyusho Hokoku (L05010), 13 Pages, 2006/07

no abstracts in English

Journal Articles

Development of metal fuel fabrication technology for irradiation tests in JOYO; Study on injection casting of U-8.5wt%Pu-10wt%Zr alloy

Nakamura, Kinya*; Kato, Tetsuya*; Iwai, Takashi; Arai, Yasuo

Denryoku Chuo Kenkyusho Hokoku (L05011), 14 Pages, 2006/07

no abstracts in English

Journal Articles

Pyrometallurgical production of U-Pu alloy and injection casting of U-Pu-Zr

Nakamura, Kinya*; Yokoo, Takeshi*; Arai, Yasuo

Transactions of the American Nuclear Society, 94(1), P. 780, 2006/06

no abstracts in English

Journal Articles

Development of metal fuel fabrication technology for irradiation test in JOYO

Nakamura, Kinya*; Iwai, Takashi; Arai, Yasuo

Denryoku Chuo Kenkyusho Hokoku (L04005), 48 Pages, 2005/04

no abstracts in English

Oral presentation

Development of fabrication technology of U-Pu-Zr alloy fuel slug, 3; Injection casting tests of U-8.5wt%-10wt%Zr alloy

Nakamura, Kinya*; Yokoo, Takeshi*; Iwai, Takashi; Arai, Yasuo

no journal, , 

Uranium-plutonium-zirconium alloy is one of the candidate nuclear fuels for fast reactors. Development of fabrication technology is progressing and irradiation tests of the U-Pu-Zr alloy fuel pins in JOYO fast reactor are scheduled in collaboration with Central Research Institute of Electric Power Industry and Japan Atomic Energy Agency. As a result of the first injection castingtest of U-Pu-Zr alloy slug in Japan, the cast alloy slug met with target specifications such as length, diameter and chemical composition.

Oral presentation

Preparation of U-Pu alloy for metallic fuel fabrication, 2; Reductive extraction of U and Pu from molten salt using Cd-Li alloys

Kato, Tetsuya*; Nakamura, Kinya*; Iwai, Takashi; Arai, Yasuo

no journal, , 

In the pyrometallugical reprocessing of spent metallic fuel, U and Pu are recovered into liquid Cd through molten salt and subsequentry purified by distillation of Cd. In this study, U and Pu in molten salt were extracted into liquid Cd that had contained Li metal as reductant. The product of Cd-U-Pu alloys was heated to distill Cd and then, the obtained U-Pu alloys were supplyed for fabrication test of U-Pu-Zr metallic fuels.

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