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Journal Articles

Concepts and basic designs of various nuclear fuels, 4; Metallic fuels for fast reactors and nitride fuels for ADS

Ogata, Takanari*; Takano, Masahide

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(7), p.541 - 546, 2021/07

This is a commentary on metallic fuels for fast reactors and nitride fuels for minor actinide transmutation in accelerator driven system, as the 4th article of serial lecture on Journal of the Atomic Energy Society of Japan; Concepts and basic designs of various nuclear fuels.

Journal Articles

Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

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 metallic fuel core and fuel concept for SFR with a 550$$^{circ}$$C core outlet temperature

Naganuma, Masayuki; Ogata, Takanari*; Mizuno, Tomoyasu

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

In the FaCT project, a design study of the metallic fuel SFR has been executed as secondary candidate. The primary interest is to achieve a core outlet temperature of 550 $$^{circ}$$C. However, the metallic fuel has a drawback that the maximum temperature of the cladding inner surface is limited to 650 $$^{circ}$$C to avoid liquid phase formation. To overcome this problem, JAEA has developed and studied the advanced core concept with single Pu-enrichment and 2 radial regions of heavy metal density. In this paper, the core and fuel design study for the middle-scale SFR applying this core concept are discussed. In addition, for the practical application of the metallic fuel to the SFR with high outlet temperature, it is necessary to expand irradiation experience under the high cladding temperature condition. Therefore, JAEA and CRIEPI planned an irradiation test of the metallic fuel in Joyo as a collaborative program. In this paper, the outline and current status of the irradiation test are reported.

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

Japanese programs in development of pyro-processing fuel cycle technology for sustainable energy supply with reduced burdens

Koyama, Tadafumi*; Ogata, Takanari*; Myochin, Munetaka; Arai, Yasuo

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

no abstracts in English

JAEA Reports

Fuel and core design studies on metal fuel sodium-cooled fast reactor, 3; Joint research report for JFY2007&2008

Okano, Yasushi; Kobayashi, Noboru*; Ogawa, Takashi; Oki, Shigeo; Naganuma, Masayuki; Okubo, Tsutomu; Mizuno, Tomoyasu; Ogata, Takanari*; Ueda, Nobuyuki*; Nishimura, Satoshi*

JAEA-Research 2009-025, 105 Pages, 2009/10

JAEA-Research-2009-025.pdf:10.45MB

A metal fuel core has specific features on high heavy metal density, hard neutron spectrum, and efficient neutron utilization. Enlarged applicable design envelops would improve core performances and features: higher breeding ratio, compacted reactor core, and, smaller amount of Pu-fissile inventory. A joint study on "Reactor Core and Fuel Design of Metal Fuel Core of Sodium Cooled Fast Reactor" by Japan Atomic Energy Agency and Central Research Institute of Electric Power Industry has been conducted during Japanese fiscal years of 2007 and 2008. This report shows the results on (1) the study on applicable design ranges of metal fuel specifications, (2) the study on conceptual core designs for high breeding ratio, and (3) the safety study on metal fuel core designed in the Fast Reactor Cycle Technology Development (FaCT) Project.

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

A Design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

Journal of Power and Energy Systems (Internet), 3(1), p.126 - 135, 2009/00

Utilizing advantages of the metal fuel core to the mixed oxide fuel one, such as its higher breeding ratio and compact core size, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8${$}$, a core height of less than 150 cm, a maximum cladding temperature of 650$$^{circ}$$C, and a fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40.

Journal Articles

Study on enhanced performance sodium-cooled metal fuel core concepts by adopting advanced fuel and flexible design criteria

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 9 Pages, 2008/05

The metal fuel core is superior to the mixed oxide fuel core because of its higher breeding ratio and compact core size resulting from neutron economics, hard neutron spectrum, and high content of heavy metal nuclides. Utilizing the advantage of the metal fuel core, conceptual sodium-cooled fast breeder reactor designs have been pursued for the attractive core properties of high breeding ratio, small inventory, compact size, low sodium void reactivity, and high transmutation ratio of the minor actinides. Among attractive cores, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void reactivity of less than 8${$}$, a core height of less than 150 cm, a maximum cladding temperature of 650 $$^circ$$C, and a fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 without blanket fuels.

JAEA Reports

None

Ota, Hirokazu*; Ogata, Takanari*; *; ; Hayashi, Hideyuki; ;

JNC TY9400 2001-015, 40 Pages, 2001/03

JNC-TY9400-2001-015.pdf:1.62MB

no abstracts in English

Oral presentation

Technical evaluations of the fuel cycle systems

Funasaka, Hideyuki; Namekawa, Takashi; Sato, Koji; Namba, Takashi; Ogata, Takanari; Yokoo, Takeshi*

no journal, , 

no abstracts in English

Oral presentation

Development of fabrication technology of U-Pu-Zr alloy fuel slug, 4; Injection casting of U-20wt%Pu-10wt%Zr alloy

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

no journal, , 

no abstracts in English

Oral presentation

A Design study of sodium cooled metal fuel core characterizing its feature, 1; Design study on high breeding ratio core without any blanket fuels

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

no journal, , 

no abstracts in English

Oral presentation

Fabrication of uranium-plutonium-zirconium alloy fuel for fast breeder reactor by low pressure casting

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

no journal, , 

Uranium-plutonium-zirconium(U-Pu-Zr) alloy is promising for fast breeder reactor. Uranium-zirconium alloy has been developed on a engineering scale in Japan. In the present experiment, U-Pu-Zr alloy fuel was fabricated by low pressure casting. It was cleared that the alloy rods had homogeneous composition and diameter, density, length, straightness and composition were satisfied with the standard of fabrication.

Oral presentation

A Design study of sodium cooled metal fuel core characterizing its feature, 2; Design study on compact core

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

no journal, , 

no abstracts in English

31 (Records 1-20 displayed on this page)