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Journal Articles

Evaluation on tolerance to failure of ODS ferritic steel claddings at the accident conditions of fast reactors

Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.

JAEA Reports

A Study on laser welding of ferritic/martensitic steel (PNC-FMS) for fast reactor fuel assemblies

Kono, Fumiaki; Sogame, Motomu; Yamada, Tomonori; Shobu, Takahisa; Naganuma, Masayuki; Ozawa, Takayuki; Muramatsu, Toshiharu

JAEA-Technology 2015-004, 57 Pages, 2015/03

JAEA-Technology-2015-004.pdf:20.87MB

Laser welding of ferritic/martensitic steel (PNC-FMS) sheets with different thicknesses (2 mm and 5 mm) was examined to investigate the weldability between the inner and outer duct in fast reactor fuel assemblies with inner duct structure (FAIDUS); the objective of the inner duct is to avoid the re-criticality in case of the core melting accident. Laser-spot and melt-run welding was performed at various laser powers, welding times and velocities to find out the appropriate welding conditions with few defects and enough penetration depth. As for the spot welding, furthermore, slow cooling rate or pulsed laser irradiation could reduce the crack and porosity in the welded zone. The strain of the welded zone almost disappeared and the hardness was comparable with that of the base metal by applying post welding heat treatment at 690 $$^{circ}$$C for 103 min. In addition, the shear strength of welded joints was confirmed to be sufficiently higher than the provisional allowance shear stress. These results indicate that laser welding would be probably applied to the PNC-FMS inner and outer ducts.

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

Fast breeder reactor core concept for heterogeneous minor actinide loading

Oki, Shigeo; Naganuma, Masayuki; Okubo, Tsutomu; Tanaka, Kenya

Journal of Nuclear Science and Technology, 50(1), p.59 - 71, 2013/01

 Times Cited Count:4 Percentile:58.85(Nuclear Science & Technology)

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

Minor actinide-bearing oxide fuel core design study for the JSFR

Naganuma, Masayuki; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Kotake, Shoji*

Nuclear Technology, 170(1), p.170 - 180, 2010/04

 Times Cited Count:9 Percentile:39.6(Nuclear Science & Technology)

In FaCT project, sodium-cooled fast reactor with mixed-oxide fuel was selected as the primary candidate. Present study focused on effects of TRU composition on the design of JSFR. In a transitional stage from LWR to FBR, there is possibility for the JSFR fuel to have high MA content due to recycle of LWR spent fuel. That affects core and fuel designs (core reactivity, material property and so on). Thus, to evaluate the effects quantitatively, design studies for the JSFR cores with two TRU compositions were conducted, one was FBR multi-recycle composition with about 1wt% of MA and the other was LWR recycle one, for which 3wt% of MA was assumed as a typical composition. The results showed that the change from FBR multi-recycle composition to LWR recycle one leads to 10% increase of sodium void reactivity, 1-2% decrease of linear power limit and 5% extension of gas plenum length. As a result, effects of TRU composition on the core and fuel designs were indicated to be benign.

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

FBR core concepts in the "FaCT" Project in Japan

Oki, Shigeo; Ogawa, Takashi; Kobayashi, Noboru; Naganuma, Masayuki; Kawashima, Katsuyuki; Maruyama, Shuhei; Mizuno, Tomoyasu; Tanaka, Toshihiko*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 10 Pages, 2008/09

Conceptual design studies of sodium-cooled fast reactor core are performed in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan. The representative MOX fuel core and the metal fuel core exert excellent performances on safety and reliability, sustainability, economic competitiveness, and nuclear non-proliferation. This paper reviews their feature in terms of reactor physics, and describes recent progress in design studies. In the recent design studies, much interest has been taken in the fuel composition change in the transition stage from light water reactors to fast breeder reactors. The core flexibility is also shown to fulfil the refined objectives such as high breeding and an enhancement of non-proliferation property.

Journal Articles

Development of advanced loop-type fast reactor in Japan, 6; Minor actinide containing oxide fuel core design study for the JSFR

Naganuma, Masayuki; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Kubo, Shigenobu*

Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.526 - 535, 2008/06

In FaCT project, sodium-cooled fast reactor with mixed-oxide fuel was selected as the primary candidate. Present study focused on effects of TRU composition on the design of JSFR. In a transitional stage from LWR to FBR, there is possibility for the JSFR fuel to have high MA content due to recycle of LWR spent fuel. That affects core and fuel designs (core reactivity, material property and so on). Thus, to evaluate the effects quantitatively, design studies for the JSFR cores with two TRU compositions were conducted, one was FBR multi-recycle composition with about 1wt% of MA and the other was LWR recycle one, for which 3wt% of MA was assumed as a typical composition. The results showed that the change from FBR multi-recycle composition to LWR recycle one leads to 10% increase of sodium void reactivity, 1-2% decrease of linear power limit and 5% extension of gas plenum length. As a result, effects of TRU composition on the core and fuel designs were indicated to be benign.

JAEA Reports

Study on reactor core and fuel design of sodium cooled fast reactor, Mixed oxide fuel core; Results in JFY2006

Ogawa, Takashi; Kobayashi, Noboru; Oki, Shigeo; Naganuma, Masayuki; Kubo, Shigenobu*; Mizuno, Tomoyasu

JAEA-Research 2007-084, 63 Pages, 2008/01

JAEA-Research-2007-084.pdf:4.3MB

The sodium-cooled large-scale "high internal conversion (HIC) type" core with MOX fuel is the most promising core concept in FaCT Project in Japan. Design study on core and fuel in JFY2006 is reported. (1) Design study of the core with MOX fuel containing MA; Based on the large-scale HIC type core in FS Phase II, we have developed a core using TRU of high MA content recovered from ALWR spent fuel. MA content in the fuel heavy metal is temporarily assumed to be 3 wt%. We have confirmed the core design feasibility with the detailed evaluations of thermal hydraulic characteristics and fuel integrity. (2) Design study of the nonproliferation core by adding Pu to the blankets; As one of the measures to enhance the intrinsic nonproliferation property of fast reactors, we have developed the nonproliferation core concept that can keep the Pu in blankets to "Reactor Grade ($$^{240}$$Pu isotope abundance ratio $$>$$ 18%)" with premixing Pu (or TRU) of core fuel to blanket fuel.

JAEA Reports

Study on influence to core and fuel design by adopting vibro-packed fuel and sphere-pac fuel

Naganuma, Masayuki; Aida, Tatsuya*; Hayashi, Hideyuki

JAEA-Research 2006-087, 72 Pages, 2007/02

JAEA-Research-2006-087.pdf:4.5MB

In the core and fuel design of sodium-cooled MOX fuel FR of FS, core designs with pellet fuel were mainly evaluated. However, vibro-packed fuel and sphere-pac fuel are also considered one of the candidates. Besides, the design must be affected by difference of the fuel behavior, however, the influence had not yet been evaluated adequately. Thus, authors examined the fuel thermal design model and evaluated the influences to the core and fuel design quantitatively. As a result, in the fuel thermal design model, selection of restructuring conditions was found to be important. Proper values were evaluated from the viewpoint of fitting PIE results. In applying this model, limitation of the stationary LHR was reduced to 390 W/cm. For FS phase-II reference cores, though it is required to modify specifications due to the decrease of LHR, the influence to nuclear performance is found to be benign. Therefore, the design that meets the requirements and targets of FS is possible for those fuels.

JAEA Reports

Study on coated layer material characteristics of coated particle fuel FBR, 3; Thick layer coating process of TiN and gas-nitriding treatment of Ti metal

Naganuma, Masayuki; Mizuno, Tomoyasu

JAEA-Research 2006-075, 65 Pages, 2006/12

JAEA-Research-2006-075.pdf:30.16MB

Helium gas cooled FBR is one of attractive core concepts, so that the design studies have been performed in the Japanese feasibility study. Here, the coated particle fuelling core is considered to be promising. One of key issues of that fuel is the coated layer material, and TiN is regarded as one of possible materials. Therefore, tests of thick coating of TiN and gaseous nitriding of Ti metal were conducted in this study. In the thick coating tests, PVD (Physical Vapor Deposition) and CVD (Chemical Vapor Deposition) are selected as coating methods. As a result, both methods are found to have capability to form 100 micrometer that is the target of the practical design. Then, as an alternative method to form thick layer, authors contrived the method to nitride Ti metal by nitrogen gas and conducted tests. As a result, the specimen with 100 micrometer thickness was found to be nitrided entirely in 1,200$$^{circ}$$C and 48 hours. However, the nitrided specimen has tendency to be brittle.

Journal Articles

Design factor using a SiC/SiC composites for core component of gas cooled fast reactor, 1; Hoop stress

Lee, J.-K.; Naganuma, Masayuki

Ceramics in Nuclear and Alternative Energy Applications, p.55 - 63, 2006/12

As a part of the design study on Gas cooled Fast Reactor (GFR), core component designs of helium gas cooled fast reactor are being researched. The most promising structural materials of core components have been identified as ceramics, especially Silicon Carbide fiber reinforced Silicon Carbide matrix composites (SiC/SiC) have an encouraging characteristics. However, the existing design factors for Fast Breeder Reactor (FBR) was based on isotropic properties of metal, like a Mises yield criterion. Therefore, the design factors which are considered characteristics of ceramic composites was investigated and refined. The difference of result was compared by FEM analysis. Refined design factors are not only to design core components of GFR but also to indicate an improvement direction of SiC/SiC composites.

Journal Articles

Preliminary calculation of stress change of fuel pin using SiC/SiC composites for GFR with changing of thermal conductivity degradation by irradiation

Lee, J.-K.; Naganuma, Masayuki

Proceedings of 15th Pacific Basin Nuclear Conference (PBNC-15) (CD-ROM), 6 Pages, 2006/10

Gas cooled Fast Reactor (GFR) is being researched as a candidate concept of Generation IV international Forum. As a main feature of GFR, it should be maintained high temperature and pressure of coolant gas for heat transfer efficiency. Present study assumed a cladding tube of fuel pin type reactor, and using of SiC/SiC composites. Thermal stress of cladding wall was estimated with a condition of linear heat rate and temperature. Through the present work, the ways to improve an applicability of SiC/SiC were clarified; they are the densification of matrix and the improvement of interphase in terms of fabrication. From the viewpoint of core design, the degradation of thermal conductivity by irradiation was identified as the most important factor. These results should be helpful not only to determine the geometries of core component but also to indicate the improvement direction of SiC/SiC.

JAEA Reports

Study on transmutation technology of Long-Lived-Fission-Products (LLFP) using commercial fast reactors; Loading type of LLFP target assembly and transmutation performances of cores designed in FS phase-II

Naganuma, Masayuki; Aida, Tatsuya*; Hayashi, Hideyuki

JAEA-Research 2006-063, 97 Pages, 2006/09

JAEA-Research-2006-063.pdf:9.19MB

In the Feasibility Study in Japan (FS), transmutation technology of LLFP using commercial fast reactors has been studied to reduce the environmental burden. In this report, loading type of LLFP target assembly, transmutation performances of FS designed cores and capability of transmutation core with high SF (transmutation / production ratio) are studied. Design studies for two loading types cores (in-core and ex-core loading type) were conducted for comparison. The in-core loading type core was found to decrease LLFP inventories significantly, thus, that was selected as the reference of FS. For FS phase-II designed cores, LLFP transmutation performances were evaluated. Every core was confirmed to have the capability to attain SF $$>$$ 1.0. Then, we conducted sensitivity evaluations of design conditions to SF for transmutation core with high SF. Since sensitivities of every condition were found to be small, we concluded that large SF may be impossible for the commercial reactors.

JAEA Reports

Study on reactor core and fuel design of sodium cooled fast reactor, Mixed oxide fuel core; Results in JFY2005

Ogawa, Takashi; Sato, Isamu; Naganuma, Masayuki; Aida, Tatsuya*; Sugino, Kazuteru; Hayashi, Hideyuki

JAEA-Research 2006-061, 54 Pages, 2006/09

JAEA-Research-2006-061.pdf:3.86MB

Sodium cooled fast reactor with mixed oxide fueled core is one of the promising candidates in "Feasibility Study on Commercialized Fast Reactor Cycle System" in Japan. The results of the study on the reactor core and fuel design in the JFY2005 are reported. (1)Design studies of high internal conversion (HIC) type core: (i)Influence of TRU composition variation on the HIC type core and fuel designs was evaluated. (ii)In adopting PNC-FMS steel as alternative cladding material of ODS steel, influence to the reactor core and fuel design was evaluated for the large-scale HIC type core. (iii)Shielding property of the large-scale HIC type core was evaluated. (iv)Some measures to extend the lifetime of control rod were studied for the large-scale HIC type core. (2)Design study on high breeding performance: The core design corresponding to a requirement of high breeding performance was studied based on the large-scale compact type core designed in the JFY2004.

JAEA Reports

Design studies on small fast reactor cores, 5; Research results in JFY2005

Uto, Nariaki; Okano, Yasushi; Naganuma, Masayuki; Mizuno, Tomoyasu; Hayashi, Hideyuki

JAEA-Research 2006-060, 68 Pages, 2006/09

JAEA-Research-2006-060.pdf:3.98MB

A design study on "Long-life Type Concept" of a 50MWe sodium-cooled metal-fueled reactor core was performed with more emphasis on irradiation results regarding fuel smear density. The concept aims at no refueling in a core life time, and achieving higher core outlet temperature such as 550$$^{circ}$$C which is advantageous to hydrogen production. The restriction of upper fuel smear density limit to 75% along with adjustments of fuel specifications showed feasibility of attaining core life time of 30 years and core outlet temperature of 550$$^{circ}$$C. No indication of occurrence of absorber-cladding mechanical interaction (ACMI) was found in the evaluation of ACMI for a control rod element. A shielding with Zr-H was selected in view of enhancement of shielding performance, and the feasibility was shown to satisfy the target allowance level of the ratio of hydrogen to zirconium, more than 1.53, with PNC316 used as the cladding material.

Journal Articles

Conceptual design study of helium cooled fast reactor in the "feasibility study" in Japan

Okano, Yasushi; Naganuma, Masayuki; Ikeda, Hirotsugu; Mizuno, Tomoyasu; Konomura, Mamoru

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

Conceptual design of gas-cooled fast reactor (GFR) have been studied for selecting applicable combinations of coolant, fuel material and configuration, and, balance of plant, as a part of feasibility study in Japan. Large-scale He-cooled GFR, employing mixed nitride fuel and achieving a high core outlet temperature of 850$$^{circ}$$C, is recognized to achieve attractive features as a future nuclear reactor system. Three fuel configurations are considered and compared in their core and safety performances; one is horizontal-flow cooling fuel assembly (F/A), another is hexagonal matrix block F/A, and the last one is sealed pin bundle F/A. The horizontal-flow and matrix block F/A cores show nearly the same neutronics performances on discharge burnup around 120GWd/ton, breeding ratio above 1.1, and, core cooling performances under depressurization condition without control rod scram or auxiliary core cooling system (ACCS) actuations; whereas around 30% smaller quantity of fissile Pu required is a merit for matrix concept. The sealed pin bundle F/A core potentially shows attractive neutronics performances on discharge burnup about 141GWd/ton with breeding ratio of 1.27, although rapid control rod scram and ACCS actuations are indispensable for core cooling under depressurization accident conditions.

Journal Articles

Conceptual design study of LLFP transmutation fast reactor cores in the "feasibility study" in Japan

Naganuma, Masayuki; Takaki, Naoyuki*; Aida, Tatsuya; Mizuno, Tomoyasu

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

In this paper, promising loading method of LLFP target assemblies and LLFP transmutation performances of typical cores designed in phase-II of FS are described. As for loading method, in-core loading type is found to be most promising in LLFP inventory. LLFP transmutation performances are evaluated by applying the in-core loading method. As a result, every core is confirmed to have capability to transmute LLFP amount more than generated one for 99Tc and 129I in keeping the breeding performance.

Journal Articles

Investigation on fabrication of SiC/SiC composite as a candidate material of fuel sub-assembly

Lee, J.-K.; Naganuma, Masayuki; Park, J.-S.*; Koyama, Akira*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The possibility of SiC/SiC (Silicon carbide fiber reinforced Silicon carbide) composites application for fuel sub-assembly of Fast Breeder Reactor was investigated. To select a raw material of SiC/SiC composites, a few kinds of SiC nano powder was estimated by SEM observation and XRD analysis. Furthermore, SiC monolithic was sintered from them and estimated by flexural test. SiC nano-powder which showed good sinterability, it was used for fabrication of SiC/SiC composites by Hot Pressing method. From the sintering condition of 1800, 1820$$^{circ}$$C temperature and 15, 20MPa pressure, SiC/SiC composite was fabricated and then estimated by tensile test. SiC/SiC composite, which made by 1820$$^{circ}$$C and 20MPa condition, showed the highest mechanical strength by the monotonic tensile test. SiC/SiC composite, which made by 1800$$^{circ}$$C and 15MPa condition, showed a stable fracture behavior at the monotonic and cyclic tensile test. And then, the hoop stress of ideal model of SiC/SiC composites was discussed. It was confirmed that applicability of SiC/SiC composites by Hot Pressing method for fuel sub-assembly structural material. To make it real attractive one, to maintain the reliability and safety as a high temperature structural material, the design and process study on SiC/SiC composites material will be continued.

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