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Journal Articles

New precise measurement of muonium hyperfine structure interval at J-PARC

Ueno, Yasuhiro*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.

Hyperfine Interactions, 238(1), p.14_1 - 14_6, 2017/11

 Times Cited Count:2 Percentile:12.39

Journal Articles

New muonium HFS measurements at J-PARC/MUSE

Strasser, P.*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.

Hyperfine Interactions, 237(1), p.124_1 - 124_9, 2016/12

 Times Cited Count:5 Percentile:6.11

Journal Articles

Effects of manufacturing process on impact properties and microstructures of ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Tanaka, Kenya

Journal of Nuclear Materials, 455(1-3), p.480 - 485, 2014/12

 Times Cited Count:9 Percentile:28.78(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) steels are noticed as an advanced alloy durable to high-temperature and high-dose neutron irradiation environment. Japan Atomic Energy Agency, 9-12Cr-ODS martensite steels have been developed as the primary candidate material for fast reactor fuel cladding tube. They would be also good candidates for fusion reactor blanket material. In this work, two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Tensile tests, creep tests, 1/3 sized Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steel was shown to be much superior than that of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores). The ductility of full pre-alloy ODS steels were better than that of pre-mix ODS steels.

Journal Articles

Microstructure characterization of oxide dispersion strengthened steels containing metallic chromium inclusions after high-temperature thermal aging

Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Tanaka, Kenya

Materials Transactions, 54(10), p.2018 - 2026, 2013/10

 Times Cited Count:4 Percentile:64.7(Materials Science, Multidisciplinary)

Microstructure characterizations of 9Cr-oxide dispersion strengthened (ODS) steels were carried out after high-temperature thermal aging to reproduce the anomalous microstructure change that occurred in the BOR-60 irradiation test-formation of abnormally coarse and irregular precipitates a few tens of micrometers in size. In the 9Cr-ODS steel containing metallic Cr inclusions, coarse and irregular precipitates were formed nearby metallic Cr inclusions after the 750$$^{circ}$$C thermal aging for 8,000h. Based on the analyses using energy dispersive X-ray spectrometry (EDX) and electron backscattered pattern (EBSP), coarse and irregular precipitates were identified as M23C6.

Journal Articles

Microstructure and high-temperature strength of high Cr ODS tempered martensitic steels

Otsuka, Satoshi; Kaito, Takeji; Tanno, Takashi; Yano, Yasuhide; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 442(1-3), p.S89 - S94, 2013/09

 Times Cited Count:11 Percentile:24.74(Materials Science, Multidisciplinary)

The manufacturing tests of 11-12Cr ODS tempered martensitic steels were carried out, and their ferritic/martensitic duplex structures were quantitatively evaluated by three types of methods, i.e. high temperature XRD, EPMA and metallography. It was demonstrated that excessive formation of residual-alpha ferrite provided by increasing Cr can be suppressed by appropriately controlling the concentration of ferrite-forming element and austenite-forming element on the basis of the parameter "chemical driving force of $$alpha$$ to $$gamma$$ reverse transformation" as a useful indication. The 11Cr-ODS steel containing a small portion of residual-alpha ferrite was successfully manufactured. In the as-received condition, this 11Cr-ODS steel is shown to have the satisfactory creep strength and ductility as high as the 9Cr-ODS steel while 0.2% proof strength at 973K is lower than in the 9Cr-ODS steel.

Journal Articles

Evaluation of mechanical properties and nano-meso structures of 9-11%Cr ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Oba, Yojiro*; Onuma, Masato*; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 440(1-3), p.568 - 574, 2013/09

 Times Cited Count:9 Percentile:31.31(Materials Science, Multidisciplinary)

This study carried out mechanical tests and microstructure characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels, and discussed the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel from the viewpoint of high-temperature strength improvement. It was shown that the residual $$alpha$$-ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel by selecting the matrix chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion condition of nano-sized oxide particles.

JAEA Reports

Irradiation behavior analyses of oxide fuel pins for SFR high breeding cores

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-012, 13 Pages, 2013/06

JAEA-Technology-2013-012.pdf:2.34MB

A mixed oxide fuel pin concept with annular pellets and an ODS cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. This fuel concept was considered with low breeding ratio as a standard, break-even breeding cores and cores with high breeding ratio (high breeding cores). Some calculations of fuel pin irradiation performance of (U,Pu) oxide fuel and minor actinides bearing oxide fuel were conducted by a fuel performance analysis code CEDAR developed in JAEA to understand the steady state irradiation behavior of fuel pins for the cores with high breeding ratio. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated. Those results show that the MOX fuel pin having the specifications and irradiation conditions used in this investigation would be irradiated moderately up to approximately 250 GWd/t with well integrity.

JAEA Reports

Fast reactor fuel pin behavior analyses in a LOF type transient event

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-011, 10 Pages, 2013/06

JAEA-Technology-2013-011.pdf:2.02MB

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, the fast reactor fuel pin performance code CEDAR was used for calculation. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross & Stoute type gap conductance model and constant gap conductance model used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross & Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of the former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross & Stoute type gap conductance model which is thought to be realistic.

JAEA Reports

Fuel temperature analyses of metallic fuel pins for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-010, 17 Pages, 2013/06

JAEA-Technology-2013-010.pdf:2.46MB

Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. Some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation grogram developed in JAEA. Axial profile of fuel pin centerline temperature calculated by using effective fuel thermal conductivity where sodium ingress into fuel was considered fits well with actual fuel micro structures after the irradiation. The effective fuel thermal conductivity with sodium ingress is suitable for the irradiation behavior investigation.

JAEA Reports

Fuel temperature analyses at overpower of metallic fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-009, 12 Pages, 2013/06

JAEA-Technology-2013-009.pdf:1.3MB

Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. As fuel temperature analyses at overpower events are also major interest, some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation program developed in JAEA. The calculated fuel temperature at the maximum power of overpower events, 110-120% of steady state power, was around 1100K in maxim. It is clear that this temperature was low enough to avoid fuel melting in the event.

JAEA Reports

Irradiation behavior analyses of oxide fuel pins for startup core of a demonstration SFR

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-007, 17 Pages, 2013/05

JAEA-Technology-2013-007.pdf:1.69MB

As a swelling resistant austenitic steel, PNC316, is candidate cladding tube material of the first core of demonstration Sodium-cooled fast reactor (SFR). The irradiation behavior of an annular MOX fuel pin with (U,Pu) oxide fuel contained in PNC 316 cladding was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin keeps its integrity at least up to 100 GWd/t of peak burnup.

JAEA Reports

Irradiation behavior analyses of MA bearing oxide fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-006, 17 Pages, 2013/05

JAEA-Technology-2013-006.pdf:1.72MB

As a fuel concept for commercialized Sodium-cooled fast reactor (SFR) system, minor actinides (MA) bearing oxide fuel with oxide dispersion strengthened martensitic steel (ODS) cladding was considered under homogeneous TRU recycling strategy. The MA content is calculated to be around 5% of heavy metal in case of trans-uranium (TRU) feed from light water reactor (LWR) spent fuel during the transition phase from LWR to fast reactor era. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated by fuel performance analytical code CEDAR developed in JAEA to investigate the irradiation behavior of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated.

JAEA Reports

Irradiation behavior analyses of oxide fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-005, 17 Pages, 2013/05

JAEA-Technology-2013-005.pdf:1.7MB

A mixed oxide fuel pin concept with annular pellets and an oxide dispersion strengthened martensitic steel (ODS) cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. The capability of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated. Those results show that the gap of the fuel pin at fabrication had enough width not to occur the considerable fuel-cladding mechanical interaction (FCMI).

JAEA Reports

Irradiation behavior analyses of metallic fuel pins for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-004, 16 Pages, 2013/05

JAEA-Technology-2013-004.pdf:1.52MB

In order to be an alternative concept to the conventional concept consisting of mixed oxide (MOX) fuel, Metallic fuel, U-Pu(TRU)-Zr metallic fuel slug and ODS cladding were considered for Sodium-cooled fast reactor (SFR) cycle system. The capability of the U-Pu(TRU)-Zr metallic fuel with ODS cladding under a high burnup condition was calculated and conducted by a simplified calculation grogram developed in JAEA. The fuel temperature profiles, gap width profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Evaluation of the profiles of plenum gas pressure and the cladding deformation after irradiation shows that the fuel pin had enough plenum volume not to cause considerable cladding deformations by plenum gas pressure. In case of 0.4% Am bearing fuel, calculation result shows that fuel centerline temperature becomes high, but increase from U-Pu-Zr fuel is insignificant.

Journal Articles

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Science and Technology, 50(5), p.470 - 480, 2013/05

 Times Cited Count:3 Percentile:67.09(Nuclear Science & Technology)

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors such as microstructure instability and fuel pin rupture occurred. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change of 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

Journal Articles

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

Kaito, Takeji; Yano, Yasuhide; Otsuka, Satoshi; Inoue, Masaki; Tanaka, Kenya; Fedoseev, A. E.*; Povstyanko, A. V.*; Novoselov, A.*

Journal of Nuclear Science and Technology, 50(4), p.387 - 399, 2013/04

 Times Cited Count:7 Percentile:40.49(Nuclear Science & Technology)

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burnup of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed indicating good application prospects for high burnup fuel. On the other hand, peculiar irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the peculiar irradiation behaviors.

Journal Articles

Irradiation performance of oxide dispersion strengthened (ODS) ferritic steel claddings for fast reactor fuels

Kaito, Takeji; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Tanaka, Kenya

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 11 Pages, 2013/03

The oxide dispersion strengthened (ODS) ferritic steel claddings developed by Japan Atomic Energy Agency were irradiated in Joyo and BOR-60 in order to confirm their irradiation performance and thus judge their applicability to high burnup and high temperature fast reactor fuels. In Joyo, material irradiation tests up to 33 dpa were carried out at in the temperature range of 693 - 1108 K. The irradiation data were obtained concerning mainly mechanical properties and of microstructure stability. In BOR-60, fuel pin irradiation tests were conducted up to burnup of 11.9 at% and neutron dose of 51 dpa. The irradiation data were obtained concerning fuel-cladding chemical interaction, dimensional stability under irradiation and so on. These results showed the superior irradiation performance of the ODS ferritic steel claddings and their application possibility as fast reactor fuels. This paper describes the evaluation of the obtained irradiation data of ODS ferritic steel claddings.

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

Effects of neutron irradiation on tensile properties of oxide dispersion strengthened (ODS) steel claddings

Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya

Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12

 Times Cited Count:12 Percentile:25.01(Materials Science, Multidisciplinary)

The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.

Journal Articles

Oxide fuel fabrication technology development of the FaCT project, 5; Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed.

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