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, Katsuhiko*; Ito, Takashi; Iwasaki, Masahiko*; et al.

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

https://doi.org/10.1007/s10751-016-1381-7

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

https://doi.org/10.1016/j.jnucmat.2014.07.075

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.

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

https://doi.org/10.1016/j.jnucmat.2013.06.010

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 to 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.

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.

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.

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.

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).

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

https://doi.org/10.1080/00223131.2013.785273

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.

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.

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

https://doi.org/10.1016/j.jnucmat.2011.05.026

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.