Microstructural evolution in tungsten binary alloys under proton and self-ion irradiations at 800C

Miyazawa, Takeshi; Kikuchi, Yuta*; Ando, Masami*; Yu, J.-H.*; Yabuuchi, Kiyohiro*; Nozawa, Takashi*; Tanigawa, Hiroyasu*; Nogami, Shuhei*; Hasegawa, Akira*

Journal of Nuclear Materials, 575, p.154239_1 - 154239_11, 2023/03

https://doi.org/10.1016/j.jnucmat.2023.154239

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*

Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06

http://doi.org/10.1109/SOFE.2015.7482333

Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.

Impacts of friction stir processing on irradiation effects in vacuum-plasma-spray coated tungsten

Ozawa, Kazumi; Tanigawa, Hiroyasu; Morisada, Yoshiaki*; Fujii, Hidetoshi*

Fusion Engineering and Design, 98-99, p.2054 - 2057, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.06.104

Reduced activation ferritic/martensitic steel, as typified by F82H, is a promising candidate for structural material of DEMO fusion reactors. To prevent plasma sputtering, tungsten (W) coating was essentially required. This study aims to examine the irradiation effects on hardness and microstructure of vacuum-plasma-spray coated W-F82H steel, with a special emphasis on the impacts of grain-refining induced by frictional stir processing (FSP). It was revealed that the hardness of the VPS-FSP W after ion-irradiation to 5.4 dpa at 800C were not remarkably changed, where bulk W usually exhibited significant irradiation hardening.

Effect of helium on irradiation creep behavior of B-doped F82H irradiated in HFIR

Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*; Myers, J.*

Fusion Science and Technology, 68(3), p.648 - 651, 2015/10

https://doi.org/10.13182/FST14-963

Pressurized tubes of F82H and B-doped F82H irradiated at 573 and 673 K up to 6dpa have been measured by a laser profilometer. The irradiation creep strain in F82H irradiated at 573 and 673 K was almost linearly dependent on the effective stress level for stresses below 260 MPa and 170 MPa, respectively. The creep strain of BN-F82H was similar to that of F82H IEA at each effective stress level except 294 MPa at 573 K irradiation. For 673 K irradiation, the creep strain of some BN-F82H tubes was larger than that of F82H tubes. It is suggested that a swelling caused in each BN-F82H because small helium babbles might be produced by a reaction of B(n, ) Li.

Evaluation of damage accumulation behavior and strength anisotropy of NITE SiC/SiC composites by acoustic emission, digital image correlation and electrical resistivity monitoring

Nozawa, Takashi; Ozawa, Kazumi; Asakura, Yuki*; Koyama, Akira*; Tanigawa, Hiroyasu

Journal of Nuclear Materials, 455(1-3), p.549 - 553, 2014/12

https://doi.org/10.1016/j.jnucmat.2014.08.043

SiC/SiC composite is a promising candidate material of fusion DEMO reactor. This paper aims to identify its damage tolerance and strength anisotropy by various characterization techniques such as acoustic emission (AE) monitoring, electrical resistivity (ER) measurement, and digital image correlation (DIC). The AE results identified that damage accumulation initiated prior to the proportional limit stress (PLS) by both tensile and compressive loadings for 2D composites. The preliminary AE waveform analysis implied that this AE detect strength corresponds to initiation of micro-cracking but the stress-strain curve shows further linearity due to the strong interfacial friction. Then fiber sliding occurred near the PLS, followed by the non-linearlity of the curve. The preliminary tensile test results using a notched specimen also suggest notch insensitivity of the composites in any loading directions. The detailed failure mechanism will eventually be discussed with ER and DIC results.

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.01.035

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

Study of safety features and accident scenarios in a fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.04.062

After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.

Key aspects of the safety study of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10

https://doi.org/10.1585/pfr.9.1405139

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.

Radiation-induced effects in physical properties of materials, 2-6; Evaluation of irradiation creep for F82H steel by using pressurized tubes

Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu

Purazuma, Kaku Yugo Gakkai-Shi, 90(1), p.64 - 67, 2014/01

http://ci.nii.ac.jp/naid/110009688791

Reduced activation ferritic/martensitic steel (RAFM) is a candidate for the material of DEMO blanket structure. The irradiation creep behavior of F82H and JLF-1 steel has been measured at 300, 400 and 500C up to 5 dpa using helium-pressurized creep tubes irradiated in HFIR. These tubes were pressurized with helium to hoop stress levels of 0400 MPa at the irradiation temperature. The results for F82H and JLF-1 with a 400 MPa hoop stress detected small creep strains ( 0.25%) after irradiation at 300C. Irradiation creep rate (creep strain/dose) was tendency to be a similar behavior for high-dose irradiated RAFM specimens in FFTF. In this paper, a procedure of irradiation creep test & evaluation was also summarized.

Microsegregation in a F82H plate

Sakasegawa, Hideo; Tanigawa, Hiroyasu

Journal of Nuclear Materials, 442(1-3), p.S18 - S22, 2013/11

https://doi.org/10.1016/j.jnucmat.2013.01.322

Through the Broader Approach (BA) activity in Japan, F82H-BA07 heat of 5 tons prepared applying electrosrag remelting (ESR) has been studied as a step toward a larger-scale melting about 20 tons. From the result of elemental mapping images using electron probe microanalysis (EPMA), micro-segregation of at least four metallic elements such as chromium, tungsten, vanadium and manganese was found as stripes parallel to the hot rolling direction. In the case of tungsten segregation, the maximum difference of content was about 1.0 wt% between the observed stripes. This difference could cause differences in nano-metric structures between stripes, and affect mechanical properties. In this presentation, we discuss how much micro-segregation should be decreased considering effects of micro-segregation on nano-metric structures and mechanical properties in addition to the result of optimization of homogenizing condition.