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Journal Articles

Mapping measurement for beam energy position monitor system for RIKEN superconducting acceleration cavity

Watanabe, Tamaki*; Toyama, Takeshi*; Hanamura, Kotoku*; Imao, Hiroshi*; Kamigaito, Osamu*; Kamoshida, Atsushi*; Kawachi, Toshihiko*; Koyama, Ryo*; Sakamoto, Naruhiko*; Fukunishi, Nobuhisa*; et al.

Proceedings of 16th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1105 - 1108, 2019/07

Upgrades for the RIKEN heavy-ion linac (RILAC) involving a new superconducting linac (SRILAC) are currently underway at the RIKEN radioactive isotope beam factory (RIBF). It is crucially important to develop nondestructive beam measurement diagnostics. We have developed a beam energy position monitor (BEPM) system which can measure not only the beam position but also the beam energy simultaneously by measuring the time of flight of the beam. We fabricated 11 BEPMs and completed the position calibration to obtain the sensitivity and offset for each BEPMs. The position accuracy has been achieved to be less than $$pm$$ 0.1 mm by using the mapping measurement.

Journal Articles

Development of beam energy position monitor system for RIKEN superconducting acceleration cavity

Watanabe, Tamaki*; Imao, Hiroshi*; Kamigaito, Osamu*; Sakamoto, Naruhiko*; Fukunishi, Nobuhisa*; Fujimaki, Masaki*; Yamada, Kazunari*; Watanabe, Yutaka*; Koyama, Ryo*; Toyama, Takeshi*; et al.

Proceedings of 15th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.49 - 54, 2018/08

no abstracts in English

Journal Articles

Development of beam energy and position monitor system at RIBF

Watanabe, Tamaki*; Fukunishi, Nobuhisa*; Fujimaki, Masaki*; Koyama, Ryo*; Toyama, Takeshi*; Miyao, Tomoaki*; Miura, Akihiko

Proceedings of 14th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1112 - 1117, 2017/12

no abstracts in English

Journal Articles

Review of five investigation committees' reports on the Fukushima Dai-ichi Nuclear Power Plant severe accident; Focusing on accident progression and causes

Watanabe, Norio; Yonomoto, Taisuke; Tamaki, Hitoshi; Nakamura, Takehiko; Maruyama, Yu

Journal of Nuclear Science and Technology, 52(1), p.41 - 56, 2015/01

 Times Cited Count:7 Percentile:60.95(Nuclear Science & Technology)

On March 11, 2011, the Tohoku District-off the Pacific Ocean Earthquake and the subsequent tsunami resulted in the severe core damage at TEPCO's Fukushima Dai-ichi Nuclear Power Plant Units 1-3, involving hydrogen explosions at Units 1, 3, and 4 and the large release of radioactive materials to the environment. Four independent committees were established by the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and TEPCO to investigate the accident and published their respective reports. Also, the Nuclear and Industrial Safety Agency carried out an analysis of accident causes to obtain the lessons learned from the accident and made its report public. This article reviews the reports and clarifies the differences in their positions, from the technological point of view, focusing on the accident progression and causes. Moreover, the undiscussed issues are identified to provide insights useful for the near-term regulatory activities including accident investigation by the Nuclear Regulation Authority.

Journal Articles

Review of five investigation committees' reports on the Fukushima Dai-ichi Nuclear Power Plant severe accident; Focusing on accident progression and causes

Watanabe, Norio; Yonomoto, Taisuke; Tamaki, Hitoshi; Nakamura, Takehiko; Maruyama, Yu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 12(2), p.113 - 127, 2013/06

On March 11, 2011, the Tohoku District - off the Pacific Ocean Earthquake and the subsequent tsunami resulted in the severe core damage at the TEPCO's Fukushima Dai-ichi Nuclear Power Station Units 1-3, involving hydrogen explosions at Units 1, 3, and 4 and the large release of radioactive materials to the environment. The four independent committees were established by the Government, the Diet of Japan and the Rebuild Japan Initiative Foundation as well as TEPCO to investigate the accident and published their respective reports. Also, the Nuclear and Industrial Safety Agency carried out the analysis of accident causes to obtain the lessons learned from the accident and made its report public. This article reviews the reports and clarifies the differences in their positions, from the technological point of view, focusing on the accident progression and causes. As well, the undiscussed issues are identified to provide insights useful for the near-term regulatory activities including accident investigation by the Nuclear Regulation Authority.

Journal Articles

Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Shibayama, Tamaki*; Watanabe, Seiichi*; Takahashi, Heishichiro

Journal of Nuclear Materials, 398(1-3), p.59 - 63, 2010/03

 Times Cited Count:9 Percentile:56.9(Materials Science, Multidisciplinary)

The in-reactor creep rupture tests of 11Cr-0.5Mo-2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using Materials Open Test Assembly in the Fast Flux Test Facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 to 1013 K to fast neutron doses ranging from 3.5 to 102 dpa. The results of post irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

Journal Articles

Performance of semiconductor radiation sensors for simple and low-cost radiation detector

Tanimura, Yoshihiko; Birumachi, Atsushi; Yoshida, Makoto; Watanabe, Tamaki*

Radioisotopes, 57(12), p.733 - 738, 2008/12

In order to develop simple but reliable radiation meters for the general public, photon detection performances of radiation sensors have been studied both experimentally in photon calibration fields and by Monte Carlo simulations. A silicon $$_{it p-i-n}$$ photodiode and a CdTe detector were selected for the low-cost sensors. Their energy responses to ambient dose equivalent H$$^*$$(10) were evaluated over the energy range from 60 keV to 2 MeV. The response of the CdTe decreases markedly with increasing photon energy. On the other hand, the photodiode has the advantage of almost flat response above 150 keV. The sensitivities of these sensors are 4 to 6 cpm for the natural radiation. Detection limits of the radiation level are low enough to know the extreme increase of radiation due to emergency situations of nuclear power plants, fuel treatment facilities and so on.

Journal Articles

Development of probabilistic safety assessment method for mixed oxide fuel fabrication facilities

Tamaki, Hitoshi; Yoshida, Kazuo; Watanabe, Norio; Muramatsu, Ken

Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(2), p.125 - 135, 2006/06

A Probabilistic Safety Assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities was developed. The procedure is a "two-part five-step" approach which takes characteristics of MOX fuel fabrication facilities into consideration. In the first part, so-called preliminary PSA, the hazard analysis approach was applied, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (FFMEA) and Risk Matrix Analysis. The FFMEA analyzes a variety of functions of equipment composing the facility to identify potential abnormal events exhaustively. In the Risk Matrix Analysis, these potential events are screened to select abnormal events as candidates to be analyzed in the second part, using two-dimensional matrix based on the likelihood evaluated by probabilistic index method and maximum unmitigated radioactive release calculated by the Five Factor Formula. For the selected abnormal events, in the second part, so-called detailed PSA, accident sequences, their occurrence frequencies and consequences are analyzed. These three analysis steps correspond to PSA procedure for nuclear power plant. The applicability of the PSA procedure was demonstrated through the trial application to model plant of MOX fuel fabrication facility.

Journal Articles

Hazard analysis approach with functional FMEA in PSA procedure for MOX fuel fabrication facility

Tamaki, Hitoshi; Yoshida, Kazuo; Watanabe, Norio; Muramatsu, Ken

Proceedings of International Topical Meeting on Probabilistic Safety Analysis (PSA '05) (CD-ROM), 11 Pages, 2005/00

A probabilistic safety assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities is being developed applicable to nuclear facilities at Japan Atomic Energy Research Institute (JAERI). As part of the PSA procedure, the approach to hazard analysis was established, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (Functional FMEA) and Risk Matrix Analysis. In the Functional FMEA, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second step, these potential events are screened to select abnormal events as candidate events to be analyzed for frequency and consequence by using two-dimensional matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. The applicability of the hazard analysis approach established was demonstrated through the trial application of the PSA procedure being developed to model plant of MOX fuel fabrication facility.

Journal Articles

Hazard identification of criticality accidents at the JCO facility

Tamaki, Hitoshi; Watanabe, Norio*; Muramatsu, Ken

Proceedings of the 2001 Topical Meeting on Practical Implementation of Nuclear Criticality Safety (CD-ROM), 10 Pages, 2001/11

no abstracts in English

JAEA Reports

Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

Watanabe, Norio; Tamaki, Hitoshi

JAERI-Review 2000-006, p.56 - 0, 2000/04

JAERI-Review-2000-006.pdf:3.78MB

no abstracts in English

Oral presentation

Fundamental study for development of austenitic ODS steel; Effect of nano-particle dispersion on radiation-induced defect formation

Yamashita, Shinichiro; Otsuka, Satoshi; Watanabe, Masashi*; Uchida, Yosuke*; Suda, Takanori*; Hashimoto, Naoyuki*; Onuki, Somei*; Shibayama, Tamaki*

no journal, , 

no abstracts in English

Oral presentation

Fundamental study for nano-particle dispersion strengthened austenitic steel creation, 3; Characterization of complex oxide precipitated during heat treatment

Oka, Hiroshi*; Watanabe, Masashi*; Hashimoto, Naoyuki*; Kinoshita, Hiroshi*; Shibayama, Tamaki*; Onuki, Somei*; Yamashita, Shinichiro; Otsuka, Satoshi

no journal, , 

In this study, ODS austenitic stainless steels based on an advanced SUS316 steel has been developed by mechanically alloying (MA) and hot extrusion with the addition of minor alloying elements. The chemical composition of ODS316 is Fe-16Cr-13Ni-0.35Y$$_{2}$$O$$_{3}$$-0.1Ti-0.6Hf. Thin foils for transmission electron microscope (TEM) examination were prepared with electro-polishing. HRTEM and EDS were used for characterization of oxide particles in ODS-316, especially precipitation behavior and chemical composition before and after heat treatment. Microstructural observation revealed that the ODS316 has grains of 0.5-1$$mu$$m in diameter and complex oxides (Y-Ti-Hf-O) distributed in matrix. The strain contrast was observed around oxide particles, suggesting coherency of oxide particles. HR-TEM observation revealed that a part of faceted particles have coherency. Interface between matrix and oxide particles after heat treatment was also carefully investigated.

Oral presentation

High-temperature tensile properties of the grain boundary engineered NIMONIC PE16

Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Tokita, Shun*; Fujii, Hiromichi*; Sato, Yutaka*; Kokawa, Hiroyuki*

no journal, , 

In order to improve ductility loss by helium embrittlement (or grain boundary embrittlement) induced under high temperature and neutron irradiation dose in nickel alloys which are expected to have high-temperature phase stability under non-irradiation, the grain boundary engineering was applied for NIMONIC PE16 to enhance the grain boundary strength. And, its high temperature tensile properties under non-irradiation were investigated as the first approach. As a result, the temperature dependence of the yield stress in the grain boundary engineered (GBE) PE16 was similar to that in NIMINIC PE16, but the yield stress was slightly lower and the uniform elongation was slightly higher at each temperature in GBE PE16 comparing to NIMINIC PE16. This would be caused by grain coarsening due to some heat treatments. If the gain size of GBE PE16 is optimized, tensile properties of GBE PE16 would be the same or more than that of NIMONIC PE16.

Oral presentation

Evaluation of mechanical property in grain boundary character distribution-optimized Ni-based alloy

Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*

no journal, , 

Recent grain boundary structure studies have shown that optimal distribution of a high frequency of coincidence site lattice boundaries and consequent discontinuity of random boundary network in the material is one of very effective methods to enhance the intergranular corrosion resistance. This advantageous property, one of important ones for structural material of nuclear reactor, can be obtained through simple thermomechanical treatment process without any change of original chemical composition. In this study, grain boundary character distribution(GBCD)-optimized Ni-based alloy (PE16) has been developed as a prospective high-performance nuclear reactor material by grain boundary engineering processing, and then tensile behavior of GBCD-optimized Ni-based alloy was investigated to evaluate the effect of grain boundary engineering processing on mechanical property. The result of tensile test at room temperature showed that tensile strength of GBCD-optimized PE16 was somewhat lower than that of as-received PE16. However, no significant change was confirmed in elongation property. Details on tensile behavior analyses would be discussed in the conference.

Oral presentation

High temperature tensile properties of the grain-boundary-engineered Ni-base alloy

Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Tokita, Shun*; Fujii, Hiromichi*; Sato, Yutaka*; Kokawa, Hiroyuki*

no journal, , 

no abstracts in English

Oral presentation

Evaluation of mechanical property in grain boundary character distribution-optimized Ni-based alloy

Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*

no journal, , 

17 (Records 1-17 displayed on this page)
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