Refine your search:     
Report No.
 - 
Search Results: Records 1-12 displayed on this page of 12
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Investigation of high flux test module for the International Fusion Materials Irradiation Facilities (IFMIF)

Miyashita, Makoto; Yutani, Toshiaki*; Sugimoto, Masayoshi

JAEA-Technology 2007-016, 62 Pages, 2007/03

JAEA-Technology-2007-016.pdf:3.2MB

This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure.

JAEA Reports

Compatibility of reduced activation ferritic/martensitic steel specimens with liquid Na and NaK in irradiation rig of IFMIF

Yutani, Toshiaki*; Nakamura, Hiroo; Sugimoto, Masayoshi

JAERI-Tech 2005-036, 10 Pages, 2005/06

JAERI-Tech-2005-036.pdf:2.06MB

In the high flux region of the International Fusion Materials Irradiation Facility (IFMIF), the neutron irradiation damage for iron-based alloys will exceed 20 dpa/ year. An accurate specimen temperature measurement under a large amount of nuclear heating is a key issue but the change of heat transfer of gap between irradiation specimens and specimen holder during irradiation test is inevitable, if gap is filled with an inert gas and temperature is monitored by a thermocouple buried in the specimen holder. A solution to make heat transfer predictable is to fill the gap with a liquid metal (sodium or sodium-potassium alloy). An issue of compatibility between Reduced Activation Ferritic/Martensitic steels and the liquid metalsis addressed in this paper, and some recommendations for designing irradiation rig are presented, such as a purification control before filling liquid metals, or a careful selection of material of rig to avoid carbon mass transfer.

Journal Articles

Removal and control of tritium in lithium target for International Fusion Materials Irradiation Facility (IFMIF)

Nakamura, Hiroo; Ida, Mizuho*; Sugimoto, Masayoshi; Yutani, Toshiaki*; Takeuchi, Hiroshi

Fusion Science and Technology, 41(3), p.845 - 849, 2002/05

This paper presents the design considerations on removal and control of tritium generated in liquid lithium target of International Fusion Materials Irradiation Facility (IFMIF). In the IFMIF, intense neutrons simulating fusion condition are produced by injecting deuterium beam with a maximum energy of 40 MeV and a maxim current of 250 mA into the liquid lithium flow with a speed of 20 m/s. Tritium is produced by direct reactions of the beam with the lithium. Total production rate is estimated to be about 10 g/year.As a reference method of the tritium removal, a cold trap with a swamping method is used. As an option, yttrium getter hot trap is considered. The concentration of hydrogen isotopes in the Li flow is detected by measuring their partial gas pressure which comes through a Nb or Nb-Zr membrane. To distinguish the isotopes from the other, a quadrupole mass spectrometer is used. The off-line sampling system is also used to measure the tritium concentration in the lithium.

Journal Articles

Tritium processing and tritium laboratory in International Fusion Materials Irradiation Facility (IFMIF)

Yutani, Toshiaki*; Nakamura, Hiroo; Sugimoto, Masayoshi; Takeuchi, Hiroshi

Fusion Science and Technology, 41(3), p.850 - 853, 2002/05

no abstracts in English

Journal Articles

Status of lithium target system for International Fusion Material Irradiation Facility (IFMIF)

Nakamura, Hiroo; Ida, Mizuho*; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki*; IFMIF International Team

Fusion Engineering and Design, 58-59, p.919 - 923, 2001/11

 Times Cited Count:9 Percentile:57.68(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Reduced cost design of liquid lithium target for International Fusion Material Irradiation Facility (IFMIF)

Nakamura, Hiroo; Ida, Mizuho*; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki*

JAERI-Tech 2000-078, 17 Pages, 2001/01

JAERI-Tech-2000-078.pdf:1.93MB

no abstracts in English

Journal Articles

Staged deployment of the international fusion materials irradiation facility (IFMIF)

Takeuchi, Hiroshi; Sugimoto, Masayoshi; Nakamura, Hiroo; Yutani, Toshiaki*; Ida, Mizuho*; Jitsukawa, Shiro; Kondo, Tatsuo; Matsuda, Shinzaburo; Matsui, Hideki*; Shannon, T. E.*; et al.

Fusion Energy 2000 (CD-ROM), 5 Pages, 2001/00

no abstracts in English

JAEA Reports

Corrosion tests for evaluation of corrosion resistance of domestic oxide dispersion strengthened ferritic steels

Yutani, Toshiaki*; *; *

PNC TJ9124 90-003, 97 Pages, 1990/09

PNC-TJ9124-90-003.pdf:20.26MB

Corrosion tests with tellurium and iodine were conducted at 500, 600 and 700$$^{circ}$$C for 100 hours, to evaluate corrosion resistance to fission products. Domestic oxide dispersion strengthened ferritic steel tubes (13Cr-0.5Y$$_{2}$$O$$_{3}$$-0.5Ti-3W steel and 11Cr-0.25Y$$_{2}$$O$$_{3}$$-0.5Ti-2W steel), high strengthened ferritic/martensitic steel tube (1FK) and modified SUS316 stainless steel tube were tested. The results were summarized as follows. (1)13Cr-0.5Y$$_{2}$$O$$_{3}$$-0.5Ti-3W steel was superior to and 11Cr-0.25Y$$_{2}$$O$$_{3}$$-0.5Ti-2W steel and 1FK were inferior to the modified SUS316 stainless steel in corrosion resistance to tellurium at 700$$^{circ}$$C. (2)It was recognized that corrosion resistance to tellurium improved with chromium content at 500 and 600$$^{circ}$$C. (3)All steel specimens had the highest weight losses at 600$$^{circ}$$C in the corrosion tests by iodine. (4)13Cr-0.5Y$$_{2}$$0$$_{3}$$-0.5Ti-3W steel, 11Cr-0.25Y$$_{2}$$O$$_{3}$$-0.5Ti-2W steel and 1FK were inferior to the modified SUS316 stainless steel in corrosion resistance to iodine at 600$$^{circ}$$C.

JAEA Reports

Corrosion tests for the evaluation of FCCI susceptibility of advanced austenitic stainless steel; Advanced austenitic stainless steel of french manufacture and modified SUS316 stainless steel fablicated by air melting method

Yutani, Toshiaki*; Wada, Takashi*; Matsuzuka, Ryuji*; Yamanaka,Tsuneyasu*; Watari, Yoshio*

PNC TJ9124 87-007, 43 Pages, 1987/09

PNC-TJ9124-87-007.pdf:3.06MB

To evaluate the corrosion resistance due to fuel-cladding chemical interaction (FCCI), corrosion tests of an advanced austenitic stainless steel of French manufacture and modified SUS316 stainless steel fabricated by air melting method and doble vacuum melting method with CsOH-CsI mixture, tellurium and iodine were conducted at 700$$^{circ}$$C for 100 hours. The result were summarized as follows. (1)The CsOH-CsI mixture (CsOH-CsI=1) produced intergranular attack, (2)Tellurium and iodine produced matrix attack. (3)There were little difference in corrosion morphology between three alloys. (4)Advanced austenitic stainless steel and modified SUS316 stainless steel farbricated by air melting method had almost same corrosion resistances to CsOH-CsI mixtures and iodine as modified SUS316 stainless steel farbricated by double vacuum melting method.

JAEA Reports

None

Yutani, Toshiaki*; *; *

PNC TJ202 85-16, 144 Pages, 1985/07

PNC-TJ202-85-16.pdf:24.41MB

no abstracts in English

Oral presentation

Decalibration characteristics of N-type thermocouples due to neutron induced transmutation under the IFMIF irradiation condition

Yutani, Toshiaki; Yamauchi, Michinori; Nakamura, Hiroo; Sugimoto, Masayoshi

no journal, , 

no abstracts in English

Oral presentation

Activities summary on lithium target system and test facilities in IFMIF/EVEDA Project, 7; Engineering design of lithium target and post irradiation examination facilities in IFMIF

Sugimoto, Masayoshi; Wakai, Eiichi; Kanemura, Takuji; Kikuchi, Takayuki; Ida, Mizuho*; Watanabe, Kazuyoshi*; Niitsuma, Shigeto*; Yamamoto, Michiyoshi*; Yutani, Toshiaki*; Hirano, Michiko*

no journal, , 

The International Fusion Materials Irradiation Facility - Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) is a EU an Japan collaborative project under the framework of the Broader Approach activities for fusion energy development and, after the Intermediate Engineering Design Report has been completed in 2013, the validation test tasks have been conducted up to 2015. About the tasks in charge of Japan were completed in March and now we continue the evaluation of the testing results and examine how to reflect them to updated engineering design. In this report, we present the technical validity of the design based on the results of various validation tests and summarize the issues to be solved in future.

12 (Records 1-12 displayed on this page)
  • 1