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JAEA Reports

Study on the evaluation method to determine the radioactivity concentration in radioactive waste generated from post-irradiation examination facilities, 2

Tsuji, Tomoyuki; Hoshino, Yuzuru; Sakai, Akihiro; Sakamoto, Yoshiaki; Suzuki, Yasuo*; Machida, Hiroshi*

JAEA-Technology 2017-010, 75 Pages, 2017/06

JAEA-Technology-2017-010.pdf:2.31MB

It is necessary for reasonable disposal to be studied on evaluation methods to determine radioactivity concentrations in the radioactive wastes, which is generated from post-irradiation examination (PIE) facilities, for establishment of reasonable confirmation methods concerning radioactive wastes generated from research, industrial, and medical facilities. It has been chosen the PIE facilities of NUCLEAR DEVELOPMENT CORPORATION as a model for this study. As a result, it has been confirmed that the theoretical methods are applied for the important nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241 and Cm-244).

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration waste generated from post-irradiation examination facilities

Hoshino, Yuzuru; Sakamoto, Yoshiaki; Muroi, Masayuki*; Mukai, Satoru*

JAEA-Technology 2015-015, 96 Pages, 2015/07

JAEA-Technology-2015-015.pdf:20.34MB

In order to dispose of the radioactive waste which generates from post-irradiation examination (PIE) facilities, the common evaluation method of radioactivity in wastes from PIE should be established by the actual data such as radioactivity values and the theoretical calculation. In this study, the radioactivity concentrations of 17 nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-244) in combustible wastes stored in NUCLEAR DEVELOPMENT CORPORATION were measured from 3 samples and the radioactivity was calculated by ORIGEN-2 based on initial contents and operation record of the spent fuel. From the comparison of the obtained data by the radiological measurement with the calculated values, the subject to be solved for establishment of the radioactivity evaluation method for PIE was extracted.

JAEA Reports

An Investigation of post-irradiation examination facility for fusion blanket materials

Yamada, Hirokazu*; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Ishitsuka, Etsuo; Uchida, Munenori*; Akutsu, Yoichi; Motoki, Ryozo; Watanabe, Wataru; Hirata, Shogo*

JAERI-Tech 2004-036, 138 Pages, 2004/03

JAERI-Tech-2004-036.pdf:4.07MB

This document shows the model case concerned with the utilization of constructed facility for the construction of post irradiation examination facility for fusion blanket. In this report, the radioisotope application and development laboratory is considered as constructed facility for post irradiation examination facility and we planned to improve the radioisotope application and development laboratory to the facility of post irradiation examination for fusion material to reduce the construction fee for this facility and promotion of efficiency by concentration of any equipment for fusion material research.

JAEA Reports

Selection of important nuclides from the viewpoint of safety assessment for disposal of radioactive waste arising from research institutes, 2; Analytical scheme for nuclide composition in radwastes arising from research and testing reactor and post-irradiation examination facilities

Asai, Shiho; Sakai, Akihiro; Yoshimori, Michiro; Kihara, Shinji

JAERI-Tech 2003-071, 46 Pages, 2003/08

JAERI-Tech-2003-071.pdf:4.31MB

As part of survey concerning radioactive inventory of radwastes arising from research institutes, an analytical scheme has been studied for verifying their nuclide composition based on calculation analysis and record, in consideration of the its characteristics. In this study, radwastes are used as samples, which arise from research and testing reactors and post-irradiation examination facilities (PIE facilities). Though separation procedures are simplified and rationalized, quantifiable values could be obtained with relative errors under 10% for almost all the samples containing nuclides like $$^{59}$$Ni and $$^{238}$$U which concentrations are low, and recoveries were high on the whole. These show that the analytical scheme is useful for chemical separation of radwastes arising from research and testing reactors and PIE facilities.

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.

Oral presentation

R&D of lithium target and test facilities for the fusion neutron source

Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Kikuchi, Takayuki; Ito, Yuzuru*; Hoashi, Eiji*; Yoshihashi, Sachiko*; Horiike, Hiroshi*; et al.

no journal, , 

no abstracts in English

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