Safety measures in the melting facilities of The Advanced Volume Reduction Facilities; Document collection of discussion meetings related to melting facilities

Iketani, Shotaro; Yokobori, Tomohiko; Ishikawa, Joji; Yasuhara, Toshiyuki*; Kozawa, Toshiyuki*; Takaizumi, Hirohide*; Momma, Takeshi*; Kurosawa, Shingo*; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.

JAEA-Review 2018-016, 46 Pages, 2018/12

JAEA-Review-2018-016.pdf:12.79MB

Japan Atomic Energy Agency (JAEA) adopts melting process for the waste processing and packaging method of radioactive miscellaneous solid waste in NSRI because melting process is effective in radioactivity evaluation, volume reduction, and stabilization treatment. Metal melting processing facilities, Incinerator, and Nonmetal melting processing facilities (hereinafter referred to as melting process facilities) have taken lots of safety measures, including measures for preventing the recurrence of the fire accidents. To exchange opinions and discuss the validity of these measures and so on with internal personnel and external experts, "Discussions on Melting Process Facilities" was held. As a document collection, this paper summarizes presentation materials of discussion meetings. Presentation materials describe "Outline of AVRF", "Safety measures in the melting facilities in WVRF", "Operation management of the melting facilities in WVRF", "Comparison of the past accident cases between facilities in and outside Japan and WVRF", and "Introduction of past accident cases and safety measures in other facilities from each committee".

Pretreatment works for disposal of radioactive wastes produced by research activities, 1

Ishihara, Keisuke; Yokota, Akira; Kanazawa, Shingo; Iketani, Shotaro; Sudo, Tomoyuki; Myodo, Masato; Irie, Hirobumi; Kato, Mitsugu; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.

JAEA-Technology 2016-024, 108 Pages, 2016/12

JAEA-Technology-2016-024.pdf:29.74MB

Radioactive isotope, nuclear fuel material and radiation generators are utilized in research institutes, universities, hospitals, private enterprises, etc. As a result, various low-level radioactive wastes (hereinafter referred to as non-nuclear radioactive wastes) are produced. Disposal site for non-nuclear radioactive wastes have not been settled yet and those wastes are stored in storage facilities of each operator for a long period. The Advanced Volume Reduction Facilities (AVRF) are built to produce waste packages so that they satisfy requirements for shallow underground disposal. In the AVRF, low-level beta-gamma solid radioactive wastes produced in the Nuclear Science Research Institute are mainly treated. To produce waste packages meeting requirements for disposal safely and efficiently, it is necessary to cut large radioactive wastes into pieces of suitable size and segregate those depending on their types of material. This report summarizes activities of pretreatment to dispose of non-nuclear radioactive wastes in the AVRF.

Evaluation of void ratio of the solidified wastes containing supercompacted wastes

Sudo, Tomoyuki; Nakashio, Nobuyuki; Osugi, Takeshi; Mimura, Ryuji; Ishihara, Keisuke; Satomi, Shinichi; Myodo, Masato; Momma, Toshiyuki; Kozawa, Kazushige

JAEA-Technology 2010-041, 38 Pages, 2011/01

JAEA-Technology-2010-041.pdf:4.73MB

The super compactor in the AVRF treats compactible metal wastes mainly generated from research reactors in the Nuclear Science Research Institute of JAEA. Those wastes are compacted with the maximum about 2,000-ton force. The supercompacted wastes are packed into the container and then immobilized with cementitious materials. The solidified wastes (containing supercompacted wastes) become an object for near surface disposal with artificial barrier. For disposal, the solidified wastes must be satisfied the technical criteria. One of the important indicators is the void ratio in the solidified wastes. In this report, we manufactured the supercompacted wastes with the ordinary treatment method for actual wastes treated in the AVRF and immobilized with a mortar grout. The void ratio of the solidified wastes were evaluated in consideration for concrete vault disposal. As a result, We confirmed the integrity of the solidified wastes from a point of view of void ratio.

Removal of penetrating pipes by dry wire-saw technology

Myodo, Masato; Kobayashi, Tadayoshi; Tomii, Hiroyuki

JAEA-Technology 2008-001, 46 Pages, 2008/03

JAEA-Technology-2008-001.pdf:19.05MB

Dry wire-saw technology was applied for removal of penetrating pipes in annex building B of JAEA's reprocessing test facility (JRTF) and concrete blocks with penetrating pipes were successfully removed. The concrete block was cracked by a silent demolition agent and was easily separated into concrete and pipes by the secondary crushing by use of a hand-breaker. Effectiveness of the dry wire-saw technology for removal of penetrating pipes was discussed and evaluated based on various work data obtained in the present activity. The use of silent demolition agent was found to be very effective for the secondary crushing of concrete blocks and its charging condition into concrete was established so as to give effective cracks for the secondary crushing. As a result of this activity, it was confirmed that the dry wire-saw technology coupled with use of a silent demolition agent was one of hopeful candidates to remove penetrating pipes safely and effectively in reprocessing facilities.

Mock-up test of remote controlled dismantling apparatus for large-sized vessels (Contract research)

Myodo, Masato; Okane, Shogo; Miyajima, Kazutoshi

JAERI-Tech 2001-025, 59 Pages, 2001/03

JAERI-Tech-2001-025.pdf:4.73MB

no abstracts in English

Application of laser to decontamination and decommissioning of nuclear facilities at JAERI

Hirabayashi, Takakuni; Kameo, Yutaka; Myodo, Masato

High-power Lasers in Civil Engineering and Architecture (Proceedings of SPIE Vol.3887), p.94 - 103, 1999/00

no abstracts in English

Characterization of sludge generated from decontamination device in Fukushima Daiichi NPS

Hinai, Hiroshi; Sato, Daisuke; Shibata, Atsuhiro; Myodo, Masato; Koma, Yoshikazu; Nomura, Kazunori

no journal, , 

We have been analyzing and characterizing the contaminated water generated in Fukushima Daiichi Nuclear Power Station (NPS) and secondary waste of the contaminated water treatment system for investigation of their waste management. As one of the contaminated water treatment system, Decontamination Device was operated from June until September 2011. The operation of this device generated secondary waste of sludge that have a high dose rate. In this device, the contaminated water was mixed with several reagents to remove some radioactive nuclides. As the result, the sludge was generated as a complex of chemically stable substances. This sludge has been stored in a concrete pit and is required to transport to another place of the site and treat for further storage. For the purposes and upcoming disposal, a small amount of the sludge was analyzed for its radioactivity, particle size distribution and others. For radiochemical analysis, the sludge was successfully dissolved into solutions and was measured for its radioactivity composition; the main nuclides were Sr-90 and Cs-137. The concentration of Sr-90 (6.6E7 [Bq/cm]) was about 10 times higher than that of Cs-137 in the sludge. And small amount of Pu-238 as alpha-ray emitting nuclide was determined. The sludge mainly contained small particles of 10 micro meter or less in diameter. The rate of sedimentation and behavior of mixing also investigated.

One piece removal of liquid waste storage tank LV-2, 1; Preparation

Kanayama, Fumihiko; Satomi, Shinichi; Myodo, Masato; Tomii, Hiroyuki; Tachibana, Mitsuo

no journal, , 

no abstracts in English

Characterization of secondary waste generated by Fukushima Daiichi Nuclear Power Station accident, 6; Analysis for characterization of sludge generated from decontamination device

Sato, Daisuke; Hinai, Hiroshi; Myodo, Masato; Koma, Yoshikazu; Shibata, Atsuhiro; Nomura, Kazunori

no journal, , 

no abstracts in English