Stabilization treatment of the sludge items containing nuclear materials at Plutonium Conversion Development Facility

Tanigawa, Masafumi; Nakamura, Daishi; Asakawa, Naoya*; Seya, Kazuhito*; Omori, Fumio*; Koiso, Katsuya*; Horigome, Kazushi; Shimizu, Yasuyuki

JAEA-Technology 2024-001, 37 Pages, 2024/05

JAEA-Technology-2024-001.pdf:2.32MB

At plutonium conversion development facility, the neutralization sedimentation and the coagulation sedimentation (sludge) items are stored in a polyethylene container packed in the plastic bag. The neutralization sedimentation items and the coagulation sedimentation items are stored in the globe box and storage room in the facility, respectively. Some sludge items generate gases, that swelled the plastic bag. We should ensure whether the bag swelling by visual confirmation. When the swelling is confirmed, those containers are transferred to the glove box to exchange the plastic bag for new one. By keeping the above procedure, those items were stored safely in the facility since its founding. The stabilization work for enhance the safe storage was planned to reduce the gas generation of the sludge items caused by the radiolysis of water. Those sludge items have the containing a sodium nitrate that has moisture-absorption characteristic. Therefore, the stabilization method aimed to remove the sodium nitrate from the items. The work was conducted from August 2018 to August 2022. The sodium concentration in items were reduced to 3 wt% or lower. Each stabilized sludge item packed in plastic bag were confirmed its swelling for over one year in the storage place. No gas generation from all item has been observed for more than the one year. And while both the neutralization and the coagulation sedimentation items were stored they were not the increasing of the moisture in the items. As a result, those items were evaluated that will not generate gases any more and confirmed to be stabilized after this treatment. Then, those neutralization sedimentation items were stored in powder cans and transferred to powder storage room as a retained waste. Based on the above results, risks of the gas generation from sludge items were decreased enough. Therefore, the safety of the stored sludge item was improved and confirmed.

Investigation of release characteristics of Kr gas arising in the reprocessing process

Otani, Takehisa; Suzuki, Kazuyuki; Hata, Katsuro; Kikuchi, Hideki; Nakamura, Daishi; Samoto, Hirotaka; Tanaka, Yukiyoshi

no journal, , 

The investigation of the behavior of krypton gas arising due to reprocessing of spent fuels has been performed at TRP. The whole amount of Kr gas transfers to the off-gas system through shearing and dissolution process, so it is applicable as an indicator to determine the progress of fuel dissolution. It is thought that the behavior of gaseous fission product, including Kr, in the spent fuels depends on burn-up and the type of spent fuels. In the reprocessing process, these deference are reflected to the migration rate of krypton gas between shearing off-gas system (SOG) and dissolver off-gas system (DOG). At TRP, four types of spent fuels (LWR; PWR, BWR and ATR; UO, MOX) were treated and examined about their release characteristics of krypton gas in order to understand the effect on burn-up and type of spent fuels. In this report, the results concerning the ATR-UO fuel and ATR-MOX fuel are discussed compared with the results of LWR fuel.

Solidification and stabilization for Pu nitrate solution at TRP, 4; Respond to equipment problems at PCDF

Numata, Shinji; Isomae, Hidemi; Omura, Masami; Tsutagi, Koichi; Kobayashi, Daisuke; Nakamura, Daishi; Nemoto, Masahiro; Iida, Masayoshi*; Tajiri, Kazuma*; Kurita, Tsutomu

no journal, , 

no abstracts in English

Design and construction of IFMIF/EVEDA Li test loop

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Tokoro, Daishiro*; Kanemura, Takuji; Ida, Mizuho; Watanabe, Kazuyoshi; Niitsuma, Shigeto; Wakai, Eiichi; et al.

no journal, , 

Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) were started from July 2007 under an international agreement called ITER Broader Approach. As a major Japanese activity, EVEDA Li test loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF has already designed and is under construction, in which feasibility of hydraulic stability of the liquid Li target, the purification systems of hot traps are major key issues to be validated in this loop. This presentation focuses on the engineering design of the ELTL, and its construction and commissioning.