Controlled release of radioactive krypton gas

Watanabe, Kazuki; Kimura, Norimichi*; Okada, Jumpei; Furuuchi, Yuta; Kuwana, Hideharu*; Otani, Takehisa; Yokota, Satoru; Nakamura, Yoshinobu

JAEA-Technology 2023-010, 29 Pages, 2023/06

JAEA-Technology-2023-010.pdf:3.12MB

The Krypton Recovery Development Facility reached an intended technical target (krypton purity of over 90% and recovery rate of over 90%) by separation and rectification of krypton gas from receiving off-gas produced by the shearing and the dissolution process in the spent fuel reprocessing at the Tokai Reprocessing Plant (TRP) between 1988 and 2001. In addition, the feasibility of the technology was confirmed through immobilization test with ion-implantation in a small test vessel from 2000 to 2002, using a part of recovered krypton gas. As there were no intentions to use the remaining radioactive krypton gas in the krypton storage cylinders, we planned to release this gas by controlling the release amount from the main stack, and conducted it from February 14 to April 26, 2022. In this work, all the radioactive krypton gas in the cylinders (about 7.110 GBq) was released at the rate of 50 GBq/min or less lower than the maximum release rate from the main stuck stipulated in safety regulations (3.710 GBq/min). Then, the equipment used in the controlled release of radioactive krypton gas and the main process (all systems, including branch pipes connected to the main process) were cleaned with nitrogen gas. Although there were delays due to weather, we were able to complete the controlled release of radioactive krypton gas by the end of April 2022, as originally targeted without any problems such as equipment failure.

Development of engineering scale extraction chromatography separation system, 2; Spray drying granulation of silica support for adsorbent

Hasegawa, Kenta; Goto, Ichiro*; Miyazaki, Yasunori; Ambai, Hiromu; Watanabe, So; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 5 Pages, 2023/05

Flexible fuel cycle system for the effective management of plutonium

Fukasawa, Tetsuo*; Hoshino, Kuniyoshi*; Yamashita, Junichi*; Takano, Masahide

Journal of Nuclear Science and Technology, 57(11), p.1215 - 1222, 2020/11

https://doi.org/10.1080/00223131.2020.1779144

The flexible fuel cycle initiative system (FFCI system) has been developed to reduce spent fuel (SF) amounts, to keep high availability factor for the reprocessing plant and to increase the proliferation resistance for the recovered Pu. The system separates most U from the SF at first, and the residual material called recycle material (RM) which contains Pu, minor actinides, fission products and remaining U will go to Pu(+U) recovery from the RM for Pu utilizing reactor in future. The Pu utilizing reactor is FBR or LWR with MOX fuel. The RM is the buffer material between SF reprocessing and Pu utilizing reactor with compact size and high proliferation resistance, which can suppress the amount of relatively pure Pu. The innovative technologies of FFCI are most U separation and temporary RM storage. They are investigated by the literature survey, fundamental experiments using simulated material and analyses using simulation code. This paper summarizes the feasibility confirmation results of FFCI.

Technological development of the particle size adjustment of dry recovered powder

Segawa, Tomoomi; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Kawaguchi, Koichi; Ishii, Katsunori; Sato, Hisato; Fukasawa, Tomonori*; Fukui, Kunihiro*

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.738 - 745, 2019/09

In the MOX fuel fabrication process, the dry grinding technology of mixed oxide pellets have been developed for the effective use of nuclear fuel materials. To develop a technology to control the particle size of dry recovered powder, the performance of the buhrstone mill and the collision plate type jet mill were studied using a simulated powder of particle size distribution about 500 m. We found that the particle size can be controlled at the range of about 250 m or less by both by adjusting the clearance between the grinding wheels of the buhrstone mill, and the clearance and elevation angle of the clarification zone of the collision plate type jet mill. And furthermore, the collision plate type jet mill is considered to be suitable for particle size control because the operating parameters of the classifier can be finely adjusted.

Development of U and Pu co-processing process; Demonstration of U, Pu and Np Co-recovery with centrifugal contactors

Kudo, Atsunari; Kurabayashi, Kazuaki; Yanagibashi, Futoshi; Sasaki, Shunichi; Sato, Takehiko; Fujimoto, Ikuo; Obu, Tomoyuki

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

The Co-processing process is the extraction process to recover Pu/U mixed product solution with given Pu/U ratio for improving of nuclear proliferation resistance. In addition, Np is also recovered with U and Pu because Np is one of minor actinides and a long-lived radionuclide and Np has the extractability into TBP solvent. Development of its flowsheet achieves to decrease environmental effect of waste materials. The orientation of development about Co-processing process is to demonstrate of reprocessing the future spent fuels from a LWR, a LWR-MOX hybrid, and a FR-MOX with one cycle. We demonstrated by use of miniature reflux-type centrifugal contactors at the partitioning unit. The test conditions of the Pu/U ratio in the loaded solvents were 1%, 3%, and 5% considering the composition of spent fuels. We used the HAN as the reductant of Np (VI) for back extraction. The results of these tests were very good. We got the prospect of U, Pu, and Np Co-processing flowsheet.

Current status of a decommissioning project in the Enrichment Engineering Facilities; Results in the first-half of the fiscal year of 2014

Matsumoto, Takashi; Morimoto, Yasuyuki; Takahashi, Nobuo; Takata, Masaharu; Yoshida, Hideaki; Nakashima, Shinichi; Ishimori, Yuu

JAEA-Technology 2015-036, 60 Pages, 2016/01

JAEA-Technology-2015-036.pdf:9.15MB

The Enrichment Engineering Facilities of the Ningyo-toge Environmental Engineering Center was constructed in order to establish the technical basis of the uranium enrichment plant in Japan. Uranium enrichment tests, using natural and reprocessed uranium, were carried out from 1979 to 1990 at two types of plants in the facilities. UF handling equipment and Supplemental equipment in these plants are intended to be dismantled by 2019 in order to make places for future projects, for example, inventory investigation, precipitation treatment, etc. This report shows the basic plan of this decommissioning project and presents the current state of dismantling in the first-half of the fiscal year of 2014, with indicating its schedule, procedure, situation, results, and so on. The dismantled materials generated amounted to 37 mesh containers and 199 drums, and the secondary waste generated amounted to 271.4 kg.

Demonstration of high-flux photon generation from an ERL-based laser Compton photon source

Nagai, Ryoji; Hajima, Ryoichi; Mori, Michiaki; Shizuma, Toshiyuki; Akagi, Tomoya*; Araki, Sakae*; Honda, Yosuke*; Kosuge, Atsushi*; Terunuma, Nobuhiro*; Urakawa, Junji*

Proceedings of 6th International Particle Accelerator Conference (IPAC '15) (Internet), p.1607 - 1609, 2015/06

http://accelconf.web.cern.ch/AccelConf/IPAC2015/papers/tupje002.pdf

Accelerator and laser technologies required for laser Compton scattering (LCS) photon source based on an energy-recovery linac (ERL) have been developed at the Compact ERL (cERL) facility. A high-flux, energy tunable, and monochromatic photon source such as the ERL-based LCS photon source is necessary for nondestructive assay of nuclear materials. For the demonstration of the ERL-based LCS photon generation, a laser enhancement cavity was installed at the recirculation loop of the cERL. The electron beam energy, the laser wavelength, and the collision angle are 20 MeV, 1064 nm, and 18 deg., respectively. The calculated maximum energy of the LCS photons is about 7 keV. A silicon drift detector (SDD) with active area of 17 mm placed 16.6 m from the collision point was used for observation of the LCS photons. As a result of the measurement, the flux on the detector, central energy, and energy width of the LCS photons were obtained as 1200 /s, 6.91 keV, and 81 eV, respectively.

Development of the beamline flux monitor for the laser Compton-scattered photon source

Nagai, Ryoji; Hajima, Ryoichi; Mori, Michiaki; Shizuma, Toshiyuki; Akagi, Tomoya*; Kosuge, Atsushi*; Honda, Yosuke*; Urakawa, Junji*

Proceedings of 11th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.839 - 842, 2014/10

http://www.pasj.jp/web_publish/pasj2014/proceedings/PDF/SAP1/SAP113.pdf

In order to demonstrate accelerator and laser technologies required for a laser Compton scattering (LCS) photon generation, a LCS photon source is under construction at the Compact ERL (cERL). We considered the flux monitors for the adjustment LCS photon source. A thin scintillator detector and a silicon drift detector are employed as flux monitors and are installed at the upstream part of the LCS beamline. The background signal level due to the bremsstrahlung of the electron beam was measured by a CsI(pure) scintillator. In the result of the measurement, the background signal is acceptable level for the flux monitors.

Estimation of the e-gun timing-jitter influence in an energy-rocovery linac

Nagai, Ryoji; Sawamura, Masaru; Hajima, Ryoichi; Kikuzawa, Nobuhiro; Nishimori, Nobuyuki; Nishitani, Tomohiro; Minehara, Eisuke

Proceedings of 2nd Annual Meeting of Particle Accelerator Society of Japan and 30th Linear Accelerator Meeting in Japan, p.376 - 378, 2005/07

no abstracts in English

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Katayama, Kazunari*; Takeishi, Toshiharu*; Manabe, Yusuke*; Nagase, Hiroyasu*; Nishikawa, Masabumi*; Miya, Naoyuki

Journal of Nuclear Materials, 340(1), p.83 - 92, 2005/04

https://doi.org/10.1016/j.jnucmat.2004.11.005

no abstracts in English