Decommissioning state of Plutonium Fuel Fabrication Facility; Dismantling the glove box W-9 and equipment interior, and a part of tunnel F1

Nagai, Yuya; Shuji, Yoshiyuki; Kawasaki, Takeshi; Aita, Takahiro; Kimura, Yasuhisa; Nemoto, Yasunori*; Onuma, Takeshi*; Tomiyama, Noboru*; Hirano, Koji*; Usui, Yasuhiro*; et al.

JAEA-Technology 2022-039, 117 Pages, 2023/06

JAEA-Technology-2022-039.pdf:11.96MB

Japan Atomic Energy Agency (JAEA) manages wide range of nuclear facilities. Many of these facilities are required to be performed adjustment with the aging and complement with the new regulatory standards and the earthquake resistant, since the Great East Japan Earthquake and the Fukushima Daiichi Nuclear Power Station accident. It is therefore desirable to promote decommissioning of facilities that have reached the end of their productive life in order to reduce risk and maintenance costs. However, the progress of facility decommissioning require large amount of money and radioactive waste storage space. In order to address these issues, JAEA has formulated a "The Medium/Long-Term Management Plan of JAEA Facilities" with three pillars: (1) consolidation and prioritization of facilities, (2) assurance of facility safety, and (3) back-end countermeasures. In this plan, Plutonium Fuel Fabrication Facility has been selected as primary decommissioned facility, and dismantling of equipment in the facilities have been underway. In this report, size reduction activities of the glove box W-9 and a part of tunnel F-1, which was connected to W-9, are presented, and the obtained findings are highlighted. The glovebox W-9 had oxidation & reduction furnace, and pellet crushing machine as equipment interior. The duration of activity took six years from February 2014 to February 2020, including suspended period of 4 years due to the enhanced authorization approval process

Validation for multi-physics simulation of core disruptive accidents in sodium-cooled fast reactors by COMPASS code

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; et al.

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09

Dispersion and freezing of molten core material was calculated by the COMPASS code to compare with the experimental data of GEYSER. Molten core material flowed up with freezing on the pipe inner surface. As a molten pool behavior, CABRI-TPA2 experiment was analyzed, where a sphere of solid steel was surrounded by solid fuel. Power was injected to cause melting and boiling of the steel sphere. SCARABEE-BE+3 test was analyzed by COMPASS as a validation of failure of duct walls.

Code development for multi-physics and multi-scale analysis of core disruptive accidents in fast reactors using particle methods

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

A computer code, named COMPASS, is being developed for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of the MPS (Moving Particle Semi-implicit) method. The project has been carried out by six organizations for five years from FY2005 to FY2009. In this paper, the outcomes of the project in FY2007 are presented. Three validation calculations were completed by following the validation plan: melt freezing and blockage formation, molten pool boiling, and duct wall failure. The COMPASS code development was supported by basic studies of the numerical method, material science for eutectic reaction of the metal fuel, and SIMMER-III analyses.

Investigation of utilizing plutonium as mixed oxide fuel(7); Evaluation of burn-up characteristics of BWR fuels

*; *

JNC TJ9420 2001-003, 72 Pages, 2001/02

JNC-TJ9420-2001-003.pdf:3.32MB

Burn-up characteristics of BWR fuels have been investigated for scenario study of FBR introduction and the other studies on practical use of FBR. Nuclear compositions, radioactivity and thermal power of actinides and fission products in the spent fuels of present burn-up (45,000 MWd/t) and high burn-up (60,000 MWd/t) have been evaluated using an burn-up calculation code SWAT. This study was carried out as a part of JNC's feasibility study on commercialized FBR cycle system. Main conditions and results of this study are follows; (1)Evaluation code. :SWAT (open code developed by JAERI) (2)Fuel enrichment. [Bum-up:45 GWd/t, UOfuel(U235 enrichment):3.8wt%, MOX fuel(Puf enrichment):3.8wt%] [Bum-up:60 GWd/t, UOfuel(U235 enrichment):4.9wt%, MOX fuel(Puf enrichment):5.0wt%] (3)Evaluation items. Decay change in nuclear compositions, radioactivity and thermal power of actinides and fission products during 10000 years (0, 1, 2, 3, 4, 5, 6, 10, 50, 100, 1000, 10000y after burn-up).

Study on high efficiency high power Klystron, 4

Hirano, Koichiro*; Yoshikawa, Kiyoshi*; Yamamoto, Yasushi*; *; *; *; *

PNC TY1604 97-001, 43 Pages, 1997/03

PNC-TY1604-97-001.pdf:1.27MB

no abstracts in English

Study on high efficiency high power Klystron

Yoshikawa, Kiyoshi*; Onishi, Masami*; Yamamoto, Yasushi*; Toku, Hisayuki*; Hirano, Koichiro; Sakuma, Minoru

PNC TY9604 94-001, 118 Pages, 1994/03

PNC-TY9604-94-001.pdf:2.89MB

None

R&D of the next generation safety analysis methods for fast reactors with new computational science and technology, 8; Status of R&D in FY2006

Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; et al.

no journal, , 

A computer code is developed based on a particle method technology in order to simulate in detail various phenomana in core disruptive accidents in fast reactors. This report is a summary of progress during FY2006 in a five-year project of the code development.

FaCT phase-I evaluation on advanced aqueous reprocessing process, 4; Co-recovery of U-Pu-Np by solvent extraction

Koma, Yoshikazu; Shibata, Atsuhiro; Nakahara, Masaumi; Ogino, Hideki; Arai, Yoichi; Onishi, Hiroyuki*; Nakajima, Yasuo; Hirano, Hiroyasu; Washiya, Tadahiro

no journal, , 

Concerning reprocessing technology for spent FBR fuel, the results of development on solvent extraction for co-recovery of U, Pu and Np will be reviewed. Chemical processing simplified by considering the condition that moderate decontamination is allowed and centrifugal contactor that provides better plant operating ratio and less Pu inventory were developed.