Development of technologies for enhanced analysis accuracy of fuel debris; Summary results of the 2020 fiscal year (Subsidy program for the project of decommissioning and contaminated water management)

Ikeuchi, Hirotomo; Koyama, Shinichi; Osaka, Masahiko; Takano, Masahide; Nakamura, Satoshi; Onozawa, Atsushi; Sasaki, Shinji; Onishi, Takashi; Maeda, Koji; Kirishima, Akira*; et al.

JAEA-Technology 2022-021, 224 Pages, 2022/10

JAEA-Technology-2022-021.pdf:12.32MB

A set of technology, including acid dissolving, has to be established for the analysis of content of elements/nuclides in the fuel debris samples. In this project, a blind test was performed for the purpose of clarifying the current level of analytical accuracy and establishing the alternative methods in case that the insoluble residue remains. Overall composition of the simulated fuel debris (homogenized powder having a specific composition) were quantitatively determined in the four analytical institutions in Japan by using their own dissolving and analytical techniques. The merit and drawback for each technique were then evaluated, based on which a tentative flow of the analyses of fuel debris was constructed.

High temperature reaction of multiple eutectic-component system; The Case of solid metallic Zr and molten stainless steel-BC

Sumita, Takehiro; Kobata, Masaaki; Takano, Masahide; Ikeda, Atsushi

Materialia, 20, p.101197_1 - 101197_11, 2021/12

https://doi.org/10.1016/j.mtla.2021.101197

Carrying-out of whole nuclear fuel materials in Plutonium Research Building No.1

Inagawa, Jun; Kitatsuji, Yoshihiro; Otobe, Haruyoshi; Nakada, Masami; Takano, Masahide; Akie, Hiroshi; Shimizu, Osamu; Komuro, Michiyasu; Oura, Hirofumi*; Nagai, Isao*; et al.

JAEA-Technology 2021-001, 144 Pages, 2021/08

JAEA-Technology-2021-001.pdf:12.98MB

Plutonium Research Building No.1 (Pu1) was qualified as a facility to decommission, and preparatory operations for decommission were worked by the research groups users and the facility managers of Pu1. The operation of transportation of whole nuclear materials in Pu1 to Back-end Cycle Key Element Research Facility (BECKY) completed at Dec. 2020. In the operation included evaluation of criticality safety for changing permission of the license for use nuclear fuel materials in BECKY, cask of the transportation, the registration request of the cask at the institute, the test transportation, formulation of plan for whole nuclear materials transportation, and the main transportation. This report circumstantially shows all of those process to help prospective decommission.

Concepts and basic designs of various nuclear fuels, 4; Metallic fuels for fast reactors and nitride fuels for ADS

Ogata, Takanari*; Takano, Masahide

Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.541 - 546, 2021/07

https://doi.org/10.3327/jaesjb.63.7_541

This is a commentary on metallic fuels for fast reactors and nitride fuels for minor actinide transmutation in accelerator driven system, as the 4th article of serial lecture on Journal of the Atomic Energy Society of Japan; Concepts and basic designs of various nuclear fuels.

Experimental evaluation of Sr and Ba distribution in ex-vessel debris under a temperature gradient

Sudo, Ayako; Sato, Takumi; Ogi, Hiroshi; Takano, Masahide

Journal of Nuclear Science and Technology, 58(4), p.473 - 481, 2021/04

https://doi.org/10.1080/00223131.2021.1879690

Dissolution behavior of Sr and Ba is crucial for evaluating secondary source terms via coolant water from ex-vessel debris accumulated at Fukushima Daiichi Nuclear Power Plant. To understand the mechanism, knowing the distribution of Sr and Ba in the ex-vessel debris is necessary. As a result of reaction tests between simulated corium and concrete materials, two layered structures were observed in the solidified sample, (A) a silicate glass-based ((Si-Al-Ca-Fe-Zr-Cr-U-Sr-Ba)-O) phase-rich layer in the upper surface region and (B) a (U,Zr)O particle-rich layer at the inner region. Measurable concentrations of Sr and Ba were observed in layer (A) (approximately 1.7 times that in the layer (B)). According to thermodynamic analysis, (U,Zr)O is predicted to solidify, in advance, in the concrete-based melt around 2177 C. Then, the residual melt is solidified as a silicate glass, and Sr and Ba are preferentially dissolved into the silicate glass. During the tests, (U,Zr)O particles sank, in advance, in the melt because of its higher density, and the silicate glass phase relocated to the surface layer. On the other hand, silicate glass containing Sr and Ba is predicted to be hardly soluble in water and chemically stable.

Survey of N isotopic enrichment plant and its cost for nitride fuel fabrication (Contract research)

Takano, Masahide

JAEA-Review 2020-080, 24 Pages, 2021/03

JAEA-Review-2020-080.pdf:1.71MB

Nitride is one of the potential fuel forms for minor actinide transmutation by the accelerator driven system. However, to avoid the C production from N by (n, p) reaction in the fuel, the special N gas highly enriched with N is needed for the fuel fabrication. To realize the availability of such gas has been an important issue. In this report, the degree of N enrichment and gas amount required for the fuel fabrication are shown first, and then among the existing isotopic enrichment methods, N cryogenic distillation is found to be a promising method from the viewpoint of constructing a huge scale plant because of its non-hazardous feature. Some commercial plants for O enrichment based on the similar method have already been operated in Japan. Its technology and components can be applied to the N enrichment plant. Assuming the supply of N gas from a cryogenic distillation plant, a series of enrichment simulation is performed to evaluate the plant size as functions of targeted degree of enrichment and annual production. By using the simulation results, the basic specifications for plant components and equipment are designed. As a result, a huge plant for annual production of 1000 kg N gas with 99% enrichment is found to be technically feasible. The N gas production cost is also evaluated to be approximately 1/30 of the current distribution price. This survey shows the availability of N gas required for the nitride fuel fabrication in both technical and economic aspects.

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.

Development of module for ADS nitride fuel performance analysis

Shibata, Hiroki; Saito, Hiroaki; Hayashi, Hirokazu; Takano, Masahide

JAEA-Data/Code 2019-023, 138 Pages, 2020/03

JAEA-Data-Code-2019-023.pdf:6.99MB

Transmutation of minor actinides in the form of nitride fuel by the accelerator driven system has been developed to reduce the radiotoxicity and volume in the radioactive wastes. Nitride fuel behavior under irradiation condition is necessary for its design and development. Nitride fuel performance analysis module based on light water reactor fuel performance code, FEMAXI-7, was developed by introducing fundamental properties of nitride pellet, 9Cr-1Mo ferrite cladding, and Pi-Bi coolant. As a result of test analysis with this module, we have understood that the nitride fuel shows excellent behavior under irradiation due to its high thermal conductivity. We found that, however, it may be a main concern that fuel cladding integrity is maintained during irradiation in which pellet-cladding mechanical interaction is increased by He gas release, low creep rate of nitride pellet at low temperatures, and high creep rate of cladding above 873 K.

Observation of simulated fuel debris using synchrotron radiation

Yoneda, Yasuhiro; Harada, Makoto; Takano, Masahide

Transactions of the Materials Research Society of Japan, 44(2), p.61 - 64, 2019/04

https://doi.org/10.14723/tmrsj.44.57

We performed three-dimensional observation of simulated fuel debris using Synchrotron Computed Tomography (CT). CT was used to make the inside of fuel debris clear. The CT observation provides that a clear contrast in the zirconia rich part and concrete rich part. Zirconia heavier than concrete moved to the lower part when crystals precipitate and aggregates near the bottom surface. As a result, phase separation occurs. The phase separation is caused by the difference in the composition ratio of zirconia, and can also be observed difference in crystal growth mode by composition ratio.

Chemical state analysis of simulated corium debris by EXAFS

Okamoto, Yoshihiro; Takano, Masahide

Progress in Nuclear Science and Technology (Internet), 5, p.200 - 203, 2018/11

https://doi.org/10.15669/pnst.5.200

Chemical state of some simulated corium debris samples containing uranium (fuel), zirconium (fuel cladding), iron (structure material), calcium (cement) and lanthanides (fission products) was investigated by synchrotron radiation based extended X-ray absorption fine structure (EXAFS) analysis. The local structure of uranium for the simulated debris was classified into fluorite UO structure and C-type structure (stabilized cubic). The UZrFeCaO sample, which consists of single phase (C-type), shows slightly shorter U-O distance. It can be concluded that the sample contains pentavalent uranium. The local structure of zirconium for U-Zr-O and U-Zr-Fe-O systems was very close to tetragonal ZrO, while that of zirconium changed to CSZ (calcia stabilized cubic) by adding calcium.