Sumita, Takehiro; Kobata, Masaaki; Takano, Masahide; Ikeda, Atsushi
Materialia, 20, p.101197_1 - 101197_11, 2021/12
Sugawara, Takanori; Moriguchi, Daisuke*; Ban, Yasutoshi; Tsubata, Yasuhiro; Takano, Masahide; Nishihara, Kenji
JAEA-Research 2021-008, 63 Pages, 2021/10
This study aims to perform the neutronics calculations for accelerator-driven system (ADS) with a new fuel composition based on the SELECT process developed by Japan Atomic Energy Agency because the previous studies had used the ideal MA (minor actinide) fuel composition without uranium and rare earth elements. Through the neutronics calculations, it is shown that two calculation cases, with/without neptunium, satisfy the design criteria. Although the new fuel composition includes uranium and rare earth elements, the ADS core with the new fuel composition is feasible and consistent with the partitioning and transmutation (P&T) cycle. Based on the new fuel composition, the heat removal during fuel powder storage and fuel assembly assembling is evaluated. For the fuel powder storage, it is found that a cylindrical tube container with a length of 500 [mm] and a diameter of 11 - 21 [mm] should be stored under water. For the fuel assembly assembling, CFD analysis indicates that the cladding tube temperature would satisfy the criterion if the inlet velocity of air is larger than 0.5 [m/s]. Through these studies, the new fuel composition which is consistent with the P&T cycle is obtained and the heat removal with the latest conditions is investigated. It is also shown that the new fuel composition can be practically handled with respect to heat generation, which is one of the most difficult points in handling MA fuel.
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
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.
Ogata, Takanari*; Takano, Masahide
Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.541 - 546, 2021/07
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.
Kusaka, Ryoji; Kumagai, Yuta; Yomogida, Takumi; Takano, Masahide; Watanabe, Masayuki; Sasaki, Takayuki*; Akiyama, Daisuke*; Sato, Nobuaki*; Kirishima, Akira*
Journal of Nuclear Science and Technology, 58(6), p.629 - 634, 2021/06
Sudo, Ayako; Sato, Takumi; Ogi, Hiroshi; Takano, Masahide
Journal of Nuclear Science and Technology, 58(4), p.473 - 481, 2021/04
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.
Iwasa, Toma; Takano, Masahide
JAEA-Technology 2020-024, 29 Pages, 2021/03
Partitioning and transmutation of minor actinides (MA) is an important issue to reduce volume and radio-toxicity of high-level radioactive wastes. In Nuclear Science Research Institute, we have been carrying out R&D on MA-bearing nitride fuel for accelerator driven system. In the actual nitride fuel fabrication process, a special nitrogen gas highly enriched with N is required to avoid C production from N by (n,p) reaction in the fuel. For the economical use of such expensive gas, we need a nitrogen circulation refining system that can remove carbon monoxide (CO) evolved by carbothermic nitridation of oxides and can use the nitrogen gas in the closed system without loss. To develop the system, at first we listed up the performance requirements, and then designed and assembled a prototype system for laboratory-scale demonstration. The system consists of CO removal unit and circulation unit that can automatically keep the system pressure and the gas flow rate constant. As a result of demonstration on the nitridation of oxide, both units completely satisfy the requirements. We confirmed that the concept can be applied to the actual fuel fabrication with further additional function such as automatic hydrogen feed for the control of decarburization.
JAEA-Review 2020-080, 24 Pages, 2021/03
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.
Sumita, Takehiro; Kitagaki, Toru; Takano, Masahide; Ikeda, Atsushi
Journal of Nuclear Materials, 543, p.152527_1 - 152527_15, 2021/01
Fukasawa, Tetsuo*; Hoshino, Kuniyoshi*; Yamashita, Junichi*; Takano, Masahide
Journal of Nuclear Science and Technology, 57(11), p.1215 - 1222, 2020/11
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.
Uno, Masayoshi*; Nishi, Tsuyoshi*; Takano, Masahide
Comprehensive Nuclear Materials, 2nd Edition, Vol.7, p.202 - 231, 2020/08
On the thermodynamic and thermophysical properties of the actinide nitrides in Comprehensive Nuclear Materials published by Elsevier as the first edition in 2012, we have revised them by adding some brand-new data. The main topics added are the solid solubility of the actinide nitrides into the zirconium nitride matrix for transmutation fuel, the lattice expansion of actinide nitrides induced by self-irradiation damage, the influence of defects accumulation on thermal conductivity, and the thermal expansion in curium nitride lattice.
Shibata, Hiroki; Saito, Hiroaki; Hayashi, Hirokazu; Takano, Masahide
JAEA-Data/Code 2019-023, 138 Pages, 2020/03
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.
Liu, J.; Miyahara, Naoya; Miwa, Shuhei; Takano, Masahide; Hidaka, Akihide; Osaka, Masahiko
Journal of Nuclear Materials, 527, p.151819_1 - 151819_7, 2019/12
To evaluate the effect of each constituent element on the evaporation rate of ruthenium (Ru) from fission-produced alloy precipitates, the oxidation and evaporation behaviors of metallic Ru, molybdenum (Mo), palladium (Pd), rhodium (Rh) and Mo-Ru-Pd-Rh alloy powders were investigated by thermogravimetric analysis under oxidizing atmospheres from 1473 to 1723 K. The findings led to the following conclusions: (1) The quick oxidation of Mo into condensed Mo oxides can effectively suppress the oxidation and evaporation of Ru in alloy powders; (2) After the complete evaporation of Mo, the evaporation loss rate of Ru would be directly influenced by the Ru activity in the Ru-Pd-Rh alloys, which is determined by the composition of alloys.
Yoneda, Yasuhiro; Harada, Makoto; Takano, Masahide
Transactions of the Materials Research Society of Japan, 44(2), p.61 - 64, 2019/04
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.
Wagakuni Shorai Sedai No Enerugi O Ninau Kakunenryo Saikuru; Datsu Tanso Shakai No Enerugi Anzen Hosho; NSA/Commentaries, No.24, p.163 - 167, 2019/03
This article summarizes R&D status of the nitride fuel cycle for minor actinides (MA) transmutation. Status of nitride fuel fabrication, material properties and fuel performance code, pyrochemical reprocessing, and nitrogen-15 enrichment are described.
Okamoto, Yoshihiro; Takano, Masahide
Progress in Nuclear Science and Technology (Internet), 5, p.200 - 203, 2018/11
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.
Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki
Journal of Nuclear Materials, 497, p.54 - 59, 2017/12
We studied oxidative dissolution of uranium and zirconium oxide [(U,Zr)O] in aqueous HO solution. The interfacial reaction is essential for anticipating how a (U,Zr)O-based molten fuel may chemically degrade after a severe accident under influence of ionizing radiation. We conducted our experiments with (U,Zr)O powder and quantitated the HO reaction via dissolved U and HO concentrations. The dissolution yield relative to HO consumption was far less for (U,Zr)O compared to that of UO. The reaction kinetics indicates that most of the HO catalytically decomposed to O at the surface of (U,Zr)O. We confirmed the HO catalytic decomposition via O production (quantitative stoichiometric agreement). In addition, post-reaction Raman scattering spectra of the undissolved (U,Zr)O showed no additional peaks (indicating a lack of secondary phase formation). The (U,Zr)O matrix is much more stable than UO against HO-induced oxidative dissolution.
Yoneda, Yasuhiro; Tsuji, Takuya; Matsumura, Daiju; Okamoto, Yoshihiro; Takaki, Seiya; Takano, Masahide
Transactions of the Materials Research Society of Japan, 42(2), p.23 - 26, 2017/04
ZnN is a possible candidate for the diluent material for nitride fuels containing transuranium elements. Pellets of inert matrix material ZrN, and surrogate nitride fuel material DyZrN, are fabricated for the purpose of investigating the crystal structure. Lattice parameters of DyZrN followed the Vegard's low, in spite of the large lattice mismatch ( 7%) between DyN and ZrN. Local structure analysis was performed by X-ray absorption fine structure (XAFS) and atomic pair-distribution function (PDF) methods. The Zr-N nearest neighbor bond distance changed as changing the Dy composition. The complex local structure of DyN and ZrN is related to the preferable effects of ZrN.
Sato, Takumi; Shibata, Hiroki; Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki
Journal of Nuclear Science and Technology, 52(10), p.1253 - 1258, 2015/10
In order to explore the applicability of the chlorination by MoCl as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO and (UZr)O simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO and (UZr)O powder were converted to UCl or UCl and ZrCl mixture at 573 K, respectively. In the case of (UZr)Osintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl were separated from UCl by volatilization at 573 K.
Sudo, Ayako; Nishi, Tsuyoshi; Shirasu, Noriko; Takano, Masahide; Kurata, Masaki
Journal of Nuclear Science and Technology, 52(10), p.1308 - 1312, 2015/10
For understanding the control blade degradation mechanism of BWR, the thermodynamic database for the fuel assembly materials is a useful tool. Although iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data is not sufficient for the region in which the boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples were analyzed by X-ray diffraction, scanning electron microscope and energy dispersed X-ray spectrometry. The results indicate that Fe(B,C) phase only exists in the intermediate region at 1273 K and that the solidus temperature widely maintains at about 1400 K for all three samples, which are different from the calculated data using previous thermodynamic database. The difference might be originated from the over-estimations of the interaction parameter between boron and carbon in Fe(B,C).