Flexible waste management system for the future application of MA P&T technology to the current high-level liquid waste

Fukasawa, Tetsuo*; Suzuki, Akihiro*; Endo, Yoichi*; Inagaki, Yaohiro*; Arima, Tatsumi*; Muroya, Yusa*; Endo, Keita*; Watanabe, Daisuke*; Matsumura, Tatsuro; Ishii, Katsunori; et al.

Journal of Nuclear Science and Technology, 61(3), p.307 - 317, 2024/03

https://doi.org/10.1080/00223131.2023.2226673

A flexible waste management system (FWM) is being developed to apply future MA partitioning and transmutation (P&T) technology to current HLLW. This FWM system will store high-level waste (HLLW) in granular form until MA partitioning and transmutation technology is realized. The feasibility of the main process was essentially confirmed by basic experiments and preliminary thermal analysis for granule production by rotary kiln from simulated HLLW and for temporary storage (50 years) of HLW granules at the HLW storage facility, respectively. The granule production experiments revealed that relatively large particles can be produced by the rotary kiln. The results of the thermal analysis showed that the small diameter canisters could be used to safely store the granules at a higher storage density than vitrified HLW. The effectiveness of the FWM system in terms of potential radiotoxicity and repository area was also evaluated, and it was shown that FWM can reduce these factors and has significant advantages in the disposal of HLW generated in current reprocessing plants. Since LWR fuel is stored for a long period of time in Japan and the operation of a reprocessing plant is expected to start soon, the FWM system is considered to be an effective system for reducing the environmental burden of HLW disposal.

Liquid decontamination using acidic electrolyzed water for various uranium-contaminated steel surfaces in dismantled centrifuge

Sakasegawa, Hideo; Nomura, Mitsuo; Sawayama, Kengo; Nakayama, Takuya; Yaita, Yumi*; Yonekawa, Hitoshi*; Kobayashi, Noboru*; Arima, Tatsumi*; Hiyama, Toshiaki*; Murata, Eiichi*

Progress in Nuclear Energy, 153, p.104396_1 - 104396_9, 2022/11

https://doi.org/10.1016/j.pnucene.2022.104396

When dismantling centrifuges in uranium-enrichment facilities, decontamination techniques must be developed to remove uranium-contaminated surfaces of dismantled parts selectively. Dismantled uranium-contaminated parts can be disposed of as nonradioactive wastes or recycled after decontamination appropriate for clearance. previously, we developed a liquid decontamination technique using acidic electrolyzed water to remove uranium-contaminated surfaces. However, further developments are still needed for its actual application. Dismantled parts have various uranium-contaminated surface features due to varied operational conditions, inhomogeneous decontamination using iodine heptafluoride gas, and changes in long-term storage conditions after dismantling. Here, we performed liquid decontamination on specimens with varying uranium-contaminated surfaces cut from a centrifuge made of low-carbon steel. From the results, the liquid decontamination can effectively remove the uranium-contaminated surfaces, and radioactive concentrations fell below the target value within twenty minutes. Although the required time should also depend on dismantled parts' sizes and shapes in their actual application, we demonstrated that it could be an effective decontamination technique for uranium-contaminated steels of dismantled centrifuges.

Separation/removal of steel surface coating film by laser cleaning

Yamane, Ikumi; Takahashi, Nobuo; Sawayama, Kengo; Nishiwaki, Hiroki; Matsumoto, Takashi; Ogawa, Jumpei; Nomura, Mitsuo; Arima, Tatsumi*

JAEA-Technology 2021-038, 18 Pages, 2022/02

JAEA-Technology-2021-038.pdf:1.61MB

We have dismantled uranium enrichment facilities in Ningyo-toge Environmental Engineering Center since their operation finished in 2001, and the total amount of metallic wastes is estimated to be about 130 thousand tons. Eighty percent of them can be disposed as nonradioactive waste (NR), but there are some steel parts possibly uranium-contaminated. We need removing painted surface of such steels and radiologically surveying to dispose them as NRs. Though painted surfaces have been conventionally removed through hand working with grinders, this manual work requires installation of green house, protective clothing, and full-face mask, in order to prevent dispersion and inhalation of airborne dusts. We desire further developments of surface cleaning techniques to reduce time, cost, workload, and secondary waste generation caused by excessive grinding. Therefore, in this study, we focused on the laser cleaning technology used for the separation and removal of paint films at construction sites. In order to improve the coating separation and removal technology for NR objects, we evaluated the coating separation and removal performance of NR steel surface by laser cleaning system, observed the coating scattering behavior by high-speed camera and investigated the coating recovery method, evaluated the laser separation and removal performance of steel surface powder, and thermodynamically evaluated the uranium compounds on steel surface. We additionally evaluated the feasibility of laser cleaning techniques in our works basing on these results, and discussed future work plans for further developments of laser cleaning techniques.

Crystal structure of ULnO (Ln = Gd, Er) solid solution

Pham, V. M.*; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Akiyama, Daisuke*; Nagai, Takayuki; Okamoto, Yoshihiro

Journal of Nuclear Materials, 556, p.153189_1 - 153189_9, 2021/12

https://doi.org/10.1016/j.jnucmat.2021.153189

The crystal structure change was evaluated for (1-x)UO-xLnO (Ln=Gd or Er; x = 0 to 0.4) samples sintered at 1973 K for 8 h under Ar and Ar-10%H atmospheres. The effect of LnO doping on the crystal structure of UO was investigated by X-ray diffraction (XRD) and X-ray absorption fine structure (XAFS). LnO doping into UO reduced the lattice parameter of UO-LnO solid solutions up to 40mol% LnO. The lattice parameters of these samples were comparable to those of stoichiometric (U,Ln)O solid solutions, that is, the O/M ratios were close to 2.00. The U L-edge XANES analysis showed that higher U oxidation states of +5 or +6 formed, in addition to + 4. The EXAFS analysis indicated that the interatomic distances of U-O and Gd-O decreased with increasing x, whereas those of Er-O may not decrease monotonically.

Molecular dynamics simulations of physical properties of water and cations in montmorillonite interlayer; Application to diffusion model

Yotsuji, Kenji*; Tachi, Yukio; Kawamura, Katsuyuki*; Arima, Tatsumi*; Sakuma, Hiroshi*

Nendo Kagaku, 58(1), p.8 - 25, 2019/00

https://doi.org/10.11362/jcssjnendokagaku.58.1_8

Molecular dynamics (MD) simulations were conducted to investigate physical properties of water and cations in montmorillonite interlayer nanopores. The swelling behaviors and hydration states were firstly evaluated as functions of interlayer cations and layer charge. The diffusion coefficients of water and cations in interlayer nanopores were decreased in comparison with those in bulk water and came closer to those in bulk water when basal spacing increased. The viscosity coefficients of interlayer water estimated indicated a significant effect of viscoelectricity at 1- and 2-layer hydration states and higher layer charge of montmorillonite. These trends from MD calculations were confirmed to be consistent with existing measured data and previous MD simulation. In addition, model and parameter related to viscoelectric effect used in the diffusion model was refined based on comparative discussion between MD simulations and measurements. The series of MD calculations could provide atomic level understanding for the developments and improvements of the diffusion model for compacted montmorillonite.

Diffusion and adsorption of uranyl ion in clays; Molecular dynamics study

Arima, Tatsumi*; Idemitsu, Kazuya*; Inagaki, Yaohiro*; Kawamura, Katsuyuki*; Tachi, Yukio; Yotsuji, Kenji

Progress in Nuclear Energy, 92, p.286 - 297, 2016/09

https://doi.org/10.1016/j.pnucene.2016.03.006

Diffusion and adsorption behavior of uranyl (UO) species is important for the performance assessment of radioactive waste disposal. The diffusion behaviors of UO, K, CO and Cl and HO in the aqueous solutions were evaluated by molecular dynamics (MD) calculations. The diffusion coefficient (De) of UO is the smallest and is 26% less than the self-diffusion coefficient of HO. For the aqueous solution with high concentration of carbonate ions, uranyl carbonate complexes: UOCO and UO(CO) can be observed. For the clay (montmorillonite or illite)-aqueous solution systems, the adsorption and diffusion behaviors of UO and K were evaluated by MD calculations. The distribution coefficients (Kd) increase with the layer charge of clay, and Kd of UO might be smaller than that of K. Further, their two-dimensional diffusion coefficients were relatively small in the adsorption layer and were extremely small for illite with higher layer charge.

Oxygen potential measurement of (PuAm)O at high temperatures

Matsumoto, Taku; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Kato, Masato; Morimoto, Kyoichi; Sunaoshi, Takeo*

Journal of Nuclear Science and Technology, 52(10), p.1296 - 1302, 2015/10

https://doi.org/10.1080/00223131.2014.986243

The oxygen potentials of (PuAm)O were measured at 1873K, 1773K and 1473K by gas equilibrium method. It was shown that following the reduction of Am at the O/M ratio above 1.96, Pu was reduced at the O/M ratio below 1.96.

Initial dissolution rate of a Japanese simulated high-level waste glass P0798 as a function of pH and temperature measured by using micro-channel flow-through test method

Inagaki, Yaohiro*; Makigaki, Hikaru; Idemitsu, Kazuya*; Arima, Tatsumi*; Mitsui, Seiichiro; Noshita, Kenji*

Journal of Nuclear Science and Technology, 49(4), p.438 - 449, 2012/04

https://doi.org/10.1080/00223131.2012.669246

Dissolution tests were performed for a simulated HLW glass by using a Micro-Channel Flow-Through (MCFT) test to evaluate the initial dissolution rate, , as a function of pH and temperature. The results indicated that the shows a "V-shaped" pH dependence at 25C, which is almost consistent with the previous results measured by using other test methods including Single Pass Flow-Through (SPFT) test. At elevated temperatures, however, the shows a "U-shaped" pH dependence with a flat bottom at neutral pH, which differs from the previous results. The results also indicated that the MCFT provided a higher value of the with a steep slope of pH dependence than the SPFT results at basic pH from 8 to 11 at 90C. With respect to the temperature dependence, the increases with temperature according to an Arrhenius law at any pH, and the apparent activation energy increases with pH, which suggests that the dissolution mechanism can change depending on pH.

Diffusion behaviors of plutonium and americium in polycrystalline urania

Sato, Isamu; Tanaka, Kosuke; Arima, Tatsumi*

IOP Conference Series; Materials Science and Engineering, 9, p.012005_1 - 012005_8, 2010/05

https://doi.org/10.1088/1757-899X/9/1/012005

UO discs and Am containing MOX discs were prepared to be diffusion couples. They were annealed for 4.5 h at 1600C in a furnace under the constant oxygen partial pressures, and then concentrations of Pu and Am across the interfaces of them were measured with EPMA. The average diffusion coefficients of Pu and Am were obtained from the concentration profiles and were between 10 and 10m/s, in which the difference between Pu and Am was undetermined. This indicates that the difference between redistributions of Pu and Am found in a PIE may not be caused by the diffusion but by the other transport phenomenon. Moreover, the diffusion behavior was affected by existence of grain boundaries and the oxygen partial pressure in the furnace during annealing. From the affection of the grain boundary to the diffusion, the "practical" diffusion estimation may need grain boundary modeling.

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.