Morita, Keisuke; Suzuki, Hideya; Matsumura, Tatsuro; Takahashi, Yuya*; Omori, Takashi*; Kaneko, Masaaki*; Asano, Kazuhito*
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.464 - 468, 2019/09
High level liquid waste (HLLW) contains several radionuclides with half-lives longer than 10 year. For reduce environmental burden of waste disposal, minor actinoids and long-lived fission products will to be partitioned and transmuted. JAEA and Toshiba developed process for recovering Se, Zr, Pd and Cs from HLLW. Solvent extraction for Zr with novel extractant, -didodecyl-2-hydroxyacetoamide (HAA) was detailed. The HAA system showed high selectivity for Zr, as indicated by the extraction order of Zr Mo Pd Ag Sb Sn Lns Fe. The extracted species was determined as Zr(HAA)(NO)(HNO). A continuous countercurrent extraction with HAA was applied to a simulated, concentrated HLLW after Pd, Se, and Cs removal, where the quantitative extraction of Zr and Mo was effectively demonstrated.
Ueda, Yuki; Kikuchi, Kei; Sugita, Tsuyoshi; Motokawa, Ryuhei
Solvent Extraction and Ion Exchange, 37(5), p.347 - 359, 2019/07
We have newly-designed fluorous phosphate(TFP) for the effective Zr(IV) ion extractant as an alternative extractant against the conventional organic phosphate, tri--butyl phosphate(TBP). Zr(IV) ion extraction system using the TBP has many problems such as the formation of the third phase during liquid-liquid extraction. Here, we develop the fluorous extraction system based on the TFP-perfluorohexane for the Zr(IV) ion extraction to improve the Zr(IV) ion extraction system with an effective extractability and without the third phase formation. Our main findings were that the significant high extraction performance of the TFP for Zr(IV) ion as compared with TBP, and the origins of the high extraction performance of the TFP are related to the water and HNO contents in the fluorous phase and the stability of the complex, Zr(No)(TFP).
Yamauchi, Akihiro*; Amaya, Masaki
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 7 Pages, 2018/10
Differential scanning calorimetry (DSC) measurements on pre-hydrided cold worked, stress relieved and recrystallized Zry-4 cladding were performed in a temperature range between 50 and 600C in order to elucidate the effect of final heat treatment at fabrication of Zircaloy-4 (Zry-4) cladding on the terminal solid solubility during the dissolution of zirconium hydrides at heating up (TSSD). Obtained DSC curves and Metallography indicate that the initial state of hydrides affects the dissolution behavior of hydride. The Arrhenius plots of the TSSD temperatures and hydrogen contents obtained from this study revealed that cold worked samples exhibited the largest TSSD and followed by stress relieved and recrystallized samples. The results of this study indicated that the difference in microstructure due to final heat treatment at fabrication of Zry-4 cladding affects the dissolution behavior of hydrides.
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.
Morita, Yasuji; Yamagishi, Isao
JAEA-Research 2017-006, 27 Pages, 2017/06
Separation of Pd by extraction with 5,8-diethyl-7-hydroxy-6-dodecanone oxime (DEHDO) was examined by batch and continuous tests for the purpose of developing Pd separation process. Batch extraction tests using n-dodecane solution of DEHDO revealed that Pd, Zr and Mo were extracted from simulated high-level radioactive liquid wastes (HLLW) and other elements were not, and also showed that the extraction rate was a little slow and a white precipitate appeared in the aqueous phase but its formation could be avoided by raising temperature. The extracted Pd was found to be back-extracted with sodium nitrite. In the continuous extraction tests with simulated HLLW without Zr and Mo, about 98% of Pd were extracted with DEHDO-n-dodecane and 95% of the extracted Pd were back-extracted with sodium nitrite and nitric acid. Continuous extraction test with simulated HLLW with Zr and Mo showed the possibility of the simultaneous separation of Pd and Mo by DEHDO extraction.
Takeuchi, Masayuki; Aihara, Haruka; Nakahara, Masaumi; Tanaka, Kotaro*
Procedia Chemistry, 21, p.182 - 189, 2016/12
A simulation technology with electrolyte thermodynamic model has been developed to evaluate the precipitation behavior in reprocessing solution based on nitric acid solution. The simulation results were compared with the experiment data from non-radioactive simulated HLLW with ten elements and Pu-Zr-Mo solution, and the reliability of the thermodynamic model was verified. Most of the precipitation species was zirconium molybdate hydrate from the both data. It is demonstrated that the chemical species and amount of the precipitation calculated by thermodynamic model reflected well that of experiments. This study has shown the thermodynamic simulation model is one of the useful tools to estimate the behavior of precipitation from the reprocessing solution.
Konda, Miki; Asai, Shiho; Hanzawa, Yukiko; Magara, Masaaki
JAEA-Technology 2015-054, 22 Pages, 2016/03
Isotope dilution mass spectrometry (IDMS) with ICP-MS is reliable method for determination of Zr-93, which is one of the long-lived fission products found in spent nuclear fuel and high-level radioactive wastes. In order to use an isotope standard solution of zirconium as the spike for IDMS, dissolving a commercially available solid isotope standard is indispensable. Prior to the dissolution of the Zr-91 isotope standard, solubility of metal zirconium in a mixture of HNO and HF was evaluated using zirconium metal chips. Then, 2 mg of the Zr-91 isotope standard was dissolved with 0.2 mL of 1 M HNO-3 v/v% HF mixed solution, followed by adjusting the concentration of Zr-91 to approximately 1,000 g/g. IDMS, in which a natural isotopic abundance standard solution of zirconium was used as the spike, was employed for the determination of the concentration of Zr-91 in the prepared Zr-91 isotope standard solution. The concentration of Zr-91 in the prepared Zr-91 isotope standard solution was (9.61.0) 10 g/g, which is in good agreement with the predicted concentration. This indicates that the Zr-91 metal isotope standard was completely dissolved with sufficient chemical stability. Additionally, no impurities were detected in the prepared Zr-91 isotope standard solution. These positive results denote that the Zr-91 isotope standard solution with the preferable quality for IDMS of Zr-93 can be obtained by the proposed dissolution procedures.
Matsumoto, Yoshinobu*; Do, Thi-Mai-Dung*; Inoue, Masao; Nagaishi, Ryuji; Ogawa, Toru
Journal of Nuclear Science and Technology, 52(10), p.1303 - 1307, 2015/10
Effects of zirconium oxides and oxidation products of zircaloy-4 on water radiolysis were investigated to predict the hydrogen generation from the water-immersed debris after a severe accident of a nuclear power plant. Observed yields of hydrogen in water containing the oxides were measured as a function of their weight fractions. Assuming that energies of Co-60 -ray deposited to water and the oxides brought about the water radiolysis to generate hydrogen independently, the radiolysis showed an additional term of hydrogen generation due to the energy deposition to the oxides. This term seemed to be dependent on the specific surface area or particle size of oxides, but not on the crystal structure of oxides in our experimental results. The oxides in distilled water gave the strong enhancement of term. The enhancement tended to saturate with increasing the weight fraction of oxides and was not apparent in the seawater.
Machida, Masahiko; Nakamura, Hiroki; Srinivasan, S. G.*; Van Duin, A. C. T.*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05
Zircalloy has been widely employed as an excellent material covering the fuel rod. The mechanical and thermal properties have been explored by various experiments. In terms of its use as the fuel cladding, its response to oxidation reactions is an important topic when it is exposed to high temperature and high pressure steam during severe accidents. Especially, the hydrogen production accompanied by the oxidation is critical because it can lead to the crisis of the hydrogen explosion, as observed in the Fukushima Nuclear Power Plant accidents. Silicon carbide (SiC) has been considered as an alternative cladding material owing to an advantage that hydrogen production is much suppressed in the equivalent condition compared to Zircalloy. Therefore, we simulate the oxidation reaction for both materials, i.e. Zirconium metal and SiC in atomistic level by using the ReaxFF reactive force field method to simulate the chemical reaction molecular dynamics. Through such comparative studies between Zirconium and SiC in the same condition, we clarify how the temperature and the steam pressure accelerates the oxidation reaction and the resultant hydrogen production in both materials at typical severe accident conditions. The advantage using ReaxFF is that it allows us to directly trace the oxygen diffusion inside the Zirconium metal and SiC depending on the temperature and vapor pressure together with the oxidation reaction. We can compare the reaction processes in both materials. Especially, we paid attention to the rate of hydrogen production in both materials.
Kuramoto, Kenichi; Shirasu, Noriko; Yamashita, Toshiyuki
Journal of Nuclear Materials, 319(1-3), p.180 - 187, 2003/06
no abstracts in English
Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Corrosion Engineering, 52(1), p.53 - 67, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Zairyo To Kankyo, 52(1), p.35 - 43, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm3 nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Kato, Hiroshige*; Mine, Tatsuya*; Mihara, Morihiro; Oi, Takao; Honda, Akira
JNC-TN8400 2001-029, 63 Pages, 2002/01
Cementitious materials will be used for the TRU waste repository as a component of engineered barrier system. The distribution coefficients which represent the retardation of radionuclides migration for the cementitious materials would be one of the important parameter for the safety assessment. The much information of radionuclide sorption onto the cementitious materials has been accumulated through the study in the world. Therefore it is necessary to compile the information and Kd of the radionuclides reported in previous studies. In this report, the Kd of the important radionuclides, such as C, Ni, Se, Sr, Zr, Nb, Mo, Tc, Sn, I, Cs, Sm, Pb, Ra, Ac, Th, Pa, U, Np, Pu, Am, Cm, for the cementitious materials were compiled as the Sorption Database (SDB). For radionuclides to be sensitive to the redox potential, e.g. Se, Tc, Pa, U, Pu and Np, some Kds measured under the controlled atmosphere had been reported, and few Kds measured under the controlled redox potential had been reported. For Se, Mo, Sm, Cm and Ac, the distribution coefficients had not been reported, therefore distribution coefficients of Se and Mo for OPC (Ordinary Portland Cement) pastes were measured by batch sorption experiments and these data were added into the SDB.
Otaka, Osamu*; Fukui, Hiroyuki*; Funakoshi, Kenichi*; Utsumi, Wataru; Irifune, Tetsuo*; Kikegawa, Takumi*
High Pressure Research, 22(1), p.221 - 226, 2002/01
no abstracts in English
; Fujisaku, Kazuhiko*; *; ; Koyama, Tomozo
JNC-TN8410 2001-013, 255 Pages, 2001/05
Mock-ups of acid recovery evaporators which are made of Ti-5% Ta alloy and Zr were tested under inactive condition for forty thousands hours to improve a corrosion resistance of acid recovery evaporator in Tokai reprocessing plant (TRP). The mock-up unit was designed and produced referring to the specification of acid recovery evaporator in TRP and the evaporation performance of the mock-up was 1/27 of TRP. A long-term durability of both evaporators was demonstrated by results of operation data, evaporation performance and corrosion resistance. The mock-up unit did not suffer from any trouble during the running test and the operation data such as temperature, flow, concentrations of nitric acid and metal ions were fairly stable within standard condition. As for the corrosion resistance, cracks and local corrosion such as intergranular attack were not observed on both evaporators after the running test, and a corrosion of weld was not selective. The average corrosion rates at measuring points were less than 0.1mm/yr, respectively, however, thickness of the Ti-5% Ta alloy evaporator was slightly reduced at all points of vapor phase region. In addition, from the result by test coupon, it is found that both materials have low susceptibility to stress corrosion cracking in this environment. The destructive inspection showed that the mechanical properties of both materials were not degraded during the running test. Finally, the total running time of the mock-up unit is much more than a maximum running time of acid recovery evaporator made of stainless steel in TRP (nearly 15,000 hours). On the basis of the test results, an excellent durability of Ti-5% Ta alloy and Zr evaporators under was successfully demonstrated throughout the mock-up test from an engineering perspective.
Yamazaki, Satoshi*; Yamashita, Toshiyuki; Matsui, Tsuneo*; Nagasaki, Takanori*
Journal of Nuclear Materials, 294(1-2), p.183 - 187, 2001/04
no abstracts in English
JNC-TN8400 2001-022, 60 Pages, 2001/03
A numerical simulation code for the TRUEX (Transuranium Extraction) process was developed. Concentration profiles of americium and europium were calculated for some experiments of the counter current extraction system those were carried out in CPF (Chemical Processing Facility) by using the code. Calculation profiles were in agreement with the experimental results. Operational conditions were also examinted for the americium recovery experiment by the TRUEX process carried out in the Plutonium Fuel Center. It was shown that lowering the concentration of nitric acid in the scrub solution and decreasing the flow rate of solvent and strip solution was effective for improving the performance of the stripping step and reducing the volume of the waste solution. In order to find the optimum conditions for various experiments, this simulation code was modified to calculate the concentration profiles of other metal elements such as zirconium and iron and the effect of oxalic acid on the extraction behavior of the metal elements. The calculated concentration profiles of americium and europium were varied by this modification. In the experiment at CPF, the calculations were carried out to obtain recovery ratio of americium in the product stream with the amount of oxalic acid added to the process. This calculation result showed that it was possible to improve the performance of decontamination of fission products by increasing oxalic acid concentration added to the process. The calculation was also carried out for finding the optimum conditions of oxalic acid concentration added to the europium recovery process.
Ioka, Ikuo; Suga, Masataka*; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi
JAERI-Tech 2001-013, 111 Pages, 2001/03
no abstracts in English
Kato, Chiaki; Kiuchi, Kiyoshi
Journal of Acoustic Emission (CD-ROM), 19, p.53 - 62, 2001/00
no abstracts in English
Bastug, T.*; Erkoc, S.*; Hirata, Masaru; Tachimori, Shoichi
Physica E, 8(3), p.223 - 229, 2000/09
no abstracts in English