Fundamental study for decorporation of plutonium by chelating agents; Coordination analyses of simulated plutonium with chelating agents by X-ray absorption spectroscopy

Uehara, Akihiro*; Shuhui, X.*; Sato, Ryotaro*; Matsumura, Daiju; Tsuji, Takuya; Yakumaru, Haruko*; Shiro, Ayumi*; Saito, Hiroyuki*; Tanaka, Izumi*; Ishihara, Hiroshi*; et al.

Advances in X-Ray Chemical Analysis, Japan, 53, p.223 - 229, 2022/03

no abstracts in English

Electrochemical recovery of Zr and Cd from molten chloride salts for reprocessing of used nitride fuels

Murakami, Tsuyoshi*; Hayashi, Hirokazu

Journal of Nuclear Materials, 558, p.153330_1 - 153330_7, 2022/01

https://doi.org/10.1016/j.jnucmat.2021.153330

Excess amounts of dissolution agents, CdCl and ZrCl, are required to dissolve transuranium (TRU: Pu and minor actinides) nitrides into LiCl-KCl melts at the chemical dissolution step, which is the first step in the reprocessing of used nitride fuels. We propose an electrochemical process where the remaining Zr and Cd are recovered from the melts to be recycled as dissolution agents for the chemical dissolution step, leaving TRU in the melts. Since the initial concentration ratio of CdCl/ZrCl remaining in the melts would depend on the condition of the chemical dissolution step and would vary during the proposed electrochemical recovery process, electrochemical behaviors of Zr and Cd were investigated in LiCl-KCl melts with various concentration ratios of CdCl/ZrCl at 723 K to confirm the basic feasibility of the proposed process. Potentiostatic electrolysis was performed using a liquid Cd cathode at -1.05 V (vs. Ag/AgCl), which was a more positive potential than the redox potentials of TRU on the liquid Cd electrode. The obtained results showed that the current efficiency for recovering Zr and Cd from the melts was as high as 100% regardless of the CdCl/ZrCl concentration ratio in the melts.

Zr separation from high-level liquid waste with a novel hydroxyacetoamide type extractant

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.

The Effect of final heat treatment at fabrication on the terminal solid solubility of hydrogen in Zry-4

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.

Development of separation process for Pd by extraction with 5,8-diethyl-7-hydroxy-6-dodecanone oxime

Morita, Yasuji; Yamagishi, Isao

JAEA-Research 2017-006, 27 Pages, 2017/06

JAEA-Research-2017-006.pdf:1.83MB

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.

Preparation of Zr-91 stable isotope standard solution for determination of Zr-93 by mass spectrometry

Konda, Miki; Asai, Shiho; Hanzawa, Yukiko; Magara, Masaaki

JAEA-Technology 2015-054, 22 Pages, 2016/03

JAEA-Technology-2015-054.pdf:3.44MB

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.

Comparative molecular simulation studies of oxidation reactions and hydrogen release for zirconium metals and silicon carbide under severe accident conditions

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.

Effects of heat-transfer on corrosion of zirconium in a boiling nitric acid solution

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.

The Sorption database of radionuclides for cementitious materials

Kato, Hiroshige*; Mine, Tatsuya*; Mihara, Morihiro; Oi, Takao; Honda, Akira

JNC TN8400 2001-029, 63 Pages, 2002/01

JNC-TN8400-2001-029.pdf:1.81MB

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.

Long-term durability test of acid recovery evaporators made of Ti-5% Ta aIloy and zirconium

; Fujisaku, Kazuhiko*; *; ; Koyama, Tomozo

JNC TN8410 2001-013, 255 Pages, 2001/05

JNC-TN8410-2001-013.pdf:24.24MB

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.