PROGRESS IN NUCLEAR ENERGY
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2023
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IF 2.7
Wei Zhang,Hongwei Zhu,Zelin Gao,Gang Xu,Xiaotong Chen,Bing Liu,Yaping Tang
As one of potential treatments for irradiated graphite (i-graphite) waste, the oxidation method has been attracting more concern due to features of high treating efficiency and significant volume reduction. The release of fission products during oxidation is of great significance for further off-gas decontamination procedures. Focusing on the release behavior of volatile cesium (Cs), Cs-loaded graphite was prepared by impregnation in this work followed by definite distribution and chemical speciation analyzing. Results showed that the increasing temperature and humidity during the oxidation were evidenced to promote the release of Cs compounds, and the chemical speciation of released compounds was prominently dependent on oxidizing temperature. When the oxidation temperature was around 700 °C, Cs 2 O was predominately released with the morphology of spherical particles. When the oxidation temperature was increased to 900 °C, the predominant releasing composition turned to Cs 2 CO 3 with morphologies of both spherical particles and nanorod-like shapes. The different predominant releasing chemical compositions at different temperatures mainly resulted from the kinetics of graphite oxidization and temperature synergistically. These findings will provide theoretical foundations to the environmental evaluation of i-graphite thermal oxidation .
研究领域:
核科学与工程
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环境科学
✓ 已验证
DOI
高影响因子
INORGANIC CHEMISTRY
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2018
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IF 4.7
Fang-Li Fan,Zhi Qin,Shi-Wei Cao,Cun-Min Tan,Qing-Gang Huang,De-Sheng Chen,Jie-Ru Wang,Xiao-Jie Yin,Chao Xu,Xiao-Gui Feng
Here, we propose the use of carboxyl-functionalized ionic liquid, [Hbet][Tf2N], to separate the fission products from spent nuclear fuels. This innovative method allows the selective dissolution of neutron poisons, lanthanides oxide, as well as some fission products with high yield, leaving most of the UO2 matrix and minor actinides behind in the spent nuclear fuel and accomplishing the actinides recovery as a group. Water-saturated [Hbet][Tf2N] can dissolve lanthanides oxide from simulated spent nuclear fuel with a dissolution ratio of 100% at 40 °C. However, the dissolution of uranium is almost negligible (<1%) under the same conditions. This big difference in dissolution provides a novel separation approach to spent nuclear fuel recycling and may open new perspectives for spent nuclear fuel reprocessing. The recovery of Nd and U from metal-loaded ionic liquids and the recyclability of the ionic liquid [Hbet][Tf2N] have also been investigated. Furthermore, a U/x value related to the lattice energy U of metal compound MxOy is used to elaborate the solubility. This work represents the first case for efficient fission products removal by selective dissolution, avoiding the complete dissolution of spent nuclear fuel, the producing of the large high-level radioactive waste, and reducing environmental hazards.
研究领域:
核化学与核燃料循环
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环境科学
✓ 已验证
产品关联
DOI
PubMed
高影响因子
SEPARATION AND PURIFICATION TECHNOLOGY
·
2023
·
IF 8.6
Hong Zhang,Qian Du,Xin Du,Zhanglian Xu,Shaoqiang Guo,Wentao Zhou
The electrorefining of spent nuclear fuels in a molten LiCl-KCl salt electrolyte recovers the valuable actinides but also releases radioactive fission products in the salt. To minimize the nuclear waste volume, dissolved fission products such as rare earths (REs) must be removed from the electrorefining salt and immobilized into proper wasteforms. Here we propose a two-step approach for treating the electrorefining waste salt by (i) the electrolytic recovery of REs from the salt using a liquid metal cathode; and subsequent (ii) oxidative extraction of REs from the liquid metal into durable wasteforms. Experimental demonstration using rare earth element samarium (Sm) and liquid metal (Bi) was successful. The Sm ions dissolved in the salt can be selectively recovered as dendrite Sm-Bi intermetallic products on the top of liquid bismuth cathode through electrolysis at appropriate applied currents or potentials. Then the Sm in the Sm-Bi alloy was efficiently extracted into both zinc-in-titania (ZIT) glass ceramic and lanthanide aluminoborosilicate (LABS) glass wasteforms through selective oxidation at Sm 2 O 3 loadings of 18–25 wt% and 50 wt%, respectively. Recycling of the electrorefining salt with minimum nuclear waste generation is foreseeable with the use of this simple, efficient, and highly compatible process.
研究领域:
核化学与技术
·
材料科学
✓ 已验证
产品关联
DOI
高影响因子
JOURNAL OF NUCLEAR MATERIALS
·
2024
·
IF 2.8
Xuejiao Li,Rongrong Cui,Yulong Song,Yu Gong
A comprehensive study by integrating first principles molecular dynamic (FPMD) simulations and differential scanning calorimetry (DSC) experiments on the physicochemical properties of multiple halophilic fission products ( M F n = RbF, SrF 2 , YF 3 , and ZrF 4 ) and molten NaF-BeF 2 (FNaBe) mixed salts is reported. The effect mechanism of product type on density and specific heat capacity has been discussed from the interionic distance and its stability, coordination number and its distribution, as well as the neighbor cluster structure and its orientation. It is concluded that volume increments of molten FNaBe+ M F n are primarily caused by Na-Na and M -F interactions, while the highest specific heat capacities ( c p ) of molten FNaBe+ZrF 4 is closely related to the stability of Zr-F bonds and its coordination structures. Furthermore, the DSC results assisted by quantitative analysis of elements indicate that the additive concentration of YF 3 is positively correlated with c p of molten FNaBe+YF 3 in a certain range.
研究领域:
材料科学
·
物理化学
✓ 已验证
产品关联
DOI
高影响因子
APPLIED SURFACE SCIENCE
·
2023
·
IF 6.7
Xudong Gong,Qiang Tian,Jiebiao Li,Ju Wang,Minhao Yan,Mark J. Henderson
Incorporation of fission products in spent nuclear fuel is of interest in the frame of safe storage of high-level nuclear waste in a deep geological repository. Uranyl oxyhydroxide films were grown on silicon wafer to represent the corrosion product formed on the surfaces of the fuel. Films were obtained by hydrolysis of uranyl ions (9 mM) in mild conditions (≤60 °C) using either Mg(OH) 2 , raw brucite, La(OH) 3 , or NH 4 OH to control the solution pH to ≈ 5.0. To identify the mineral phases we used the grazing-incidence X-ray technique. Atomic force microscopy data indicated that the surfaces of metaschoepite films prepared using either Mg(OH) 2 or brucite at 50 °C for 3 h were self-affine fractal ( D f = 2.5). The concentration (wt.%) of reactant ions detected in a metaschoepite film after reaction at 70 °C for three days as determined by X-ray fluorescence was: La (8.90) > Nd (5.6) > Ce (3.2) > Eu (2.5) > Pb (0.7) > Sr (0.14). Metaschoepite and sodium uranate hydrate were identified in a film formed in Cl-SO 4 -Na type groundwater from the Beishan area (China), the preselected area of a radioactive waste disposal repository.
研究领域:
核废料管理
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材料科学
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环境科学
✓ 已验证
产品关联
DOI
高影响因子
危险品化学品经营许可证（带存储）