Natural Clay Minerals as a Starting Material for Matrices for the Immobilization of Radioactive Waste from Pyrochemical Processing of SNF

Matveenko, A. V.; Varlakov, A. P.; Zherebtsov, A. A.; Germanov, A. V.; Mikheev, Ivan V.; Kalmykov, Stepan N.; Петров, В. Г.

doi:10.3390/su131910780

Cited by 3 publications

(5 citation statements)

References 13 publications

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“…Ceramics based on bentonite clay represent polyphase materials with the following main observed phases (Figure 1c): KAlSi 3 O 8 (Phase #7, orthoclase, main peaks at 3.31 and 3.77 Å), LiAlSi 3 O 8 (Phase #8, main peaks at 3.47 and 3.89 Å), SrAl 2 Si 2 O 8 (Phase #9, main peaks at 3.27 and 3.22 Å). Additionally, some minor phases ((Na,K)AlSi 3 O 8 , CsAlSiO 4 , CsAlSi 2 O 6 ) can present in the samples as it was shown in [17]. Thus, it can be suggested that fission products (primarily cesium and strontium) enter the crystal lattice of ceramics based on bentonite clay, that improves the immobilization (fixation) of RW in the samples.…”

Section: Characterization Of the Synthesised Compounds Containing Chl...mentioning

confidence: 73%

“…Imitators of the spent electrolyte formed as a result of pyrochemical processing of MNUP SNF with the composition Li 0.4 K 0. [17] were used to synthesize compounds based on MPP matrix, NAFP glass, and bentonite ceramics. The conditions for the synthesis of the studied compounds are presented below.…”

Section: Chemicals and Proceduresmentioning

confidence: 99%

“…To study the hydrolytic stability of NAFP glass, the samples were prepared containing radioactive tracers 137 [25]. The synthesis of the samples was performed according to [17]. The mixture of chlorides was ground, dehydrated in an oven at 250 • C for 2 h, and then stored in a desiccator over P 2 O 5 .…”

Section: Synthesis Of Nafp Glassmentioning

confidence: 99%

“…Sodium aluminophosphate (NAFP) glass is a promising matrix for the immobilization of HLW, since its synthesis temperature is lower than that of aluminophosphate glass, which is the only industrial matrix for HLW immobilization used in Russia [2]. Natural clay minerals (bentonite clay) can be used as a possible matrix material for HLW immobilization because the matrix material, in its chemical and phase composition, will be similar to the host rocks of the waste repository [17]. At present, a magnesium potassium phosphate (MPP) matrix, obtained at room temperature and being a synthetic analogue of the natural mineral K-struvite, is considered as an alternative to glass for HLW immobilization.…”

Section: Introductionmentioning

confidence: 99%

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Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

et al. 2021

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Immobilization of spent electrolyte–radioactive waste (RW) generated during the pyrochemical processing of mixed nitride uranium–plutonium spent nuclear fuel is an acute task for further development of the closed nuclear fuel cycle with fast neutron reactors. The electrolyte is a mixture of chloride salts that cannot be immobilized directly in conventional cement or glass matrix. In this work, a low-temperature magnesium potassium phosphate (MPP) matrix and two types of high-temperature matrices (sodium aluminoironphosphate (NAFP) glass and ceramics based on bentonite clay) were synthesized. Two systems (Li0.4K0.28La0.08Cs0.016Sr0.016Ba0.016Cl and Li0.56K0.40Cs0.02Sr0.02Cl) were used as spent electrolyte imitators. The phase composition and structure of obtained materials were studied by XRD and SEM-EDS methods. The differential leaching rate of Cs from MPP compound and ceramic based on bentonite clay was about 10−5 g/(cm2·day), and the rate of Na from NAFP glass was about 10−6 g/(cm2·day). The rate of 239Pu from MPP compound (leaching at 25 °C) and NAFP glass (leaching at 90 °C) was about 10−6 and 10−7 g/(cm2·day), respectively. All the synthesized materials demonstrated high hydrolytic, mechanical compression strength (40–50 MPa) even after thermal (up to 450 °C) and irradiation (up to 109 Gy) tests. The characteristics of the studied matrices correspond to the current requirements to immobilized high-level RW, that allow us to suggest these materials for industrial processing of the spent electrolyte.

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Section: Characterization Of the Synthesised Compounds Containing Chl...mentioning

confidence: 73%

Section: Chemicals and Proceduresmentioning

confidence: 99%

Section: Synthesis Of Nafp Glassmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

et al. 2021

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“…There are several studies on the handling and storage of spent fuel [1][2][3][4][5][6][7][8][9][10]. After remaining in interim wet storage for more than 4 years, spent fuel can be moved to dry cask storage.…”

Section: Introductionmentioning

confidence: 99%

Theoretical and Experimental Analysis on Influence of Natural Airflow on Spent Fuel Heat Removal in Dry Cask Storage

et al. 2022

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A key issue contributing to the success of NPP technology is the safe handling of radioactive waste, particularly spent nuclear fuel. According to the IAEA safety standard, the spent fuel must be stored in interim wet storage for several years so the radiation and the decay heat of the spent fuel will decrease to the safe limit values, after which the spent fuel can be moved to dry storage. In this study, we performed a theoretical analysis of heat removal by natural convection airflow in spent nuclear fuel dry storage. The temperature difference between the air inside and outside dry storage produces an air density difference. The air density difference causes a pressure difference, which then generates natural airflow. The result of the theoretical analysis was validated with simulation software and experimental investigation using a reduced-scale dry storage prototype. The dry storage prototype consisted of a dry cask body and two canisters stacked to store materials testing reactor (MTR) spent fuel, which generates decay heat. The cask body had four air inlet vents on the bottom and four air outlet vents at the top. To simulate the decay heat from the spent fuel in the two canisters, the canisters were wrapped with an electric wire heater that was connected to a voltage regulator to adjust the heat power. The theoretical analysis results of this study are relatively consistent with the experimental results, with the mean relative deviation (MRD) values for the prediction of air velocity, the heat rate using natural airflow, and the heat rate using the thermal resistance network equation are +0.76, −23.69, and −29.54%, respectively.

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Application of Clay Materials for Sorption of Radionuclides from Waste Solutions

Singh

2023

Minerals

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The wide application of nuclear resources in various fields has resulted in the production of radioactive waste, which poses a serious threat to lives and the environment. Nuclear waste contains long-lived radionuclides and, due to its mobility in environments, the proper management of generated waste is necessary. To impede the mobility of radionuclides in environments, various materials have been tested as suitable sorbents under different experimental conditions. In this review, we thoroughly discuss some key and recent contributions to the application of natural clays (NCs) and modified/functionalized clays (MCs) for the sorption of various radionuclides in their cationic and anion forms from (simulated) waste solutions under different experimental conditions. More specifically, we discuss the key developments toward the use of natural clays for the efficient sorption of various radioactive contaminates. Later, this review targets the modification/functionalization of natural clays using various organic moieties to improve their removal capacities for various radionuclides/hazardous ions present in waste solutions. Finally, we summarize the major aspects and highlight the key challenges to be addressed in future studies to further enhance the application of clays and clay-based materials for selective and effective removal of various radionuclides from waste solutions.

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Natural Clay Minerals as a Starting Material for Matrices for the Immobilization of Radioactive Waste from Pyrochemical Processing of SNF

Cited by 3 publications

References 13 publications

Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

Theoretical and Experimental Analysis on Influence of Natural Airflow on Spent Fuel Heat Removal in Dry Cask Storage

Application of Clay Materials for Sorption of Radionuclides from Waste Solutions

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