РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor

Shadrin, A. Yu.; Dvoeglazov, K. N.; Maslennikov, A. G.; Kashcheev, V. A.; Tret’yakova, S. G.; Shmidt, O. V.; Vidanov, V. L.; Ustinov, O. A.; Volk, V. I.; Veselov, S. N.; Ishunin, V. S.

doi:10.1134/s1066362216030085

Cited by 35 publications

(5 citation statements)

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“…This technology allows recycling of MNUP SNF with a short exposure time and provides a high degree of purification of fissile materials. The anodic dissolution of (U,Pu)N in the melt of alkali metal chlorides and the cathodic deposition of fissile materials using a liquid cadmium cathode are the main processes in the pyrochemical recycling of MNUP SNF [1]. Various systems are used as electrolytes, including NaCl-2CsCl, NaCl-KCl, LiCl-4.53NaCl-4.88KCl-0.66CsCl, 3LiCl-2KCl and…”

Section: Introductionmentioning

confidence: 99%

Conditioning of Spent Electrolyte Surrogate LiCl-KCl-CsCl Using Magnesium Potassium Phosphate Compound

et al. 2020

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The current work was aimed at developing a new conditioning method of spent electrolyte-radioactive waste (RW) generated during the pyrochemical reprocessing of mixed nitride uranium-plutonium spent nuclear fuel. Magnesium potassium phosphate (MPP) compound samples were synthesized under solidification of the electrolyte surrogate solution in a LiCl-KCl-CsCl system. The phase composition and structure of obtained compounds were studied by XRD and SEM-EDS methods. It was found that the compounds possessed a high compressive strength of 17–26 MPa. Hydrolytic stability of the compounds was evaluated in accordance with the long semi-dynamic test GOST R 52126-2003 and with the static PCT test. The 137Cs content in the leached solutions was determined by gamma-ray spectrometry, and other compound components were determined by ICP–AES and ICP–MS methods. The differential leaching rate of Cs at 25 °C from monolithic samples on the 91st day of samples contact with water was 5–11 × 10−5 g/(cm2·day) (GOST R 52126-2003), and was 4–29 × 10−7 g/(cm2∙day) on the 7th day at 90 °C from crushed samples (PCT). The thermal stability of the compound at 180 °C and 450 °C was shown. The characteristics of the obtained MPP compound correspond to the current regulatory requirements for materials for RW conditioning.

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Section: Introductionmentioning

confidence: 99%

Conditioning of Spent Electrolyte Surrogate LiCl-KCl-CsCl Using Magnesium Potassium Phosphate Compound

et al. 2020

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“…Reprocessing with using different solvents. The chosen spent fuel reprocessing technologies: Molten chloride salts electrolysis, including actinides regeneration to metal on a solid-state cathode together with the water technology with organic extraction agents [10], [11]. The advantages of decentralized spent fuel reprocessing (within the SNFC) are:…”

Section: Fuel Reprocessing For "Brest" Reactorsmentioning

confidence: 99%

On-Site Nuclear Fuel Cycle of “BREST” Reactors

Glazov¹

2018

IJEPE

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Dynamic developing of modern nuclear industry demands meeting the following requirements: improved safety, reduced capital costs, radioactive waste (RW) management issues, independence of limited resources. Efficiency of uranium resources used in "BREST" reactors based on a closed fuel cycle is about 160 times higher than for VVER, RBMK reactors [1], which makes it possible to stop searching for new deposits and uranium mining. The need for periodical fuel regeneration and fabrication in a closed cycle includes reproduction of plutonium in the core without the uranium containing screens (breeding ratio is approximately 1,05, ensuring a high level of safety and support of the non-proliferation regime) and transmutation of the most dangerous long-lived actinides and high refining of RW, achieving the radiation balance of buried RW and extracted uranium ore. The manufacturing is located directly at the NPP to avoid transportation of fissile materials. This approach provides economic efficiency of the entire complex.

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“…Such wastes are generated in the USA after the pyrochemical processing of SNF from the research reactor EBR-II [20]. In Russia, chloride waste is generated during the processing of mixed uranium-plutonium SNF (SNUP SNF) in the BREST-OD-300 reactor with lead coolant [21].…”

Section: Introductionmentioning

confidence: 99%

“…In some works [21][22][23], it has been shown that matrices of the "zeolite/sodalite in glass" type can be used for this type of waste. The sintering of chlorapatite and borosilicate glass [24] and the sintering of glass powders to obtain glass-composite materials [24][25][26] were also proposed, since it is known that glass composite materials have lower synthesis temperatures as well as the possibility of forming target stable phases.…”

Section: Introductionmentioning

confidence: 99%

Medium-Temperature Glass-Composite Phosphate Materials for the Immobilization of Chloride Radioactive Waste

Frolova,

Belova,

Vinokurov

2023

J. Compos. Sci.

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Among the many radiochemical problems, the search for new materials and technologies for the immobilization of radioactive waste remains relevant, and the range continues to change and expand. The possibility of immobilizing the spent chloride electrolyte after the pyrochemical processing of the mixed uranium-plutonium spent nuclear fuel of the new fast reactor BREST-OD-300 on lead coolant into glass-composite phosphate materials synthesized at temperatures of 650–750 °C was studied. The structure of the obtained samples was studied using XRD and SEM/EDS methods. It has been shown that the spent electrolyte simulator components create stable mixed pyrophosphate phases in the glass composite structure. The materials were found to have high hydrolytic stability. This indicates the promise of using phosphate glass composites as materials for the reliable immobilization of the spent electrolyte.

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РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor

Cited by 35 publications

References 10 publications

Conditioning of Spent Electrolyte Surrogate LiCl-KCl-CsCl Using Magnesium Potassium Phosphate Compound

Conditioning of Spent Electrolyte Surrogate LiCl-KCl-CsCl Using Magnesium Potassium Phosphate Compound

On-Site Nuclear Fuel Cycle of “BREST” Reactors

Medium-Temperature Glass-Composite Phosphate Materials for the Immobilization of Chloride Radioactive Waste

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