RU94037117A - Process of pyrochemical recovery of nuclear fuel - Google Patents

Process of pyrochemical recovery of nuclear fuel

Info

Publication number
RU94037117A
RU94037117A RU94037117/25A RU94037117A RU94037117A RU 94037117 A RU94037117 A RU 94037117A RU 94037117/25 A RU94037117/25 A RU 94037117/25A RU 94037117 A RU94037117 A RU 94037117A RU 94037117 A RU94037117 A RU 94037117A
Authority
RU
Russia
Prior art keywords
uranium
fuel
plutonium
container
electrolyte
Prior art date
Application number
RU94037117/25A
Other languages
Russian (ru)
Other versions
RU2079909C1 (en
Inventor
О.Н. Дубровин
В.В. Орлов
Б.Д. Рогозкин
А.Г. Сила-Новицкий
В.В. Шантяков
А.И. Филин
Original Assignee
Научно-исследовательский и конструкторский институт энерготехники
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Научно-исследовательский и конструкторский институт энерготехники filed Critical Научно-исследовательский и конструкторский институт энерготехники
Priority to RU94037117A priority Critical patent/RU2079909C1/en
Application granted granted Critical
Publication of RU2079909C1 publication Critical patent/RU2079909C1/en
Publication of RU94037117A publication Critical patent/RU94037117A/en

Links

Classifications

    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
    • Y02W30/00—Technologies for solid waste management
    • Y02W30/50—Reuse, recycling or recovery technologies

Landscapes

  • Electrolytic Production Of Metals (AREA)

Abstract

FIELD: processing of irradiated and rejected nuclear fuel. SUBSTANCE: invention refers to processing of mononitride uranium-plutonium fuel. Nitride is loaded into container and electrolyte carrying not less than 15.0% of uranium thrichloride by mass is used. Electrolyte is heated to temperature not below 600 C and current having density not more than 0.3 A/sq.cm is created on container and not more than 0.4 A/sq.cm - on electrode. Current with density not more than 0.1 A/sq.cm is formed across fusible electrode. Melt is heated to temperature not higher than 700 C. Fusible metal or alloy which is more electrically positive than uranium nitride is also placed into container with nitride fuel. This metal alloy is solved in container and uranium and plutonium are precipitated on fusible electrode - cathode. EFFECT: decomposition of mononitride fuel and transition of uranium, plutonium and actinides in the form of chlorides into electrolyte, reduction and extraction of metallic uranium and plutonium, recovery of fission products from nuclear fuel.

Claims (1)

Изобретение относится к области переработки облученного и бракованного ядерного топлива, в частности мононитридного уранплутониевого топлива. Задача, на решение которой направлен предлагаемый способ, заключается в получении металлических урана и плутония из облученного и/или бракованного ядерного топлива, в частности мононитридного уранплутониевого топлива, для производства из них ядерного топлива, годного для нужд атомной промышленности. В результате решения этой задачи обеспечивается разложение мононитридного топлива и переход урана, плутония и актинидов в виде хлоридов в электролит, восстановление и выделение металлического урана и плутония, удаление продуктов деления из ядерного топлива. Согласно изобретению в контейнер загружают нитридное топливо и используют электролит, в состав которого входит не менее 15 мас. % трихлорида урана, нагревают электролит до температуры не менее 600oС и создают на контейнере ток плотностью не более 0,3 А/см2, на электроде - не более 0,4 А/см2 кроме того, на легкоплавком электроде создают ток плотностью не более 0,1 А/см2 и расплав нагревают до температуры не более 700oC, а также в контейнер с нитридным топливом помещают легкоплавкий металл или сплав, более электроположительный, чем нитрид урана, растворяют в контейнере этот металл или сплав и осаждают уран и плутоний на легкоплавком электродекатоде.The invention relates to the field of processing of irradiated and defective nuclear fuel, in particular mononitride uranium-plutonium fuel. The problem to which the proposed method is aimed is to obtain metallic uranium and plutonium from irradiated and / or defective nuclear fuel, in particular mononitride uranium-plutonium fuel, for the production of nuclear fuel suitable for the needs of the nuclear industry. As a result of solving this problem, the decomposition of mononitride fuel and the transition of uranium, plutonium and actinides in the form of chlorides into an electrolyte, the reduction and release of metallic uranium and plutonium, and the removal of fission products from nuclear fuel are ensured. According to the invention, nitride fuel is loaded into a container and an electrolyte is used, which contains at least 15 wt. % uranium trichloride, heat the electrolyte to a temperature of at least 600 o C and create a current on the container with a density of not more than 0.3 A / cm 2 , on the electrode - not more than 0.4 A / cm 2 in addition, create a current with a density on the low-melting electrode not more than 0.1 A / cm 2 and the melt is heated to a temperature of not more than 700 o C, as well as a fusible metal or alloy more electropositive than uranium nitride is placed in a container with nitride fuel, this metal or alloy is dissolved in the container and uranium is deposited and plutonium on a low-melting electrode cathode.
RU94037117A 1994-09-27 1994-09-27 Method of nuclear fuel pyrochemical regeneration RU2079909C1 (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
RU94037117A RU2079909C1 (en) 1994-09-27 1994-09-27 Method of nuclear fuel pyrochemical regeneration

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
RU94037117A RU2079909C1 (en) 1994-09-27 1994-09-27 Method of nuclear fuel pyrochemical regeneration

Publications (2)

Publication Number Publication Date
RU2079909C1 RU2079909C1 (en) 1997-05-20
RU94037117A true RU94037117A (en) 1997-05-27

Family

ID=20161226

Family Applications (1)

Application Number Title Priority Date Filing Date
RU94037117A RU2079909C1 (en) 1994-09-27 1994-09-27 Method of nuclear fuel pyrochemical regeneration

Country Status (1)

Country Link
RU (1) RU2079909C1 (en)

Families Citing this family (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
RU2562809C1 (en) * 2014-05-30 2015-09-10 Открытое акционерное общество "Новосибирский завод химконцентратов" (ОАО "НЗХК") Method of regeneration of nuclear fuel powders from fuel elements and dispersion compositions based on aluminium and aluminium alloys
RU2603844C1 (en) * 2015-10-01 2016-12-10 Федеральное государственное бюджетное учреждение науки Институт высокотемпературной электрохимии Уральского отделения Российской Академии наук Method of nitride spent nuclear fuel recycling in salt melts
EP3734615A4 (en) 2017-12-29 2021-08-11 State Atomic Energy Corporation "Rosatom" on Behalf of The Russian Federation Method for reprocessing nitride spent nuclear fuel in molten salts
RU2758450C1 (en) * 2020-08-16 2021-10-28 Акционерное общество «Прорыв» Method for processing nitride snf in salt melts with removal of the residual amount of the chlorinating agent
RU2766563C2 (en) * 2020-08-16 2022-03-15 Акционерное общество «Прорыв» Method of processing nitride snf in molten salt with extraction of the target component using a precipitator

Also Published As

Publication number Publication date
RU2079909C1 (en) 1997-05-20

Similar Documents

Publication Publication Date Title
US4880506A (en) Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels
CA1337740C (en) Process to separate transuranic elements from nuclear waste
US2951793A (en) Electrolysis of thorium and uranium
RU94037117A (en) Process of pyrochemical recovery of nuclear fuel
FR2761804A1 (en) PROCESS FOR RECOVERY OF URANIUM AND TRANSURANIAN ELEMENTS FROM WASTE NUCLEAR FUEL
US3669832A (en) Coated-particle fuel and fertile element for nuclear reactors and method of operating same
JPH03123896A (en) Recovery of actinides
EP1240647B1 (en) Actinide production
AU628468B2 (en) Method of treatment of high-level radioactive waste
CA1172438A (en) Treatment of fuel pellets
EP1393324B1 (en) Actinide production
JPH08122487A (en) Reprocessing method of spent fuel and annihilationmethod of element having long half-life
DE2820060C2 (en)
JPS57191940A (en) Negative hydrogen or heavy hydrogen ion source using semiconductor
US2903402A (en) Recovery of valuable material from graphite bodies
US2930738A (en) Regeneration of reactor fuel elements
GB1176024A (en) Method of Re-Processing Nuclear Fuel or Fertile Material or Nuclear Fuel or Fertile Elements Coated With or Containing Graphite.
Böck Investigations on irradiated fission chamber electrodes
JP2809733B2 (en) Molten salt electrorefining equipment
US3446713A (en) Separation of the core of a uranium fuel element from its envelope
JPH0373896A (en) Fused salt electrolysis refining device
US3567648A (en) Dissolution of stainless steel clad nuclear fuel elements
Williamson et al. Improved Apparatus for Electrolytic Reaction
Rodrigues Dissolution and clarification of RERTR silicide fuels
GB1194624A (en) Oven for Processing Irradiated Nuclear Fuels

Legal Events

Date Code Title Description
HK4A Changes in a published invention