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WO2016114685A1 - Procédé de retraitement de combustible nucléaire irradié - Google Patents

Procédé de retraitement de combustible nucléaire irradié Download PDF

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Publication number
WO2016114685A1
WO2016114685A1 PCT/RU2015/000623 RU2015000623W WO2016114685A1 WO 2016114685 A1 WO2016114685 A1 WO 2016114685A1 RU 2015000623 W RU2015000623 W RU 2015000623W WO 2016114685 A1 WO2016114685 A1 WO 2016114685A1
Authority
WO
WIPO (PCT)
Prior art keywords
stage
nuclear fuel
temperature
irradiated nuclear
gas
Prior art date
Legal status (The legal status 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 status listed.)
Ceased
Application number
PCT/RU2015/000623
Other languages
English (en)
Russian (ru)
Inventor
Петр Михайлович ГАВРИЛОВ
Игорь Александрович МЕРКУЛОВ
Владимир Викторович БОНДИН
Игорь Евгеньевич ПОЛЯКОВ
Павел Николаевич ДАРОВСКИХ
Владимир Иванович ВОЛК
Константин Николаевич ДВОЕГЛАЗОВ
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
FEDERAL STATE UNITARY ENTERPRISE "MINING AND CHEMICAL COMBINE" (FSUE "MCC")
Original Assignee
FEDERAL STATE UNITARY ENTERPRISE "MINING AND CHEMICAL COMBINE" (FSUE "MCC")
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 FEDERAL STATE UNITARY ENTERPRISE "MINING AND CHEMICAL COMBINE" (FSUE "MCC") filed Critical FEDERAL STATE UNITARY ENTERPRISE "MINING AND CHEMICAL COMBINE" (FSUE "MCC")
Publication of WO2016114685A1 publication Critical patent/WO2016114685A1/fr
Anticipated expiration legal-status Critical
Ceased legal-status Critical Current

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Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/28—Treating solids
    • G21F9/30—Processing

Definitions

  • the invention relates to methods for reprocessing irradiated nuclear fuel (SNF) from VVER-1000 reactors with the aim of localizing tritium, which is a beta-active emitting nuclide, on the head operations of SNF reprocessing and can be used in atomic energy during the reprocessing of nuclear reactors.
  • SNF irradiated nuclear fuel
  • the disadvantage of the prototype lies in the increased sublimation of cesium in the first stage of the process of high-temperature oxidative processing of SNF, significant entrainment of the highly active Cs 137 isotope from the fuel composition, which leads to additional problems of decontamination of the equipment used, as well as a significant load on the tritium-trapping system carried out with the gas using NaA zeolites followed by isolation of zeolites.
  • the technical result of the invention is to reduce the entrainment of cesium from the volatirising SNF, reducing the load on the system for capturing tritium from the gas stream and reducing the amount of tritium-containing solid radioactive waste (SRW).
  • the technical result is achieved by the proposed method, which includes two-stage SNF-oxidation with SNF treatment at the first stage with air, additionally containing carbon dioxide, for 60-K360 minutes at a temperature of 40 (H650 ° C and treatment at the second stage with air or an oxygen-air mixture containing additionally water vapor in an amount corresponding to the dew point at a temperature of 3 (H40 ° C, at a temperature of 35 (H450 ° C and constant mechanical activation of the reaction mass at each of the stages, while the carbon dioxide content in g the first stage gas mixture is 4- ⁇ % vol., the processing time for spent fuel in the second stage is 60 ⁇ -360 minutes, the gas stream removed from the reaction chamber is cooled, the condensate is separated and sent to produce a cement compound, and the non-condensable gas stream is sent to the gas cleaning system .
  • the gas stream is cooled to a temperature of 0 - 5 ° C.
  • only the second stage gas stream is directed to cooling and condensation.
  • T2O tritium water
  • NTO tritium water
  • TrO (TB) + H 2 O (g) H 2 O (TB) + T 2 O (g) .
  • Both stages are carried out at a constant mechanical activation of the reaction mass, providing improved access of the reagent gas to the fuel due to surface renewal.
  • the gas flow rate at each stage corresponds to 1 (R50 full exchanges of the volume of the reaction chamber per hour.
  • the gas flow before entering the reaction chamber is heated to the temperature of the internal chamber volume, ie up to 400 650 ° ⁇ - at the first stage, and up to 350 450 ° ⁇ - at the second stage, respectively.
  • the modes of irradiated fuel fibsidation were checked using fragments of fuel elements (fuel rods) 32 mm long with VFA-1000 SFA of the Balakovo NPP with a burnup of 51.89 GW per day / ton of uranium after 10 years of exposure.
  • the degree of fibration was determined by the gravimetric method, determining the mass of the destroyed fuel. Determination of tritium concentration was performed using the SKS-07P-B11 liquid scintillation complex. Cesium concentration was determined using the SCS-07P-G7 gamma-spectrometry complex.
  • the oxidation of fragments of fuel elements was carried out according to the following mode: the first stage was carried out at a temperature of 550 ⁇ 50 ° C in air, additionally containing carbon dioxide in an amount of 1 to 4% by volume. for 360 minutes, with preheating the mixture of air and carbon dioxide to 550 ⁇ 50 ° C, the second stage was carried out at a temperature of 350 + 450 ° C in air containing steam in an amount corresponding to the dew point of the vapor-gas mixture at a temperature of 3 (H40 ° C for 360 minutes, in the second stage, the gas-vapor mixture was heated before entering the reaction chamber up to 350- ⁇ 450 ° C. The gas flow rate at each stage was maintained for about 30 complete exchanges of the volume of the reaction chamber per hour.
  • the oxidation of fragments of fuel elements was carried out according to the following mode: the first stage was carried out at a temperature of 550 ⁇ 50 ° C in air, additionally containing carbon dioxide in an amount of 4-40% by volume. for 360 minutes, with preheating of the mixture of air and carbon dioxide to 550 ⁇ 50 ° C, the second stage was carried out under the same conditions as in the first experiment.
  • the gas flow rate at each stage was maintained at about 30 full volume exchanges of the reaction chamber per hour.
  • the volume of the reaction chamber in both the first and second experience was equal to
  • the hot gas stream, after fiberization, was passed through a metal cloth filter to clean the stream from aerosol entrainment.
  • experiment 1 the gas stream was passed through a cooled layer of zeolite NaA, and then sent to the gas cleaning system.
  • the gas stream was pre-cooled to 0 ° C, the condensate formed was separated, and the non-condensable gas stream was passed through a fresh layer of zeolite, and then sent to the gas cleaning system.
  • the degree of SNF fiberization in both experiments was at least 99%. Extraction of tritium in both experiments was not less than 99.97%.
  • the cesium ablation in experiment 1 was 0.1%, in experiment 2 - 0.01%.
  • the gain of the zeolite layer due to water absorption was 5.5 g in experiment 1 and 0.67 g in experiment 2. Accordingly, the volume of condensate in experiment 2 was ⁇ 6 ml.

Landscapes

  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Inorganic Compounds Of Heavy Metals (AREA)

Abstract

L'invention concerne des procédés de retraitement de combustible nucléaire irradié (CNI) des réacteurs VVER-1000 visant à localiser le tritium qui est un nucléide rayonnant béta-actif, lors des opérations principales de retraitement de CNI et peut être utilisée en énergie nucléaire pour le retraitement du CNI de réacteur nucléaire. Le procédé consiste en un retraitement en deux stades par oxydation du combustible nucléaire usé (voloxydation de CNI) en dioxyde d'uranium, qui comprend au premier stade un traitement thermique des fragments du CNI à une température de 400÷650°C dans un milieu d'air qui comprend en outre du gaz carbonique, pendant 60-360 min et à un deuxième stade un traitement thermique à une température de 350÷450°C dans un milieu d'air ou enrichi en oxygène, qui comprend en outre de la vapeur d'eau dans des quantités correspondant au point de rosée à 30-40°, les deux stades s'effectuant lors d'une activation mécanique constante ou périodique de la masse réactionnelle, qui se distingue en ce que la teneur en gaz carbonique du mélange gazeux au premier stade correspond à 4÷10 % en volume, la durée de traitement du CNI au deuxième stade est de 60-360 min, le flux gazeux évacué de la chambre de réaction est refroidi, le condensat est séparé et utilisé pour obtenir du compound de ciment, et le flux gazeux non condensé est dirigé dans le système de purification de gaz. Le résultat technique consiste en ce que lors du traitement par oxydation selon le procédé proposé, avec un degré de voloxydation identique, le degré d'entraînement de césium est divisé par 10, et la quantité de déchets radioactifs solides contenant du tritium est divisée par 8.
PCT/RU2015/000623 2015-01-12 2015-09-29 Procédé de retraitement de combustible nucléaire irradié Ceased WO2016114685A1 (fr)

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
RU2015100955/07A RU2579753C1 (ru) 2015-01-12 2015-01-12 Способ переработки облученного ядерного топлива
RU2015100955 2015-01-12

Publications (1)

Publication Number Publication Date
WO2016114685A1 true WO2016114685A1 (fr) 2016-07-21

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RU (1) RU2579753C1 (fr)
WO (1) WO2016114685A1 (fr)

Families Citing this family (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
RU2716137C1 (ru) * 2019-10-18 2020-03-06 Федеральное государственное унитарное предприятие "Горно-химический комбинат" (ФГУП "ГХК") Установка для волоксидации отработавшего ядерного топлива

Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
RU2268515C1 (ru) * 2004-05-11 2006-01-20 Открытое акционерное общество "Новосибирский завод химконцентратов" Способ переработки металлических отходов, содержащих радионуклиды
RU2459299C1 (ru) * 2011-04-20 2012-08-20 Федеральное государственное унитарное предприятие "Горно-химический комбинат" Способ переработки облученного ядерного топлива
US20130197293A1 (en) * 2012-02-01 2013-08-01 Dimitre S. Assenov Nano flex hlw/spent fuel rods recycling and permanent disposal

Family Cites Families (2)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US6635232B1 (en) * 1999-05-13 2003-10-21 Kabushiki Kaisha Toshiba Method of chemically decontaminating components of radioactive material handling facility and system for carrying out the same
JP2001318194A (ja) * 2000-05-09 2001-11-16 Mitsubishi Heavy Ind Ltd 放射性廃棄物溶融方法

Patent Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
RU2268515C1 (ru) * 2004-05-11 2006-01-20 Открытое акционерное общество "Новосибирский завод химконцентратов" Способ переработки металлических отходов, содержащих радионуклиды
RU2459299C1 (ru) * 2011-04-20 2012-08-20 Федеральное государственное унитарное предприятие "Горно-химический комбинат" Способ переработки облученного ядерного топлива
US20130197293A1 (en) * 2012-02-01 2013-08-01 Dimitre S. Assenov Nano flex hlw/spent fuel rods recycling and permanent disposal

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