WO2016114685A1 - Method for reprocessing irradiated nuclear fuel - Google Patents
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- 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
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- 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
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- 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.
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Abstract
Description
СПОСОБ ПЕРЕРАБОТКИ ОБЛУЧЕННОГО ЯДЕРНОГО ТОПЛИВА METHOD OF PROCESSING OF IRRADIATED NUCLEAR FUEL
Изобретение относится к способам переработки облученного ядерного топлива (ОЯТ) реакторов ВВЭР-1000 с целью локализации трития, являющегося бета-активным излучающим нуклидом, на головных операциях переработки ОЯТ и может быть использовано в атомной энергетике при переработке ОЯТ ядерных реакторов. 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.
В настоящее время для переработки ОЯТ используют водно-экстракционные технологии, наиболее острой проблемой которых является наличие больших объемов тритийсодержащих растворов на всех этапах переработки, что существенно усложняет и удорожает переработку жидких радиоактивных отходов (ЖРО). Поэтому до проведения основной водно-экстракционной переработки целесообразно проводить предварительную обработку ОЯТ для локализации трития и других летучих продуктов деления. Предварительная обработка ОЯТ проводится различными окислительными высокотемпературными способами. At present, water-extraction technologies are used for the reprocessing of SNF, the most acute problem of which is the presence of large volumes of tritium-containing solutions at all stages of reprocessing, which significantly complicates and increases the cost of processing liquid radioactive waste (LRW). Therefore, prior to the main water-extraction processing, it is advisable to pre-process SNF to localize tritium and other volatile fission products. Pre-treatment of spent nuclear fuel is carried out by various oxidizing high-temperature methods.
Известны способы высокотемпературной окислительной обработки фрагментов с ОЯТ при температуре от 480 до 600°С в присутствии воздуха или кислорода. При этом степень удаления трития из ОЯТ составляет 99 %.(G.D.DelCui, R.D.Hunt, J.A.Jonsonandother.Advanced head end for the treatment of LWR fuel. OECD Nuclear Energy Agency. 11- th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation Hyatt at Fisherman's Wharf, San Francisco, California, 1-5 November 2010). Known methods of high-temperature oxidative treatment of fragments with spent fuel at a temperature of from 480 to 600 ° C in the presence of air or oxygen. At the same time, the degree of removal of tritium from SNF is 99%. (GDDelCui, RDHunt, JAJonsonandother.Advanced heading for LWR fuel. OECD Nuclear Energy Agency. 11th Information Exchange Meeting and Transmutation Hyatt at Fisherman's Wharf, San Francisco, California, 1-5 November 2010).
Известен способ двухстадийной высокотемпературной окислительной обработки фрагментов ОЯТ, по которому первую стадию проводят при температуре 400 650 °С в воздушной среде, дополнительно содержащей углекислый газ в количестве 1+4 % об. в течение 60+360 минут, вторую стадию проводят при температуре 350+450 °С в воздушной или обогащенной по кислороду среде, содержащей водяной пар в количестве, соответствующему точке росы парогазовой смеси при температуре 30+40 °С в течение 30+120 минут (Патент RU 2459299, МПК G21F9/30, 2006.01, «СПОСОБ ПЕРЕРАБОТКИ ОБЛУЧЕННОГО ЯДЕРНОГО ТОПЛИВА»). По технической сущности и достигаемому положительному эффекту этот способ является наиболее близким к заявляемому способу и выбран в качестве прототипа. The known method of two-stage high-temperature oxidative processing of SNF fragments, in which the first stage is carried out at a temperature of 400 650 ° C in air, additionally containing carbon dioxide in an amount of 1 + 4% by volume. for 60 + 360 minutes, the second stage is carried out at a temperature of 350 + 450 ° C in air or oxygen-enriched medium containing water vapor in an amount corresponding to the dew point of the gas-vapor mixture at a temperature of 30 + 40 ° C for 30 + 120 minutes ( Patent RU 2459299, IPC G21F9 / 30, 2006.01, “METHOD OF PROCESSING OF IRRADIATED NUCLEAR FUEL”). The technical essence and the achieved positive effect, this method is closest to the claimed method and is selected as a prototype.
Недостаток прототипа заключается в повышенной возгонке цезия на первой стадии процесса высокотемпературной окислительной обработки ОЯТ, значительным уносом из топливной композиции высокоактивного изотопа Cs137, что приводит к дополнительным проблемам дезактивации используемого оборудования, а также в значительной нагрузке на систему улавливания трития из газовой фазы, осуществляемую с помощью цеолитов NaA с последующей изоляцией цеолитов. 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.
Техническим результатом предлагаемого изобретения является снижении уноса цезия из волоксидируемого ОЯТ, снижение нагрузки на систему улавливания трития из газового потока и снижение количества тритийсодержащих твердых радиоактивных отходов (ТРО). 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).
Технический результат достигается предложенным способом, который включает двустадийную волоксидацию ОЯТ с обработкой ОЯТ на первой стадии воздухом, дополнительно содержащим углекислый газ, в течение 60-К360 минут при температуре 40(Н650 °С и обработкой на второй стадии воздухом или кислородно-воздушной смесью, содержащей дополнительно пары воды в количестве, соответствующему точке росы при температуре 3(Н40 °С, при температуре 35(Н450 °С и постоянной механоактивации реакционной массы на каждой из стадий, при этом содержание углекислого газа в газовой смеси первой стадии составляет 4-НО % об., время обработки ОЯТ на второй стадии составляет 60^-360 минут, выводимый из реакционной камеры газовый поток охлаждают, конденсат отделяют и направляют на получение цементного компаунда, а неконденсируемый газовый поток направляют в систему газоочистки. 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-НО% 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 .
В частном варианте газовый поток охлаждают до температуры 0 - 5 °С. In the private version, the gas stream is cooled to a temperature of 0 - 5 ° C.
В другом частном варианте на охлаждение и конденсацию направляют только газовый поток второй стадии. In another private variant, only the second stage gas stream is directed to cooling and condensation.
На первой стадии происходит разрушение структуры диоксида урана, окисление трития до тритиевой воды и удаление основной массы трития, включая удаление из третированного гидроксида цезия по реакции: 2CsOT + C02 = Cs2C03 + T20 At the first stage, the structure of uranium dioxide is destroyed, the oxidation of tritium to tritium water and the removal of the main mass of tritium, including the removal of tertiary cesium hydroxide by the reaction: 2CsOT + C0 2 = Cs 2 C0 3 + T 2 0
На второй стадии удаляется особопрочно адсорбированная тритиевая вода (Т2О; НТО), удерживаемая на дефектах и дислокациях решетки окисленных форм по реакциям изотопного обмена типа: At the second stage, the highly adsorbed tritium water (T2O; NTO), which is retained on defects and lattice dislocations of the oxidized forms, is removed according to isotope exchange type:
НТО(та) + Н2О(Г) = Н2О(ТВ) + НТО(Г) NTR (ta) + H 2 O (G) = H 2 O (TV) + NTR (G)
ТгО(ТВ) + Н2О(г) = Н2О(ТВ) + Т2О(г). TrO (TB) + H 2 O (g) = H 2 O (TB) + T 2 O (g) .
Обе стадии проводят при постоянной механоактивации реакционной массы, обеспечивающей улучшенный доступ газа-реагента к топливу за счет обновления поверхности. Расход газового потока на каждой стадии соответствует 1(R50 полным обменам объема реакционной камеры в час. Для уменьшения общей продолжительности обработки и достижения требуемой степени волоксидации ОЯТ газовый поток перед входом в реакционную камеру подогревается до температуры внутреннего объема камеры, т.е. до 400 650 °С - на первой стадии, и до 350 450 °С - на второй стадии соответственно. 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. To reduce the total processing time and to achieve the required degree of SNF fiberization, the gas flow before entering the reaction chamber is heated to the temperature of the internal chamber volume, ie up to 400 650 ° С - at the first stage, and up to 350 450 ° С - at the second stage, respectively.
Пример осуществления способа. An example of the method.
Проверку режимов волоксидации облученного топлива проводили с использованием фрагментов тепловыделяющих элементов (твэлов) длиной 32 мм ОТВС ВВЭР-1000 Балаковской АЭС с выгоранием 51,89 ГВт сут/т урана после 10 - летней выдержки. Степень волоксидации определяли весовым методом, определяя массу разрушенного топлива. Определение концентрации трития выполняли с использованием жидко-сцинтилляционного комплекса СКС-07П-Б11. Определение концентрации цезия выполняли с использованием гама- спектрометрического комплекса СКС-07П-Г7. 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.
Для сравнения прототипа и заявленного способа проведено два опыта при одинаковой продолжительности стадий обработки в течение 360 мин. For comparison of the prototype and the claimed method, two experiments were conducted with the same duration of processing stages for 360 minutes.
В первом опыте волоксидацию фрагментов твэлов проводили по следующему режиму: первая стадия проводилась при температуре 550 ± 50 °С в воздушной среде, дополнительно содержащей углекислый газ в количестве 1 4 % об. в течение 360 минут, при предварительном подогреве смеси воздуха и углекислого газа до 550 ± 50 °С, вторая стадия проводилась при температуре 350+450 °С в воздушной среде содержащей водяной пар в количестве, соответствующем точке росы парогазовой смеси при температуре з 3(H40 °C в течение 360 минут, на второй стадии парогазовую смесь подогревали перед вводом в реакционную камеру до 350-^450 °С. Расход газового потока на каждой стадии поддерживали около 30 полных обменов объема реакционной камеры в час. In the first experiment, 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.
Во втором опыте волоксидацию фрагментов твэлов проводили по следующему режиму: первая стадия проводилась при температуре 550 ± 50 °С в воздушной среде, дополнительно содержащей углекислый газ в количестве 4-40 % об. в течение 360 минут, при предварительном подогреве смеси воздуха и углекислого газа до 550 ± 50 °С, вторая стадия проводилась в тех же условиях, что и в первом опыте. Расход газового потока на каждой стадии поддерживали около 30 полных обменов объема реакционной камеры в час. In the second experiment, 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.
Масса ОЯТ как в первом, так и во втором опыте составляла 250 г. The mass of SNF, both in the first and in the second experiment, was 250 g.
Объем реакционной камеры как в первом, так и во втором опыте был равен The volume of the reaction chamber in both the first and second experience was equal to
0,75 л. 0.75 l.
Как в первом, так и во втором опыте горячий газовый поток после волоксидации пропускали через металлотканевый фильтр для очистки потока от аэрозольного уноса. Both in the first and in the second experiment, the hot gas stream, after fiberization, was passed through a metal cloth filter to clean the stream from aerosol entrainment.
Далее, в опыте 1 газовый поток пропускали через охлаждаемый слой цеолита NaA, после чего направляли в систему газоочистки. Further, in experiment 1, the gas stream was passed through a cooled layer of zeolite NaA, and then sent to the gas cleaning system.
В опыте 2 газовый поток предварительно охлаждали до 0 °С, отделяли образующийся конденсат, а неконденсируемый газовый поток пропускали через свежий слой цеолита, после чего направляли в систему газоочистки. In experiment 2, 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.
Результаты опытов были следующими. The results of the experiments were as follows.
Степень волоксидации ОЯТ в обоих опытах составила не менее 99%. Извлечение трития в обоих опытах составило не менее 99,97 %. Унос цезия в опыте 1 составил 0,1%, в опыте 2 - 0,01 %. Привес слоя цеолита, обусловленный поглощением воды, составил 5,5 г в опыте 1 и 0,67 г в опыте 2. Соответственно, объем конденсата в опыте 2 составил ~ 6 мл. 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.
Таким образом, проведение окислительной обработки по предполагаемому способу позволяет понизить унос цезия в 10 раз и снизить нагрузку на систему улавливания трития и, соответственно, снизить количество тритийсодержащих ТРО более чем в 8 раз. Thus, carrying out the oxidation treatment according to the proposed method makes it possible to reduce the cesium ablation by 10 times and reduce the load on the tritium trapping system and, accordingly, reduce the amount of tritium-containing SRW by more than 8 times.
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| RU2268515C1 (en) * | 2004-05-11 | 2006-01-20 | Открытое акционерное общество "Новосибирский завод химконцентратов" | Method of processing of the metallic waste containing radionuclides |
| RU2459299C1 (en) * | 2011-04-20 | 2012-08-20 | Федеральное государственное унитарное предприятие "Горно-химический комбинат" | Method of processing irradiated fuel |
| 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)
| Publication number | Priority date | Publication date | Assignee | Title |
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| 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 (en) * | 2000-05-09 | 2001-11-16 | Mitsubishi Heavy Ind Ltd | Melting method for radioactive waste |
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2015
- 2015-01-12 RU RU2015100955/07A patent/RU2579753C1/en not_active IP Right Cessation
- 2015-09-29 WO PCT/RU2015/000623 patent/WO2016114685A1/en not_active Ceased
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| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| RU2268515C1 (en) * | 2004-05-11 | 2006-01-20 | Открытое акционерное общество "Новосибирский завод химконцентратов" | Method of processing of the metallic waste containing radionuclides |
| RU2459299C1 (en) * | 2011-04-20 | 2012-08-20 | Федеральное государственное унитарное предприятие "Горно-химический комбинат" | Method of processing irradiated fuel |
| 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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| RU2579753C1 (en) | 2016-04-10 |
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