EP0180308A1 - Zéolite an borosilicate pour l'élimination des déchets nucléaires - Google Patents

Zéolite an borosilicate pour l'élimination des déchets nucléaires Download PDF

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Publication number
EP0180308A1
EP0180308A1 EP85306341A EP85306341A EP0180308A1 EP 0180308 A1 EP0180308 A1 EP 0180308A1 EP 85306341 A EP85306341 A EP 85306341A EP 85306341 A EP85306341 A EP 85306341A EP 0180308 A1 EP0180308 A1 EP 0180308A1
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Prior art keywords
zeolite
silica
radioactive
boron
waste
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EP85306341A
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German (de)
English (en)
Inventor
Pochen Chu
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ExxonMobil Oil Corp
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Mobil Oil Corp
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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
    • G21F9/04—Treating liquids
    • G21F9/06—Processing
    • G21F9/12—Processing by absorption; by adsorption; by ion-exchange
    • 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
    • G21F9/301—Processing by fixation in stable solid media
    • G21F9/302—Processing by fixation in stable solid media in an inorganic matrix
    • G21F9/305—Glass or glass like matrix

Definitions

  • This invention relates to a process for removing ions from highly alkaline waste liquors. More particularly, this invention relates to a process for removing radioactive cesium and strontium from highly alkaline waste liquors.
  • Radioactive waste material is in liquid form and is obtained, for example, by dissolving spent fuel in nitric acid.
  • the fuel element is dissolved in nitric acid for subsequent separation of uranium and plutonium from the fission products, whereby the fission products are retained in nitric acid solution which is a waste liquor.
  • This waste liquor is then neutralized by the addition of excess sodium hydroxide, sufficient excess of sodium hydroxide being added to render the waste liquor highly alkaline.
  • Radioisotopic elements in liquid waste are ultimately safely disposed of by burial in the ground, in the ocean, or in geological formations of suitable composition.
  • concentration of radioisotopic elements in liquid waste, especially low-level waste is too low to permit the economic disposal of the entire liquid by direct burial. Therefore, the liquid waste must undergo preliminary treatments so as to concentrate the radioactive waste's values in the liquid.
  • Various methods have been suggested to effect this result, examples being: evaporation of the liquids; fixation of radioisotopic elements by solids; precipitation of radioisotopic elements from the waste liquids; and calcination of the waste liquids.
  • waste liquids especially waste liquids containing a low level of radioisotopic elements
  • concentration of waste by evaporation is not practical for handling large volumes of low-level radioactive waste.
  • Commonly used ion exchangers include organic resins such as styrene-divinyl benzene polmers containing sulfonic, phosphonic, carboxylic, phenolic, thiol, amino, or chelating groups for ion exchange.
  • Commercially available resinous ion exchangers include lonac C-240 and A-540 (ionac Chemical Corp.); Amberlite IRC-120, IRC-938, and Amberlyst A-21 (Rohm & Haas); Dowex 50, Retardion (Dow Chemical); and Ductile C-10, A-101 (Diamond Shamrock Co.).
  • Inorganic ion exchangers are also known and include sulfonated coal, mineral loughlimite, talc, kaolin, bentonite, vermiculite, amorphous silica-alumina, and crystalline aluminosilicates including such zeolites as NaA, chabasite, NaX, NaY, etc.
  • the inorganic ion exchangers have greater stability under the conditions of heat and radioactivity encountered during waste treatment as compared to the resinous ion exchange materials.
  • the zeolitic ion exchangers also have improved selectivity for cesium and strontium radioactive isotopes, thus, enabling a more efficient removal of these radioactive ions as is achieved with amorphous clay ion exchangers.
  • the most acceptable and widely practiced way of disposal at the present time comprises removing radioactive nucleides from liquid or gaseous wastes by adsorption or ion exchange with a naturally-occurring or synthetic zeolite, mixing the radwaste-loaded zeolite with glass particles, and firing the mixture at 1050°C to form leach-resistant and corrosion-resistant glass which can be stored in geologically stable sites.
  • U.S. Patent 3,017, 242 describes dinoptifolite, a sodium-aluminium silicate type of zeolite, which has preferential affinity for radioactive cesium-137 when contacted by a radioactive waste solution.
  • U.S. Patent 3,167,504 describes a solid synthetic amorphous hydrous zeolite which is effective for removing radiocesium from radioactive liquors.
  • the synthetic crystalline silicate has the approximate empirical formula Na,O (1-2)AI,O, (4-14) SiO,.
  • U.S. Patent 3,262,885 describes a process for consolidating fission-products-containing zeolites by mixing the zeolite with lithium flouride, silica and boron oxide, heating the mixture at about 800°C and cooling to room temperature.
  • U.S. Patent 3,380,916 teaches the removal of radioactive cesium and strontium from highly alkaline waste liquor by ion exchange with a zeolite which is sodium, potassium, calcium, cadmium, strontium, copper, zinc, cobalt, iron, silver, or nickel aluminosilicate.
  • U. S. Patent 4,087,375 describes the capture of corrosive radionucleides, including Cs(137) and Sr(89-90) by ion exchange or mechanical adsorption with the mordenite form of zeolite, sintering of the zeolite to a ceramic form, and thereby sealing and fixing the nuclides in sintered mordenite.
  • U. S. Patent 4,097,401 relates to a borosilicate glass which is particularly suitable as the solidification matrix for highly radioactive wastes by denitrating and calcining the radioactive waste solution, melting the calcined waste product together with a borosilicate frit containing nucleation agents at a temperature of 1050-1200°c, maintaining the melted frit for 3-5 hours, and subsequently cooling and then heating the frit to permit nucleation to occur.
  • the crystalline zeolites used for ion exchange of radioactive solutions are preferably partially in the hydrogen form.
  • the original cations associated therewith may be replaced by a wide variety of other cations according to techniques well known in the art.
  • typical replacing cations include hydrogen, ammonium, and metal cations, including mixtures of the same, particular preference being given to cations of metals such as rare earth metals, manganese, and calcium, as well as metals of Group 11 of the Periodic Table.
  • all dissolved cations are acceptable to, and are received within, the zeolite structure.
  • Typical ion-exchange techniques involve contacting the zeolite with a solution of a salt of the desired replacing cation or cations.
  • a salt of the desired replacing cation or cations can be employed, particular preference is giving to chlorides, nitrates, and sulfates. In radioactive waste solutions, nitrates are the predominant and indeed almost the sole anion.
  • Ordinary zeolites or aluminosilicates although very effective in adsorbing radioactive nucleides from liquid or gaseous wastes, contain substantial amounts of aluminum.
  • the high aluminum content increases the melting point and viscosity of the glass mix used to contain the radioactive- containing zeolite, thus making the disposal process extremely difficult and uneconomical to operate.
  • the present invention provides a process for removing radioactive ions from waste water comprising: contacting said radioactive ion-containing-waste water with a zeolite which contains boron within the crystal framework thereof for a contact time sufficient for said zeolite to adsorb said radioactive ions, admixing said radioactive ion-containing zeolite with a soft glass powder to form a uniform mixture; and sintering said mixture to form a glass like melt.
  • the boron-containing zeolite is surprisingly effective for removing both strontium and cesium from radioactive waste liquors and due to the replacement of aluminum in the zeolite framework the radwaste-loaded zeolite can be consolidated with glass at lower temperatures than has been previously feasible.
  • zeolites which determine their performance in nuclear waste treatment are: (1) structural stability under high temperature and intense radioactivity; (2) composition that is compatible with the subsequent glass making step; and (3) high selectivity for the ions of strontium and cesium.
  • high-silicate borosilicates have structures which are thermally more stable than low SiO,/AI,O, aluminosilicates, and that boron in a glass mix tends to lower the melting point of the mix and is compatible therewith
  • the use of boron-containing zeolite adsorbents has not been suggested for removing radioactive nuclides from waste liquors. It has been found that the substitution of boron for framework aluminum does not effect selectivities of the zeolite for individual ions and thus the boron sites exhibit ion exchange capacity equivalent to that of the aluminum sites.
  • the figure is a comparison of various ion exchange isotherms for aluminosilicate zeolite beta and borosilicate zeolite beta.
  • the process of this invention for treating waste water containing radio cesium and radio strontium in solution comprises the following steps:
  • the melt begins to soften at about 850°C and is completely vitrified at about 950°C.
  • the contacting step of the process is conducted by:
  • the borosilicate zeolites useful in the present invention can be any zeolite which can be formed with boron atoms . in the zeolite crystal framework.
  • useful zeolites can be true borosilicates with only impurity levels of aluminum present in the zeolite framework or boron-containing aluminosilicate zeolites in which a portion of the framework aluminum atoms are replaced by boron atoms.
  • the useful zeolites will contain sufficient boron such that the firing temperature of the radwaste-loaded zeolite and glass mixture can be lowered from the typical 1050°C.
  • Zeolites which contain a relatively high aluminum content such as to provide a silica/alumina mole ratio of at least 5 can have incorporated in the zeolite framework a boron content sufficient to lower the melting temperature of the glass matrix or at least lower the viscosity of the mix.
  • a relatively high aluminum content zeolites will contain sufficient boron to provide a silica/boria ratio of at most about 1000.
  • the zeolite ion exchanger will have a silica to alumina ratio of 12 to 1000 and contain sufficient boron to provide a silica/boria ratio of 10 to 500.
  • the zeolite ion exchanger will be free of alumina providing a silica/alumina ratio of at least 200.
  • the boron content should be such to provide a silica/boria ratio of no greater than 1000.
  • such aluminum-free zeolites will have a framework silica/boria ratio of 5 to 500.
  • the boron framework sites of the zeolite are just as capable of ion exchange as the aluminum sites such that there is no loss in effectiveness in regards to the ion exchange capacity and ion-exchange selectivity of the borosilicate for the radioactive cesium and strontium ions.
  • aluminum-free zeolites which contain framework boron are effecrive ion exchangers and additionally substantially lower the temperatures required for fixing the radwaste loaded zeolite into the glass matrix for disposal.
  • borosilicates or boron containing zeolites wherein at least a portion of the aluminum framework atoms of the zeolite are replaced with boron is known and described in the patent literature.
  • U.S. Paten No. 3,328, 1 19 discloses a synthetic crystalline aluminosilicate zeolite containing boria as an intricate part of the crystal framework in general and, more specifically, discloses boria-substituted synthetic zeolite A, synthetic faujasite and synthetic mordenite.
  • U.S. Patent Nos. 4,268,429; 4,269,813; and 4,285,919; issued to Klotz disclose the formation of borosilicate zeolites.
  • the medium and large pore zeolites are preferred in this invention.
  • Such zeolites include ZSM-5, ZSM-11, ZSM-12, ZSM-23, ZSM-35, ZSM-38, ZSM-48, and zeolite beta.
  • Aluminosilicate zeolite beta is disclosed in U.S. Patent No. Re 28,341.
  • the borosilicate zeolites useful in the present invention can be prepared from reaction mixtures containing a source of cations, such as, for example, organic nitrogen-containing cations, an alkali or alkaline earth metal ion source, a source of silicon, such as an oxide of silicon, a source of aluminum such as an oxide of aluminum, water and a source of boron, such as, for example, an oxide of boron.
  • a source of cations such as, for example, organic nitrogen-containing cations, an alkali or alkaline earth metal ion source
  • silicon such as an oxide of silicon
  • a source of aluminum such as an oxide of aluminum
  • water a source of boron, such as, for example, an oxide of boron.
  • boron such as, for example, an oxide of boron.
  • the reaction mixtures will have compositions, in terms of mole ratios of oxides, within the following ranges: wherein R represents organic cations and M represents alkali or al
  • Reaction conditions consist of heating the foregoing reaction mixtures to a temperature of 80°C to 200°C for 24 hours to 90 days.
  • a more preferred temperature range is 100°C to 180°C for 24 hours to 21 days.
  • the digestion of the gel particles is carried out until crystals of the desired boroaluminosilicate form.
  • the crystalline product is separated from the reaction medium, as by cooling the whole to room temperature, tittering and water washing at conditions including a pH above 7.
  • the above reaction mixture compositons can be prepared utilizing materials which supply the appropriate oxides.
  • Such compositons may include sodium silicate, silica hydrosol, silica gel, silicic acid, sodium hydroxide, a source of aluminum, a source of boron and an appropriate organic compound.
  • the source of aluminum way be an added aluminum-containing compound or silica-containing materials or alkali metal-containing materials containing aluminum.
  • the organic compounds contain an element of Group VA, such as nitrogen or phosphorus.
  • the organic compound selected may direct synthesis toward one or another zeolite structure for the boroaluminosilicate material prepared.
  • primary organic amines containing from 2 to about 10 carbon atoms or organic ammonium compounds such as tetraalkylammonium compounds in which the alkyl contains from 2 to 5 carbon atoms will direct the formation of boroaluminosilicate having the structure of zeolite ZSM-5 from the above reaction mixture under appropriate conditions.
  • organic ammonium compounds such as tetraalkylammonium compounds in which the alkyl contains from 2 to 5 carbon atoms
  • the quaternary compounds of tetrabutylammonium chloride or hydroxide may be used to direct synthesis under appropriate conditions of boroaluminosilicate have the structure of ZSM-5/ZSM-11 intermediate or ZSM- 11 .
  • Tetraethylammonium cation sources may be used to direct synthesis of boroaluminosilicate having the structure of ZSM-12 under appropriate conditions.
  • a boroaluminosilicate having the structure of ZSM-23 may be directed from the reaction mixture to using pyrrolidine as the organic compound. Ethylenediamine as well as pyrrolidine will direct a ZSM-35 structure and 2-(hydroxyalkyl)trialkylammonium compounds such as 2-(hydroxyethyl)trimethylammoninm chloride will direct a ZSM-38 structure. If the reaction mixture contains a molar ratio of C,-C I2 alkylamine/tetramethylammonium compound within the range of from 1/1 to 10/1, a boroaluminosilicate having the structure of ZSM-48 may be formed under appropriate conditions.
  • the reaction mixture from which it is to be crystallized will have a composition, in terms of mole ratios of oxides, within the following ranges: wherein R and M are as above defined.
  • Reaction conditions here include a temperature of 80°C to 200°C, preferably 100°C to 180°C, for 40 hours to 30 days, preferably from 60 hours to 15 days.
  • a boroaluminosilicate having the structure of zeolite ZSM-11 is desired, for example, the reaction mixture from which it is to be crystallized will have a composition, in terms of mole ratios of oxides, within the following ranges: wherein R and M are as above defined.
  • Crystallization temperatures and times are as indicated above for preparation of boroaluminosilicate ZSM-5.
  • the reaction mixture from which it is to be crystallized will have a composition, in terms of mole ratios of oxides, within the following ranges: wherein R and M are as above defined.
  • Reaction conditions here include a temperature of 85°C to 175°C, preferably from 130°C to 150°C, for 24 hours to 50 days, preferably from 24 hours to 21 days.
  • the reaction mixture from which it is to be crystallized will have a composition, in terms of mole ratios of oxides, within the following ranges: wherein R and M are as above defined.
  • Reaction conditions here include a temperature of 85°C to 200°C, preferably from 130°C to 175°C, for 24 hours to 90 days, preferably from 24 hours to 21 days.
  • the reaction mixture from which it is to be crystallized will have a composition, in terms of mole ratios of oxides, within the following ranges: wherein R and M are as above defined.
  • Crystallization temperatures and times are as listed above for preparation of boroaluminosilicate ZSM-11.
  • Another way to direct synthesis of the present boroaluminosilicate molecular sieve having a particular crystal structure is to provide seed crystals of the desired structure, e.g. boroaluminosilicate zeolite of ZSM-5 structure, in the reaction mixture initially. This may be facilitated by providing at least about 0.01 percent, preferably at least 0.1 percent and still more preferably at least 1 percent seed crystals of the desired boroaluminosilicate (based on total reaction mixture weight).
  • This example compares the ion exchange capacity of borosilicate zeolite beta with aluminosilicate zeolite beta.
  • the borosilicate zeolite beta samples were converted to the sodium form by conventional nitrogen precalcination followed by sodium back exchange.
  • all exchange solutions in this example were kept at a pH higher than 9 by the addition of sodium hydroxide.
  • the properties of borosilicate zeolite beta and aluminosilicate zeolite beta are shown in Table I below.
  • the borosilicate zeolite beta was ion exchanged with borosilicate or aluminosilicate zeolite beta are indistinguish- various alkali metal ions and strontium. able. This can be seen from Figure 1, where the solid lines
  • the chemical compositon analysis and cation balance 25 are ion exchange isotherms for the aluminosilicate zeolite of the exchanged zeolite beta indicate that the boron, in- beta and data points are those for the boron containing deed, is incorporated within the framework and has ion sample.
  • the equivalent fractions Z M , S M can be defined exchange capacity.
  • the borosilicate zeolite beta of Example 1 was compared with sodium exchanged zeolite X and zeolite A for strontium and cesium ion exchange capacity.
  • the temperature of ion exchange was 25°C with a pH of the exchange solution kept higher than 9. The results are shown in Table 2.
  • alpha A/B Relative ion exchange selectivity of zeolites for various cations has been obtained and expressed as the separation factor, alpha A/B, which is defined as:
  • alpha A/B is the separation factor of ion A over ion B and Z A
  • Za, S A and S B are ionic fractions of A and B in the zeolite and solution phase, respectively.
  • a borosilicate beta was used for ion exchanging strontium ion.
  • the zeolite When saturated with strontium ion, the zeolite was dried and one part thereof was mixed with four parts of type 0220 soft glass powder from Owens Corning Illinois Company to form a uniform mixture. This mixture was then heated in an electric furnace. The mixture started to soften at about 850°C and completely vitrified into a glass-like melt at about 950°C, a temperature that is considerably lower than 1050°C which is the meting temperature of a mix of the same glass powder and aluminosilicate.
  • the glass melt was cooled to room temperature. Fifty parts of water were then added to one part of the glass solids. The aqueous mixture was heated at 100°C for 24 hours. The glass solids, after cooling and after leaching in water, were analyzed. The strontium contents in the glass solids before and after the water-leaching step were found to be exactty the same, indicating that the strontium ion containing Beta was fixed and formed an integral part of the leach-resisting glass compound.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Inorganic Chemistry (AREA)
  • Solid-Sorbent Or Filter-Aiding Compositions (AREA)
  • Silicates, Zeolites, And Molecular Sieves (AREA)
EP85306341A 1984-10-25 1985-09-06 Zéolite an borosilicate pour l'élimination des déchets nucléaires Withdrawn EP0180308A1 (fr)

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Cited By (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
FR2927725A1 (fr) * 2008-02-18 2009-08-21 Commissariat Energie Atomique Procede de decontamination d'un effluent liquide en un ou plusieurs elements chimiques par extraction solide-liquide mettant en oeuvre une boucle de recyclage
CN102208223A (zh) * 2011-04-29 2011-10-05 清华大学 一种锶铯共固化体的制备方法
RU2499309C2 (ru) * 2011-07-20 2013-11-20 Елена Михайловна Евсина Сорбент для удаления радионуклидов из воды
CN107188533A (zh) * 2017-06-07 2017-09-22 西南科技大学 一种地聚合物陶瓷固化高放废液的方法
US20200030771A1 (en) * 2018-07-26 2020-01-30 Korea Atomic Energy Research Institute Radionuclide adsorbent, method for preparing the same and method for removing radionuclide using the same
WO2020055337A1 (fr) * 2018-09-14 2020-03-19 Vuje, A.S. Additifs pour la vitrification de déchets contenant des radionucléides de césium radioactif liquide ayant une efficacité de rétention élevée desdits radionucléides sur toute la plage de température de vitrification, leur procédé de préparation et leur utilisation
CN113053554A (zh) * 2021-04-21 2021-06-29 西南交通大学 一种通过水热-烧结联用固化放射性元素的方法

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* Cited by examiner, † Cited by third party
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JP6385141B2 (ja) * 2014-05-30 2018-09-05 株式会社 エー・イー・エル 放射能汚染水中の放射性汚染物質の除去方法

Citations (5)

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FR2291583A1 (fr) * 1974-11-15 1976-06-11 Atomenergi Ab Procede pour l'elimination et la neutralisation d'un isotope radioactif a partir d'une solution aqueuse
FR2310616A1 (fr) * 1975-05-07 1976-12-03 Shin Tohoku Chemical Ind Co Lt Procede de traitement des eaux residuaires radio-actives
GB2025685A (en) * 1978-07-18 1980-01-23 Nukem Gmbh A process for solidifying radioactive fission products
FR2440778A1 (fr) * 1978-11-09 1980-06-06 Macedo Pedro Fixation par echange d'ions de matieres toxiques dans une matrice de verre
EP0049936A1 (fr) * 1980-10-13 1982-04-21 European Atomic Energy Community (Euratom) Procédé pour capsuler des matériaux dans un zéolite, d'une manière stable

Patent Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
FR2291583A1 (fr) * 1974-11-15 1976-06-11 Atomenergi Ab Procede pour l'elimination et la neutralisation d'un isotope radioactif a partir d'une solution aqueuse
FR2310616A1 (fr) * 1975-05-07 1976-12-03 Shin Tohoku Chemical Ind Co Lt Procede de traitement des eaux residuaires radio-actives
GB2025685A (en) * 1978-07-18 1980-01-23 Nukem Gmbh A process for solidifying radioactive fission products
FR2440778A1 (fr) * 1978-11-09 1980-06-06 Macedo Pedro Fixation par echange d'ions de matieres toxiques dans une matrice de verre
EP0049936A1 (fr) * 1980-10-13 1982-04-21 European Atomic Energy Community (Euratom) Procédé pour capsuler des matériaux dans un zéolite, d'une manière stable

Cited By (13)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
FR2927725A1 (fr) * 2008-02-18 2009-08-21 Commissariat Energie Atomique Procede de decontamination d'un effluent liquide en un ou plusieurs elements chimiques par extraction solide-liquide mettant en oeuvre une boucle de recyclage
WO2009103703A1 (fr) * 2008-02-18 2009-08-27 Commissariat A L'energie Atomique Procede de decontamination d'un effluent liquide radioactif en un ou plusieurs elements chimiques radioactifs par extraction solide-liquide mettant en oeuvre une boucle de recyclage
CN101952898B (zh) * 2008-02-18 2014-03-26 法国原子能与替代能源委员会 使用再循环回路通过固-液萃取法对放射性液态流出物净化的方法
US8696911B2 (en) 2008-02-18 2014-04-15 Commissariat A L'energie Atomique Et Aux Energies Alternatives Decontamination of radioactive liquid effluent by solid-liquid extraction using a recycle loop
CN102208223A (zh) * 2011-04-29 2011-10-05 清华大学 一种锶铯共固化体的制备方法
RU2499309C2 (ru) * 2011-07-20 2013-11-20 Елена Михайловна Евсина Сорбент для удаления радионуклидов из воды
CN107188533A (zh) * 2017-06-07 2017-09-22 西南科技大学 一种地聚合物陶瓷固化高放废液的方法
CN107188533B (zh) * 2017-06-07 2020-08-11 西南科技大学 一种地聚合物陶瓷固化高放废液的方法
US20200030771A1 (en) * 2018-07-26 2020-01-30 Korea Atomic Energy Research Institute Radionuclide adsorbent, method for preparing the same and method for removing radionuclide using the same
WO2020055337A1 (fr) * 2018-09-14 2020-03-19 Vuje, A.S. Additifs pour la vitrification de déchets contenant des radionucléides de césium radioactif liquide ayant une efficacité de rétention élevée desdits radionucléides sur toute la plage de température de vitrification, leur procédé de préparation et leur utilisation
JP2022502678A (ja) * 2018-09-14 2022-01-11 ブイエ，アー．エス． 全ガラス化温度範囲にわたり放射性核種の高い保持効率を有する、液体で放射性のセシウム放射性核種含有廃棄物のガラス化のための添加剤、それらの調製方法およびそれらの使用
JP7114816B2 (ja) 2018-09-14 2022-08-08 ブイエ，アー．エス． 全ガラス化温度範囲にわたり放射性核種の高い保持効率を有する、液体で放射性のセシウム放射性核種含有廃棄物のガラス化のための添加剤、それらの調製方法およびそれらの使用
CN113053554A (zh) * 2021-04-21 2021-06-29 西南交通大学 一种通过水热-烧结联用固化放射性元素的方法

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