USH1013H - Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing - Google Patents
Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing Download PDFInfo
- Publication number
- USH1013H USH1013H US07/392,327 US39232789A USH1013H US H1013 H USH1013 H US H1013H US 39232789 A US39232789 A US 39232789A US H1013 H USH1013 H US H1013H
- Authority
- US
- United States
- Prior art keywords
- thorium
- mixture
- weight
- low level
- immobilization
- 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.)
- Abandoned
Links
- 238000000034 method Methods 0.000 title claims abstract description 21
- 239000002925 low-level radioactive waste Substances 0.000 title claims abstract description 5
- ZSLUVFAKFWKJRC-IGMARMGPSA-N 232Th Chemical compound [232Th] ZSLUVFAKFWKJRC-IGMARMGPSA-N 0.000 title claims description 12
- 229910052776 Thorium Inorganic materials 0.000 title claims description 12
- 229910052770 Uranium Inorganic materials 0.000 title claims description 11
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 title claims description 11
- 238000011038 discontinuous diafiltration by volume reduction Methods 0.000 title abstract description 6
- VYPSYNLAJGMNEJ-UHFFFAOYSA-N Silicium dioxide Chemical compound O=[Si]=O VYPSYNLAJGMNEJ-UHFFFAOYSA-N 0.000 claims abstract description 18
- 239000002699 waste material Substances 0.000 claims abstract description 18
- 239000000203 mixture Substances 0.000 claims abstract description 17
- 239000003795 chemical substances by application Substances 0.000 claims abstract description 10
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 claims abstract description 9
- 239000004576 sand Substances 0.000 claims abstract description 6
- 239000000377 silicon dioxide Substances 0.000 claims abstract description 6
- 238000010438 heat treatment Methods 0.000 claims abstract description 4
- 229910052761 rare earth metal Inorganic materials 0.000 claims abstract description 4
- 150000002910 rare earth metals Chemical class 0.000 claims abstract description 4
- IKNAJTLCCWPIQD-UHFFFAOYSA-K cerium(3+);lanthanum(3+);neodymium(3+);oxygen(2-);phosphate Chemical compound [O-2].[La+3].[Ce+3].[Nd+3].[O-]P([O-])([O-])=O IKNAJTLCCWPIQD-UHFFFAOYSA-K 0.000 claims description 6
- 229910052590 monazite Inorganic materials 0.000 claims description 6
- 239000007787 solid Substances 0.000 claims description 6
- 229910000442 triuranium octoxide Inorganic materials 0.000 claims description 5
- 229910004748 Na2 B4 O7 Inorganic materials 0.000 claims description 4
- 229910003252 NaBO2 Inorganic materials 0.000 claims description 2
- 229910052586 apatite Inorganic materials 0.000 claims description 2
- 238000007496 glass forming Methods 0.000 claims description 2
- VSIIXMUUUJUKCM-UHFFFAOYSA-D pentacalcium;fluoride;triphosphate Chemical compound [F-].[Ca+2].[Ca+2].[Ca+2].[Ca+2].[Ca+2].[O-]P([O-])([O-])=O.[O-]P([O-])([O-])=O.[O-]P([O-])([O-])=O VSIIXMUUUJUKCM-UHFFFAOYSA-D 0.000 claims description 2
- NVIFVTYDZMXWGX-UHFFFAOYSA-N sodium metaborate Chemical compound [Na+].[O-]B=O NVIFVTYDZMXWGX-UHFFFAOYSA-N 0.000 claims description 2
- -1 xenotine Inorganic materials 0.000 claims description 2
- 238000001816 cooling Methods 0.000 claims 1
- 238000001035 drying Methods 0.000 claims 1
- 239000006060 molten glass Substances 0.000 abstract description 8
- 238000011084 recovery Methods 0.000 abstract description 2
- CDBYLPFSWZWCQE-UHFFFAOYSA-L Sodium Carbonate Chemical compound [Na+].[Na+].[O-]C([O-])=O CDBYLPFSWZWCQE-UHFFFAOYSA-L 0.000 abstract 2
- 229910021538 borax Inorganic materials 0.000 abstract 1
- 229910000029 sodium carbonate Inorganic materials 0.000 abstract 1
- 229960001922 sodium perborate Drugs 0.000 abstract 1
- 235000010339 sodium tetraborate Nutrition 0.000 abstract 1
- YKLJGMBLPUQQOI-UHFFFAOYSA-M sodium;oxidooxy(oxo)borane Chemical compound [Na+].[O-]OB=O YKLJGMBLPUQQOI-UHFFFAOYSA-M 0.000 abstract 1
- BSVBQGMMJUBVOD-UHFFFAOYSA-N trisodium borate Chemical compound [Na+].[Na+].[Na+].[O-]B([O-])[O-] BSVBQGMMJUBVOD-UHFFFAOYSA-N 0.000 abstract 1
- 239000002901 radioactive waste Substances 0.000 description 9
- 239000011521 glass Substances 0.000 description 4
- 230000008018 melting Effects 0.000 description 3
- 238000002844 melting Methods 0.000 description 3
- 206010073306 Exposure to radiation Diseases 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 239000002002 slurry Substances 0.000 description 2
- 229910018404 Al2 O3 Inorganic materials 0.000 description 1
- 229910000831 Steel Inorganic materials 0.000 description 1
- 229910004369 ThO2 Inorganic materials 0.000 description 1
- 238000009933 burial Methods 0.000 description 1
- 239000003638 chemical reducing agent Substances 0.000 description 1
- 229910052681 coesite Inorganic materials 0.000 description 1
- 239000012141 concentrate Substances 0.000 description 1
- 238000010924 continuous production Methods 0.000 description 1
- 238000005336 cracking Methods 0.000 description 1
- 229910052906 cristobalite Inorganic materials 0.000 description 1
- 238000009792 diffusion process Methods 0.000 description 1
- 230000008020 evaporation Effects 0.000 description 1
- 238000001704 evaporation Methods 0.000 description 1
- 230000003100 immobilizing effect Effects 0.000 description 1
- 229910052500 inorganic mineral Inorganic materials 0.000 description 1
- 239000000155 melt Substances 0.000 description 1
- 229910044991 metal oxide Inorganic materials 0.000 description 1
- 150000004706 metal oxides Chemical class 0.000 description 1
- 239000011707 mineral Substances 0.000 description 1
- 239000000843 powder Substances 0.000 description 1
- 230000002285 radioactive effect Effects 0.000 description 1
- 229910052704 radon Inorganic materials 0.000 description 1
- SYUHGPGVQRZVTB-UHFFFAOYSA-N radon atom Chemical compound [Rn] SYUHGPGVQRZVTB-UHFFFAOYSA-N 0.000 description 1
- 239000002900 solid radioactive waste Substances 0.000 description 1
- 238000007711 solidification Methods 0.000 description 1
- 230000008023 solidification Effects 0.000 description 1
- 229910001220 stainless steel Inorganic materials 0.000 description 1
- 239000010935 stainless steel Substances 0.000 description 1
- 239000010959 steel Substances 0.000 description 1
- 229910052682 stishovite Inorganic materials 0.000 description 1
- 238000003860 storage Methods 0.000 description 1
- ZCUFMDLYAMJYST-UHFFFAOYSA-N thorium dioxide Chemical compound O=[Th]=O ZCUFMDLYAMJYST-UHFFFAOYSA-N 0.000 description 1
- 231100000331 toxic Toxicity 0.000 description 1
- 230000002588 toxic effect Effects 0.000 description 1
- 239000010891 toxic waste Substances 0.000 description 1
- 229910052905 tridymite Inorganic materials 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
- 230000004584 weight gain Effects 0.000 description 1
- 235000019786 weight gain Nutrition 0.000 description 1
- 230000004580 weight loss Effects 0.000 description 1
- 238000003466 welding Methods 0.000 description 1
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
- 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 of treating radioactive waste, and more specifically to a process for the immobilization and volume reduction of low level radioactive wastes produced by Rare Earth recovery processes.
- pitchblende or monazite produces low level radioactive waste residues which bear thorium, and/or uranium and their natural decay daughters.
- these thorium and/or uranium bearing wastes are in the form of slurries and, as such, present disposal problems due to volume and containment considerations.
- U.S. Pat. No. 4,725,383 teaches a process for volume reduction and solidification of a radioactive waste solution by adding ZnO or a mixture of ZnO with Al 2 O 3 and/or CaO, dehydrating the mixture, and melting to produce a vitrified solid.
- Another process for volume reduction and immobilization of waste is taught by U.S. Pat. No. 4,395,367 wherein fission waste is treated by mixing a glass forming agent, a metal oxide and a reducing agent with the fission waste, and heating the mixture until melted.
- a method of treating the radioactive waste residues which result from the processing of pitchblende or monazite or other thorium or uranium bearing minerals which comprises forming a mixture of a dried radioactive thorium and/or uranium containing waste residue and a fluxing agent, melting the mixture to form molten glass, and pouring the molten glass into a suitable container.
- the present invention is directed to a process of treating the solid thorium and/or uranium containing radioactive waste obtained from the processing of monazite, pitchblende, xenotine, apatite, bastnasite or other rare-earth bearing ores.
- the process of this invention comprises mixing the dried waste residue with about 0.1 to 50% by weight of a fluxing agent, optionally about 0.1 to 50% by weight SiO 2 , and heating the mixture to a molten glassy state, and pouring the molten glass into a suitable container for burial or storage.
- the waste residues capable of being treated under this process are any solid radioactive waste residues which bear thorium or uranium, and their natural decay daughters.
- Appropriate fluxing agents for use in the invention include, but are not limited to, NaOH, Na 2 CO 3 , NaBO 2 , Na 2 B 4 O 7 or mixtures thereof.
- Silica sand may also be added to the mixture if necessary.
- the specific proportions of fluxing agent and silica sand to be added to the mixture can vary depending on the desired melt viscosity and are not per se critical to the invention provided that adequate pour viscosities are developed in the mixture.
- the preferred melt viscosities of this invention are in the range 1000 to 3000 centipoise at pour temperatures of 1093° to 1260° C.
- the dried radioactive waste residues, together with the fluxing agent and optionally silica sand, may be added batchwise or in a continuous process to an appropriate glass furnace.
- the furnace is typically heated to a temperature in the range 1200° to 1800° C., and the glass residence times in the melter are between 2 and 24 hours. Furnace temperatures and glass residence times are generally interdependent. Therefore, if a temperature higher than 1800° C. is used, a correspondingly shorter residence time will be required to achieve the desired molten glass state.
- the molten glass waste residues are then poured into steel or stainless steel, or other similar containers in which the melt is cooled to form a solid, vitrified mass, whereupon the containers can be sealed by welding or other suitable method.
- the final percentage of thorium and uranium present in the reduced volume vitrified waste is about 0.1 to 50% by weight and 0.01 to 10% by weight respectively.
- Thorium bearing radioactive waste residues in the form of a thick slurry were obtained from a monazite cracking plant. The initial density of this residue was 1.85 g/cc.
- the waste residues were dried, ground to a fine powder, and mixed with 20% by weight Na 2 B 4 O 7 . The mixture was heated to 1540° C. for two hours to produce a molten glass that was readily pourable. The glass residue was cooled to room temperature and had a density of 4.0 g/cc. After accounting for the weight loss due to water evaporation, and weight gain from the addition of the Na 2 B 4 O 7 , this represents approximately a 70% volume reduction.
Landscapes
- Chemical & Material Sciences (AREA)
- Inorganic Chemistry (AREA)
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Processing Of Solid Wastes (AREA)
Abstract
A process for the immobilization and volume reduction of low level radioactive wastes produced from the processing of rare earth recovery processes comprising mixing the waste residue with 0.1 to 50% of a fluxing agent and optionally with 0.1 to 50% silica sand, heating the mixture to a temperature in the range of about 1200° to 1800° C. to form molten glass, and pouring the molten glass into a suitable container to cool and solidify into a vitrified mass. Suitable fluxing agents include sodium hydroxide, sodium carbonate, sodium borate, sodium perborate, or mixtures thereof.
Description
1. Field of Invention
This invention relates to a process of treating radioactive waste, and more specifically to a process for the immobilization and volume reduction of low level radioactive wastes produced by Rare Earth recovery processes.
2. Description of the Prior Art
The processing of pitchblende or monazite produces low level radioactive waste residues which bear thorium, and/or uranium and their natural decay daughters. Typically, these thorium and/or uranium bearing wastes are in the form of slurries and, as such, present disposal problems due to volume and containment considerations.
Various processes are known in the prior art to concentrate and immobilize toxic and/or radioactive wastes. For example, U.S. Pat. No. 4,725,383 teaches a process for volume reduction and solidification of a radioactive waste solution by adding ZnO or a mixture of ZnO with Al2 O3 and/or CaO, dehydrating the mixture, and melting to produce a vitrified solid. Another process for volume reduction and immobilization of waste is taught by U.S. Pat. No. 4,395,367 wherein fission waste is treated by mixing a glass forming agent, a metal oxide and a reducing agent with the fission waste, and heating the mixture until melted.
However, none of the known processes are effective on thorium and/or uranium bearing waste residues due to the extremely high melting point of ThO2 (3200° C.) and UO2 (2500° C.). Under the process of this invention, it has now been discovered that the volume of these thorium and/or uranium bearing waste residues can be reduced by as much as 60 to 80 percent and that it is possible to immobilize these wastes into a solid, vitrified mass.
It is an object of this invention to provide a process for the concentration and immobilization of radioactive waste residues.
It is another object of this invention to provide a process for vitrifying the waste residues obtained from processing pitchblende or monazite.
It is another object of this invention to provide a process for reducing the volume of radioactive waste residues and immobilizing these residues to produce a dustless environmentally safe form.
It is another object of this invention to reduce the diffusion of radon into the environment, reduce alpha and beta radiation exposures, and maintain gamma radiation exposure within reasonable limits.
Under the process of the present invention, there has been provided a method of treating the radioactive waste residues which result from the processing of pitchblende or monazite or other thorium or uranium bearing minerals, which comprises forming a mixture of a dried radioactive thorium and/or uranium containing waste residue and a fluxing agent, melting the mixture to form molten glass, and pouring the molten glass into a suitable container.
The present invention is directed to a process of treating the solid thorium and/or uranium containing radioactive waste obtained from the processing of monazite, pitchblende, xenotine, apatite, bastnasite or other rare-earth bearing ores. The process of this invention comprises mixing the dried waste residue with about 0.1 to 50% by weight of a fluxing agent, optionally about 0.1 to 50% by weight SiO2, and heating the mixture to a molten glassy state, and pouring the molten glass into a suitable container for burial or storage.
The waste residues capable of being treated under this process are any solid radioactive waste residues which bear thorium or uranium, and their natural decay daughters. Appropriate fluxing agents for use in the invention include, but are not limited to, NaOH, Na2 CO3, NaBO2, Na2 B4 O7 or mixtures thereof. Silica sand may also be added to the mixture if necessary. The specific proportions of fluxing agent and silica sand to be added to the mixture can vary depending on the desired melt viscosity and are not per se critical to the invention provided that adequate pour viscosities are developed in the mixture. The preferred melt viscosities of this invention are in the range 1000 to 3000 centipoise at pour temperatures of 1093° to 1260° C.
The dried radioactive waste residues, together with the fluxing agent and optionally silica sand, may be added batchwise or in a continuous process to an appropriate glass furnace. The furnace is typically heated to a temperature in the range 1200° to 1800° C., and the glass residence times in the melter are between 2 and 24 hours. Furnace temperatures and glass residence times are generally interdependent. Therefore, if a temperature higher than 1800° C. is used, a correspondingly shorter residence time will be required to achieve the desired molten glass state.
The molten glass waste residues are then poured into steel or stainless steel, or other similar containers in which the melt is cooled to form a solid, vitrified mass, whereupon the containers can be sealed by welding or other suitable method.
The final percentage of thorium and uranium present in the reduced volume vitrified waste is about 0.1 to 50% by weight and 0.01 to 10% by weight respectively.
Without further elaboration, it is believed that one skilled in the art, using the preceding detailed description can utilize the present invention to its fullest extent.
The following example is provided to illustrate the invention in accordance with the principles of this invention, but is not to be construed as limiting the invention in any way except as indicated in the appended claims. All parts and percentages are by weight unless otherwise indicated.
Thorium bearing radioactive waste residues in the form of a thick slurry were obtained from a monazite cracking plant. The initial density of this residue was 1.85 g/cc. The waste residues were dried, ground to a fine powder, and mixed with 20% by weight Na2 B4 O7. The mixture was heated to 1540° C. for two hours to produce a molten glass that was readily pourable. The glass residue was cooled to room temperature and had a density of 4.0 g/cc. After accounting for the weight loss due to water evaporation, and weight gain from the addition of the Na2 B4 O7, this represents approximately a 70% volume reduction.
Claims (3)
1. A process for treating low level radioactive waste comprising:
(a) drying a waste residue obtained from the removal of rare-earths from a rare-earth bearing ore selected from the group consisting of monazite, pitchblende, xenotine, apatite and bastnasite to obtain a dried waste residue which contains 0.1 to 50% by weight thorium and/or 0.01 to 10% by weight uranium;
(b) mixing the dried residue with silica sand and a glass forming agent selected from the group consisting of NaOH, Na2 CO3, Na2 B4 O7 and NaBO2 and mixtures thereof to obtain a mixture which contains 0.1 to 50% by weight of said agent;
(c) heating the mixture to a temperature of 1200° to 1800° C. to obtain a molten mass; and
(d) cooling said molten mass to obtain a solid, vitrified mass having a volume 60 to 80 percent less than said waste residue.
2. The process of claim 1 wherein the sand is added in amounts ranging from 0.1 to 50 percent by weight of the total mixture.
3. A solid vitrified product prepared by the process of claim 2.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US07/392,327 USH1013H (en) | 1989-08-11 | 1989-08-11 | Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing |
| AU60176/90A AU6017690A (en) | 1989-08-11 | 1990-08-06 | Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing |
| BR909003923A BR9003923A (en) | 1989-08-11 | 1990-08-09 | PROCESS FOR IMMOBILIZATION AND REDUCING THE VOLUME OF LOW LEVEL RADIOACTIVITY WASTE CONTAINING TORIO AND URANIUM, COMPOSITION OF VITRIFIED AND REDUCED VOLUME RADIOACTIVE WASTE |
| FR9010270A FR2652193A1 (en) | 1989-08-11 | 1990-08-10 | Method for immobilisation and volume reduction of low-level radioactive waste in the processing of thorium and uranium |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US07/392,327 USH1013H (en) | 1989-08-11 | 1989-08-11 | Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| USH1013H true USH1013H (en) | 1992-01-07 |
Family
ID=23550155
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| US07/392,327 Abandoned USH1013H (en) | 1989-08-11 | 1989-08-11 | Process for the immobilization and volume reduction of low level radioactive wastes from thorium and uranium processing |
Country Status (4)
| Country | Link |
|---|---|
| US (1) | USH1013H (en) |
| AU (1) | AU6017690A (en) |
| BR (1) | BR9003923A (en) |
| FR (1) | FR2652193A1 (en) |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5350569A (en) * | 1993-03-30 | 1994-09-27 | The United States Of America As Represented By The United States Department Of Energy | Storage of nuclear materials by encapsulation in fullerenes |
| RU2176416C1 (en) * | 2000-07-31 | 2001-11-27 | Институт структурной макрокинетики и проблем материаловедения РАН | Radioactive waste immobilization process |
| US6454695B1 (en) * | 1998-02-05 | 2002-09-24 | Fumie Morishige | Therapeutic instrument for treating or relieving psoriasis, atopic dermatitis, articular rheumatism and/or cancer or preventing the progress of these diseases and method of utilization thereof |
| US6635796B2 (en) | 1990-03-16 | 2003-10-21 | Sevenson Environmental Services, Inc. | Reduction of leachability and solubility of radionuclides and radioactive substances in contaminated soils and materials |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| DE4427179A1 (en) * | 1994-08-01 | 1996-02-08 | Siemens Ag | Process for recycling metal parts that are radioactively contaminated by uranium |
Citations (13)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4020004A (en) | 1975-11-21 | 1977-04-26 | The United States Of America As Represented By The United States Energy Research And Development Administration | Conversion of radioactive ferrocyanide compounds to immobile glasses |
| US4312774A (en) | 1978-11-09 | 1982-01-26 | Pedro B. Macedo | Immobilization of radwastes in glass containers and products formed thereby |
| US4354954A (en) | 1978-04-29 | 1982-10-19 | Kernforschungszentrum Karlsruhe Gesellschaft Mit Beschrankter Haftung | Method for solidifying aqueous radioactive wastes for noncontaminating storage |
| US4376070A (en) | 1980-06-25 | 1983-03-08 | Westinghouse Electric Corp. | Containment of nuclear waste |
| US4382974A (en) | 1981-03-19 | 1983-05-10 | Westinghouse Electric Corp. | Synthetic monazite coated nuclear waste containing glass |
| US4395367A (en) | 1981-11-17 | 1983-07-26 | Rohrmann Charles A | Process for treating fission waste |
| US4422965A (en) | 1980-08-11 | 1983-12-27 | Westinghouse Electric Corp. | Nuclear waste encapsulation in borosilicate glass by chemical polymerization |
| US4514329A (en) | 1981-07-06 | 1985-04-30 | Agency Of Industrial Science & Technology | Process for vitrifying liquid radioactive waste |
| US4528011A (en) | 1979-04-30 | 1985-07-09 | Pedro B. Macedo | Immobilization of radwastes in glass containers and products formed thereby |
| US4534893A (en) | 1982-04-17 | 1985-08-13 | Kernforschungszentrum Karlsruhe Gmbh | Method for solidifying radioactive wastes |
| US4626382A (en) | 1983-07-06 | 1986-12-02 | Deutsche Gesellschaft Fur Wiederaufarbeitung Von Kernbrennstoffen Mbh | Method of producing a glass block containing radioactive fission products and apparatus therefor |
| US4661291A (en) | 1984-09-25 | 1987-04-28 | Mitsui Engineering & Shipbuilding Co., Ltd. | Method for fixation of incinerator ash or iodine sorbent |
| US4725383A (en) | 1983-08-09 | 1988-02-16 | Ebara Corporation | Process for volume reduction and solidification of a radioactive sodium borate waste solution |
Family Cites Families (7)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5343640A (en) * | 1976-10-01 | 1978-04-19 | Hitachi Denkaihaku Kenkyusho | Electrolytic etching method for aluminum |
| DE2731327C3 (en) * | 1977-07-12 | 1981-01-22 | Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe | Process for filtering dust from radioactive waste gases and equipment for carrying out the process |
| JPS58115066A (en) * | 1981-12-25 | 1983-07-08 | 動力炉・核燃料開発事業団 | Artificial ore |
| JPS6038700A (en) * | 1983-08-10 | 1985-02-28 | 東京電力株式会社 | Method of melting and solidifying radioactive waste incinerated ash |
| JPS60236098A (en) * | 1984-05-09 | 1985-11-22 | 日本碍子株式会社 | Method of treating radioactive waste |
| JPS6186692A (en) * | 1984-10-05 | 1986-05-02 | 株式会社日立製作所 | Method for solidifying used radioactive ion exchange resin |
| JPS61132899A (en) * | 1984-12-03 | 1986-06-20 | 明星工業株式会社 | Volume reducing method of radioactive heat-insulating waste |
-
1989
- 1989-08-11 US US07/392,327 patent/USH1013H/en not_active Abandoned
-
1990
- 1990-08-06 AU AU60176/90A patent/AU6017690A/en not_active Abandoned
- 1990-08-09 BR BR909003923A patent/BR9003923A/en unknown
- 1990-08-10 FR FR9010270A patent/FR2652193A1/en active Pending
Patent Citations (13)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4020004A (en) | 1975-11-21 | 1977-04-26 | The United States Of America As Represented By The United States Energy Research And Development Administration | Conversion of radioactive ferrocyanide compounds to immobile glasses |
| US4354954A (en) | 1978-04-29 | 1982-10-19 | Kernforschungszentrum Karlsruhe Gesellschaft Mit Beschrankter Haftung | Method for solidifying aqueous radioactive wastes for noncontaminating storage |
| US4312774A (en) | 1978-11-09 | 1982-01-26 | Pedro B. Macedo | Immobilization of radwastes in glass containers and products formed thereby |
| US4528011A (en) | 1979-04-30 | 1985-07-09 | Pedro B. Macedo | Immobilization of radwastes in glass containers and products formed thereby |
| US4376070A (en) | 1980-06-25 | 1983-03-08 | Westinghouse Electric Corp. | Containment of nuclear waste |
| US4422965A (en) | 1980-08-11 | 1983-12-27 | Westinghouse Electric Corp. | Nuclear waste encapsulation in borosilicate glass by chemical polymerization |
| US4382974A (en) | 1981-03-19 | 1983-05-10 | Westinghouse Electric Corp. | Synthetic monazite coated nuclear waste containing glass |
| US4514329A (en) | 1981-07-06 | 1985-04-30 | Agency Of Industrial Science & Technology | Process for vitrifying liquid radioactive waste |
| US4395367A (en) | 1981-11-17 | 1983-07-26 | Rohrmann Charles A | Process for treating fission waste |
| US4534893A (en) | 1982-04-17 | 1985-08-13 | Kernforschungszentrum Karlsruhe Gmbh | Method for solidifying radioactive wastes |
| US4626382A (en) | 1983-07-06 | 1986-12-02 | Deutsche Gesellschaft Fur Wiederaufarbeitung Von Kernbrennstoffen Mbh | Method of producing a glass block containing radioactive fission products and apparatus therefor |
| US4725383A (en) | 1983-08-09 | 1988-02-16 | Ebara Corporation | Process for volume reduction and solidification of a radioactive sodium borate waste solution |
| US4661291A (en) | 1984-09-25 | 1987-04-28 | Mitsui Engineering & Shipbuilding Co., Ltd. | Method for fixation of incinerator ash or iodine sorbent |
Non-Patent Citations (1)
| Title |
|---|
| Bruce et al., Progress in Nuclear Energy, 1958, p. 443-455. |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US6635796B2 (en) | 1990-03-16 | 2003-10-21 | Sevenson Environmental Services, Inc. | Reduction of leachability and solubility of radionuclides and radioactive substances in contaminated soils and materials |
| US5350569A (en) * | 1993-03-30 | 1994-09-27 | The United States Of America As Represented By The United States Department Of Energy | Storage of nuclear materials by encapsulation in fullerenes |
| US6454695B1 (en) * | 1998-02-05 | 2002-09-24 | Fumie Morishige | Therapeutic instrument for treating or relieving psoriasis, atopic dermatitis, articular rheumatism and/or cancer or preventing the progress of these diseases and method of utilization thereof |
| RU2176416C1 (en) * | 2000-07-31 | 2001-11-27 | Институт структурной макрокинетики и проблем материаловедения РАН | Radioactive waste immobilization process |
Also Published As
| Publication number | Publication date |
|---|---|
| BR9003923A (en) | 1991-09-03 |
| AU6017690A (en) | 1991-02-14 |
| FR2652193A1 (en) | 1991-03-22 |
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