GB2073477A - Waste treatment - Google Patents
Waste treatment Download PDFInfo
- Publication number
- GB2073477A GB2073477A GB8108260A GB8108260A GB2073477A GB 2073477 A GB2073477 A GB 2073477A GB 8108260 A GB8108260 A GB 8108260A GB 8108260 A GB8108260 A GB 8108260A GB 2073477 A GB2073477 A GB 2073477A
- Authority
- GB
- United Kingdom
- Prior art keywords
- effluent
- precipitates
- radionuclides
- liquid radioactive
- ions
- 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.)
- Withdrawn
Links
- 239000002699 waste material Substances 0.000 title abstract description 3
- 239000002244 precipitate Substances 0.000 claims abstract description 34
- 230000002285 radioactive effect Effects 0.000 claims abstract description 20
- 239000007788 liquid Substances 0.000 claims abstract description 17
- 239000006228 supernatant Substances 0.000 claims abstract description 5
- 238000000034 method Methods 0.000 claims description 16
- -1 sulphate ions Chemical class 0.000 claims description 7
- 150000002500 ions Chemical class 0.000 claims description 3
- XLJKHNWPARRRJB-UHFFFAOYSA-N cobalt(2+) Chemical compound [Co+2] XLJKHNWPARRRJB-UHFFFAOYSA-N 0.000 claims description 2
- 229910021653 sulphate ion Inorganic materials 0.000 claims description 2
- 229910001429 cobalt ion Inorganic materials 0.000 claims 1
- 238000005202 decontamination Methods 0.000 abstract description 17
- 230000003588 decontaminative effect Effects 0.000 abstract description 16
- TZCXTZWJZNENPQ-UHFFFAOYSA-L barium sulfate Chemical compound [Ba+2].[O-]S([O-])(=O)=O TZCXTZWJZNENPQ-UHFFFAOYSA-L 0.000 abstract description 12
- 229910052792 caesium Inorganic materials 0.000 abstract description 6
- 238000001556 precipitation Methods 0.000 abstract description 3
- 229910052712 strontium Inorganic materials 0.000 abstract description 3
- XLYOFNOQVPJJNP-UHFFFAOYSA-M hydroxide Chemical compound [OH-] XLYOFNOQVPJJNP-UHFFFAOYSA-M 0.000 abstract description 2
- 238000011065 in-situ storage Methods 0.000 abstract description 2
- 238000000926 separation method Methods 0.000 abstract description 2
- 230000007704 transition Effects 0.000 abstract description 2
- VRRFSFYSLSPWQY-UHFFFAOYSA-N sulfanylidenecobalt Chemical compound [Co]=S VRRFSFYSLSPWQY-UHFFFAOYSA-N 0.000 abstract 1
- 238000007792 addition Methods 0.000 description 7
- 238000002474 experimental method Methods 0.000 description 6
- 239000000243 solution Substances 0.000 description 6
- 239000011777 magnesium Substances 0.000 description 5
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 3
- 229910052770 Uranium Inorganic materials 0.000 description 3
- 229910052782 aluminium Inorganic materials 0.000 description 3
- 239000000446 fuel Substances 0.000 description 3
- 238000005342 ion exchange Methods 0.000 description 3
- 229910052742 iron Inorganic materials 0.000 description 3
- 229910052749 magnesium Inorganic materials 0.000 description 3
- 238000005259 measurement Methods 0.000 description 3
- 229910017604 nitric acid Inorganic materials 0.000 description 3
- 229910052707 ruthenium Inorganic materials 0.000 description 3
- 229910052791 calcium Inorganic materials 0.000 description 2
- 238000005253 cladding Methods 0.000 description 2
- 238000000975 co-precipitation Methods 0.000 description 2
- 238000004519 manufacturing process Methods 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
- 239000000047 product Substances 0.000 description 2
- 238000000638 solvent extraction Methods 0.000 description 2
- 238000001179 sorption measurement Methods 0.000 description 2
- 238000003860 storage Methods 0.000 description 2
- 229910018626 Al(OH) Inorganic materials 0.000 description 1
- 229910052684 Cerium Inorganic materials 0.000 description 1
- CWYNVVGOOAEACU-UHFFFAOYSA-N Fe2+ Chemical compound [Fe+2] CWYNVVGOOAEACU-UHFFFAOYSA-N 0.000 description 1
- FYYHWMGAXLPEAU-UHFFFAOYSA-N Magnesium Chemical compound [Mg] FYYHWMGAXLPEAU-UHFFFAOYSA-N 0.000 description 1
- 229910000861 Mg alloy Inorganic materials 0.000 description 1
- UCKMPCXJQFINFW-UHFFFAOYSA-N Sulphide Chemical compound [S-2] UCKMPCXJQFINFW-UHFFFAOYSA-N 0.000 description 1
- 229910001093 Zr alloy Inorganic materials 0.000 description 1
- 229910045601 alloy Inorganic materials 0.000 description 1
- 239000000956 alloy Substances 0.000 description 1
- 229910021502 aluminium hydroxide Inorganic materials 0.000 description 1
- WNROFYMDJYEPJX-UHFFFAOYSA-K aluminium hydroxide Chemical compound [OH-].[OH-].[OH-].[Al+3] WNROFYMDJYEPJX-UHFFFAOYSA-K 0.000 description 1
- 229910052787 antimony Inorganic materials 0.000 description 1
- 239000007864 aqueous solution Substances 0.000 description 1
- 238000009924 canning Methods 0.000 description 1
- 229910052923 celestite Inorganic materials 0.000 description 1
- 239000003153 chemical reaction reagent Substances 0.000 description 1
- 150000001805 chlorine compounds Chemical class 0.000 description 1
- 239000012141 concentrate Substances 0.000 description 1
- 239000000356 contaminant Substances 0.000 description 1
- 238000004090 dissolution Methods 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 229910001679 gibbsite Inorganic materials 0.000 description 1
- 239000010763 heavy fuel oil Substances 0.000 description 1
- 150000004679 hydroxides Chemical class 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 230000003647 oxidation Effects 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- 230000001376 precipitating effect Effects 0.000 description 1
- 238000012958 reprocessing Methods 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 238000010561 standard procedure Methods 0.000 description 1
- 238000012360 testing method Methods 0.000 description 1
- 229910052723 transition metal Inorganic materials 0.000 description 1
- 150000003624 transition metals Chemical class 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 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
- G21F9/04—Treating liquids
- G21F9/06—Processing
- G21F9/10—Processing by flocculation
Landscapes
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Removal Of Specific Substances (AREA)
Abstract
In the treatment of wastes, such as liquid radioactive effluents, it is known to remove radionuclides by successive in situ precipitation of cobalt sulphide, an hydroxide, barium sulphate and a transition element ferrocyanide, followed by separation of the thereby decontaminated effluent. In this invention, use is made of precipitates such as obtained above in the treatment of further fresh liquid radioactive effluent, when it is found that the precipitates have additional capacity for extracting radionuclides. The resulting supernatant liquor may then be subjected to a further precipitation treatment such as above. Decontamination factors for radionuclides of Ce, Ru, Sr and Cs have been considerably enhanced.
Description
SPECIFICATION
Waste treatment
This invention relates to the decontamination of liquid radioactive effluents.
Radio-nuclides present in liquid radioactive effluents such as effluents arising from the reprocessing of irradiated nuclear fuel have been precipitated for two purposes, firstly to reduce the radioactivity in the effluent to a level at which it may be discharged, for example into the sea, and secondly to concentrate the radio-nuclides into as small a volume as possible for storage purposes. British Patent Specification No 1,211,816 describes a method of decontaminating liquid radioactive effluents by, after optional oxidation of the effluent, forming successively in situ and without separation of the respective precipitates, precipitates of ferrous sulphide and hydroxide, barium sulphate and a ferrocyanide of a transition element, the pH being between 7 and 9 at the end of the treatment, and the combined precipitates then separated from the decontaminated effluent.Part of the radio-nuclide content of the effluent thereby becomes incorporated into the precipitates by processes such as adsorption on the surface of the precipitates, co-precipitation by isomorphism and ion-exchange on the precipitates.
When the above process is used to treat an effluent comprising a solution produced by dissolution of irradiated Mg/Zr alloy clad fuel elements in nitric acid, it is generally found that decontamination factors of about 150 may be achieved for py-emitting radionuclides and that decontamination factors of > 1000 may be achieved for a-emitting radionuclides. However, in order to discharge treated effluent into the sea, decontamination factors of the order of 1000 are currently required.
It has now been surprisingly found according to the present invention that precipitates obtained in decontamination processes such as described before have additional capacity for incorporating radionuclides therein and that use may be made of this capacity in the decontamination of further liquid radioactive effluent in order to meet the aforementioned requirements.
Thus, the present invention provides a method of decontaminating liquid radioactive effluents comprising
(i) forming precipitates in the effluent to extract radionuclides from the effluent into the precipitates:
(ii) separating the combined precipitates from the effluent;
(iii) contacting the separated combined precipitates with fresh liquid radioactive effluent thereby to
extract radionuclides from the fresh effluent into the combined precipitates; and
(iv) removing the supernatant effluent resulting from step (iii) and forming precipitates therein to
extract radionuclidesfrom the effluent into the precipitates.
The precipitates may then be separated to give a decontaminated effluent.
In certain test experiments, Decontamination Factors of radionuclides of Cs, Sr and Ce were improved by a factor of 2-30 by the above method in comparison with the prior art method. The invention therefore gives rise to a treated effluent which is more suitable for safe discharge than hitherto. Also, the invention makes more effective use of the capacity of the precipitates to contain radio-nuclides and is therefore capable of giving rise to a reduced volume of precipitates, for example half that obtained hitherto. This is a desirable aim since, ultimately, such precipitates have to be treated and stored in a safe environment.
An example of a liquid radioactive effluent to which the present invention may be applied is the effluent arising as a result of the handling of irradiated fuel from the first generation of British nuclear power stations.
Thus, irradiated uranium metal fuel encapsulated in a magnesium (Magnox) alloy is decanned and the separated Magnox cladding dissolved in nitric acid; solids are removed and a large proportion of the U and
Pu content of the resulting solution is extracted by solvent extraction in accordance with methods known in the art. The remaining aqueous solution constitutes a liquid radioactive effluent and may include for example, ions of Mg, Al, Fe, Ca, U, Pu, Sb, Ce, Sr, Cs and Ru wherein the radionuclides Cue144, Sr90, Cs137 and
Ru106 are particularly noteworthy as radioactive contaminants.
The Mg and Al arise from the canning material. The Fe and Ca arise from incidental additions during storage and the U and Pu are constituted by residual fuel which adhered to the Magnox cladding during the mechanical decanning process and which was incompletely extracted during the above mentioned solvent extraction process. The Sb, Ce, Sr, Cs and Ru arise as fission products of the original irradiation process.
The precipitates produced in step (i) and step (iv) may be precipitates known in the art for removing radioactive nuclides from liquid radioactive effluents and, as is known in the art, their order of production and the conditions under which they are produced (such as pH) may be selected to optimise the removal of particular radionuclides from the effluent. The selection of precipitates and their conditions of production will be determined by the composition of the liquid radioactive effluent. Where the effluent arises from the first generation of British nuclear power stations and has a constitution as indicated above, sulphate ions may be added initially so that BaSO4 may subsequently be precipitated. A transition metal (eg Ni) ferrocyanide may then be precipitated; this removes Cs, probably by an ion exchange mechanism.Sulphide ions may next be provided and Ba++ ion added to precipitate BaSO4 and remove SR by isomorphous precipitation of SrSO4. Co++ ions may be added to precipitate CoS which removes Ru by ion exchange.
Al(OH)3will be precipitated at some stage during addition of OH ions and it is believed that this probably removes Ce, possibly by a co-precipitation or by adsorption onto the Al(OH)3 It has been found that increasing the pH to about 6.5 during step (iii) increases the decontamination factors of Ce, Sr, Cs and Ru.
The invention will now be particularly described by way of example only as follows:
Example
A volume (100 to 200 ml) of simulated liquid radioactive effluent solution in 0.5M HNO3 and of the following composition was prepared:
Mg 20 g/l Cs1.16'mg/l Al 0.14 girl SrO.35mg/l
Fe 0.35 g/l Ru 0.95 mg/l
Ca 0.5 g/l Ce 0.97 mg/l
The listed elements were, of course, present in ionic form. Small quantities of the radioactive nuclides Cs'37, Sr85, Ru103 and Ce'44 were added in the form of their chlorides, ie the solution was 'spiked'.
Step (i) The simulated solution was then treated by the following additions in sequence: -SO4=, OH-, Ni2+ and Fe(CN)B,OH-, S=, Ba2+, Co2+, OH--l floculating aid. This gave precipitates including Ni2Fe(CN)6, BaSO4 and
CoS.
After 11/2 hours settling, the decontamination factors in respect of Ce, Ru, Sr and Cs were determined by standard methods and then additional Ba2+ together with a floculating aid added to precipitate further
BaSO4. After 11/2 hours settling, the decontamination factors were again determined.
Step (iij The supernatant liquor was drawn off to separate the precipitates.
Step (ill) The precipitates were stirred for 1 hour with a similar volume of fresh simulated liquid radioactive effluent, the pH having been adjusted to 6.5 to precipitate hydroxides of Al and Fe. After 2 hours settling, the decontamination factors were again determined.
Step (ivJ The supernatant liquor from step (iii) was drawn off, re-spiked and subjected to the same treatment as described above in step (i),
A final measurement of decontamination factors was carried out after a still further settling of 2 hours.
The decontamination factor (or D F) values obtained are summarised in the table below in respect of two separate experiments, designated experiments 1 and 2 respectively. A decontamination factor in respect of a
particular nuclide is the ratio of its total activity present before and after a particular treatment. It should be
noted that, in view of the low concentration of the radionuclides in solution, the statistical accuracy of the
decontamination factors is not high and that there is necessarily some scatter of results from one experiment to another. The steps referred to in the table correspond to those in the above description, and the
measurements were taken at the stages described above.
Radionuclide and Experiment Number
Ce Ru Sr Cs
Step
1 2 1 2 1 2 1 2
(i)
before Ba2+ addition 263 383 1600 1145 44 52 740 926
after Ba2+ addition 1380 1400 1910 859 1560 1390 2130 1900
(iii) 102 103 491 554 28 25 487 584
(iv)
before Ba2+ addition 66 101 115 97 87 53 227 245
after Ba2+ addition 322 277 124 85 326 245 157 ' 123
final measurement 372 319 124 91 304 318 167 140
Overall OF
Product of steps 35,000 50,000 8,000 75,000 (iii) and (iv)
Additional experiments have been carried out which show that in a process according to the present invention, considerable economies can be made in the use of precipitating reagents, whilst still producing a final effluent with decontamination factors within permissable limits for discharge of the effluent into the sea.
Claims (5)
1. A method of decontaminating liquid radioactive effluents comprising the steps of
(i) forming precipitates in the effluent to extract radionuclides from the effluent into the precipitates;
(ii) separating the combined precipitates from the effluent;
(iii) contacting the separated combined precipitates with fresh liquid radioactive effluent thereby to
extract radionuclides from the fresh effluent into the combined precipitates; and
(iv) removing the supernatant effluent resulting from step (iii) and forming precipitates therein to
extract radionuclides from the effluent into the precipitates.
2. A method according to claim 1 further comprising separating decontaminated effluent from the precipitates formed in step (iv).
3. A method according to claim 1 or 2 comprising adding sulphate ions, sulphide ions, cobalt ions, hydroxide ions and ferrocyanide ions to the effluent in steps (i) and (iv).
4. A method according to claim 1,2 or 3 comprising adjusting pH to about 6.5 during step (iii).
5. A method of decontaminating liquid radioactive effluents according to claim 1 and substantially as herein described.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB8108260A GB2073477A (en) | 1980-03-28 | 1981-03-17 | Waste treatment |
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB8010559 | 1980-03-28 | ||
| GB8108260A GB2073477A (en) | 1980-03-28 | 1981-03-17 | Waste treatment |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| GB2073477A true GB2073477A (en) | 1981-10-14 |
Family
ID=26275016
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| GB8108260A Withdrawn GB2073477A (en) | 1980-03-28 | 1981-03-17 | Waste treatment |
Country Status (1)
| Country | Link |
|---|---|
| GB (1) | GB2073477A (en) |
Cited By (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO1988005204A2 (en) * | 1986-12-24 | 1988-07-14 | Martin Ganter | Process for decontaminating radioactively polluted liquids |
| FR2669460A1 (en) * | 1990-11-16 | 1992-05-22 | Commissariat Energie Atomique | PROCESS FOR THE SEPARATION OF RUTHENIUM AND, POSSIBLY, CESIUM AND COBALT PRESENT IN AQUEOUS SOLUTION SUCH AS AN EFFLUENT FROM AN IRRADY FUEL PURIFYING PLANT. |
| GB2437864A (en) * | 2007-03-20 | 2007-11-07 | Malcolm Brody | Rapid and Selective Dissolution of Magnesium Alloy |
| GB2448208A (en) * | 2007-07-26 | 2008-10-08 | Malcolm Brody | The Removal of Radionuclide Contaminants after the Dissolutionof Magnox |
| RU2620259C1 (en) * | 2016-01-19 | 2017-05-24 | Общество С Ограниченной Ответственностью "Наука - Технологии - Производство" | Sorption material for the selective recovery of radionuclides strontium from solutions with high content of hardness solutions, the method of its production, its application and method of radionuclides extraction from structures with high content of hardness solutions |
| RU2698800C1 (en) * | 2018-07-02 | 2019-08-30 | Федеральное государственное унитарное предприятие "Предприятие по обращению с радиоактивными отходами "РосРАО" (ФГУП "РосРАО") | Method of producing strontium sorbent for solutions containing hardness salts |
-
1981
- 1981-03-17 GB GB8108260A patent/GB2073477A/en not_active Withdrawn
Cited By (10)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO1988005204A2 (en) * | 1986-12-24 | 1988-07-14 | Martin Ganter | Process for decontaminating radioactively polluted liquids |
| WO1988005204A3 (en) * | 1986-12-24 | 1988-08-25 | Martin Ganter | Process for decontaminating radioactively polluted liquids |
| US4925597A (en) * | 1986-12-24 | 1990-05-15 | Martin Ganter | Method for the decontamination of radioactively contaminated liquids |
| FR2669460A1 (en) * | 1990-11-16 | 1992-05-22 | Commissariat Energie Atomique | PROCESS FOR THE SEPARATION OF RUTHENIUM AND, POSSIBLY, CESIUM AND COBALT PRESENT IN AQUEOUS SOLUTION SUCH AS AN EFFLUENT FROM AN IRRADY FUEL PURIFYING PLANT. |
| GB2437864A (en) * | 2007-03-20 | 2007-11-07 | Malcolm Brody | Rapid and Selective Dissolution of Magnesium Alloy |
| GB2437864B (en) * | 2007-03-20 | 2008-05-14 | Malcolm Brody | The rapid and selective dissolution of magnesium alloy |
| GB2448208A (en) * | 2007-07-26 | 2008-10-08 | Malcolm Brody | The Removal of Radionuclide Contaminants after the Dissolutionof Magnox |
| GB2448208B (en) * | 2007-07-26 | 2009-08-12 | Malcolm Brody | Removal of radionuclide contaminants after the dissolution of magnox |
| RU2620259C1 (en) * | 2016-01-19 | 2017-05-24 | Общество С Ограниченной Ответственностью "Наука - Технологии - Производство" | Sorption material for the selective recovery of radionuclides strontium from solutions with high content of hardness solutions, the method of its production, its application and method of radionuclides extraction from structures with high content of hardness solutions |
| RU2698800C1 (en) * | 2018-07-02 | 2019-08-30 | Федеральное государственное унитарное предприятие "Предприятие по обращению с радиоактивными отходами "РосРАО" (ФГУП "РосРАО") | Method of producing strontium sorbent for solutions containing hardness salts |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US4269706A (en) | Method of decontaminating radioactive process waste waters | |
| US5217585A (en) | Transition metal decontamination process | |
| GB2073477A (en) | Waste treatment | |
| EP0649560B1 (en) | Process for selectively concentrating the radioactivity of thorium containing magnesium slag | |
| US2766204A (en) | Method for decontamination of radioactively contaminated aqueous solution | |
| JPS6116957B2 (en) | ||
| CA1154885A (en) | Method of decontaminating nuclear power plant waste liquids or removing radioactive components therefrom | |
| Liljenzin et al. | Reducing the long-term hazard of reactor waste through actinide removal and destruction in nuclear reactors | |
| GB1193596A (en) | Ion Exchange Process for Recovering Americium and Curium. | |
| Davies et al. | Waste treatment | |
| Bokelund et al. | Recovery and chemical purification of actinides at JRC, Karlsruhe | |
| US4582637A (en) | Reprocessing of irradiated nuclear fuel | |
| US3029200A (en) | Removal of radioactive ions from waters | |
| US3748104A (en) | Extraction of transplutonium elements from an aqueous solution containing rare earths and fission products | |
| GB1563967A (en) | Process for recovering molybdenum-99 from a matrix containing neutron irradiated fissionable materials and fisson products | |
| US3343914A (en) | Recovering strontium from reactor fuel reprocessing solutions as strontium sulfate | |
| JP3144160B2 (en) | How to remove radioactive materials in water | |
| US2865705A (en) | Improvement upon the carrier precipitation of plutonium ions from nitric acid solutions | |
| EP0619044B1 (en) | The treatment of solid organic wastes | |
| JPS62176913A (en) | Process for separation and recovery of cesium from treating liquid containing sodium salt | |
| Harjula et al. | Development of a selective cesium and strontium removal system for the JAERI Tokai-Mura site-laboratory tests | |
| US4756853A (en) | Process for the conversion into usable condition of actinide ions contained in the solid residue of a sulfate reprocessing process for organic, actinide-containing radioactive solid waste | |
| KR102100873B1 (en) | Decontamination method with reduced radioactive waste | |
| JPS63188796A (en) | Method of processing decontaminated waste liquor | |
| US3275421A (en) | Recovery of actinides from halite debris |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| WAP | Application withdrawn, taken to be withdrawn or refused ** after publication under section 16(1) |