CN116564572A - Method for removing alpha nuclide in target treatment waste liquid in medical isotope production - Google Patents
Method for removing alpha nuclide in target treatment waste liquid in medical isotope production Download PDFInfo
- Publication number
- CN116564572A CN116564572A CN202310597184.3A CN202310597184A CN116564572A CN 116564572 A CN116564572 A CN 116564572A CN 202310597184 A CN202310597184 A CN 202310597184A CN 116564572 A CN116564572 A CN 116564572A
- Authority
- CN
- China
- Prior art keywords
- waste liquid
- alpha
- nuclides
- treatment
- floccules
- 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.)
- Pending
Links
- 239000007788 liquid Substances 0.000 title claims abstract description 172
- 239000002699 waste material Substances 0.000 title claims abstract description 156
- 238000000034 method Methods 0.000 title claims abstract description 71
- 238000011282 treatment Methods 0.000 title claims abstract description 69
- 238000004519 manufacturing process Methods 0.000 title claims abstract description 22
- MUBZPKHOEPUJKR-UHFFFAOYSA-N Oxalic acid Chemical compound OC(=O)C(O)=O MUBZPKHOEPUJKR-UHFFFAOYSA-N 0.000 claims abstract description 45
- 238000000926 separation method Methods 0.000 claims abstract description 26
- 230000003647 oxidation Effects 0.000 claims abstract description 22
- 238000007254 oxidation reaction Methods 0.000 claims abstract description 22
- 239000002244 precipitate Substances 0.000 claims abstract description 16
- 239000006228 supernatant Substances 0.000 claims abstract description 15
- 235000006408 oxalic acid Nutrition 0.000 claims abstract description 13
- 239000001488 sodium phosphate Substances 0.000 claims abstract description 10
- 229910000162 sodium phosphate Inorganic materials 0.000 claims abstract description 10
- 229910052979 sodium sulfide Inorganic materials 0.000 claims abstract description 10
- GRVFOGOEDUUMBP-UHFFFAOYSA-N sodium sulfide (anhydrous) Chemical compound [Na+].[Na+].[S-2] GRVFOGOEDUUMBP-UHFFFAOYSA-N 0.000 claims abstract description 10
- RYFMWSXOAZQYPI-UHFFFAOYSA-K trisodium phosphate Chemical compound [Na+].[Na+].[Na+].[O-]P([O-])([O-])=O RYFMWSXOAZQYPI-UHFFFAOYSA-K 0.000 claims abstract description 10
- PMZURENOXWZQFD-UHFFFAOYSA-L Sodium Sulfate Chemical compound [Na+].[Na+].[O-]S([O-])(=O)=O PMZURENOXWZQFD-UHFFFAOYSA-L 0.000 claims abstract description 9
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- YVBOZGOAVJZITM-UHFFFAOYSA-P ammonium phosphomolybdate Chemical compound [NH4+].[NH4+].[NH4+].[NH4+].[O-]P([O-])=O.[O-][Mo]([O-])(=O)=O YVBOZGOAVJZITM-UHFFFAOYSA-P 0.000 claims abstract description 7
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- 235000003891 ferrous sulphate Nutrition 0.000 claims abstract description 4
- 239000011790 ferrous sulphate Substances 0.000 claims abstract description 4
- BAUYGSIQEAFULO-UHFFFAOYSA-L iron(2+) sulfate (anhydrous) Chemical compound [Fe+2].[O-]S([O-])(=O)=O BAUYGSIQEAFULO-UHFFFAOYSA-L 0.000 claims abstract description 4
- 229910000359 iron(II) sulfate Inorganic materials 0.000 claims abstract description 4
- KRKNYBCHXYNGOX-UHFFFAOYSA-N citric acid Chemical compound OC(=O)CC(O)(C(O)=O)CC(O)=O KRKNYBCHXYNGOX-UHFFFAOYSA-N 0.000 claims description 33
- 238000001556 precipitation Methods 0.000 claims description 21
- 230000000694 effects Effects 0.000 claims description 16
- 229910052776 Thorium Inorganic materials 0.000 claims description 14
- ZSLUVFAKFWKJRC-IGMARMGPSA-N 232Th Chemical compound [232Th] ZSLUVFAKFWKJRC-IGMARMGPSA-N 0.000 claims description 13
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- 150000001255 actinides Chemical class 0.000 claims description 10
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- HZEBHPIOVYHPMT-UHFFFAOYSA-N polonium atom Chemical compound [Po] HZEBHPIOVYHPMT-UHFFFAOYSA-N 0.000 claims description 10
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- 239000003054 catalyst Substances 0.000 claims description 9
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- NBIIXXVUZAFLBC-UHFFFAOYSA-K phosphate Chemical compound [O-]P([O-])([O-])=O NBIIXXVUZAFLBC-UHFFFAOYSA-K 0.000 claims description 4
- 239000010452 phosphate Substances 0.000 claims description 4
- 230000001105 regulatory effect Effects 0.000 claims description 4
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- IEECXTSVVFWGSE-UHFFFAOYSA-M iron(3+);oxygen(2-);hydroxide Chemical compound [OH-].[O-2].[Fe+3] IEECXTSVVFWGSE-UHFFFAOYSA-M 0.000 claims description 3
- 229910052745 lead Inorganic materials 0.000 claims description 3
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- 235000018660 ammonium molybdate Nutrition 0.000 claims 1
- 239000011609 ammonium molybdate Substances 0.000 claims 1
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- 230000002829 reductive effect Effects 0.000 abstract description 16
- 239000000047 product Substances 0.000 abstract description 6
- RUTXIHLAWFEWGM-UHFFFAOYSA-H iron(3+) sulfate Chemical compound [Fe+3].[Fe+3].[O-]S([O-])(=O)=O.[O-]S([O-])(=O)=O.[O-]S([O-])(=O)=O RUTXIHLAWFEWGM-UHFFFAOYSA-H 0.000 abstract description 2
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Classifications
-
- C—CHEMISTRY; METALLURGY
- C02—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F9/00—Multistage treatment of water, waste water or sewage
-
- 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
-
- C—CHEMISTRY; METALLURGY
- C02—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F1/00—Treatment of water, waste water, or sewage
- C02F1/52—Treatment of water, waste water, or sewage by flocculation or precipitation of suspended impurities
- C02F1/5236—Treatment of water, waste water, or sewage by flocculation or precipitation of suspended impurities using inorganic agents
-
- C—CHEMISTRY; METALLURGY
- C02—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F1/00—Treatment of water, waste water, or sewage
- C02F1/66—Treatment of water, waste water, or sewage by neutralisation; pH adjustment
-
- C—CHEMISTRY; METALLURGY
- C02—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F—TREATMENT OF WATER, WASTE WATER, SEWAGE, OR SLUDGE
- C02F1/00—Treatment of water, waste water, or sewage
- C02F1/72—Treatment of water, waste water, or sewage by oxidation
- C02F1/722—Oxidation by peroxides
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Abstract
The invention discloses a method for removing alpha nuclides in target treatment waste liquid in medical isotope production. Comprising the following steps: fenton oxidation treatment is carried out on target treatment waste liquid produced by medical isotopes to be treated; adding a carrier BaCl into the treated waste liquid 2 Adding one or more of oxalic acid, sodium phosphate, sodium sulfide, ammonium phosphomolybdate and sodium sulfate for coprecipitation, separating out precipitate, drying, barreling, welding and sealing or solidifying the precipitate; supernatant fluidAdjusting pH to neutral or weak acidity, adding ferric sulfate, ferrous sulfate flocculant or coagulant aid, separating out floccules, drying, barreling, welding and sealing or solidifying the floccules; the waste liquid after the separation of the floccules realizes non-alpha. The method can remove most of alpha nuclides in the waste liquid, so that the volume of the obtained alpha precipitate dried product is reduced by at least two orders of magnitude compared with the original product, thereby greatly reducing the quantity of alpha wastes and greatly reducing the final disposal cost.
Description
Technical Field
The invention belongs to the field of target treatment waste liquid in medical isotope production, and particularly relates to a method for removing alpha nuclides in target treatment waste liquid in medical isotope production.
Background
The key to nuclear medicine is the radiation beam and radioisotope, which currently has irreplaceable effects on radiopharmaceuticals for the treatment of diseases such as cancer. In diagnostic terms, radiopharmaceuticals may provide information on blood flow, function, and metabolism at the molecular level of the human body; in the aspect of treatment, the radiopharmaceuticals can be used for killing pathological tissues by utilizing the radioactivity of the radiopharmaceuticals, so that the accurate removal of micro focus is realized. The key to radioisotope pharmaceuticals is the radioisotope.
The current production methods of medical radioactive isotopes are three: 1. conventional methods for producing radioisotopes by a reactor, including the fission method and the neutron activation method; 2. separating and purifying the high-level radioactive waste liquid to obtain partial nuclides; 3. the target is generated by bombarding the isotope with an accelerator to generate nuclear reaction, and then the medical isotope is obtained through separation and purification. When the accelerator is used for producing isotopes, radioactive nuclides are generated in the target prepared by beam bombardment, radioactive waste liquid is generated in the process of separating the target nuclides, and sources of the waste liquid mainly comprise the waste liquid generated in the processes of dissolving, separating and the like of the target and the waste liquid generated in the processes of isotope preparation and use. These radioactive waste solutions need to be treated and disposed of to avoid harm to the environment and human body.
Patent CN 113436771B discloses a radioactive waste liquid treatment method, an outlet at the top of a secondary side of a steam generator is communicated with an inlet of a condenser, an outlet at the bottom of the secondary side of the steam generator is communicated with a secondary side inlet of a residual liquid collecting tank, a secondary side outlet of the residual liquid collecting tank is communicated with a compression packer, an outlet of the condenser is respectively communicated with an inlet of the waste liquid collecting tank and an inlet of a radioactive waste liquid tank, an outlet of the waste liquid collecting tank is communicated with a primary side inlet of the steam generator through a primary heater, a primary side outlet of the steam generator is communicated with a primary side inlet of a primary heat exchanger, and a primary side outlet of the primary heat exchanger is communicated with an inlet of the condenser; the outlet of the circulating fan is communicated with the inlet of the circulating fan through the primary side of the secondary heater and the residual liquid distillation collecting tank and the primary side of the secondary heat exchanger; the system can solve the problems of larger steam pressure fluctuation and more frequent temperature regulation in the waste liquid evaporation process. But is a treatment method mainly aiming at radioactive waste liquid of a nuclear power station, and the general characteristics of the radioactive waste liquid are that a large amount of radioactive waste liquid generated by the nuclear power station is treated by an evaporation method, so that the decontamination factor is high and the volume reduction effect is good. The method has the advantages of high construction cost, high operation and maintenance cost in the later period, low waste liquid amount, high specific activity and high content of volatile gas nuclides.
A radioactive waste treatment system as described in patent CN 111768885B, comprising: an ultrafiltration unit comprising an ultrafiltration membrane for removing colloidal state nuclides from the radioactive waste; the concentration unit comprises a reverse osmosis device, a water inlet of the reverse osmosis device is connected with a purified liquid outlet of the ultrafiltration unit, and the reverse osmosis device receives the purified liquid of the ultrafiltration unit so as to perform reverse osmosis concentration on the purified liquid of the ultrafiltration unit; the ion exchange unit comprises an ion exchange bed, the water inlet of the ion exchange unit is connected with the concentrated solution outlet of the reverse osmosis device, and the ion exchange unit receives the concentrated solution of the concentration unit so as to extract the radionuclide enriched in the concentrated solution of the concentration unit to a solid phase. The radioactive waste treatment system is mainly aimed at radioactive waste generated by nuclear power plant and is mainly characterized by low radioactivity (its specific activity is generally 10 5 ～10 7 Bq/L), radioactive species containedLess% 137 Cs、 89 Sr、 90 Sr、 58 Co、 110m Ag、 3 H, etc.), the liquid quantity is large (several tens of m 3 And a), performing ultrafiltration reverse osmosis and ion exchange to generate concentrated solution and a large amount of secondary radioactive waste, discharging the liquid treatment end result to reach the coastal discharge limit requirement, and finally performing solidification treatment on the concentrated solution and the secondary solid waste to achieve the purpose of volume reduction. The method is not suitable for the annual low yield (0.5-1 m) of waste liquid 3 ) Has relatively high specific activity (10) 11 ～10 12 Bq/L), radionuclide species (half-life of about 300 or more in the range of 1 hour to several decades), on the one hand, the investment cost is excessive, on the other hand, ultrafiltration, reverse osmosis and ion exchange cannot completely remove alpha nuclides in the waste liquid, the alpha waste amount cannot be reduced, and the waste liquid is not suitable for medium-high radioactive waste liquid with complex nuclides.
The high-power proton beam current is generated based on the accelerator, and the isotope production target is bombarded, so that the medical radioisotope is generated with high yield and high efficiency, and then is efficiently separated and purified, thereby providing good support and guarantee for preparing the isotope targeted therapeutic drug. The radioactive waste, in particular radioactive waste liquid, produced in the production of medical isotopes is different from the waste liquid produced by traditional nuclear medicine, nuclear technology applications and nuclear power. The radioactive waste liquid source produced in the separation process has specificity, and the activity of the waste liquid after separation is about 10 10 -10 12 Bq/L, the radioactive element variety contained after cooling for 1 year is up to 50, the nuclide variety is up to 300, and the annual yield is more than 1m 3 The half-life of these nuclides ranges from minutes to thousands of years, with 37.70% half-life less than 1 hour, about 14.98% half-life greater than or equal to 1 hour and less than 1 day, about 36.44% half-life greater than or equal to 1 day less than 30 days, about 10.21% half-life greater than or equal to 30 days less than 1 year, and about 0.67% half-life greater than 1 year; the waste liquid has less kinds of non-radioactive impurities, low concentration and low acidity, and the generated medium-high level waste liquid has the following characteristics relative to the waste liquid generated by the post-treatment of spent fuel from the viewpoint of waste liquid treatmentProprietary characteristics: alpha nuclides contained in the waste liquid are different, alpha nuclides in the waste liquid generated by post-treatment of spent fuel are mainly actinides, non-actinides such as Ra, po and Pb are also contained in the waste liquid generated by medical isotope production except actinides, the service lives of other alpha nuclides in the waste liquid after dissolving targets are shorter except long-life nuclides such as Th and Pa (T1/2 is more than 104 a), and compared with the high-emission waste liquid generated by post-treatment of spent fuel, the activity concentration of the alpha nuclides is lower by 1-2 orders of magnitude; the total activity concentration of the waste liquid generated by dissolving the target is 3-4 orders of magnitude lower than that of the high-level waste liquid generated by post-treatment of spent fuel.
Due to these characteristics of nuclides, the waste liquid cannot be discharged after temporary decay as in common nuclear medicine and nuclear technology utilization projects. The radioactive waste liquid generated by nuclear power plants is generally 10 5 ～10 7 The Bq/L level generally adopts flocculation, filtration, ion exchange or filtration evaporation and other process routes, and the process mainly aims at radioactive waste liquid with less nuclide species, low activity and larger annual yield, and the construction and operation cost of treatment facilities are larger.
Disclosure of Invention
The invention aims to provide a method for removing alpha nuclides in target treatment waste liquid (waste liquid remained after target nuclides are separated after target dissolution) in medical isotope production. The method adopts a separation and preparation method to treat the source item characteristics in the carding waste liquid, and adopts a process of 'special coprecipitation and flocculation precipitation' to separate alpha parent nuclide and part of alpha nuclide in the waste liquid, so that after the parent nuclide is separated, the half life period of the daughter nuclide is shorter, and the daughter nuclide can decay rapidly, thereby leading the waste liquid to be 'non-alpha', realizing the separation of the alpha nuclide, further reducing the quantity of alpha waste and achieving the purpose of volume reduction.
The method for removing alpha nuclides in target treatment waste liquid in medical isotope production provided by the invention comprises the following steps:
1) Fenton oxidation treatment is carried out on target treatment waste liquid produced by medical isotopes to be treated;
2) Adding a carrier BaCl into the waste liquid after Fenton oxidation treatment 2 Adding oxalic acid, sodium phosphate, sodium sulfide and phosphomolybdic acidCoprecipitation is carried out on one or a combination of a plurality of ammonium and sodium sulfate, so that most of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra) and thorium (Th) in the waste liquid are deposited from the waste liquid in a precipitation form, the precipitate is separated, dried, barreled and welded and sealed or the precipitate is solidified;
3) Regulating the pH value of the supernatant fluid after precipitation and separation in the step 2) to be neutral or weak acid, adding a flocculating agent or coagulant aid, flocculating to separate other easily-hydrolyzed nuclides, easily-adsorbed nuclides and residual particles in the waste liquid along with the settling of the floccules, separating the floccules, drying, barreling, welding and sealing or solidifying the floccules; the waste liquid after the separation of the floccules realizes non-alpha.
In the step 1), the target treatment waste liquid is the radioactive waste liquid which is remained after the target is dissolved and the target nuclide is separated after being bombarded;
after being bombarded by proton beams, the metal target is dissolved, target nuclides are separated through a series of separation processes, and the residual radionuclides are basically fully contained in waste liquid, namely target treatment waste liquid produced by medical isotopes to be treated;
in one embodiment of the invention, the protons bombard the thorium target to produce alpha-medical isotopes 225 Ac and 223 ra), the metal Th-232 target is dissolved after being bombarded by proton beams, the target nuclide is separated through a series of separation processes, the generated residual radionuclide is basically fully contained in the waste liquid, and the obtained waste liquid is the target treatment waste liquid produced by the medical isotope to be treated.
The activity of the waste liquid is 10 10 ～10 12 Bq/L, up to 300 species of nuclides, and alpha radionuclides contained in the Bq/L contain non-actinides such as Ra, po and Pb besides actinides.
Oxidative destruction of citric acid or its complexes by Fenton oxidation treatment, because citric acid in the waste liquid may form complexes with certain radionuclides (especially alpha nuclides);
the Fenton oxidation treatment adopts H 2 O 2 As an oxidant, a trace amount of catalyst is added,
the catalyst is an oxide of Fe, cu or Mn, such as Fe 3 O 4 ，MnO 2 CuO, etc.;
h is added to each liter of waste liquid according to the amount of citric acid contained in the waste liquid (added in the separation process) 2 O 2 The amount of (2) is 0.01mol to 0.05mol, specifically may be 0.03mol;
based on H 2 O 2 The molar amount of the catalyst is H 2 O 2 From 0.1% to 20%.
The Fenton oxidation process has the advantages that the oxidant is H 2 O 2 The reduction products are water and citrate ions, and besides adding a trace amount of catalyst, no extra salt impurities are added into the waste liquid, and the key nuclides in the waste liquid can be oxidized into common stable states while destroying the citric acid.
In the above method step 2), based on the kind and total amount of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra), thorium (Th) and the like to be removed in the waste liquid, one or a combination of several of oxalic acid, sodium phosphate, sodium sulfide, ammonium phosphomolybdate, sodium sulfate and the like is selected,
based on the total activity of the nuclides, at least 0.01mol of BaCl is needed per liter of waste liquid 2 0.01mol of oxalic acid, sodium phosphate, sodium sulfide, ammonium phosphomolybdate and sodium sulfate, wherein when the combination of the two or more substances is selected, the amount of each substance is 0.01mol respectively;
the oxalic acid can be solid oxalic acid or saturated oxalic acid solution;
most of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra) and thorium (Th) in the wastewater treated by coprecipitation are deposited from the wastewater in a precipitated form, and the decontamination factor of the nuclides can reach more than hundred times.
In the above method step 3), the addition of the flocculant is selected according to the characteristics of the nuclide,
the flocculant can be one or a combination of a plurality of ferric salt, aluminum salt, phosphate and oxalate,
specifically, the flocculant is ferrous sulfate or ferric hydroxide;
adding at least 0.01mol of flocculant per liter of waste liquid based on the volume of the waste liquid to be treated;
flocculating agent is added to remove part of actinides, and the decontamination factor reaches more than 1000.
According to the method, firstly, radioactive waste liquid generated in the separation process of target production medical isotopes is generated according to the bombardment of isotopes by an accelerator, the types of nuclides in the waste liquid are multiple and complex, the waste liquid contains alpha nuclides with higher toxicity, the alpha nuclides need to be subjected to the de-alpha conversion, and the final yield of alpha waste is reduced.
The scheme of the invention is different from the original process scheme of flocculation precipitation, the source item characteristics in the waste liquid are treated by adopting a separation preparation method, alpha parent nuclide and partial alpha nuclide in the waste liquid are separated by adopting a process of Fenton oxidation, special coprecipitation and flocculation precipitation, and after the parent nuclide is separated, the half life period of the daughter nuclide is shorter, and the daughter nuclide can decay quickly, so that the waste liquid is non-alpha, the separation of alpha nuclide is achieved, the quantity of alpha waste is further reduced, and the purpose of volume reduction is achieved. After treatment, the composition components of nuclides in the waste liquid can be changed, and necessary pretreatment is provided for the subsequent treatment process.
The invention belongs to a method and a process for treating radioactive waste liquid generated by producing medical isotopes based on an accelerator, which are mainly applied to the treatment of the radioactive waste liquid generated by producing medical isotopes. The method is mainly used for removing most of alpha nuclides in the waste liquid by providing a special coprecipitation and flocculation precipitation method aiming at the special source item, so that the volume of the obtained alpha precipitation dried matter is reduced by at least two orders of magnitude compared with the original one, thereby greatly reducing the quantity of alpha waste and greatly reducing the final disposal cost. The nuclide components in the treated waste liquid are greatly changed, and the subsequent purification treatment is utilized, so that the sustainable production of medical isotopes is finally ensured.
The overall treatment route is as shown in fig. 1, and the overall system method is the combined use of the waste liquid treatment methods of Fenton oxidation, special coprecipitation and flocculation precipitation. In the process of the waste liquid treatment process route, the implementation technology is simple and effective, the equipment is mature and reliable, the alpha waste is reduced in volume and quantity, and meanwhile, the waste treatment and disposal are facilitated without adding basic principles such as new waste. The main purpose of the various processes for simultaneously carrying out the treatment of the liquid alpha waste is to remove alpha pollutants in the waste liquid or separate alpha nuclides into a small volume so as to facilitate subsequent preparation and disposal, thereby realizing the idea of minimizing the waste, and having convenient and simple operation without increasing the construction cost. The treatment method mainly comprises the separation and sediment preparation of Fenton oxidation, special coprecipitation and flocculation precipitation.
In the "flocculation + precipitation" treatment of radioactive waste water, flocculation is usually carried out first, followed by co-precipitation. However, in this case, since the source water is acidic, a method of coprecipitating and then flocculating is adopted in order to avoid repeated adjustment of acidity and alkalinity, simplification of the process, and increase of secondary waste amount due to introduction of excessive salt into the wastewater. Firstly, adding a carrier BaCl into the wastewater 2 And then adding different combinations of solid oxalic acid (or saturated oxalic acid solution), sodium phosphate, sodium sulfide, ammonium phosphomolybdate, sodium sulfate and the like to carry out coprecipitation. From the description of the physicochemical properties of the key nuclides described above, it is known that most of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra) and thorium (Th) in wastewater will be deposited from the waste liquid in the form of precipitate by coprecipitation of oxalic acid, phosphate and sulfide. The deposited precipitate is separated and dried, and then barreled and welded and sealed (the precipitate can be solidified).
Adjusting pH value of supernatant to neutral or weak acidity, adding flocculant or coagulant aid such as ferric salt and aluminum salt or phosphate and oxalate, flocculating, and settling together with floccules other easily hydrolyzed nuclides, easily adsorbed nuclides and residual particulate matters in waste liquid, thereby realizing separation from water body. The floccules are separated out and dried, and then barreled, welded and sealed (the floccules can be solidified). So far, most of the extremely toxic group nuclides and alpha nuclides in the wastewater are purified and removed, the rest alpha nuclides are separated from the parent, the half life period is short, decay is very short, the alpha radioactivity level of the wastewater is greatly reduced, the non-alpha conversion of the wastewater is realized, the volume of alpha wastes is obviously reduced, and the rest waste liquid is more beneficial to the next purification treatment due to the change of nuclide components.
The process flow has the advantages that an innovative treatment idea for separating alpha nuclides from other nuclides is provided, and the production of alpha wastes is reduced; the wastewater from which alpha nuclides are separated can be treated by adopting a mature technology; the secondary waste is less.
The invention takes the main principle of reducing the volume of waste and removing alpha radionuclides from the waste, separates alpha parent nuclides and part of alpha nuclides from the waste liquid by adopting a special multiple combined process, thereby reducing the quantity of the alpha waste, achieving the purpose of reducing the volume of the alpha waste, providing necessary pretreatment for the subsequent treatment process, changing the composition components of nuclides in the waste liquid and facilitating the subsequent treatment.
Drawings
FIG. 1 is a schematic flow diagram of the present invention, wherein 0-wastewater; 1-Fenton oxidation; 2-coprecipitation; 3-flocculating and settling; 4-the treated supernatant; 5-solids removal servicing.
FIG. 2 is a treatment plant kit of the present invention wherein 0-wastewater; 101-sampling groove; 102-Fenton oxidation reagent adding box; 103-a waste liquid conveying pump; 2-coprecipitation; 201-a coprecipitation reagent addition box; 202-a mixer; 203-solids removal servicing; 3-flocculating and settling; 301-a flocculation reagent adding box; 302-a mixer; 303-solids removal servicing; 4-supernatant after treatment.
Detailed Description
The following detailed description of the invention is provided in connection with the accompanying drawings that are presented to illustrate the invention and not to limit the scope thereof. The examples provided below are intended as guidelines for further modifications by one of ordinary skill in the art and are not to be construed as limiting the invention in any way.
The experimental methods in the following examples, unless otherwise specified, are conventional methods, and are carried out according to techniques or conditions described in the literature in the field or according to the product specifications. Materials, reagents and the like used in the examples described below are commercially available unless otherwise specified.
The invention belongs to a method and a process for treating radioactive waste liquid generated by producing medical isotopes based on an accelerator, which are mainly applied to the treatment of the radioactive waste liquid generated by producing medical isotopes. The method is mainly used for removing most of alpha nuclides in the waste liquid by providing a special coprecipitation and flocculation precipitation method aiming at the special source item, so that the volume of the obtained alpha precipitation dried matter is reduced by at least two orders of magnitude compared with the original one, thereby greatly reducing the quantity of alpha waste and greatly reducing the final disposal cost. The nuclide components in the treated waste liquid are greatly changed, and the subsequent purification treatment is utilized, so that the sustainable production of medical isotopes is finally ensured.
And (3) generating a waste liquid source item and impurity analysis based on an accelerator production isotope separation process:
alpha-medical isotope production by proton bombardment of thorium target 225 Ac and 223 ra), the isotope production target (thorium target) is bombarded with a proton beam, and the energetic protons undergo spallation reaction with the target, producing about several hundred fission products. After the thorium target is irradiated for 10 days and cooled for 10 hours, the thorium target is sent to a separation hot room by an automatic transfer device, then the target is dissolved by a wet method for separation, and aiming at the current domestic requirements, the target medical isotopes of the patent are Ra-223 and Ac-225, and other produced nuclides are impurity nuclides and enter radioactive waste liquid. The main elements include about 57 kinds of elements, such As Ac (actinium), ag (silver), ar (argon), as (arsenic), at (astatine), ba (barium), be (beryllium), bi (bismuth), br (bromine), C (carbon), cd (cadmium), ce (cerium), co (cobalt), cs (cesium), dy (dysprosium), eu (europium), fe (iron), fr (francium), ga (gallium), gd (gadolinium), ge (germanium), H (hydrogen), hg (mercury), I (iodine), in (indium), K (potassium), kr (krypton), la (lanthanum), mo (molybdenum), nb (niobium), nd (neodymium), ni (nickel), pa (protactinium), pb (lead), pd (palladium), pm (promethium), po (polonium), pr (praseodymium), ra (radium), rb (rhodium), rn (rhodium (ruthenium), sb (antimony), se (samarium), sn (tin), sr (strontium), tb (terbium), tc (Te), te (radon), tl (neodymium (Tl), tl (yttrium (Tl) (Y), zr (zirconium) has an isotope content of about 300 or more.
Low level waste liquid: the low-level waste liquid mainly comes from a separation hot chamber, separation equipment and pipeline cleaningThe nuclides contained in the laboratory operation waste liquid are all nuclides or partial nuclides of the separated medium-level waste liquid, the exact type is variable, and the total specific activity is not more than 10 6 Bq/L, and contains a small amount of acidic substances such as sulfuric acid, hydrofluoric acid, citric acid, etc. The radioactive waste liquid is collected to a low-level waste liquid storage tank.
Waste liquid from medium-level discharge: the radioactive waste liquid is mainly the radioactive waste liquid produced in the separation process, and is produced in the processes of dissolving, separating, concentrating and purifying thorium target, and the radioactivity specific activity of the waste liquid after extracting target nuclide is about 10 11 Bq/L contains sulfuric acid, hydrofluoric acid, citric acid and other acidic substances, and its chemical components contain nitric acid HNO per liter 3 About 0.78mol/L, ammonium sulfate (NH) 4 ) 2 SO 4 About 0.16mol/L and about 0.028mol/L. The waste liquid contains a great amount of actinides, lanthanoids and other nuclides. The radioactive waste liquid is collected to a temporary storage tank for the medium-level waste liquid.
The overall treatment route is as shown in fig. 1, and the overall system method is the combined use of the waste liquid treatment methods of Fenton oxidation, special coprecipitation and flocculation precipitation. In the process of the waste liquid treatment process route, the implementation technology is simple and effective, the equipment is mature and reliable, the waste volume is reduced and the waste is reduced, and meanwhile, the further treatment and disposal of the waste are facilitated without adding basic principles such as new waste. The main purpose of the various processes for simultaneously carrying out the treatment of the liquid alpha waste is to remove alpha pollutants in the waste liquid or separate alpha nuclides into a small volume so as to facilitate subsequent preparation and disposal, thereby realizing the idea of minimizing the waste, and having convenient and simple operation without increasing the construction cost. The treatment method mainly comprises the separation and sediment preparation of Fenton oxidation, special coprecipitation and flocculation precipitation.
Separation and preparation system:
the important alpha nuclides that need to be of interest in the waste solution to solubilize the target are shown in table 1. In Table 1, the remaining alpha nuclides are mostly daughter of several parent nuclei, except for those of parent nuclei Th-228, th-229, th-230, pa-231. The nuclides of the parent nuclides and the polar group are separated from the waste liquid, so that most of the alpha nuclides remained in the waste liquid can be decayed in a short time because of the short half-life period (the nuclides with longer half-life periods such as Ra-224, po-208, po-209, ac-225 and the like are nuclides of the same element as the nuclides in the polar group and are removed when the nuclides in the polar group are separated), thus the high-level waste liquid can be degraded, namely non-alpha waste liquid is realized, and the rest waste liquid becomes non-alpha waste liquid, so that the treatment is easy.
The invention provides an innovative treatment idea for separating alpha nuclides from other nuclides aiming at the radioactive waste liquid generated by accelerator production of medical isotopes, mainly aiming at special radioactive waste liquid generated by production of Ra-223 and Ac-225, which is beneficial to reducing the generation amount of alpha wastes, can meet the requirements of continuous production in a field, simplify construction cost and operation cost, facilitate operation and maintenance, reduce the generation of secondary wastes, and simultaneously meet the requirements of safe discharge or off-site safe transportation of wastes, and ensure the continuous production of medical isotopes.
The whole treatment route is as shown in figure 1, the radioactive wastewater 0 enters a wastewater collector 1, and a Fenton oxidation process is adopted in the wastewater collector; performing 2 coprecipitation technology on the waste liquid subjected to primary treatment; and 5, separating the deposited precipitate, drying, and then barreling, welding and sealing (the precipitate can be solidified). The supernatant fluid enters the next step 3 flocculation precipitation process, thereby realizing separation from the water body, 5 separating out the floccules for drying, and then barreling, welding and sealing (the floccules can be solidified). And finally discharging the treated supernatant into a 4 temporary storage tank for further treatment.
As shown in fig. 2, the 0 radioactive waste liquid is collected to the 1 waste liquid temporary storage tank through a pipeline, the physical and chemical properties of the waste liquid are monitored through the 101 sampling tank, and the Fenton oxidation process is carried out in the 1 according to the monitoring result, so that citric acid or a complex thereof is damaged by oxidation, and the citric acid in the waste liquid can possibly form a complex with certain radionuclides (especially alpha nuclides) to influence the coprecipitation and flocculation effects. By Fenton oxidationThe process has the advantages that the oxidant is H 2 O 2 The reduction product is water, and besides the addition of a trace amount of catalyst, no extra salt impurities are added into the waste liquid, and the key nuclides in the waste liquid can be oxidized into common stable states while the citric acid is destroyed. After Fenton technology, firstly adding carrier BaCl into the wastewater 2 Then 201 adding solid oxalic acid (or saturated oxalic acid solution), sodium phosphate, sodium sulfide or ammonium phosphomolybdate, sodium sulfate and the like into an adding box, mixing by a mixer, and inputting into a 2 coprecipitation box for coprecipitation. By co-precipitation, most of actinium (Ac), bismuth (Bi), polonium (Pa), polonium (Po), radium (Ra) and thorium (Th) in the wastewater will be deposited from the waste liquid in the form of precipitate, 203 the deposited precipitate is separated and dried, and then the precipitate can be sealed by barrel welding (the precipitate can also be solidified). After coprecipitation process, the flocculation precipitation process is carried out again, the pH value is regulated to be neutral or weak acid through a 301 adding box, then flocculating agent is added for flocculation, the adding of the flocculating agent is selected according to the characteristics of nuclides, such as ferrous sulfate, ferric hydroxide and the like, part of actinides nuclides are removed, the decontamination factor reaches more than 1000, the waste liquid is conveyed to a 3 flocculation precipitation box after passing through a 302 mixer, colloid and dispersion particles of other suspended matters in the waste liquid are settled together with the flocculating body under the related action of molecular force through flocculation, the flocculating body is separated out for drying through 303, and then barreled welding and sealing are carried out (the flocculating body can be solidified). So far, most of the extremely toxic group nuclides and alpha nuclides in the wastewater are purified and removed, and the rest alpha nuclides are separated from the parent, so that the half life is very short, the decay is very short, the alpha radioactivity level of the wastewater is greatly reduced, the non-alpha conversion of the wastewater is realized, and the volume of alpha waste is remarkably reduced. Finally, the supernatant is transported to a waste liquid temporary storage tank for subsequent purification treatment
Examples
Taking radioactive waste liquid generated by accelerator production of medical isotopes (Ra-223 and Ac-225) as an example, 100MeV500 mu A protons bombard Th-232 targets for 10 days, the mass of the targets is about 166g, the targets are cooled for 10 days and separated, and part of target nuclides (Ra-223 and Ac-225) are separated, and the rest other radionuclides are leftThe elements contained In the waste liquid are about 57 kinds, ac (actinium), ag (silver), ar (argon), as (arsenic), at (astatine), ba (barium), be (beryllium), bi (bismuth), br (bromine), C (carbon), cd (cadmium), ce (cerium), co (cobalt), cs (cesium), dy (dysprosium), eu (europium), fe (iron), fr (francium), ga (gallium), gd (gadolinium), ge (germanium), H (hydrogen), hg (mercury), I (iodine), in (indium), K (potassium), kr (krypton), la (lanthanum), mo (molybdenum), nb (niobium), nd (neodymium), ni (nickel), pa (protactinium), pb (lead), pd (palladium), pm (Pm), po (polonium), pr (praseodymium), radium (radium), rb (rubidium), rn (radon), rh (rhodium), ru (ruthenium), sb (antimony), se (selenium), sm (samarium), sn (tin), sr (Tb), tb (terbium), tc (Te), tl (Tl) and Tl (Tl) of radioactive liquid, Y (yttrium) and Zr (zirconium) with isotope content of about 300, the final content of the separated waste liquid is about 42L, and the waste liquid contains sulfuric acid, hydrofluoric acid, citric acid and other acidic substances, and the chemical components of the waste liquid contain nitric acid HNO per liter 3 About 0.78mol/L, ammonium sulfate (NH) 4 ) 2 SO 4 About 0.16mol/L and about 0.028mol/L. These waste liquids 42L were collected in a temporary storage container 1, and were subjected to sampling analysis by 101 to analyze the components and physicochemical properties of the initial waste liquid, and 1.3mol of H was added by 102 2 O 2 (while adding 0.26mol MnO through 102) 2 As catalyst) is subjected to Fenton oxidation. After sufficient oxidation (60 minutes), the mixture was pumped to 202 a mixer by 103 waste liquid transfer pump, and simultaneously co-precipitant was added in 201 reagent addition tank, and carrier 1mol BaCl was added 2 Mixing solid oxalic acid, sodium phosphate and sodium sulfide in 0.42mol each by a 202 mixer, inputting into a 2 coprecipitation box for coprecipitation, separating deposited precipitate, drying 203, and barreling for solidification. After the coprecipitation process, sampling and monitoring are carried out on the supernatant after the preliminary treatment through 205, and the supernatant is used for the next step of acid-base adjustment and the addition amount of the flocculating agent, and then the flocculation precipitation process is carried out. The waste liquid is sent to a 302 mixer through a 204 delivery pump, the pH value is regulated to be neutral or weak acid (the pH value is between 5.5 and 7) through a 301 adding box, 0.42mol ferric sulfate flocculating agent is added for flocculation, the colloid and the dispersed particles of other suspended matters in the waste liquid are settled together with the flocculating body under the related action of molecular force through a 3 flocculation precipitation box, and the flocculating body is separated out and enters into the mixer through a 303Drying, and barreling to wait for subsequent curing treatment. By coprecipitation and flocculation, most of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra) and thorium (Th) in the wastewater are deposited from the waste liquid in a precipitated form, and the decontamination factor reaches more than 1000. The supernatant in the 3 settling tank is communicated with a 304 centrifugal machine, the effect of the 3 settling tank is that the high-speed rotation generates strong centrifugal force, the settling velocity of particles in liquid is accelerated, settled substances in the liquid are separated, then the separated wet solid waste liquid is separated out and enters 303 to be dried and barreled, the liquid from the centrifugal machine enters 4 to be processed supernatant temporary storage tank, the processed supernatant is monitored through 401, and the subsequent other processing is waited. The vast majority of the extremely toxic group nuclides and alpha nuclides (in table 1) in the wastewater are purified and removed, and the rest alpha nuclides are separated from the parent, so that the half life period is very short, decay is very short, the alpha radioactivity level of the wastewater is greatly reduced, the non-alpha conversion of the wastewater is realized, and the volume of alpha waste is remarkably reduced. And finally, conveying the supernatant to a waste liquid temporary storage tank for subsequent purification treatment.
The present invention is described in detail above. It will be apparent to those skilled in the art that the present invention can be practiced in a wide range of equivalent parameters, concentrations, and conditions without departing from the spirit and scope of the invention and without undue experimentation. While the invention has been described with respect to specific embodiments, it will be appreciated that the invention may be further modified. In general, this application is intended to cover any variations, uses, or adaptations of the invention following, in general, the principles of the invention and including such departures from the present disclosure as come within known or customary practice within the art to which the invention pertains.
Claims (6)
1. A method for removing alpha nuclides in target treatment waste liquid of medical isotope production comprises the following steps:
1) Fenton oxidation treatment is carried out on target treatment waste liquid produced by medical isotopes to be treated;
2) Adding a carrier BaCl into the waste liquid after Fenton oxidation treatment 2 Then adding oxalic acid, sodium phosphate, sodium sulfide and phosphorusCo-precipitation is carried out on one or a combination of more of ammonium molybdate and sodium sulfate, the precipitate is separated out, and the solution is dried, barreled, welded and sealed or solidified;
3) Regulating the pH value of the supernatant fluid after precipitation and separation in the step 2) to be neutral or weak acid, adding a flocculating agent or a coagulant aid, separating out floccules, drying, barreling, welding and sealing or solidifying the floccules; the waste liquid after the separation of the floccules realizes non-alpha.
2. The method according to claim 1, characterized in that: the activity of the waste liquid is 10 10 ～10 12 Bq/L, the species of nuclides reaches 300, and alpha radionuclides contained in the Bq/L contain Ra, po and Pb non-actinides besides actinides.
3. The method according to claim 1 or 2, characterized in that: oxidative destruction of citric acid or its complex by Fenton oxidation treatment;
the Fenton oxidation treatment adopts H 2 O 2 As an oxidant, a trace amount of catalyst is added,
the catalyst is an oxide of Fe, cu or Mn;
adding H into each liter of waste liquid 2 O 2 The amount of (2) is 0.01mol to 0.05mol;
based on H 2 O 2 The molar amount of the catalyst is H 2 O 2 From 0.1% to 20%.
4. A method according to any one of claims 1-3, characterized in that: in the step 2), one or a combination of a plurality of oxalic acid, sodium phosphate, sodium sulfide, ammonium phosphomolybdate and sodium sulfate is selected based on the types and the total amount of actinium (Ac), bismuth (Bi), protactinium (Pa), polonium (Po), radium (Ra) and thorium (Th) in the waste liquid,
at least 0.01mol of BaCl is added into each liter of waste liquid 2 0.01mol of oxalic acid, sodium phosphate, sodium sulfide, ammonium phosphomolybdate and sodium sulfate.
5. The method according to any one of claims 1-4, wherein: in the step 3), the flocculant is one or a combination of a plurality of ferric salt, aluminum salt, phosphate and oxalate.
6. The method according to claim 5, wherein: the flocculant is ferrous sulfate or ferric hydroxide;
at least 0.01mol of flocculant per liter of waste liquid is added based on the volume of waste liquid to be treated.
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