CN116356145A - Method for extracting technetium from nuclear fuel post-treatment waste liquid by extraction method - Google Patents
Method for extracting technetium from nuclear fuel post-treatment waste liquid by extraction method Download PDFInfo
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- 229910052713 technetium Inorganic materials 0.000 title claims abstract description 124
- GKLVYJBZJHMRIY-UHFFFAOYSA-N technetium atom Chemical compound [Tc] GKLVYJBZJHMRIY-UHFFFAOYSA-N 0.000 title claims abstract description 124
- 238000000605 extraction Methods 0.000 title claims abstract description 72
- 238000000034 method Methods 0.000 title claims abstract description 67
- 239000007788 liquid Substances 0.000 title claims abstract description 60
- 238000011282 treatment Methods 0.000 title claims abstract description 52
- 239000002699 waste material Substances 0.000 title claims abstract description 44
- 239000003758 nuclear fuel Substances 0.000 title claims abstract description 23
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims abstract description 52
- 229910052726 zirconium Inorganic materials 0.000 claims abstract description 52
- 239000012074 organic phase Substances 0.000 claims abstract description 42
- 238000000658 coextraction Methods 0.000 claims abstract description 36
- CLLJENYXWWLLIS-UHFFFAOYSA-N [Zr].[Tc] Chemical compound [Zr].[Tc] CLLJENYXWWLLIS-UHFFFAOYSA-N 0.000 claims abstract description 27
- 239000003350 kerosene Substances 0.000 claims abstract description 16
- 230000000536 complexating effect Effects 0.000 claims abstract description 11
- 239000003795 chemical substances by application Substances 0.000 claims description 35
- 239000000047 product Substances 0.000 claims description 17
- 238000005406 washing Methods 0.000 claims description 14
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims description 12
- 229910017604 nitric acid Inorganic materials 0.000 claims description 12
- OAKJQQAXSVQMHS-UHFFFAOYSA-N hydrazine group Chemical group NN OAKJQQAXSVQMHS-UHFFFAOYSA-N 0.000 claims description 10
- HDZGCSFEDULWCS-UHFFFAOYSA-N monomethylhydrazine Chemical compound CNN HDZGCSFEDULWCS-UHFFFAOYSA-N 0.000 claims description 10
- OERNJTNJEZOPIA-UHFFFAOYSA-N zirconium nitrate Chemical compound [Zr+4].[O-][N+]([O-])=O.[O-][N+]([O-])=O.[O-][N+]([O-])=O.[O-][N+]([O-])=O OERNJTNJEZOPIA-UHFFFAOYSA-N 0.000 claims description 10
- 239000003599 detergent Substances 0.000 claims description 7
- 238000012958 reprocessing Methods 0.000 claims description 7
- 239000003153 chemical reaction reagent Substances 0.000 claims description 6
- 238000001704 evaporation Methods 0.000 claims description 6
- 239000002253 acid Substances 0.000 claims description 5
- 229920000962 poly(amidoamine) Polymers 0.000 claims description 3
- 239000013589 supplement Substances 0.000 claims description 3
- 230000008020 evaporation Effects 0.000 claims 1
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 abstract description 51
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 abstract description 16
- 239000002904 solvent Substances 0.000 abstract description 11
- 239000002901 radioactive waste Substances 0.000 abstract description 6
- 229910021645 metal ion Inorganic materials 0.000 abstract description 5
- 238000007711 solidification Methods 0.000 abstract description 5
- 230000008023 solidification Effects 0.000 abstract description 5
- 239000011521 glass Substances 0.000 abstract description 4
- 230000007774 longterm Effects 0.000 abstract description 4
- 238000011084 recovery Methods 0.000 description 7
- -1 N235 Chemical compound 0.000 description 6
- GKLVYJBZJHMRIY-OUBTZVSYSA-N Technetium-99 Chemical compound [99Tc] GKLVYJBZJHMRIY-OUBTZVSYSA-N 0.000 description 5
- WJWSFWHDKPKKES-UHFFFAOYSA-N plutonium uranium Chemical compound [U].[Pu] WJWSFWHDKPKKES-UHFFFAOYSA-N 0.000 description 5
- 229910052778 Plutonium Inorganic materials 0.000 description 4
- 150000002500 ions Chemical class 0.000 description 4
- 239000000203 mixture Substances 0.000 description 4
- 238000005192 partition Methods 0.000 description 4
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 4
- 230000000694 effects Effects 0.000 description 3
- 238000001179 sorption measurement Methods 0.000 description 3
- ZWEHNKRNPOVVGH-UHFFFAOYSA-N 2-Butanone Chemical compound CCC(C)=O ZWEHNKRNPOVVGH-UHFFFAOYSA-N 0.000 description 2
- LFQSCWFLJHTTHZ-UHFFFAOYSA-N Ethanol Chemical group CCO LFQSCWFLJHTTHZ-UHFFFAOYSA-N 0.000 description 2
- JUJWROOIHBZHMG-UHFFFAOYSA-N Pyridine Chemical compound C1=CC=NC=C1 JUJWROOIHBZHMG-UHFFFAOYSA-N 0.000 description 2
- 238000006243 chemical reaction Methods 0.000 description 2
- 238000010668 complexation reaction Methods 0.000 description 2
- 230000007547 defect Effects 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 229910052739 hydrogen Inorganic materials 0.000 description 2
- 239000001257 hydrogen Substances 0.000 description 2
- 230000007062 hydrolysis Effects 0.000 description 2
- 238000006460 hydrolysis reaction Methods 0.000 description 2
- 239000010815 organic waste Substances 0.000 description 2
- 239000012071 phase Substances 0.000 description 2
- 238000006116 polymerization reaction Methods 0.000 description 2
- 238000001556 precipitation Methods 0.000 description 2
- 230000002285 radioactive effect Effects 0.000 description 2
- 230000001502 supplementing effect Effects 0.000 description 2
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 2
- OOYDTEGVMRPGRK-UHFFFAOYSA-N (9-octanoyl-8,10-dioxoheptadecan-9-yl)azanium;chloride Chemical compound [Cl-].CCCCCCCC(=O)C([NH3+])(C(=O)CCCCCCC)C(=O)CCCCCCC OOYDTEGVMRPGRK-UHFFFAOYSA-N 0.000 description 1
- LJHFIVQEAFAURQ-ZPUQHVIOSA-N (NE)-N-[(2E)-2-hydroxyiminoethylidene]hydroxylamine Chemical compound O\N=C\C=N\O LJHFIVQEAFAURQ-ZPUQHVIOSA-N 0.000 description 1
- UWFPQDYNARKOHK-UHFFFAOYSA-N 2,3,4,5-tetraphenylpyridine Chemical compound C1=CC=CC=C1C(C(=C1C=2C=CC=CC=2)C=2C=CC=CC=2)=CN=C1C1=CC=CC=C1 UWFPQDYNARKOHK-UHFFFAOYSA-N 0.000 description 1
- VYQLYCLTFJDXGV-UHFFFAOYSA-N 2-[[2-(dioctylamino)-2-oxoethyl]-methylamino]-n,n-dioctylacetamide Chemical compound CCCCCCCCN(CCCCCCCC)C(=O)CN(C)CC(=O)N(CCCCCCCC)CCCCCCCC VYQLYCLTFJDXGV-UHFFFAOYSA-N 0.000 description 1
- NEAQRZUHTPSBBM-UHFFFAOYSA-N 2-hydroxy-3,3-dimethyl-7-nitro-4h-isoquinolin-1-one Chemical compound C1=C([N+]([O-])=O)C=C2C(=O)N(O)C(C)(C)CC2=C1 NEAQRZUHTPSBBM-UHFFFAOYSA-N 0.000 description 1
- 241001657948 Midea Species 0.000 description 1
- KJTLSVCANCCWHF-UHFFFAOYSA-N Ruthenium Chemical compound [Ru] KJTLSVCANCCWHF-UHFFFAOYSA-N 0.000 description 1
- 206010041662 Splinter Diseases 0.000 description 1
- 229910052770 Uranium Inorganic materials 0.000 description 1
- 230000002411 adverse Effects 0.000 description 1
- 150000001450 anions Chemical class 0.000 description 1
- 230000009286 beneficial effect Effects 0.000 description 1
- 229910052792 caesium Inorganic materials 0.000 description 1
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 description 1
- 150000001768 cations Chemical class 0.000 description 1
- 239000003638 chemical reducing agent Substances 0.000 description 1
- 239000008139 complexing agent Substances 0.000 description 1
- 125000004122 cyclic group Chemical group 0.000 description 1
- 238000005202 decontamination Methods 0.000 description 1
- 230000003588 decontaminative effect Effects 0.000 description 1
- 239000007857 degradation product Substances 0.000 description 1
- 230000001627 detrimental effect Effects 0.000 description 1
- 238000004821 distillation Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- ZZUFCTLCJUWOSV-UHFFFAOYSA-N furosemide Chemical compound C1=C(Cl)C(S(=O)(=O)N)=CC(C(O)=O)=C1NCC1=CC=CO1 ZZUFCTLCJUWOSV-UHFFFAOYSA-N 0.000 description 1
- 238000009375 geological disposal Methods 0.000 description 1
- 239000002927 high level radioactive waste Substances 0.000 description 1
- 239000012535 impurity Substances 0.000 description 1
- 238000005342 ion exchange Methods 0.000 description 1
- 239000003456 ion exchange resin Substances 0.000 description 1
- 229920003303 ion-exchange polymer Polymers 0.000 description 1
- 239000008204 material by function Substances 0.000 description 1
- 230000005012 migration Effects 0.000 description 1
- 238000013508 migration Methods 0.000 description 1
- XTAZYLNFDRKIHJ-UHFFFAOYSA-N n,n-dioctyloctan-1-amine Chemical compound CCCCCCCCN(CCCCCCCC)CCCCCCCC XTAZYLNFDRKIHJ-UHFFFAOYSA-N 0.000 description 1
- 235000019353 potassium silicate Nutrition 0.000 description 1
- 239000002244 precipitate Substances 0.000 description 1
- 238000011112 process operation Methods 0.000 description 1
- 238000000746 purification Methods 0.000 description 1
- UMJSCPRVCHMLSP-UHFFFAOYSA-N pyridine Natural products COC1=CC=CN=C1 UMJSCPRVCHMLSP-UHFFFAOYSA-N 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 230000008929 regeneration Effects 0.000 description 1
- 238000011069 regeneration method Methods 0.000 description 1
- 229910052707 ruthenium Inorganic materials 0.000 description 1
- 238000000926 separation method Methods 0.000 description 1
- 239000002910 solid waste Substances 0.000 description 1
- 239000002915 spent fuel radioactive waste Substances 0.000 description 1
- 229910052712 strontium Inorganic materials 0.000 description 1
- CIOAGBVUUVVLOB-UHFFFAOYSA-N strontium atom Chemical compound [Sr] CIOAGBVUUVVLOB-UHFFFAOYSA-N 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 1
- 229940054541 urex Drugs 0.000 description 1
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B7/00—Working up raw materials other than ores, e.g. scrap, to produce non-ferrous metals and compounds thereof; Methods of a general interest or applied to the winning of more than two metals
- C22B7/006—Wet processes
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B3/00—Extraction of metal compounds from ores or concentrates by wet processes
- C22B3/20—Treatment or purification of solutions, e.g. obtained by leaching
- C22B3/26—Treatment or purification of solutions, e.g. obtained by leaching by liquid-liquid extraction using organic compounds
- C22B3/38—Treatment or purification of solutions, e.g. obtained by leaching by liquid-liquid extraction using organic compounds containing phosphorus
- C22B3/384—Pentavalent phosphorus oxyacids, esters thereof
- C22B3/3846—Phosphoric acid, e.g. (O)P(OH)3
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B34/00—Obtaining refractory metals
- C22B34/10—Obtaining titanium, zirconium or hafnium
- C22B34/14—Obtaining zirconium or hafnium
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B61/00—Obtaining metals not elsewhere provided for in this subclass
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/04—Treating liquids
- G21F9/06—Processing
- G21F9/12—Processing by absorption; by adsorption; by ion-exchange
- G21F9/125—Processing by absorption; by adsorption; by ion-exchange by solvent extraction
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- Mechanical Engineering (AREA)
- Materials Engineering (AREA)
- Organic Chemistry (AREA)
- Manufacturing & Machinery (AREA)
- Metallurgy (AREA)
- Environmental & Geological Engineering (AREA)
- Life Sciences & Earth Sciences (AREA)
- Geology (AREA)
- General Life Sciences & Earth Sciences (AREA)
- Geochemistry & Mineralogy (AREA)
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- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Extraction Or Liquid Replacement (AREA)
Abstract
The invention relates to a method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method, which comprises the following steps: preparing technetium-zirconium co-extraction feed, co-extracting technetium-zirconium by taking TBP-kerosene organic solution as an extractant, reducing and back-extracting technetium in TBP organic phase co-extraction liquid by adopting a reducing back-extractant, and back-extracting zirconium in the residual TBP organic phase by using a complexing back-extractant; the method improves the current situation that more organic phase radioactive waste liquid except tributyl phosphate is generated when the extraction method is adopted to extract technetium from nuclear fuel post-treatment waste liquid in the prior art, and a set of solvent treatment system is additionally required to be added, can avoid introducing extra solvent and metal ion groups into the post-treatment system, and can be better matched with the main process flow and facilities of a post-treatment plant adopting a PUREX flow, thereby enabling the technetium extraction process to be easier to implement in the post-treatment plant. The method provided by the invention can avoid technetium from entering glass solidification and geological treatment, and prevent the risk of causing long-term damage to the biosphere.
Description
Technical Field
The invention belongs to the technical field of nuclear fuel post-treatment, and particularly relates to a method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method.
Background
Technetium-99 is a long-life fission product with a relatively high yield in nuclear reactors, a fission yield of about 6% and a half-life of about 2.11×10 5 Year after year. In the high-level radioactive waste liquid glass solidification process of the PUREX post-treatment flow, technetium-99 is easy to form a co-volatile with gamma radionuclide Cs-137 in the main section, and has very adverse effect on process operation; in geological treatments, technetium-99 is susceptible to migration, with a potential long-term radiation hazard to the biosphere. Therefore, in spent fuel post-treatment, it is desirable to extract technetium from the radioactive waste liquid, avoiding its entry into glass solidification and geological disposal.
Technetium extraction from post-treatment processes or waste streams is a reported or patented technology of several types:
(1) The urex+ procedure proposed in the united states, the PUREX procedure for separating technetium proposed in Liu Fang, etc. The method adopts hydrophilic ethylhydroxamic acid (AHA) as a key reagent, greatly changes the main process flow frame of the post-treatment, and is not suitable for the traditional PUREX flow post-treatment factory; the technetium product obtained by the method has poor decontamination factors on splinter elements and uranium plutonium.
(2) Adsorption and ion exchange processes. Various ion exchange resins and COF and MOF functional materials are developed at home and abroad for technetium extraction. The concentration of technetium in the post-treatment waste liquid is higher, the waste liquid has complex composition, and contains organic reagent degradation products, so that the service life of the ion exchanger is easily influenced, and the adsorption method is not suitable for extracting a large amount of technetium. The adsorption method is relatively suitable for trapping technetium with low concentration in the discharged waste liquid. While the ion exchanger will produce a significant amount of radioactive solid waste after discharge, which is detrimental to minimizing the radioactive waste.
(3) Trioctylamine, N235, aliquat336, pyridine and its derivatives, butanone, NTAamide, MIDOA, etc. are used as extractant, and extracted by extraction method. Extraction, washing and back extraction are needed in the extraction method, so that the high recovery rate of technetium and the purification factor can be realized. However, this type of method generates more radioactive waste liquid of organic phase other than tributyl phosphate, and requires an additional solvent treatment system, which is very disadvantageous for post-treatment plants.
(4) Precipitation. Large cation reagents such as tetraphenylpyridine acetate or tetrafluoroborate can form a precipitate with pertechnetate anions, and have strong selectivity. The precipitation method adopted in the strong radioactive operation has a plurality of operation defects in engineering, and macromolecular organic matters such as tetraphenylpyridine and the like are introduced into the waste liquid.
It can be seen that the above existing solutions have some difficulties in terms of practicality for the post-treatment plants using the PUREX process. Thus, there is a need for a technetium extraction solution that can be implemented to better match the main process flow and facilities of the reprocessing plant employing the PUREX process.
Disclosure of Invention
Aiming at the defects existing in the prior art, the invention aims to provide a method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method, tributyl phosphate (TBP) is adopted as an extracting agent, technetium-99 is extracted from the waste liquid in a PUREX flow, the introduction of solvents except TBP in the technetium extraction process is avoided, and the extra organic waste types are not greatly increased, so that the method can be better matched with the main process flow and facilities of a post-treatment plant adopting the PUREX flow.
In order to achieve the above purpose, the invention adopts the technical scheme that: a method for extracting technetium from nuclear fuel reprocessing waste liquid by extraction, the method comprising the steps of:
s1, preparing technetium-zirconium co-extraction feeding: evaporating and concentrating technetium extraction source waste liquid, adding a certain amount of zirconium nitrate, and preparing technetium-zirconium co-extraction feed;
s2, co-extracting technetium and zirconium: extracting technetium and zirconium in the technetium-zirconium co-extraction feed by adopting an extracting agent, and washing the co-extraction liquid by using a washing agent to obtain TBP organic phase co-extraction liquid;
the extractant is TBP-kerosene organic solution;
s3, back extraction of technetium: reducing and stripping technetium in the TBP organic phase co-extract by adopting a reducing and stripping agent to obtain a stripping technetium product;
s4, back-extracting zirconium in the residual TBP organic phase: and stripping zirconium in the residual TBP organic phase by using a complexing stripping agent.
In step S1, evaporating and concentrating the technetium-extracted source item waste liquid by more than 20 times;
in the prepared technetium-zirconium co-extraction feed, the concentration of nitric acid is 4-5 mol/L.
Further, in the extractant, the mass percentage of TBP is 15-50%.
Further, the detergent is nitric acid, and the molar concentration of the detergent is 4-5 mol/L.
In the step S2, the process of co-extracting technetium and zirconium adopts a mode that a 6-10-level extraction working section and a 1-8-level washing working section are connected in series.
Further, in the step S3, the reducing stripping agent is hydrazine or methyl hydrazine.
Further, in the step S3, after the technetium in the stripping TBP organic phase co-extract is reduced by adopting a reducing stripping agent, the method further includes the following steps:
and (3) carrying out the supplementary extraction on zirconium in the back-extracted technetium product by adopting a supplementary extraction agent to obtain the technetium product.
Further, the stripping agent is TBP-kerosene organic solution, wherein the mass percentage of TBP is 15-50%.
In the step S3, a 4-10-level back extraction technetium working section and a 4-8-level zirconium supplement working section are adopted to be connected in series.
Further, the complexing back-extraction agent adopts ethylhydroxamic acid or hydrophilic polyamidoamine reagent.
The invention has the beneficial effects that: the method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method provided by the invention comprises the following steps: preparing technetium-zirconium co-extraction feed, co-extracting technetium-zirconium by taking TBP-kerosene organic solution as an extractant, reducing and back-extracting technetium in TBP organic phase co-extraction liquid by adopting a reducing back-extractant, and back-extracting zirconium in the residual TBP organic phase by using a complexing back-extractant; the method improves the current situation that more organic phase radioactive waste liquid except tributyl phosphate is generated when the extraction method is adopted to extract technetium from nuclear fuel post-treatment waste liquid in the prior art, and a set of solvent treatment system is additionally required to be added, can avoid introducing extra solvent and metal ion groups into the post-treatment system, and can be better matched with the main process flow and facilities of a post-treatment plant adopting a PUREX flow, thereby enabling the technetium extraction process to be easier to implement in the post-treatment plant. The method provided by the invention can extract technetium from nuclear fuel post-treatment waste liquid, reduce alpha and gamma activities in technetium products, and enable the technetium to be safely and conveniently stored for a long time; technetium can be prevented from entering glass solidification and geological treatment, and long-term hazard risks to biospheres are prevented.
Drawings
FIG. 1 is a schematic flow chart of a method for extracting technetium from nuclear fuel reprocessing waste liquid by an extraction method according to an embodiment of the present invention;
FIG. 2 is a schematic illustration of a PUREX post-treatment process flow according to an embodiment of the present invention;
FIG. 3 is a graphical representation of the acidity curve of the partition ratio in a technetium-zirconium co-extraction feed system, according to an embodiment of the present invention;
FIG. 4 is a diagram of the reduction of TcO with methylhydrazine (MMH) according to an embodiment of the invention 4 － Is a reaction kinetics curve of (2);
fig. 5 is a flow chart of a technetium extraction process using TBP-kerosene organic solution as extractant according to an embodiment of the present invention.
Detailed Description
The technical solutions of the embodiments of the present invention will be further clearly and completely described below with reference to the accompanying drawings and examples, and it is apparent that the described embodiments are only some embodiments of the present invention, but not all embodiments, and all other examples obtained by those skilled in the art without making creative efforts based on the examples in the present invention are included in the protection scope of the present invention.
Considering that several existing solutions have some difficulties that cannot be overcome when applied to a post-treatment plant that adopts the PUREX process. The inventor believes that the best solution is to propose a technetium extraction process based on tributyl phosphate (TBP) as an extractant, extract technetium-99 from the waste liquid in the PUREX process, avoid introducing solvents other than TBP in the technetium extraction process, and not greatly increase the types of extra organic waste, thus being better matched with the main process flow and facilities of a post-treatment plant adopting the PUREX process.
As shown in fig. 1, the method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method provided by the embodiment of the invention comprises the following steps:
s1, preparing technetium-zirconium co-extraction feed AF: evaporating and concentrating the technetium extraction source waste liquid by more than 20 times, and adding a certain amount of zirconium nitrate to assist the extraction of technetium, so as to prepare and form technetium-zirconium co-extraction feed AF;
specifically, in the PUREX post-treatment process shown in fig. 2, technetium mainly enters a tail liquid 1AXXW of technetium washing and uranium-plutonium supplementary extraction and a tail liquid 2AW of plutonium wires. The main compositions of the two technetium-extracted source waste solutions (technetium washing and uranium-plutonium supplementary extraction tail solution 1AXXW and plutonium line tail solution 2 AW) are shown in Table 1, and the nitric acid concentrations are relatively high (> 3.5 mol/L). In the technological process of extracting technetium, the main decontaminating object is trace uranium plutonium, strontium, cesium, ruthenium and other components.
TABLE 1 main composition of two technetium-extracting source waste solutions
In the technetium extraction source waste liquid (1 AXXW, 2 AW), the concentration of technetium is relatively low, which is unfavorable for the extraction process of the technetium-zirconium co-extraction step, in the embodiment, the concentration of technetium needs to be about 20 times, and the concentration of nitric acid is controlled to be 4-5 mol/L, so that the concentrated technetium extraction source waste liquid is obtained and is used for preparing technetium-zirconium co-extraction feed AF.
Extraction of technetium (TcO) directly from nitric acid medium using TBP 4 － ) The partition ratio of (a) is very low (less than 0.6), but the inventors found that the metal ions having high valence (such as Zr 4+ 、Pu 4+ ) When present, technetium can form co-extraction with high valence metal ions, resulting in higher partition ratios, up to approximately 2, as shown by the acidity profile of the partition ratio in the technetium zirconium co-extraction feed system of fig. 3. Therefore, a certain amount of stable zirconium isotope is added into the concentrated technetium extraction source waste liquid, so that the extraction of technetium can be enhanced. Zirconium can also be extracted under the condition of high acidity, only trace zirconium enters into tail liquid, namely concentrated technetium extraction source waste liquid and added zirconium are prepared to form technetium-zirconium co-extraction feed AF.
S2, co-extracting technetium and zirconium (A process section): extracting technetium and zirconium in the technetium-zirconium co-extraction feed AF by adopting an extractant AX, and washing the co-extraction liquid by using a detergent AS to obtain TBP organic phase co-extraction liquid BF;
specifically, in step S2, the extractant AX is TBP-kerosene organic solution; the detergent AS is nitric acid.
Optionally, in the extractant AX, the mass percentage of TBP is 15-50%.
Optionally, the molar concentration of the nitric acid is 4-5 mol/L.
Specifically, the process of co-extracting technetium and zirconium (A process section) adopts a mode that a 6-10-level extraction working section and a 1-8-level washing working section are connected in series; in the finally obtained TBP organic phase co-extract BF, the recovery rate of technetium reaches more than 98 percent; the recovery rate of zirconium reaches more than 98 percent.
S3, back extraction technetium (B process section): reducing and stripping technetium in TBP organic phase co-extract BF by adopting a reducing and stripping agent BS to obtain a stripping technetium product;
specifically, in step S3, the reducing stripping agent BS is hydrazine or methyl hydrazine.
The TBP organic phase co-extract BF contains technetium and zirconium, and technetium needs to be selectively back extracted. Technetium can be extracted when present at +7 valency, whereas technetium of lower valency (+6, +5, +4) is not extracted by TBP. Therefore, technetium can be reduced and back extracted by using organic reducing agents such as hydrazine, methyl hydrazine and the like under the acidity condition that the concentration of hydrogen ions is 2-3 mol/L, zirconium and trace uranium and plutonium still remain in a TBP organic phase, and separation of technetium and the impurities is realized. Reduction of TcO with methyl hydrazine (MMH) 4 － The reaction kinetics curves of (2) are shown in FIG. 4.
Optionally, in order to prevent zirconium from entering the technetium product stream, after the technetium in the TBP organic phase co-extract BF is reduced with the reducing stripping agent BS in step S3, the steps of: and (3) carrying out the supplementary extraction on zirconium in the back-extracted technetium product by adopting a supplementary extraction agent BX to obtain the technetium product BP.
Specifically, the stripping agent BX is TBP-kerosene organic solution, wherein the mass percentage of TBP is 15-50%.
Specifically, step S3 adopts a 4-10-level back extraction technetium working section and a 4-8-level zirconium supplementing working section to be connected in series, and the technetium yield of technetium product BP is more than 99.9%. After the steps of back extraction of technetium and zirconium supplement extraction in the step S3, most of technetium in the TBP organic phase co-extract BF is extracted, but the residual zirconium and TBP-kerosene organic solution in the TBP organic phase CF can be recycled.
S4, back-extracting zirconium in the residual TBP organic phase CF (C process section): back-extracting zirconium in the residual TBP organic phase CF by adopting a complexing back-extracting agent CX;
wherein, the zirconium-containing water phase is concentrated after destroying the complexing agent, and the mixture is returned to the step S1 to prepare technetium-zirconium co-extraction feed AF; the residual TBP-kerosene organic solution in the TBP organic phase CF is sent to a solvent treatment unit of a post-treatment plant for regeneration.
Specifically, in the remaining TBP organic phase CF, the TBP-kerosene organic solution and zirconium need to be recycled. The cyclic multiplexing method is as follows: firstly, back-extracting zirconium in the residual TBP organic phase CF by using a complexation back-extracting agent CX, concentrating the back-extracted zirconium in the water phase, and then preparing technetium-zirconium co-extraction feed AF; then, the remaining TBP organic phase is sent to a solvent treatment unit of a post-treatment plant, and after being regenerated through washing, distillation and other procedures, the TBP organic phase is used for preparing the extractant AX in the step S2 and the extractant BX in the step S3.
Alternatively, to avoid zirconium hydrolysis and polymerization, the zirconium is stripped by complexation under high acidity conditions when stripping the remaining TBP organic phase CF of zirconium in step S4. Specifically, the complexing stripping agent CX is ethanol hydroxamic acid (AHA) or hydrophilic polyamidoamine reagents (NTAamide (C2 HE), MIDEA and the like), so that the tetravalent element can be selectively complexed and stripped, and the recovery of zirconium can be realized.
A complete technetium extraction process flow diagram using TBP-kerosene organic solution as extractant is shown in fig. 5.
Example 1
In this example 1, a method for extracting technetium from nuclear fuel post-treatment waste liquid by using an extraction method is provided in this embodiment, and technetium in the technetium extraction source waste liquid is extracted by using a tail liquid 1AXXW and a plutonium tail liquid 2AW of technetium washing and uranium plutonium supplementary extraction in a PUREX post-treatment process flow as the technetium extraction source waste liquid, and the method includes the following steps:
s1, preparing technetium-zirconium co-extraction feed AF: evaporating and concentrating the technetium extraction source waste liquid by more than 20 times, and adding a certain amount of zirconium nitrate for assisting the extraction of technetium;
the technetium-zirconium co-extraction feed AF is prepared from concentrated technetium-extraction source waste liquid and added zirconium nitrate, and the main components of the prepared technetium-zirconium co-extraction feed AF comprise: zirconium with a mass concentration of 1.8g/L, technetium with a mass concentration of 0.4g/L, nitric acid with a molar concentration of 4.5 mol/L.
S2, co-extracting technetium and zirconium (A process section): adopting TBP-kerosene organic solution as extractant AX to extract technetium and zirconium in the technetium-zirconium co-extraction feed AF; then, nitric acid is used AS a detergent AS to wash the co-extract to obtain TBP organic phase co-extract BF;
wherein, adopting a mode of connecting an 8-level extraction working section and a 4-level washing working section in series, the ratio of the feed flows of the technetium-zirconium co-extraction feed AF, the extractant AX and the detergent AS is 1:1:0.25; in the final TBP organic phase co-extract BF, the recovery rate of technetium is more than 98%, and the recovery rate of zirconium is more than 98%.
Specifically, in the TBP-kerosene organic solution, the mass percentage of TBP is 30%.
Specifically, the molar concentration of the nitric acid is 4-5 mol/L.
S3, back extraction technetium (B process section): firstly, adopting hydrazine or methyl hydrazine as a reduction stripping agent BS to reduce technetium in a stripping TBP organic phase co-extract BF; then, to prevent zirconium from entering the technetium product stream, zirconium is stripped with stripping agent BX, which is a TBP-kerosene organic solution.
Specifically, under the acidity condition that the concentration of hydrogen ions is 2-3 mol/L, 0.5mol/L methyl hydrazine is adopted as a reduction stripping agent BS, and the mass percentage of TBP in the stripping agent BX is 30%.
Adopting a mode that a 10-level back extraction technetium working section and a 6-level zirconium supplementing working section are connected in series, and adopting TBP organic phase co-extraction liquid BF: stripping agent BS: the ratio of the feed flow of the stripping agent BX is 1:0.25:0.25, and the technetium yield of technetium product BP is more than 99.9%.
S4, back-extracting zirconium in the residual TBP organic phase (C process section): stripping zirconium in the residual TBP organic phase CF with a complexing stripping agent CX;
specifically, to avoid zirconium hydrolysis and polymerization, the stripping of the remaining TBP organic phase CF in step S4 is performed at high acidity conditions using 0.2mol/L of glyoxime acid (AHA) as complexing stripping agent CX, wherein the remaining TBP organic phase CF: the ratio of the feed flow rates of the complexing stripping agents CX is 1:0.5, and the recovery rate of zirconium is more than 99.9 percent.
The extraction method for extracting technetium from the nuclear fuel post-treatment waste liquid provided by the embodiment can extract technetium from the nuclear fuel post-treatment waste liquid, reduce the alpha and gamma activities in technetium products, and ensure that the technetium can be safely and conveniently stored for a long time; technetium can be prevented from entering glass solidification and geological treatment, and long-term hazard risks to biospheres are prevented. According to the embodiment, the current situation that more organic phase radioactive waste liquid except tributyl phosphate is generated when the extraction method is adopted to extract technetium from the nuclear fuel post-treatment waste liquid in the prior art and a set of solvent treatment system is required to be additionally added is changed, so that the introduction of additional solvents and metal ion groups into the post-treatment system can be avoided, the main process flow and facilities of a post-treatment plant adopting a PUREX flow can be better matched, and the technetium extraction process is easier to implement in the post-treatment plant.
The method of the present invention is not limited to the specific embodiments described above, but the above examples are merely illustrative of the present invention, and the present invention may be embodied in other specific forms or in other specific forms without departing from the spirit or essential characteristics thereof. The described embodiments are, therefore, to be considered in all respects as illustrative and not restrictive. The scope of the invention should be indicated by the appended claims, and any changes that are equivalent to the intent and scope of the claims are intended to be encompassed within the scope of the invention.
Claims (10)
1. A method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method, which is characterized by comprising the following steps:
s1, preparing technetium-zirconium co-extraction feeding: evaporating and concentrating technetium extraction source waste liquid, adding a certain amount of zirconium nitrate, and preparing technetium-zirconium co-extraction feed;
s2, co-extracting technetium and zirconium: extracting technetium and zirconium in the technetium-zirconium co-extraction feed by adopting an extracting agent, and washing the co-extraction liquid by using a washing agent to obtain TBP organic phase co-extraction liquid;
the extractant is TBP-kerosene organic solution;
s3, back extraction of technetium: reducing and stripping technetium in the TBP organic phase co-extract by adopting a reducing and stripping agent to obtain a stripping technetium product;
s4, back-extracting zirconium in the residual TBP organic phase: and stripping zirconium in the residual TBP organic phase by using a complexing stripping agent.
2. The method for extracting technetium from nuclear fuel reprocessing waste liquid according to claim 1, wherein in step S1, the technetium-extracted source waste liquid is concentrated by evaporation more than 20 times;
in the prepared technetium-zirconium co-extraction feed, the concentration of nitric acid is 4-5 mol/L.
3. The method for extracting technetium from nuclear fuel reprocessing waste liquid according to claim 1, wherein the mass percentage of TBP in the extractant is 15-50%.
4. The method for extracting technetium from nuclear fuel reprocessing waste liquid according to claim 1, wherein the detergent is nitric acid with a molar concentration of 4-5 mol/L.
5. The method according to claim 1, wherein in step S2, the co-extraction of technetium and zirconium is performed by connecting a 6-10 stage extraction section and a 1-8 stage washing section in series.
6. The method according to claim 1, wherein in step S3, the reducing stripping agent is hydrazine or methyl hydrazine.
7. The method according to claim 1, wherein in step S3, after reducing technetium in the stripping TBP organic phase co-extract with a reducing stripping agent, the method further comprises the steps of:
and (3) carrying out the supplementary extraction on zirconium in the back-extracted technetium product by adopting a supplementary extraction agent to obtain the technetium product.
8. The method for extracting technetium from nuclear fuel post-treatment waste liquid by adopting an extraction method according to claim 7, wherein the extraction supplement is TBP-kerosene organic solution, and the mass percentage of TBP is 15-50%.
9. The method according to claim 7, wherein in step S3, a 4-10 stage technetium stripping section and a 4-8 stage zirconium stripping section are connected in series.
10. The method for extracting technetium from nuclear fuel reprocessing waste liquid according to any one of claims 1 to 9, wherein: the complexing back-extraction agent adopts ethylhydroxamic acid or hydrophilic polyamidoamine reagent.
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