CN1472347A - Method for separating uranium and plutonium in Pretz process - Google Patents
Method for separating uranium and plutonium in Pretz process Download PDFInfo
- Publication number
- CN1472347A CN1472347A CNA021258708A CN02125870A CN1472347A CN 1472347 A CN1472347 A CN 1472347A CN A021258708 A CNA021258708 A CN A021258708A CN 02125870 A CN02125870 A CN 02125870A CN 1472347 A CN1472347 A CN 1472347A
- Authority
- CN
- China
- Prior art keywords
- plutonium
- uranium
- solution
- hno
- organic phase
- 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.)
- Granted
Links
- 238000000034 method Methods 0.000 title claims abstract description 37
- 229910052778 Plutonium Inorganic materials 0.000 title claims abstract description 25
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 title claims abstract description 25
- 229910052770 Uranium Inorganic materials 0.000 title claims abstract description 24
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 title claims abstract description 21
- 229940053390 pretz Drugs 0.000 title 1
- 239000012074 organic phase Substances 0.000 claims abstract description 18
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 claims abstract description 14
- 239000003795 chemical substances by application Substances 0.000 claims abstract description 14
- VSNHCAURESNICA-UHFFFAOYSA-N Hydroxyurea Chemical compound NC(=O)NO VSNHCAURESNICA-UHFFFAOYSA-N 0.000 claims abstract description 9
- 229960001330 hydroxycarbamide Drugs 0.000 claims abstract description 9
- 238000003756 stirring Methods 0.000 claims abstract description 4
- 239000008346 aqueous phase Substances 0.000 claims abstract 2
- 230000002829 reductive effect Effects 0.000 claims description 20
- 239000000243 solution Substances 0.000 claims description 20
- 238000000605 extraction Methods 0.000 claims description 18
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 16
- 239000003153 chemical reaction reagent Substances 0.000 claims description 4
- 239000003350 kerosene Substances 0.000 claims description 2
- 239000011259 mixed solution Substances 0.000 claims description 2
- 238000000926 separation method Methods 0.000 abstract description 7
- 229910017604 nitric acid Inorganic materials 0.000 abstract description 5
- 239000003638 chemical reducing agent Substances 0.000 abstract description 4
- 239000003758 nuclear fuel Substances 0.000 abstract description 2
- 150000001224 Uranium Chemical class 0.000 abstract 1
- 238000012805 post-processing Methods 0.000 abstract 1
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 11
- 230000000694 effects Effects 0.000 description 4
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 3
- 150000001335 aliphatic alkanes Chemical class 0.000 description 3
- 239000003085 diluting agent Substances 0.000 description 3
- OAKJQQAXSVQMHS-UHFFFAOYSA-N Hydrazine Chemical compound NN OAKJQQAXSVQMHS-UHFFFAOYSA-N 0.000 description 2
- AMIMRNSIRUDHCM-UHFFFAOYSA-N Isopropylaldehyde Chemical compound CC(C)C=O AMIMRNSIRUDHCM-UHFFFAOYSA-N 0.000 description 2
- 238000010790 dilution Methods 0.000 description 2
- 239000012895 dilution Substances 0.000 description 2
- 238000012958 reprocessing Methods 0.000 description 2
- 239000002915 spent fuel radioactive waste Substances 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- UNBYXRSBNHGVLA-UHFFFAOYSA-N 2-[ethyl(hydroxy)amino]ethanol Chemical compound CCN(O)CCO UNBYXRSBNHGVLA-UHFFFAOYSA-N 0.000 description 1
- AVXURJPOCDRRFD-UHFFFAOYSA-N Hydroxylamine Chemical compound ON AVXURJPOCDRRFD-UHFFFAOYSA-N 0.000 description 1
- 229910052781 Neptunium Inorganic materials 0.000 description 1
- IYQHAABWBDVIEE-UHFFFAOYSA-N [Pu+4] Chemical compound [Pu+4] IYQHAABWBDVIEE-UHFFFAOYSA-N 0.000 description 1
- 239000002253 acid Substances 0.000 description 1
- 229910052799 carbon Inorganic materials 0.000 description 1
- 230000007812 deficiency Effects 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 150000002429 hydrazines Chemical class 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- -1 neptunium (VI) ions Chemical class 0.000 description 1
- LFNLGNPSGWYGGD-UHFFFAOYSA-N neptunium atom Chemical compound [Np] LFNLGNPSGWYGGD-UHFFFAOYSA-N 0.000 description 1
- 238000005025 nuclear technology Methods 0.000 description 1
- 230000001590 oxidative effect Effects 0.000 description 1
- 238000002360 preparation method Methods 0.000 description 1
- 230000000717 retained effect Effects 0.000 description 1
- 239000003643 water by type Substances 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Inorganic Compounds Of Heavy Metals (AREA)
Abstract
The invention belongs to the technical field of nuclear fuel post-processing, and particularly relates to a method for separating uranium and plutonium (Pu) in a Purex flow. The method is in HNO3Co-extracting the co-decontaminated uranium and plutonium into an organic phase TBP solution in a medium, adding a reducing agent, fully stirring to reduce Pu (IV) into Pu (III), and back-extracting the Pu (III) into an aqueous phase HNO3In the solution, the reducing agent used is hydroxyurea HU, a salt-free agent. Due to the adoption of hydroxyurea HU as a reducing agent, the process conditions are simplified, and the separation speed is increasedAnd the efficiency is obviously improved.
Description
Technical field
The invention belongs to the nuclear fuel reprocessing technical field, particularly relate to uranium in a kind of PUREX process (U), the isolating method of plutonium (Pu).
Background technology
General rex (Purex) flow process is a kind ofly to make extraction agent with tributyl phosphate (TBP), separates and reclaim the technical process of uranium, plutonium from the reactor spent fuel.In the Purex flow process, U after codecontamination and Pu come together in the organic phase TBP/ alkane diluent altogether.U exists with the chemical combination attitude of U (VI) in TBP/ alkane diluent solution, and Pu exists with the chemical combination attitude of Pu (IV).Pu (IV) easily is reduced to Pu (III) than the oxidizing potential height of Pu (III).Adopt suitable reductive agent Pu (IV) can be reduced to Pu (III), the distribution of Pu (III) in the TBP/ alkane diluent is minimum, and therefore, Pu (IV) is reduced into Pu (III) back and is arrived water by back extraction.And U (VI) is not reduced, and still is retained in the organic phase, thereby reaches uranium, the isolating effect of plutonium.Aftertreatment factory all adopts the Purex flow process at present, and uranium, plutonium separate how employing U (IV) make reductive agent on stream.Dilution U electrolytic preparation U (IV) generally adopts in aftertreatment factory, and the increase of U amount can increase the weight of the burden of flow process SEPARATION OF URANIUM, plutonium, moreover dilution U can dilute in the U product
235The concentration of U.In order to stablize U (IV) and Pu (III), need to add a large amount of hydrazines and do to support reductive agent, the use of hydrazine will produce HN in solution
3, NH
4 +Be harmful to the material of flow process, and the technical process complexity.Make reductive agent to SEPARATION OF URANIUM in order to overcome with U (IV), the deficiency that the plutonium flow process is brought, in recent years, bibliographical information has been developed many salt-free organic reagents, for example: azanol and derivative thereof [VS Koltunov, RJ Taylor, SM Baranov, et al., Thereduction of plutonium (IV) and neptunium (VI) ions byN, N-ethyl (hydroxyethyl) hydroxylamine in nitric acid, Radiochim.Acta.1999,86:115-21], isobutyric aldehyde [G Uchiyama, S Fujine, S Hotoku et al., Newseparation process for neptunium, plutonium and uranium usingbutyraldehydes as reductant in reprocessing, Nuclear technology.1993,102,341-351], low-carbon (LC) hydroximic acid [B.Ya.zilberman, A.N.Mashkin, A.K.Nardova.et al. " Method of spent nuclear fuel reprocessing " .] etc.
Summary of the invention
The object of the present invention is to provide utilize in a kind of Purex flow process novel salt-free reductive agent fast, the method for high efficiency separation uranium, plutonium.
Technical scheme of the present invention is as follows: uranium, the isolating method of plutonium in the Purex flow process, and at HNO
3Uranium that will be after codecontamination in the medium and plutonium come together in the organic phase TBP solution altogether, add reductive agent and fully stir, and make Pu (IV) be reduced into Pu (III), Pu (III) by back extraction to water HNO
3In the solution, wherein the reductive agent that is adopted is salt-free reagent hydroxyurea HU.
Uranium, the isolating method of plutonium, wherein water HNO in the Purex flow process as mentioned above
3The concentration of hydroxyurea HU is 10～50 times of Pu concentration in the organic phase TBP solution in the solution, water HNO
3The volume ratio of solution and organic phase TBP solution is 1: 1～1: 5.
Because present method adopts salt-free reagent hydroxyurea HU as reductive agent, U (VI) is not being constituted under the prerequisite of influence, can in minutes Pu (IV) be reduced into Pu (III), and can in several minutes, water be arrived in Pu (III) back extraction, and HU can destroy HNO
3The HNO that produces in the system
2Thereby, can stablize Pu (III) under the condition that support reductive agent not adding, simplified the flow process condition, single-stage uranium/plutonium separation factor (removing plutonium in the uranium) can be up to 10
3More than, velocity of separation and efficient all are significantly improved.
Embodiment
Uranium, the isolating method of plutonium in the Purex flow process are at the HNO of 0.1～4.0mol/L
3Uranium that will be after codecontamination in the medium and plutonium come together in the 30%TBP/ kerosene mixed solution altogether, add reductive agent hydroxyurea HU, make water HNO
3The concentration of hydroxyurea HU is 10～50 times of Pu concentration in the organic phase TBP solution in the solution, water HNO
3The volume ratio of solution and organic phase TBP solution is 1: 1～1: 5, fully stirs, and makes Pu (IV) be reduced into Pu (III), and Pu (III) is arrived water HNO by back extraction
3In the solution, thus the separation of realization uranium, plutonium, and the temperature condition of reaction is 15～50 ℃.
Listed the reduction reextraction effect of different HU and Pu concentration ratio in the table 1.
Table 1
Water: C
HNO3=1.0mol/L; Organic phase: C
Pu=0.000234mol/L; V
Organic phaseV
Water=1: 1; Temperature=15 ℃
| ????HU/Pu | The back extraction ratio % of back extraction 1min | The back extraction ratio % of back extraction 3min | The back extraction ratio % of back extraction 5min | The back extraction ratio % of back extraction 30min |
| ????10 | ????90.8％ | ????95.8％ | ????95.7％ | ????96.2％ |
| ????20 | ????93.8％ | ????95.4％ | ????96.0％ | ????96.3％ |
| ????50 | ????94.2％ | ????96.7％ | ????96.3％ | ????97.5％ |
Listed the reduction reextraction effect of different waters and organic phase volume ratio in the table 2.
Table 2
Water: C
HNO3=1.0mol/L; Organic phase: C
HU: C
Pu=50; Temperature: 15 ℃
| Compare (V Water∶V Organic phase) | The back extraction ratio % of back extraction 1min | The back extraction ratio % of back extraction 3min |
| ????1∶1 | ????94.2％ | ????96.7％ |
| ????1∶3 | ????92.5％ | ????95.8％ |
| ????1∶5 | ????92.1％ | ????95.3％ |
Listed water in the table 3 and the organic phase volume ratio is, different HNO under the differing temps at 1: 1 o'clock
3The reduction reextraction effect of concentration.
Table 3
Claims (4)
1. uranium, the isolating method of plutonium in the Purex flow process are at HNO
3Uranium that will be after codecontamination in the medium and plutonium come together in the organic phase TBP solution altogether, add reductive agent and fully stir, and make Pu (IV) be reduced into Pu (III), Pu (III) by back extraction to water HNO
3In the solution, it is characterized in that: the reductive agent that is adopted is salt-free reagent hydroxyurea HU.
2. uranium, the isolating method of plutonium is characterized in that: water HNO in a kind of Purex flow process as claimed in claim 1
3The concentration of hydroxyurea HU is 10～50 times of Pu concentration in the organic phase TBP solution in the solution, water HNO
3The volume ratio of solution and organic phase TBP solution is 1: 1～1: 5.
3. uranium, the isolating method of plutonium is characterized in that: aqueous phase HNO in a kind of Purex flow process as claimed in claim 1 or 2
3Concentration be 0.1～4.0mol/L, organic phase TBP solution is 30%TBP/ kerosene mixed solution.
4. uranium, the isolating method of plutonium in a kind of Purex flow process as claimed in claim 3, it is characterized in that: the temperature condition of reaction is 15～50 ℃.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CNB021258708A CN1229814C (en) | 2002-07-31 | 2002-07-31 | Method for separating uranium and plutonium in Pretz process |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CNB021258708A CN1229814C (en) | 2002-07-31 | 2002-07-31 | Method for separating uranium and plutonium in Pretz process |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN1472347A true CN1472347A (en) | 2004-02-04 |
| CN1229814C CN1229814C (en) | 2005-11-30 |
Family
ID=34143118
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CNB021258708A Expired - Fee Related CN1229814C (en) | 2002-07-31 | 2002-07-31 | Method for separating uranium and plutonium in Pretz process |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN1229814C (en) |
Cited By (9)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN102290111A (en) * | 2011-06-28 | 2011-12-21 | 中国原子能科学研究院 | Method for purifying and circulating uranium in PUREX process |
| CN102352436A (en) * | 2011-07-21 | 2012-02-15 | 中国原子能科学研究院 | Method for separating U (uranium) from Pu (plutonium) in Purex process |
| CN101484226B (en) * | 2006-07-03 | 2012-06-27 | 阿海珐核循环公司 | Method for separating a chemical element from uranium (VI) using an aqueous nitric phase, in a uranium extraction cycle |
| CN102778522A (en) * | 2012-08-08 | 2012-11-14 | 中国原子能科学研究院 | Method for detecting radiolysis behavior of 30% TBP (Tri-Butyl-Phosphate)-kerosene |
| CN103103369A (en) * | 2011-11-14 | 2013-05-15 | 中国原子能科学研究院 | Catalytic oxidation method of plutonium in nitric acid system |
| CN103820656A (en) * | 2014-01-28 | 2014-05-28 | 中国原子能科学研究院 | Uranium and plutonium separation technology in Purex process |
| CN106893878A (en) * | 2017-03-02 | 2017-06-27 | 中国原子能科学研究院 | A kind of method that plutonium is reclaimed in the spentnuclear fuel from radioactivity |
| CN107130121A (en) * | 2017-05-09 | 2017-09-05 | 中国原子能科学研究院 | Neptunium, the uranium purification process of plutonium are removed simultaneously in a kind of nuclear fuel Purex post processings flow |
| CN112194551A (en) * | 2020-09-03 | 2021-01-08 | 中国原子能科学研究院 | Diluent and hydrogenation preparation method and composition thereof |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN102206753B (en) * | 2011-04-01 | 2013-07-10 | 中国原子能科学研究院 | Method for improving 2A process plutonium concentration multiple in Purex process |
-
2002
- 2002-07-31 CN CNB021258708A patent/CN1229814C/en not_active Expired - Fee Related
Cited By (13)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN101484226B (en) * | 2006-07-03 | 2012-06-27 | 阿海珐核循环公司 | Method for separating a chemical element from uranium (VI) using an aqueous nitric phase, in a uranium extraction cycle |
| CN102290111B (en) * | 2011-06-28 | 2014-06-04 | 中国原子能科学研究院 | Method for purifying and circulating uranium in PUREX process |
| CN102290111A (en) * | 2011-06-28 | 2011-12-21 | 中国原子能科学研究院 | Method for purifying and circulating uranium in PUREX process |
| CN102352436A (en) * | 2011-07-21 | 2012-02-15 | 中国原子能科学研究院 | Method for separating U (uranium) from Pu (plutonium) in Purex process |
| CN103103369A (en) * | 2011-11-14 | 2013-05-15 | 中国原子能科学研究院 | Catalytic oxidation method of plutonium in nitric acid system |
| CN103103369B (en) * | 2011-11-14 | 2014-03-05 | 中国原子能科学研究院 | Catalytic oxidation method of plutonium in nitric acid system |
| CN102778522A (en) * | 2012-08-08 | 2012-11-14 | 中国原子能科学研究院 | Method for detecting radiolysis behavior of 30% TBP (Tri-Butyl-Phosphate)-kerosene |
| CN103820656A (en) * | 2014-01-28 | 2014-05-28 | 中国原子能科学研究院 | Uranium and plutonium separation technology in Purex process |
| CN106893878A (en) * | 2017-03-02 | 2017-06-27 | 中国原子能科学研究院 | A kind of method that plutonium is reclaimed in the spentnuclear fuel from radioactivity |
| WO2018157424A1 (en) * | 2017-03-02 | 2018-09-07 | 中国原子能科学研究院 | Method for recycling plutonium from spent radioactive fuel |
| CN106893878B (en) * | 2017-03-02 | 2018-11-30 | 中国原子能科学研究院 | A method of recycling plutonium from radioactivity spentnuclear fuel |
| CN107130121A (en) * | 2017-05-09 | 2017-09-05 | 中国原子能科学研究院 | Neptunium, the uranium purification process of plutonium are removed simultaneously in a kind of nuclear fuel Purex post processings flow |
| CN112194551A (en) * | 2020-09-03 | 2021-01-08 | 中国原子能科学研究院 | Diluent and hydrogenation preparation method and composition thereof |
Also Published As
| Publication number | Publication date |
|---|---|
| CN1229814C (en) | 2005-11-30 |
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Granted publication date: 20051130 Termination date: 20110731 |