CN104004927A - Process for purifying zirconium in Purex flow plutonium purification cycle - Google Patents
Process for purifying zirconium in Purex flow plutonium purification cycle Download PDFInfo
- Publication number
- CN104004927A CN104004927A CN201410236283.XA CN201410236283A CN104004927A CN 104004927 A CN104004927 A CN 104004927A CN 201410236283 A CN201410236283 A CN 201410236283A CN 104004927 A CN104004927 A CN 104004927A
- Authority
- CN
- China
- Prior art keywords
- plutonium
- purification
- zirconium
- concentration
- purification cycle
- 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
Landscapes
- Inorganic Compounds Of Heavy Metals (AREA)
Abstract
The invention belongs to the technical field of spent fuel post-treatment, and discloses a process for purifying zirconium in a Purex flow plutonium purification cycle. 30% tri-butyl-phosphate (TBP)-kerosene is adopted as an extraction agent 2AX in a 2A extractor, so as to recover uranium and plutonium and purify fission fragments. The process comprises an extraction section and a washing section; and the concentration of a nitric acid in a 2AF feed liquid is 3.5-3.7mol/L, and 2AS is an HNO3 solution of which the concentration is 0.8-0.85mol/L. The process has the advantages that the plutonium yield is greater than 99.9%, and the purification coefficient of zirconium is greater than 100.
Description
Technical field
The invention belongs to spent fuel post-processing technology field, be specifically related to the purification process of zirconium in Purex flow process plutonium purification cycle.
Background technology
Spent fuel refers in nuclear reactor, and irradiation draws off after reaching the specific burn up of plan discharging from heap, and the nuclear fuel of use in this heap no longer.It is carried out to chemical treatment, to remove the impurity such as fission product and to reclaim fissile nuclide and fertile nuclide and some can be utilized the process of material, be called spent fuel aftertreatment.At present, various countries' spent fuel aftertreatment flow process mainly adopts Purex flow process or its improved flow process.It mainly comprises codecontamination partitioning cycle, plutonium purification cycle and three circulations of uranium purification cycle.
In the fission product of spent fuel, zirconium (Zr) yield is higher, account for gross activity than great, as
235u initial enrichment be 3.25% UO
2nuclear fuel element, when irradiation burnup is 33GWd/tU, cooling 3 years, the content of Zr was 3500g/tU.In spent fuel last handling process, after processing, codecontamination section (df of Zr is generally greater than to 10
3), a small amount of Zr is extracted by tributyl phosphate (TBP) together with plutonium with uranium, and in uranium plutonium separating technology, disperses to enter uranium purifying cycle and plutonium purification cycle.Because flow process is higher to the purity requirement of plutonium in final plutonium product, the whole process gross decontamination coefficient general requirement of the fission products such as Zr, Ru is greater than to 10
6.Therefore,, in plutonium purification cycle, on the basis of plutonium yield, need to further separate zirconium guaranteeing.
In current open source literature, relevant Zr is at HNO
3extraction behavior in solution has been reported, but is the experimental study of the above zirconium extraction process of gram liter level, research be the behavior of Zr in uranium plutonium coextraction process.And have not yet to see the report of Zr in plutonium purification cycle.
In the research work of plutonium purification cycle, find that Zr is difficult to reach the decontamination index (DF of expection
zr>100).And in plutonium purification cycle and codecontamination cycle, the purification of Zr has some serious difference: 1) in codecontamination cycle, Zr concentration (0.35g/gPu) is far above the Zr concentration (0.22mg/gPu) in Pu purification cycle, and concentration differs approximately 1600 times; 2) organic phase uranium plutonium saturation ratio higher (saturation ratio approximately 75%) in codecontamination cycle, be beneficial to the purification of Zr, and organic phase uranium plutonium saturation ratio lower (saturation ratio is less than 15%) in Pu purification cycle is unfavorable for the purification of Zr; 3) there is Zr (NO in codecontamination cycle
3)
4, TcO
4-coextraction, there is not TcO in Pu purification cycle
4-.Therefore, in plutonium purification cycle, the purification difficulty of micro-Zr is also increased greatly.
Summary of the invention
(1) goal of the invention
According to the existing problem of prior art, the invention provides the purification process of trace zirconium in a kind of Purex flow process, this technique is for plutonium purification cycle, and has plutonium yield and be greater than 99.9% and the df of the zirconium feature that is greater than 100.
(2) technical scheme
In order to solve the existing problem of prior art, the present invention is achieved by the following technical solutions:
The purification process of zirconium in plutonium purification cycle, this technique is with 30%TBP-kerosene, as extraction agent 2AX, to realize the recovery of uranium plutonium and the purification of fission product in 2A extractor, it comprises extraction section and washing section, and key is, in 2AF feed liquid, the concentration of nitric acid is 3.5～3.7mol/L; 2AS is that concentration is the HNO of 0.8～0.85mol/L
3solution.
Preferably, the extraction progression of described extraction section is 8～10 grades, and the washing progression of washing section is 6～8 grades.
Preferably, in described extraction section, the stream ratio of 2AF:2AX is 4～6.
Preferably, this technique at room temperature operates.
(3) beneficial effect
The purification process of zirconium in plutonium purification cycle, this technique is difficult to meet the high uranium plutonium rate of recovery and high zirconium df simultaneously.Adopt the purification process of zirconium in plutonium purification cycle provided by the invention, its have plutonium yield higher than 99.9% and the df of zirconium higher than 100 feature, this be mainly due to:
(1) in 2AF feed liquid, the concentration of acid is chosen as 3.5～3.7mol/L, though the range of choice of this concentration is narrow, only in this concentration range, guarantee uranium plutonium yield and zirconium enter water.Because the high extraction that is conducive to plutonium and uranium of acid concentration in 2AF feed liquid, but can increase the reservation of zirconium in organic phase simultaneously, therefore, concentration is chosen as 3.5～3.7mol/L.
(2) concentration of 2AS nitric acid is 0.8～0.85mol/L, in this concentration range, concentration can guarantee the purification to fission products such as zirconium and rutheniums lower than 0.85mol/L on the one hand, also can not cause because concentration of nitric acid is too low on the other hand the hydrolysis of plutonium in washing section and follow-up 2B technique.
(3) the few and stream of washing progression than minor benefit in uranium plutonium extraction, but also can increase the reservation of zirconium in organic phase, therefore washing progression is chosen as 6～8 grades, the stream of 2AF:2AX is than 4～6.
Accompanying drawing explanation
Fig. 1 is Zr concentration profiles at different levels.
Embodiment
Below in conjunction with specification drawings and specific embodiments, the present invention is further elaborated.
Embodiment 1
In temperature, be under the condition of 25 ℃, adopt mixing section volume 12ml, the mixer-settler that clarifying chamber's volume is 18ml is as 2A extraction tank, and organic phase is about 2:1 with the ratio of water, simulation plutonium purification cycle.The Pu of consisting of of 2AF feed liquid: 6.28g/L wherein
4+, the UO2 of 0.74g/L
2+, the HNO of 3.5mol/L
3and the Zr of 0.3073mg/L (adding U target to facilitate the analysis of carrying out Zr); 2AX is 30%TBP-kerosene; 2AS is that concentration is the nitric acid of 0.8mol/L; Feed liquid is 1min in the mixing section residence time, and extraction progression is 8 grades, and washing progression is 7 grades, and the stream of 2AF:2AX:2AS is than being 4:1:0.45.
During experiment, first with not uranium-bearing, plutonium and zirconium feed liquid, fill groove two hours, then add the feed liquid of uranium-bearing, plutonium and zirconium, after operation 5.0h, each gets up-to-date style at a flash for 2 hours, after operation 16h, stops groove analysis.
After balance, the yield of plutonium is 99.96%, and the material balance of uranium is 97.4%, and the material balance of nitric acid is 105.8%, and the material balance of plutonium is 98.2%, shows experimentation operational excellence, and experimental result is reliable.
In the instantaneous sample of 2AW and 2AP, Zr concentration analysis, in Table 1, owing to containing uranium, plutonium in test, can not adopt ICP-MS to analyze, therefore sneak into radioactivity in Zr
95zr, counts and determines that Zr concentration in solution, its specific radioactivity are 251.67Bq/mg by measuring its γ.Material balance that after balance, instantaneous sample Zr is 90.6%, Zr in Table the material balance of 1, Zr is poor is that it is right due to the Pu that contains higher concentration in sample
95the accuracy tool that the gamma spectrum of Zr is measured has a certain impact, but according to the material balance of uranium, acid, plutonium, thinks that Zr approaches balance, data are reliable and have certain representativeness.
The instantaneous sample concentration of Zr after table 1 balance
| The instantaneous sample of water | c(Zr)/(μg/L) | The instantaneous sample of organic phase | c(Zr)/(μg/L) |
| 2AW1 | 250.00 | 2AP1 | 5.24 |
| 2AW2 | 247.58 | 2AP2 | 6.24 |
| 2AW3 | 243.73 | 2AP3 | 6.20 |
| 2AW4 | 239.16 | 2AP4 | 6.16 |
| 2AW5 | 243.45 | 2AP5 | 7.15 |
| 2AW6 | 244.60 | 2AP6 | 8.38 |
Its concentration distribution at different levels are shown in Fig. 1, and as can be seen from Figure, few at organic phase exit end Zr content, these processing condition have realized the purification to sliver element zirconium preferably.And according to instantaneous sample zirconium concentration after balance, calculate the df DF of Zr
zr=163.
By the thermal test of plutonium purification cycle stand, show: in the situation that guaranteeing that Pu yield is greater than 99.9%, the df of Zr is greater than 100, higher than design objective, meets the needs of spent fuel aftertreatment flow process.
Embodiment 2
Method, the step used with embodiment 1 are identical, and different is that in 2AF feed liquid, concentration of nitric acid is 3.7mol/L, and 2AS is that concentration is the nitric acid of 0.85mol/L; The extraction progression of extraction section is 10 grades, and the washing progression of washing section is 8 grades; The stream ratio of 2AF:2AX is 6.
The demonstration of plutonium purification cycle stand thermal test result, the df that Pu yield is 99.97%, Zr is 159.
Claims (4)
- The purification process of zirconium in 1.Purex flow process plutonium purification cycle, this technique is with 30%TBP-kerosene, as extraction agent 2AX, to realize the recovery of uranium plutonium and the purification of fission product in 2A extractor, it comprises extraction section and washing section, it is characterized in that, in 2AF feed liquid, the concentration of nitric acid is 3.5～3.7mol/L; 2AS is that concentration is the HNO of 0.8～0.85mol/L 3solution.
- 2. the purification process of zirconium in Purex flow process plutonium purification cycle according to claim 1, is characterized in that, the extraction progression of described extraction section is 8～10 grades, and the washing progression of washing section is 6～8 grades.
- 3. the purification process of zirconium in Purex flow process plutonium purification cycle according to claim 1, is characterized in that, in described extraction section, the stream of 2AF:2AX ratio is 4～6.
- 4. the purification process of zirconium in Purex flow process plutonium purification cycle according to claim 1, is characterized in that, the service temperature of this technique is room temperature.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201410236283.XA CN104004927B (en) | 2014-05-30 | 2014-05-30 | The purification technique of zirconium in Purex flow plutonium purification cycle |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201410236283.XA CN104004927B (en) | 2014-05-30 | 2014-05-30 | The purification technique of zirconium in Purex flow plutonium purification cycle |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN104004927A true CN104004927A (en) | 2014-08-27 |
| CN104004927B CN104004927B (en) | 2016-09-14 |
Family
ID=51365833
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201410236283.XA Active CN104004927B (en) | 2014-05-30 | 2014-05-30 | The purification technique of zirconium in Purex flow plutonium purification cycle |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN104004927B (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN1277725A (en) * | 1997-10-31 | 2000-12-20 | 英国核燃料公共有限公司 | Nuclear fuel reprocessing |
| CN101484226A (en) * | 2006-07-03 | 2009-07-15 | 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 |
| CN103339269A (en) * | 2010-11-25 | 2013-10-02 | 原子能和替代能源委员会 | Process for separating americium from other metallic elements present in an acidic aqueous or organic phase and applications thereof |
-
2014
- 2014-05-30 CN CN201410236283.XA patent/CN104004927B/en active Active
Patent Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN1277725A (en) * | 1997-10-31 | 2000-12-20 | 英国核燃料公共有限公司 | Nuclear fuel reprocessing |
| CN101484226A (en) * | 2006-07-03 | 2009-07-15 | 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 |
| CN103339269A (en) * | 2010-11-25 | 2013-10-02 | 原子能和替代能源委员会 | Process for separating americium from other metallic elements present in an acidic aqueous or organic phase and applications thereof |
Non-Patent Citations (1)
| Title |
|---|
| 吴华武: "《核燃料化学工艺学》", 30 June 1986, article "purex流程及其工艺条件分析" * |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
| 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 |
Also Published As
| Publication number | Publication date |
|---|---|
| CN104004927B (en) | 2016-09-14 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| Alcock et al. | Tri-n-butyl phosphate as an extracting solvent for inorganic nitrates—V: Further results for the tetra-and hexavalent actinide nitrates | |
| CN103325431B (en) | A kind of PUREX flow process being separated technetium | |
| CN103305702B (en) | Process for recovering and purifying neptunium from waste liquor discharged from 2AW+2DW in Purex flow | |
| CN106893878B (en) | A method of recycling plutonium from radioactivity spentnuclear fuel | |
| ZA200809415B (en) | Process for reprocessing a spent nuclear fuel and of preparing a mixed uranium-plutonium oxide | |
| CN107130121B (en) | The uranium purification process of neptunium, plutonium is removed in a kind of nuclear fuel Purex post-processing process simultaneously | |
| CN104004928B (en) | A kind of purifying method for concentration of plutonium | |
| CN110144471B (en) | Method for extracting technetium from nuclear fuel post-treatment waste liquid | |
| RU2012157573A (en) | METHOD FOR PROCESSING WASTE NUCLEAR FUEL, NOT REQUIRING REDUCTIVE RE-EXTRACTION OF PLUTONIUM | |
| CN108389634A (en) | A kind of molten salt reactor thorium fuel circulation utilization method | |
| CN106555062B (en) | Uranium recovery method | |
| CN105734310B (en) | A kind of plutonium using dioxime Asia acid amides as reduction reextraction agent purifies method for concentration | |
| CN102943183B (en) | Method for extracting uranium and thorium from southern ionic rare-earth lixivium | |
| CN101471150B (en) | Uniform water solution nuclear reactor fuel solution purification process | |
| CN111863298B (en) | Deep purification method of PUREX process polluted solvent | |
| CN104004927A (en) | Process for purifying zirconium in Purex flow plutonium purification cycle | |
| CN111485125B (en) | Method for recovering technetium from spent fuel post-treatment waste liquid | |
| Vanel et al. | Modeling and flowsheet design of an Am separation process using TODGA and H4TPAEN | |
| CN103451428B (en) | Ruthenium purifying technology in plutonium purification cycle of Purex process | |
| CN108031288B (en) | Method for simultaneously producing radioactive strontium-90 and cesium-137 by adopting centrifugal extractor | |
| CN102936655A (en) | Technology for recovery of uranium in ceramic UO2 pellet by supercritical CO2 extraction | |
| Kumari et al. | Solvent extraction studies of protactinium for its recovery from short-cooled spent fuel and high-level waste solutions in thorium fuel cycle using diisobutyl carbinol (DIBC) as extractant | |
| Zverev et al. | Possible options for uranium-carbide SNF processing | |
| CN103345956A (en) | Preparation method of liquid alpha irradiation source | |
| CN104073656A (en) | Method for extracting strontium by sulfo-calixarene under acidic condition |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| C06 | Publication | ||
| PB01 | Publication | ||
| C10 | Entry into substantive examination | ||
| SE01 | Entry into force of request for substantive examination | ||
| C14 | Grant of patent or utility model | ||
| GR01 | Patent grant |