CN103345956A - Preparation method of liquid alpha irradiation source - Google Patents
Preparation method of liquid alpha irradiation source Download PDFInfo
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- CN103345956A CN103345956A CN2013102557373A CN201310255737A CN103345956A CN 103345956 A CN103345956 A CN 103345956A CN 2013102557373 A CN2013102557373 A CN 2013102557373A CN 201310255737 A CN201310255737 A CN 201310255737A CN 103345956 A CN103345956 A CN 103345956A
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- alpha irradiation
- irradiation source
- liquid alpha
- radiolysis
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Abstract
The invention relates to a preparation method of a liquid alpha irradiation source to solve the problem that an existing alpha irradiation source used for alpha radiolysis behavior research is low in dosage rate, can not meet the requirement for immediate measurement and can not reflect the true radiolysis condition of a measured object. The preparation method of the liquid alpha irradiation source includes a first step of conducting electrochemistry dissolution, namely, utilizing an electrolytic cell which is divided into a cathode chamber and an anode chamber through a partition film, and utilizing silver nitrate to conduct catalyzing, a second step of purifying, namely, removing silver through an anion exchange method, and a third step of calibrating the dosage rate. The liquid alpha irradiation source prepared through the preparation method of the liquid alpha irradiation source is high in dosage rate and capable of meeting the requirement of the alpha radiolysis behavior research of chemical reagents for immediacy of measurement in a post processing procedure, and reflects the radiolysis condition of the measured object in a relatively real mode. The liquid alpha irradiation source can also be suitable for the radiolysis research on other reagents in an aqueous phase and on various organic solvents.
Description
Technical field
The present invention relates to a kind of preparation method of radioactive source, particularly the preparation method in a kind of liquid alpha irradiation source.
Background technology
The object of spentnuclear fuel aftertreatment mainly is the spentnuclear fuel that nuclear power plant reactor draws off, and the characteristics of spentnuclear fuel maximum are to have extremely strong radioactivity.When spentnuclear fuel draws off in the reactor, no matter whether reach the burn-up level of design, always contain a certain amount of not fission nuclide and newly-generated nucleic.When spentnuclear fuel was carried out aftertreatment, the radioactivity in the aftertreatment flow process derived from the spentnuclear fuel
235U,
238U,
239Pu,
237Np,
241Am, and long-life of producing of nuclear fission, the fission product of high yield, as
95Zr,
95Nb (
95The decay daughter of Zr), ruthenium (
103Ru,
106Ru), some alkaline metal (as
137Cs), earth alkali metal (as
90Sr) and rare earth element (as
144Ce,
147Pm) etc.Because the existence of these nucleic, cause α, β, gamma activity in the aftertreatment flow process all very strong, therefore radiolysis can take place in the solvent that adopts and other chemical reagent usually, brings some harmful effects for the technological process of aftertreatment, as: the loss of (1) U, Pu; (2) decontamination factor of U, Pu product reduces; (3) extraction stage efficiency reduces, and precipitation, third phase, emulsification occur and is difficult to the equal situation of branch; (4) flow path efficiency reduces; (5) the flow process instability, have borderline risk etc.Because these influence the effect that direct relation aftertreatment, so the radiolysis research of chemical reagent has very important meaning in the aftertreatment technology.
After the Purex flow process that with TBP/ kerosene is extractant was born, the variation of essence did not take place in aftertreatment flow process so far, moved in the world at present and adopt this class flow process in the aftertreatment factory that founds a capital.The Purex flow process is extractant with tributyl phosphate (TBP), and kerosene or n-dodecane are thinning agent, HNO
3Be salting-out agents, the difference of utilizing U, Pu and fission product to extract behavior each other realizes separation and the purification of uranium plutonium.The Purex flow process comprises the purification of uranium plutonium codecontamination separation, uranium and three extraction cycle of purification of plutonium.In the codecontamination separating cycle, utilize extractant to realize that uranium, plutonium separate with fission product, and the separation between the uranium, plutonium.The decontamination cycle of uranium is finished the further purification of uranium product liquid to remove fission product and trace plutonium.The plutonium lean solution of being come by the codecontamination separating cycle enters plutonium behind condiment decontamination cycle utilizes the further purification of reduction reextraction agent realization plutonium product liquid with concentrated, and therefore in the plutonium purification cycle, the irradiation that solution is subjected to mainly is alpha activity.Along with the raising of reactor fuel enrichment, the raising of average specific power and the intensification of burnup, the specific radioactivity in the spentnuclear fuel, decay heat, plutonium content and other transuranic element content all significantly increase.Therefore, based on the demand of reality, the α radiolysis behavioral study of chemical reagent is demanded urgently carrying out in the aftertreatment flow process.
At present, both at home and abroad for the most airtight solid-state alpha irradiation sources of adopting of the research of α radiolysis behavior, as
244Cm,
210Po or
238Pu microballoon etc.Yet the sealing means that adopt in airtight solid-state alpha irradiation source can reduce the energy of α particle significantly, and sealant is more thick, and the alpha-particle energy that discharges is more low, and for example thickness is that 0.17 millimeter platinum coating can make
210The energy of the α particle that Po produces reduces only about half of, and therefore the dose rate in airtight solid-state alpha irradiation source is very low.If use it for the α radiolysis behavioral study of chemical reagent in the aftertreatment flow process, for satisfying the requirement of irradiation dose, then must need a large amount of radiolysis time, can not satisfy instant Testing requirement like this, can be subjected to serious chemical reaction again disturbs, therefore the true radiolysis situation that can't reflect measurand needs a kind of new alpha irradiation source of development badly to satisfy the α radiolysis behavioral study of chemical reagent in the aftertreatment flow process.
Summary of the invention
Low for solving existing alpha irradiation source dose rate for α radiolysis behavioral study, can not satisfy instant Testing requirement, the problems such as true radiolysis situation that can't reflect measurand the invention provides the preparation method in a kind of liquid alpha irradiation source, and this method may further comprise the steps:
1) electrochemical dissolution
Employing is the electrochemical dissolution device with barrier film with the electrolytic tank that cathode chamber and anode chamber separate, and described barrier film is for can guarantee Pu (NO in the electrochemical dissolution process
3)
2- 6, Ag
2+And Ag
+Can not pass through, and NO
3 -And H
+The barrier film that can pass through; With solid
238PuO
2Put into the anode chamber, and the indoor adding of anode contains the nitric acid of silver nitrate, described silver nitrate is used for the catalysis of reaction, and its concentration is greater than 0.05mol/L, and the concentration of described nitric acid is 6-8mol/L; Adding concentration in the cathode chamber is the nitric acid of 6-8mol/L; Energising makes solid
238PuO
2Contained after the dissolving
238Pu solution;
2) purifying
With containing of obtaining
238Pu solution feeds anion-exchange column, because
238Pu is present in the complex anion, therefore be attracted on the anion-exchange column, and silver flows out with cationic form, thereby silver is removed; Nitric acid with 0.25-0.5mol/L will be adsorbed on the anion-exchange column then
238Pu elutes, and is contained
238The Pu storing solution;
3) dose rate is demarcated
To containing
238The dose rate of Pu storing solution is demarcated, and obtains the liquid alpha irradiation source that dose rate is determined.
Described electrolytic tank is preferably H type electrolytic tank;
Described barrier film is preferably and adopts specification is the barrier film that the glass sand of G5 is made;
The packing material of described anion-exchange column is preferably 2606 type anion exchange resins;
Described containing
238The demarcation of the dose rate of Pu storing solution adopts Fricke dosage to count preferably.
In the electrochemical dissolution process, Ag
+Under the effect of electric current, generate solid in the anode chamber
238PuO
2Dissolving have the Ag of catalytic action
2+, under catalytic action and acid condition, solid
238PuO
2Reaction generates Pu (NO
3)
2- 6Thereby complex anion is dissolved.
In purge process, plutonium is with Pu (NO
3)
2- 6The form of complex anion is adsorbed on the anion-exchange column, and silver can not form complex anion in anion exchange procedures, only can be with Ag
+Form flows out, thereby realizes the separation of the two.
The preparation method in liquid alpha irradiation of the present invention source is with solid
238PuO
2Be the liquid alpha irradiation of feedstock production source, prepared liquid alpha irradiation source dose rate height, and can adjust dose rate according to practical application easily, can satisfy the instantaneity requirement of α radiolysis behavioral study to measuring of chemical reagent in the aftertreatment flow process, the radiolysis time can be very short in addition, suffered chemical reaction disturbs less, therefore can reflect the radiolysis situation of measurand comparatively really.The α radiolysis behavioral study of chemical reagent, liquid alpha irradiation of the present invention source can also be applicable to the radiolysis research of other reagent of aqueous phase and various organic solvents in the aftertreatment flow process.
Embodiment
Below in conjunction with specific embodiment embodiments of the present invention are described further.
Embodiment
The preparation method in a kind of liquid alpha irradiation source, this method may further comprise the steps:
1) electrochemical dissolution
Employing is the electrochemical dissolution device with barrier film with the H type electrolytic tank that cathode chamber and anode chamber separate, and it is the barrier film that the glass sand of G5 is made that described barrier film adopts specification; With solid
238PuO
2Put into the anode chamber, and the indoor adding of anode contains the nitric acid of silver nitrate, described silver nitrate is used for the catalysis of reaction, and its concentration is 0.1mol/L, and the concentration of described nitric acid is 6.0mol/L; Adding concentration in the cathode chamber is the nitric acid of 6.12mol/L; Energising makes solid
238PuO
2Contained after the dissolving
238Pu solution;
2) purifying
With containing of obtaining
238Pu solution feeds anion-exchange column, and the packing material of anion-exchange column is 2606 type anion exchange resins, because
238Pu is present in the complex anion, therefore be attracted on the anion-exchange column, and silver flows out with cationic form, thereby silver is removed; Nitric acid with 0.35mol/L will be adsorbed on the anion-exchange column then
238Pu elutes, and is contained
238The Pu storing solution;
3) dose rate is demarcated
Adopt the Fricke quantimeter to containing
238The dose rate of Pu storing solution is demarcated, and obtains the liquid alpha irradiation source that dose rate is determined.
Liquid alpha irradiation of the present invention source successfully has been used for solvent alpha irradiation dosage in the research of α radiolysis, organic phase of aftertreatment flow process chemical reagent influences many-sided research work such as research to what plutonium kept, has obtained good effect.
Claims (5)
1. the preparation method in a liquid alpha irradiation source is characterized in that this method may further comprise the steps:
1) electrochemical dissolution
Employing is the electrochemical dissolution device with barrier film with the electrolytic tank that cathode chamber and anode chamber separate, and described barrier film is for can guarantee Pu (NO in the electrochemical dissolution process
3)
2- 6, Ag
2+And Ag
+Can not pass through, and NO
3 -And H
+The barrier film that can pass through; With solid
238PuO
2Put into the anode chamber, and the indoor adding of anode contains the nitric acid of silver nitrate, described silver nitrate is used for the catalysis of reaction, and its concentration is greater than 0.05mol/L, and the concentration of described nitric acid is 6-8mol/L; Adding concentration in the cathode chamber is the nitric acid of 6-8mol/L; Energising makes solid
238PuO
2Contained after the dissolving
238Pu solution;
2) purifying
With containing of obtaining
238Pu solution feeds anion-exchange column, because
238Pu is present in the complex anion, therefore be attracted on the anion-exchange column, and silver flows out with cationic form, thereby silver is removed; Nitric acid with 0.25-0.5mol/L will be adsorbed on the anion-exchange column then
238Pu elutes, and is contained
238The Pu storing solution;
3) dose rate is demarcated
To containing
238The dose rate of Pu storing solution is demarcated, and obtains the liquid alpha irradiation source that dose rate is determined.
2. the preparation method in liquid alpha irradiation as claimed in claim 1 source, it is characterized in that: described electrolytic tank is H type electrolytic tank.
3. the preparation method in liquid alpha irradiation as claimed in claim 1 source is characterized in that: described barrier film is the barrier film that the glass sand of G5 is made for adopting specification.
4. the preparation method in liquid alpha irradiation as claimed in claim 1 source, it is characterized in that: the packing material of described anion-exchange column is 2606 type anion exchange resins.
5. the preparation method in liquid alpha irradiation as claimed in claim 1 source is characterized in that: described containing
238The Fricke quantimeter is adopted in the demarcation of the dose rate of Pu storing solution.
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Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104237449A (en) * | 2014-08-21 | 2014-12-24 | 浙江大学 | Barium carbonate electrolytic dissolution device |
| CN108977659A (en) * | 2018-08-20 | 2018-12-11 | 中国原子能科学研究院 | Plutonium, palladium, silver, cadmium, tin and antimony group separation method |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3976775A (en) * | 1975-05-06 | 1976-08-24 | The United States Of America As Represented By The United States Energy Research And Development Administration | Method for dissolving plutonium dioxide |
| WO1989010981A1 (en) * | 1988-05-03 | 1989-11-16 | Battelle Memorial Institute | Apparatus and method for dissolving hazardous waste materials by catalyzed electrochemical dissolution |
-
2013
- 2013-06-26 CN CN2013102557373A patent/CN103345956A/en active Pending
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3976775A (en) * | 1975-05-06 | 1976-08-24 | The United States Of America As Represented By The United States Energy Research And Development Administration | Method for dissolving plutonium dioxide |
| WO1989010981A1 (en) * | 1988-05-03 | 1989-11-16 | Battelle Memorial Institute | Apparatus and method for dissolving hazardous waste materials by catalyzed electrochemical dissolution |
Non-Patent Citations (3)
| Title |
|---|
| SIMON COTTON: "《Lanthanide and Actinide Chemistry》", 31 December 2006, JOHN WILEY & SONS, LTD * |
| 王兴尧等: "催化电解法溶解钚的氧化物", 《原子能科学技术》 * |
| 罗文宗，张文青: "《钚的分析化学》", 30 June 1991, 原子能出版社 * |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104237449A (en) * | 2014-08-21 | 2014-12-24 | 浙江大学 | Barium carbonate electrolytic dissolution device |
| CN104237449B (en) * | 2014-08-21 | 2016-06-08 | 浙江大学 | Brium carbonate electrodissolution device |
| CN108977659A (en) * | 2018-08-20 | 2018-12-11 | 中国原子能科学研究院 | Plutonium, palladium, silver, cadmium, tin and antimony group separation method |
| CN108977659B (en) * | 2018-08-20 | 2020-04-14 | 中国原子能科学研究院 | Method for separating plutonium, palladium, silver, cadmium, tin and antimony |
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Application publication date: 20131009 |