CN115385302A - Tritium recovery and purification system and method for waste neutron target - Google Patents
Tritium recovery and purification system and method for waste neutron target Download PDFInfo
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- CN115385302A CN115385302A CN202210898869.7A CN202210898869A CN115385302A CN 115385302 A CN115385302 A CN 115385302A CN 202210898869 A CN202210898869 A CN 202210898869A CN 115385302 A CN115385302 A CN 115385302A
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- 229910052722 tritium Inorganic materials 0.000 title claims abstract description 342
- YZCKVEUIGOORGS-NJFSPNSNSA-N Tritium Chemical compound [3H] YZCKVEUIGOORGS-NJFSPNSNSA-N 0.000 title claims abstract description 339
- 238000011084 recovery Methods 0.000 title claims abstract description 42
- 238000000746 purification Methods 0.000 title claims abstract description 39
- 238000000034 method Methods 0.000 title claims abstract description 27
- 239000002699 waste material Substances 0.000 title claims abstract description 26
- 238000000926 separation method Methods 0.000 claims abstract description 158
- 239000007789 gas Substances 0.000 claims abstract description 108
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 claims abstract description 64
- 229910052739 hydrogen Inorganic materials 0.000 claims abstract description 64
- 239000001257 hydrogen Substances 0.000 claims abstract description 64
- 238000003860 storage Methods 0.000 claims abstract description 37
- 239000002245 particle Substances 0.000 claims abstract description 30
- 239000011232 storage material Substances 0.000 claims abstract description 30
- 229910001252 Pd alloy Inorganic materials 0.000 claims abstract description 29
- 239000012528 membrane Substances 0.000 claims abstract description 22
- 238000000197 pyrolysis Methods 0.000 claims abstract description 18
- 238000000576 coating method Methods 0.000 claims description 25
- 239000011248 coating agent Substances 0.000 claims description 22
- 238000010438 heat treatment Methods 0.000 claims description 20
- 229910045601 alloy Inorganic materials 0.000 claims description 18
- 239000000956 alloy Substances 0.000 claims description 18
- 239000000463 material Substances 0.000 claims description 15
- 238000004891 communication Methods 0.000 claims description 9
- 239000011261 inert gas Substances 0.000 claims description 6
- 239000012535 impurity Substances 0.000 claims description 5
- KDLHZDBZIXYQEI-UHFFFAOYSA-N palladium Substances [Pd] KDLHZDBZIXYQEI-UHFFFAOYSA-N 0.000 claims description 4
- 238000005524 ceramic coating Methods 0.000 claims description 3
- 238000002955 isolation Methods 0.000 claims description 3
- 150000004767 nitrides Chemical class 0.000 claims description 3
- 239000011224 oxide ceramic Substances 0.000 claims description 3
- 229910052574 oxide ceramic Inorganic materials 0.000 claims description 3
- 238000010926 purge Methods 0.000 claims description 3
- 230000008569 process Effects 0.000 abstract description 6
- 230000005445 isotope effect Effects 0.000 abstract description 3
- 230000035699 permeability Effects 0.000 abstract description 2
- 238000001179 sorption measurement Methods 0.000 abstract description 2
- 239000002808 molecular sieve Substances 0.000 description 10
- URGAHOPLAPQHLN-UHFFFAOYSA-N sodium aluminosilicate Chemical compound [Na+].[Al+3].[O-][Si]([O-])=O.[O-][Si]([O-])=O URGAHOPLAPQHLN-UHFFFAOYSA-N 0.000 description 10
- YZCKVEUIGOORGS-OUBTZVSYSA-N Deuterium Chemical compound [2H] YZCKVEUIGOORGS-OUBTZVSYSA-N 0.000 description 8
- 229910052805 deuterium Inorganic materials 0.000 description 8
- 230000009286 beneficial effect Effects 0.000 description 7
- 239000010936 titanium Substances 0.000 description 7
- 229910052719 titanium Inorganic materials 0.000 description 7
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 6
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 description 5
- 239000007788 liquid Substances 0.000 description 5
- 239000000758 substrate Substances 0.000 description 5
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- 238000012544 monitoring process Methods 0.000 description 3
- 229910052757 nitrogen Inorganic materials 0.000 description 3
- YZCKVEUIGOORGS-IGMARMGPSA-N Protium Chemical compound [1H] YZCKVEUIGOORGS-IGMARMGPSA-N 0.000 description 2
- 241000720974 Protium Species 0.000 description 2
- 238000006243 chemical reaction Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 239000001307 helium Substances 0.000 description 2
- 229910052734 helium Inorganic materials 0.000 description 2
- SWQJXJOGLNCZEY-UHFFFAOYSA-N helium atom Chemical compound [He] SWQJXJOGLNCZEY-UHFFFAOYSA-N 0.000 description 2
- 238000010884 ion-beam technique Methods 0.000 description 2
- 238000005372 isotope separation Methods 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
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- 230000002285 radioactive effect Effects 0.000 description 2
- 238000004064 recycling Methods 0.000 description 2
- 239000002912 waste gas Substances 0.000 description 2
- 229910000619 316 stainless steel Inorganic materials 0.000 description 1
- RYGMFSIKBFXOCR-UHFFFAOYSA-N Copper Chemical compound [Cu] RYGMFSIKBFXOCR-UHFFFAOYSA-N 0.000 description 1
- 229910001182 Mo alloy Inorganic materials 0.000 description 1
- 206010028980 Neoplasm Diseases 0.000 description 1
- 229910001069 Ti alloy Inorganic materials 0.000 description 1
- 229910052770 Uranium Inorganic materials 0.000 description 1
- 229910000756 V alloy Inorganic materials 0.000 description 1
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 1
- 229910001093 Zr alloy Inorganic materials 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- 238000000516 activation analysis Methods 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 238000009835 boiling Methods 0.000 description 1
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- 238000009833 condensation Methods 0.000 description 1
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- 229910052802 copper Inorganic materials 0.000 description 1
- 239000008358 core component Substances 0.000 description 1
- 238000013461 design Methods 0.000 description 1
- 125000004431 deuterium atom Chemical group 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000007710 freezing Methods 0.000 description 1
- 230000008014 freezing Effects 0.000 description 1
- 125000004435 hydrogen atom Chemical group [H]* 0.000 description 1
- 238000007689 inspection Methods 0.000 description 1
- 238000005259 measurement Methods 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 239000012466 permeate Substances 0.000 description 1
- 238000005371 permeation separation Methods 0.000 description 1
- 239000000047 product Substances 0.000 description 1
- 238000005086 pumping Methods 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 150000003384 small molecules Chemical class 0.000 description 1
- 239000002910 solid waste Substances 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
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Classifications
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- C—CHEMISTRY; METALLURGY
- C01—INORGANIC CHEMISTRY
- C01B—NON-METALLIC ELEMENTS; COMPOUNDS THEREOF; METALLOIDS OR COMPOUNDS THEREOF NOT COVERED BY SUBCLASS C01C
- C01B4/00—Hydrogen isotopes; Inorganic compounds thereof prepared by isotope exchange, e.g. NH3 + D2 → NH2D + HD
-
- 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
- Y02E30/10—Nuclear fusion reactors
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- Chemical & Material Sciences (AREA)
- Organic Chemistry (AREA)
- Health & Medical Sciences (AREA)
- General Health & Medical Sciences (AREA)
- Inorganic Chemistry (AREA)
- Separation Using Semi-Permeable Membranes (AREA)
Abstract
The invention relates to the technical field of radiochemistry, in particular to a tritium recovery and purification system and method for a waste neutron target. The system comprises a main pipeline, wherein an air inlet and an air outlet are respectively arranged at two ends of the main pipeline, and the middle part of the main pipeline is respectively communicated with a pyrolysis cavity and a storage cavity; the gas outlet of the main pipeline is communicated with the tritium separation system; the middle part of the main pipeline is also communicated with a vacuum pump set; hydrogen storage material particles are arranged in the storage cavity; the tritium separation system comprises at least one tritium separation unit, and a palladium alloy membrane is fixedly arranged in each tritium separation unit; the palladium alloy membrane separates the inner cavity of the tritium separation unit into a tritium-rich cavity and a tritium-poor cavity. The system adopts hydrogen storage material particles to only adsorb hydrogen isotope gas and directly obtains high-purity hydrogen isotope gas through a pyrolysis-adsorption circulation process; meanwhile, the hydrogen permeability and isotope effect of the palladium alloy are adopted to realize the enrichment of tritium. The method greatly improves the separation coefficient and the recovery efficiency of tritium.
Description
Technical Field
The invention relates to the technical field of radiochemistry, in particular to a tritium recovery and purification system and method for a waste neutron target.
Background
Neutron technology is widely applied to the fields of neutron photography, oil field well logging, cancer treatment, neutron security inspection, activation analysis, irradiation breeding, irradiation damage research, nuclear measurement and the like. The major neutron sources currently available are nuclear reactors, radioactive sources and accelerator type neutron sources. The accelerator type deuterium-tritium neutron source bombards a neutron target through accelerating deuterium ion beams to generate a fusion neutron of 14.1MeV through a deuterium-tritium fusion reaction. The accelerator type deuterium-tritium neutron source can generate neutrons only by electrifying, so that the safety is high. In addition, the neutron source has the advantages of miniaturization, portability, low cost and the like, and has wide application prospect. In an accelerator-type deuterium-tritium neutron source, the neutron target is the core component that directly generates neutrons. Because titanium has higher tritium absorption density and stability, the most used neutron target at home and abroad is a titanium film plated on a substrate material, and tritium is absorbed by the titanium film to obtain the tritium-titanium target.
Tritium is an important strategic resource, which is very scarce and expensive, and the price of 1 g of tritium is as high as ten million. In the process of using the tritium-titanium target on an accelerator neutron source, only the place bombarded by the deuterium ion beam participates in the fusion reaction of deuterium and tritium, so that the consumed tritium is far lower than the stored tritium in the neutron target. The neutron target reaching the service life is often treated as radioactive solid waste, so that the environment is polluted, and tritium resources are wasted.
In the recovery and purification of tritium, purification and isotope separation are key steps.
In the prior art, the purification method mainly utilizes the difference of condensation temperatures of different gas molecules to obtain high-purity hydrogen isotope gas. The low-temperature molecular sieve separation system mainly comprises a normal-temperature molecular sieve adsorber, a liquid nitrogen molecular sieve adsorber and a liquid helium temperature molecular sieve adsorber. Specifically, by utilizing a low-temperature molecular sieve separation system, gas released by heating of the waste neutron target is adsorbed by a molecular sieve through a low-temperature molecular sieve condensed by liquid nitrogen due to low-temperature liquefaction of oxygen, nitrogen, water vapor and the like; the residual gas passes through the molecular sieve condensed by liquid helium, hydrogen isotopes are condensed to liquid state and adsorbed in the molecular sieve, and two low-temperature molecular sieves with cooling systems are needed for one-time purification. In order to obtain hydrogen isotope gas with higher purity, a plurality of low-temperature classifying screens are required to be connected in series to realize multi-stage purification, the equipment is large, the process is complex, and the separation coefficient is low. And because the boiling point of the hydrogen isotope is lower, the low-temperature freezing cost is higher.
For isotope separation, a common solution in the prior art is cryogenic rectification to separate hydrogen isotopes. The cryogenic rectification system mainly comprises a refrigerator, a rectification column, a circulating pump, a balancer and the like, wherein the most critical part is the rectification column. In order to improve efficiency and purity, the rectification column generally needs more than 2 columns, each 2-3m. Therefore, the whole equipment is huge and complex to operate.
Disclosure of Invention
The invention aims to solve the technical problem of providing a tritium recovery and purification system and method for a waste neutron target.
The technical scheme for solving the technical problems is as follows:
the invention provides a tritium recovery and purification system for a waste neutron target, which comprises a main pipeline, wherein the two ends of the main pipeline are respectively provided with an air inlet and an air outlet, and the middle part of the main pipeline is respectively communicated with a pyrolysis cavity and a storage cavity; the gas outlet of the main pipeline is communicated with a tritium separation system; the middle part of the main pipeline is also communicated with a vacuum pump set; hydrogen storage material particles are arranged in the storage cavity; the tritium separation system comprises at least one tritium separation unit, and a palladium alloy film is fixedly installed in each tritium separation unit; the palladium alloy membrane separates the inner cavity of the tritium separation unit into a tritium-rich cavity and a tritium-poor cavity.
The technical scheme of the invention has the beneficial effects that:
the invention adopts the principle that hydrogen storage material particles only adsorb hydrogen isotope gas, only a storage cavity with a built-in hydrogen storage alloy is designed in the system, and the high-purity hydrogen isotope gas is directly obtained through a pyrolysis-adsorption circulation process; meanwhile, by adopting the hydrogen permeability and isotope effect of palladium alloy, according to different degrees of easiness of protium, deuterium and tritium for permeating a palladium alloy film, protium and deuterium of small molecules preferentially pass through, and tritium of larger molecules is difficult to permeate through, so that tritium enrichment is realized. Meanwhile, by combining the multi-stage circulating permeation separation design, the separation coefficient and the recovery efficiency of tritium are greatly improved.
The invention can also be implemented by the following technical scheme:
further, the tritium separation system includes at least two tritium separation units; the gas outlet of the main pipeline is communicated with a tritium-rich cavity of one tritium separation unit; the tritium-rich cavity of one tritium separation unit is communicated with the tritium-rich cavity of the other tritium separation unit through a pipeline, and the tritium-poor cavity of the other tritium separation unit is also communicated with the tritium-rich cavity of the one tritium separation unit through another pipeline; the tritium-poor cavity of one tritium separation unit is also communicated with an exhaust gas collection tank; the tritium-rich chamber of the other tritium separation unit is also in communication with the tritium bed.
The beneficial effect of adopting the above further scheme is: the hydrogen isotope mixed gas can be circularly separated for many times between two tritium separation units, so that the separation effect is better, and the purity of the recovered tritium gas is higher.
Furthermore, a one-way valve is arranged on a pipeline for communicating the tritium-rich cavity of one tritium separation unit with the tritium-rich cavity of the other tritium separation unit.
The beneficial effect of adopting the further scheme is that: through setting up the check valve, make the tritium-containing gas only can flow to the rich tritium chamber of next-level tritium separation unit from last one-level poor tritium chamber, can not the backward flow.
Furthermore, a valve and a tritium pump are arranged on a pipeline for communicating the tritium-rich cavity of one tritium separation unit with the tritium-rich cavity of the other tritium separation unit; the hydrogen storage material particles are arranged in the tritium pump; and a tritium measuring instrument is arranged on each tritium-rich cavity.
The beneficial effect of adopting the above further scheme is: the tritium measuring instrument can monitor the tritium content in the tritium-rich cavity in real time, after the tritium content reaches a set limit, the tritium measuring instrument feeds a measured signal back to the control system, the control system opens a corresponding valve, and tritium-containing gas in the tritium-rich cavity of the separation unit is absorbed through a hydrogen storage material in the tritium pump; then, closing the valve, and heating the tritium pump to release all tritium adsorbed in the tritium pump to the tritium-rich cavity of the last-stage separation unit for further separation; the whole process can be fully automatically controlled by feeding back a monitoring signal of the tritium content through the tritium measuring instrument.
Further, the palladium alloy membrane comprises a support material and an alloy coating coated on the support material; the alloy coating is made of PdAuNiCuY alloy; wherein, the Ag content is 15-30 a t.%, the Au content is 0-10 at.%, the Cu content is 0-10 at.%, the Ni content is 0-2 at.%, the Y content is 0-2 at.%, and the rest is Pd.
The beneficial effect of adopting the further scheme is that: the palladium alloy membrane of the above composition has good separation efficiency and strength.
Further, the thickness of the alloy coating is 1-1000 μm.
Further, the inner surfaces of the main pipeline, the pyrolysis cavity, the storage cavity, the vacuum pump set and the tritium separation system are coated with tritium-resistant coatings; the tritium-resistant coating comprises one of an oxide ceramic coating, a diamond-like coating, a SiC coating and a nitride coating.
The beneficial effect of adopting the above further scheme is: by applying a tritium-resistant coating, damage to the system by tritium in the gas can be prevented.
Further, the particle size of the hydrogen storage material particles is 1-5 mm; an isolation net of 100-500 meshes is fixed at the communication part of the storage cavity and the main pipeline.
The beneficial effect of adopting the further scheme is that: by providing the separation net, the particles of the hydrogen storage material can be prevented from entering the main pipeline or other structures.
The invention provides a method for recovering and purifying tritium of a waste neutron target by adopting the system, which comprises the following steps:
s1, closing an air inlet and an air outlet of the main pipeline; putting the waste neutron target into the pyrolysis cavity for heating to obtain gas released by the waste neutron target;
s2, the gas enters the storage cavity through the main pipeline, and the hydrogen storage material particles in the storage cavity adsorb hydrogen isotope mixed gas in the gas;
s3, starting the vacuum pump set, recovering impurity gases which are not adsorbed by the hydrogen storage material particles in the main pipeline and the storage cavity, and closing the vacuum pump set after recovery;
s4, heating the storage cavity to 300-500 ℃, opening a gas outlet of the main pipeline, and preserving heat for 10-40 min to enable the hydrogen storage material particles to release the hydrogen isotope gas;
s5, the hydrogen isotope gas enters one tritium separation unit of the tritium separation system through the main pipeline and is separated into tritium-rich gas and tritium-poor gas by the palladium alloy membrane; the tritium-rich gas flows out through the tritium-rich cavity and is recycled, and the tritium-poor gas flows out through the tritium-poor cavity and is recycled.
The recovery and purification method can effectively solve the problem of recycling tritium in the waste neutron target, the recovery rate reaches more than 90%, and the purity of the recovered tritium gas reaches more than 99.9%.
Further, in the step S1, the heating temperature of the waste neutron target is 300-600 ℃, the heating rate is 3-10 ℃/min, and the heat preservation time is 40-80 min.
Further, in the step S3, the vacuum pump set is started, and when the vacuum degree in the system is smaller than 10 -6 Pa, continuously keeping the vacuum pump group on for 30min, and then closing the vacuum pump group.
Further, after the step S5 is finished, the air inlet of the main pipeline is opened, and the air outlet of the main pipeline is closed; starting the vacuum pump set; and introducing inert gas into the system for purging through the gas inlet of the main pipeline, so that the inert gas carries the residual tritium gas in the system to flow out through the vacuum pump set, and purifying the flowing gas.
Drawings
FIG. 1 is a schematic diagram of the tritium recovery and purification system for spent neutron targets of the present invention;
fig. 2 is a schematic structural diagram of a five-stage tritium separation system in a tritium recovery and purification system for a spent neutron target according to an embodiment of the present invention.
In the drawings, the components represented by the respective reference numerals are listed below:
1. a pyrolysis chamber; 2. a storage chamber;
3. a tritium separation system; 30. a palladium alloy film; 31. a first tritium separation unit; 32. a second tritium separation unit; 33. a third tritium separation unit; 34. a fourth tritium separation unit; 35. a fifth tritium separation unit;
4. a waste gas collection tank; 5. a tritium bed;
6. a valve;
7. a vacuum pump set;
8. a purification device;
9. a one-way valve;
10. a tritium pump;
11. a tritium measuring instrument;
12. and (4) a main pipeline.
Detailed Description
The principles and features of this invention are described below in conjunction with the following drawings, which are set forth to illustrate, but are not to be construed to limit the scope of the invention.
As shown in fig. 1, the device comprises a main pipeline 12, wherein both ends of the main pipeline 12 are respectively provided with an air inlet and an air outlet, and the middle part of the main pipeline 12 is respectively communicated with a pyrolysis cavity 1 and a storage cavity 2; the air outlet of the main pipeline 12 is communicated with the tritium separation system 3; the middle part of the main pipeline 12 is also communicated with a vacuum pump set 7; hydrogen storage material particles are arranged in the storage cavity 2; the tritium separation system 3 comprises at least one tritium separation unit, and a palladium alloy membrane 30 is fixedly arranged in each tritium separation unit; a palladium alloy membrane 30 separates the inner cavity of the tritium separation unit into a tritium-rich chamber and a tritium-poor chamber.
According to the tritium recovery and purification system, the pyrolysis cavity 1 is used for heating the waste neutron target, so that the waste neutron target releases tritium-containing mixed gas; the tritium-containing mixed gas enters the storage cavity 2 through the main pipeline 12, and hydrogen isotopes in the mixed gas can be absorbed by hydrogen storage material particles in the storage cavity 2; the unabsorbed impurities in the mixed gas can be separated out of the system by the vacuum pump unit 7 through vacuum pumping. Then, heating the storage cavity 2 to make the hydrogen storage material particles release hydrogen isotope mixed gas; the hydrogen isotope mixed gas enters the tritium separation system 3 through the gas outlet of the main pipeline 12 and is separated into tritium-rich gas and tritium-poor gas by a palladium alloy membrane 30 in the tritium separation unit; in this way, tritium in the spent neutron target can be recovered and purified.
The tritium recovery and purification system has the advantages of simple structure, low cost, high tritium recovery efficiency and the like, can effectively solve the problem of recycling tritium in the waste neutron target, and has the recovery rate of more than 90 percent and the purity of the recovered tritium gas of more than 99.9 percent.
According to the palladium alloy membrane 30 adopted by the invention, due to the isotope effect, in the hydrogen isotope mixed gas, lighter hydrogen atoms and deuterium atoms can more easily penetrate through the palladium alloy membrane 30, and heavier tritium atoms can not easily penetrate through the palladium alloy membrane 30, so that hydrogen and deuterium can be enriched in a tritium-poor cavity through the palladium alloy membrane 30, and tritium can be enriched in a tritium-rich cavity.
The hydrogen storage material particles adopted by the invention are hydrogen storage alloy, and specifically are one or the combination of more than two of uranium, zirconium alloy, titanium alloy, laNi alloy, vanadium alloy and palladium alloy; these alloys can effectively absorb hydrogen isotopes in a mixed gas.
Preferably, the pyrolysis chamber 1 is close to the inlet of the main conduit 12 and the storage chamber 2 is close to the outlet of the main conduit 12.
Preferably, the storage cavity 2 can be a plurality of cavities, and a plurality of storage cavities 2 can be connected to the main pipeline 12 in parallel; the hydrogen storage material particles in the plurality of storage chambers 2 may be of different materials.
In the above scheme, it is preferable that the tritium separation system 3 includes at least two tritium separation units.
Specifically, taking two tritium separation units as an example, the gas outlet of the main pipeline 12 is communicated with the tritium-rich cavity of the first tritium separation unit; the tritium-rich cavity of the first tritium separation unit is also communicated with the tritium-rich cavity and the tritium-poor cavity of the second tritium separation unit through two pipelines respectively; the tritium-poor cavity of the first tritium separation unit is also communicated with an exhaust gas collection tank 4; the tritium-rich chamber of the second tritium separation unit is also in communication with tritium bed 5. The hydrogen isotope mixed gas entering the first tritium separation unit through the main pipeline 12 firstly enters a tritium-rich cavity of the tritium separation unit, the tritium-rich gas enriched in the tritium-rich cavity of the first tritium separation unit circulates to the tritium-rich cavity of the second tritium separation unit through a pipeline, and is separated again by a palladium alloy membrane 30 in the second tritium separation unit; tritium-poor gas enriched in the tritium-poor cavity of the second tritium separation unit is continuously circulated to the first tritium separation unit through a pipeline and is continuously separated by the palladium alloy membrane 30 in the first tritium separation unit.
Therefore, the hydrogen isotope mixed gas can be circularly separated for multiple times between the two tritium separation units, so that the separation effect is better, and the purity of the recovered tritium gas is higher.
It should be noted that the gas outlet of the main conduit 12 may be directly connected to the tritium-enriched chamber of the first tritium separation unit.
Preferably, a valve 6 and a tritium pump 10 are arranged on a pipeline for communicating the tritium-rich cavity of one tritium separation unit with the tritium-rich cavity of the other tritium separation unit; hydrogen storage material particles are provided within the tritium pump 10. A one-way valve 9 is arranged on a pipeline for communicating the tritium-rich cavity of one tritium separation unit with the tritium-rich cavity of the other tritium separation unit, and each tritium-rich cavity is provided with a tritium measuring instrument 11; and each valve 6, each one-way valve 9, each tritium pump 10 and each tritium measuring instrument 11 are all connected with a control system circuit.
Through setting up check valve 9, make the tritium-containing gas only can flow to the rich tritium chamber of next-level tritium separation unit from last one-level poor tritium chamber, can not the backward flow. Meanwhile, because the one-way valve 9 is arranged between the two separation units, the gas output by the main pipeline 12 can only enter one separation unit through the one-way valve 9, and can not be shunted and can not enter the previous separation unit.
The tritium measuring instrument 11 can monitor the tritium content in the tritium-rich cavity in real time, after the tritium content reaches a set limit, the tritium measuring instrument 11 feeds a measured signal back to the control system, the control system opens the corresponding valve 6, and tritium-containing gas in the tritium-rich cavity of the separation unit is absorbed through a hydrogen storage material in the tritium pump 10; then, closing the valve 6, heating the tritium pump 10 to 200-450 ℃, and releasing all tritium adsorbed in the tritium pump 10 to a tritium-rich cavity of the upper-stage separation unit for further separation; the whole process can be fully automatically controlled by the tritium measuring instrument 11 through feeding back the monitoring signal of the tritium content.
Preferably, the set limit of the tritium content in the tritium-rich cavity is that the tritium content is reduced by 1-5% according to the purity of the final required product.
In the system of the present invention, the tritium separation system 3 having two tritium separation units is a two-stage separation system; the higher the grade number of the separation system is, the better the recovery and purification effect of the system is, and the higher the purity of the recovered tritium gas is.
In theory, there may be a variety of communication means between the tritium separation units of the present invention. As long as tritium-rich gas and tritium-poor gas can respectively flow through a plurality of tritium separation units and are separated for a plurality of times, the tritium separation units can be arranged.
In particular, as shown in FIG. 2, in one embodiment of the present invention, tritium separation system 3 is a five-stage separation system. The tritium separation system 3 includes a first tritium separation unit 31, a second tritium separation unit 32, a third tritium separation unit 33, a fourth tritium separation unit 34, and a fifth tritium separation unit 35. The tritium-poor cavity of the first tritium separation unit 31 is connected with the tritium-rich cavity of the second tritium separation unit 32 in series through a pipeline, a one-way valve 9 is arranged on the pipeline in series, and the one-way valve 9 can control gas in the pipeline to flow into the tritium-rich cavity of the second tritium separation unit 32 from the tritium-poor cavity of the first tritium separation unit 31 and cannot flow in the reverse direction. In this way, until the tritium-poor cavity of the fifth tritium separation unit 35 is communicated with the exhaust gas collection tank 4; in this way, the tritium-depleted gas separated by each tritium separation unit passes through each separation unit in sequence and unidirectionally, and is finally collected in the off-gas collection tank 4.
And a tritium measuring instrument 11 is fixedly arranged on each tritium-enriched cavity. The tritium-rich cavity of the second tritium separation unit 32 is communicated with the tritium-rich cavity of the first tritium separation unit 31 in parallel, and a valve 6 and a tritium pump 10 are respectively arranged on the parallel pipelines. When a tritium measuring instrument 11 monitors that the tritium content in a tritium-rich cavity of the tritium measuring instrument is reduced by 1% -5%, transmitting a monitoring signal to a control system, controlling to open a valve 6 on a parallel pipeline between a first tritium separation unit 31 and a second tritium separation unit 32 by the control system, and absorbing tritium-containing gas by a hydrogen storage material in a tritium pump 10 on the pipeline; then, the valve 6 is closed, the tritium pump 10 is heated to 200 to 450 ℃, tritium adsorbed in the tritium pump 10 is completely released into the tritium-rich cavity of the first tritium separation unit 31, and the gases are further separated through the palladium alloy film 30 in the first tritium separation unit 31 and the serial pipeline with the second tritium separation unit 32.
A pipeline between the poor tritium cavity of the second tritium separation unit 32 and the tritium-rich cavity of the third tritium separation unit 33 is communicated with an air outlet of the main pipeline 12, and hydrogen isotope mixed gas enters from the pipeline. The tritium-rich chamber of the first tritium separation unit 31 is in communication with the tritium bed 5. Thus, the tritium-rich gas primarily separated in the third tritium separation unit 33 passes through the second tritium separation unit 32 and the first tritium separation unit 31 in sequence to continuously remove hydrogen or deuterium therein, and finally enters the tritium bed 5 to be recovered and fixed.
The tritium-rich cavity of the fifth tritium separation unit 35 is communicated with the tritium-rich cavity of the fourth tritium separation unit 34, the tritium-rich cavity of the fourth tritium separation unit 34 is communicated with the tritium-rich cavity of the third tritium separation unit 33 in the above manner, so that tritium in the tritium-poor gas can be continuously recovered when the tritium-poor gas passes through each separation unit, and finally enters the tritium bed 5, and the tritium recovery rate is improved.
The palladium alloy membrane 30 of the present invention includes a support material and an alloy coating layer coated on the support material; the material of the alloy coating is PdAuNiCuY alloy; wherein, the content of Ag is 15-30 at.%, the content of Au is 0-10 at.%, the content of Cu is 0-10 at.%, the content of Ni is 0-2 at.%, the content of Y is 0-2 at.%, and the rest is Pd; the palladium alloy membrane 30 of the above composition has good separation efficiency and strength.
Preferably, in the PdAgAuNiCuY alloy, the Pd content is 65at.%, the Ag content is 20at.%, the Au content is 8at.%, the Ni content is 1.5at.%, the Cu content is 5at.%, and the Y content is 0.5at.%.
Preferably, the thickness of the alloy coating is 1 to 1000 μm.
Preferably, the inner surfaces of the main pipeline 12, the pyrolysis cavity 1, the storage cavity 2, the vacuum pump set 7 and the tritium separation system 3 are coated with tritium-resistant coatings; the tritium-resistant coating comprises one of an oxide ceramic coating, a diamond-like coating, a SiC coating and a nitride coating; by applying a tritium-resistant coating, damage to the system by tritium in the gas can be prevented.
Preferably, the particle size of the hydrogen storage material particles is 1-5 mm; an isolation net of 100-500 meshes is fixed at the communication part of the storage cavity 2 and the main pipeline 12; by providing the separation net, the particles of hydrogen storage material can be prevented from entering the main pipeline 12 or other structures.
Preferably, the vacuum pump unit 7 is connected to the purification device 8, and the impurity gas can be collected in the purification device 8 and purified.
Preferably, in the system of the present invention, the air inlet of the main pipeline 12 is provided with a valve 6; a valve 6 is arranged on a pipeline for communicating the pyrolysis cavity 1 with the main pipeline 12, a pipeline for communicating the storage cavity 2 with the main pipeline 12, an air outlet of the main pipeline 12, an air inlet of the waste gas collecting tank 4 and an air inlet of the tritium bed 5; through setting up above-mentioned valve, can control each corresponding each structure to make the use of this system more nimble controllable.
Preferably, all of the lines and devices in the system of the present invention are 316 stainless steel.
Preferably, the system leakage rate of the invention is better than 1 x 10 -10 Pa·m 3 /s。
The tritium recovery and purification method of the waste neutron target adopts the recovery and purification system to recover and purify tritium, and specifically comprises the following steps:
s1, closing an air outlet of a main pipeline 12; and (3) removing the waste neutron target with the service life exceeding the service life from the neutron source device, and putting the waste neutron target into the pyrolysis cavity 1 for heating to obtain the gas released by the waste neutron target.
Preferably, the heating temperature is 300-600 ℃, the heating rate is 3-10 ℃/min, and the heat preservation time is 40-80 min
S2, gas enters the storage cavity 2 through the main pipeline 12, and hydrogen isotopes in the gas are adsorbed by hydrogen storage material particles in the storage cavity 2;
s3, closing the communication between the pyrolysis cavity 1 and the main pipeline 12, and starting the vacuum pump set 7; recovering impurity gas which is not adsorbed by hydrogen storage material particles in the gas until the vacuum degree in the system is less than 10 -6 And Pa, continuously keeping the vacuum pump unit 7 open for 30min, and closing the vacuum pump unit 7 after recovery.
S4, heating the storage cavity 2 to enable the hydrogen storage material particles to release hydrogen isotope gas containing deuterium gas and tritium gas;
s5, opening an air outlet of the main pipeline 12, enabling the hydrogen isotope mixed gas to enter a tritium separation unit of the tritium separation system 3 through the main pipeline 12, and separating the hydrogen isotope mixed gas into tritium-rich gas and tritium-poor gas through a palladium alloy membrane 30; tritium-rich gas flows out through the tritium-rich cavity and is recycled, and tritium-poor gas flows out through the tritium-poor cavity and is recycled.
Preferably, after step S5 is finished, the air outlet of the main pipeline 12 is closed. Introducing inert gas into the system through a gas inlet of the main pipeline 12 for purging, so that the inert gas carrying residual tritium gas in the system is discharged out of the system; the gas exiting the system is collected and purified.
The invention is illustrated by the following specific examples:
example 1
The waste neutron target treated by the embodiment is a neutron target taking oxygen-free copper as a substrate material, the diameter of the neutron target is 100mm, and the thickness of a titanium film of the neutron target is 3 microns.
The tritium recovery and purification method is adopted for treatment. In the system adopted in this embodiment, the tritium separation system 3 is a three-stage separation system, that is, has three separation units.
In the method adopted in this embodiment, in step S1, the pyrolysis chamber 1 is heated to 450 ± 50 ℃, the heating rate is 3 ± 1 ℃/min, and the temperature is maintained for 40min.
After recovery and purification, the purity of the finally obtained tritium gas reaches 99.59 percent by utilizing gas chromatography detection, and the purity meets the use requirement.
Example 2
The waste neutron target treated in the embodiment is a neutron target with CrZrCu alloy as a substrate material, the diameter of the neutron target is 260mm, and the thickness of a titanium film is 2 μm.
The tritium recovery and purification method is adopted for treatment. In the system adopted in this embodiment, the tritium separation system 3 is a two-stage separation system, that is, has two separation units.
In the method adopted in this embodiment, in step S1, the pyrolysis chamber 1 is heated to 350 ± 50 ℃, the heating rate is 5 ± 1 ℃/min, and the temperature is maintained for 40min.
The purity of the finally obtained tritium gas reaches 99.2 percent by utilizing gas chromatography detection, and the purity meets the use requirement.
Example 3
The waste neutron target treated in the embodiment is a neutron target with a molybdenum alloy as a substrate material, the diameter of the neutron target is 150mm, and the thickness of a zirconium film is 3 μm.
The tritium recovery and purification method is adopted for treatment. In the system adopted in this embodiment, the tritium separation system 3 is a four-stage separation system, that is, has four separation units.
In the method adopted in this embodiment, in step S1, the pyrolysis chamber 1 is heated to 450 ± 50 ℃, the heating rate is 5 ± 1 ℃/min, and the temperature is maintained for 40min.
The purity of the finally obtained tritium gas reaches 99.99 percent by utilizing gas chromatography detection, and the purity meets the use requirement.
It can be seen from the above embodiments that the system and method of the present invention have good tritium recovery efficiency and purification effect for the waste neutron targets of different substrate materials. The purity of tritium recovered in each embodiment is more than 99%, and the requirement of tritium recovery is met.
The terms "first" and "second" in the description of the present invention are used for descriptive purposes only and are not to be construed as indicating or implying relative importance or implicitly indicating the number of technical features indicated. Thus, a feature defined as "first" or "second" may explicitly or implicitly include at least one such feature. In the description of the present invention, "a plurality" means at least two, e.g., two, three, etc., unless specifically limited otherwise.
In the description of the specification, reference to the description of "one embodiment," "some embodiments," "an example," "a specific example," or "some examples" or the like means that a particular feature, structure, material, or characteristic described in connection with the embodiment or example is included in at least one embodiment or example of the invention. In this specification, the schematic representations of the terms used above are not necessarily intended to refer to the same embodiment or example. Furthermore, the particular features, structures, materials, or characteristics described may be combined in any suitable manner in any one or more embodiments or examples. Furthermore, various embodiments or examples and features of different embodiments or examples described in this specification can be combined and combined by one skilled in the art without contradiction.
In the description of the present invention, it should be noted that, unless otherwise explicitly specified or limited, the terms "mounted," "connected," and "connected" are to be construed broadly, e.g., as meaning either a fixed connection, a removable connection, or an integral connection; can be mechanically or electrically connected; they may be connected directly or indirectly through intervening media, or they may be interconnected between two elements. The specific meanings of the above terms in the present invention can be understood in specific cases to those skilled in the art.
The above description is only for the purpose of illustrating the preferred embodiments of the present invention and is not to be construed as limiting the invention, and any modifications, equivalents, improvements and the like that fall within the spirit and principle of the present invention are intended to be included therein.
Claims (12)
1. A tritium recovery and purification system for a waste neutron target is characterized by comprising a main pipeline (12), wherein the two ends of the main pipeline (12) are respectively provided with an air inlet and an air outlet, and the middle part of the main pipeline (12) is respectively communicated with a pyrolysis cavity (1) and a storage cavity (2); the air outlet of the main pipeline (12) is communicated with the tritium separation system (3); the middle part of the main pipeline (12) is also communicated with a vacuum pump set (7);
hydrogen storage material particles are arranged in the storage cavity (2);
the tritium separation system (3) comprises at least one tritium separation unit, and a palladium alloy membrane (30) is fixedly arranged in each tritium separation unit; the palladium alloy membrane (30) separates the inner cavity of the tritium separation unit into a tritium-rich chamber and a tritium-poor chamber.
2. A tritium recovery and purification system for spent neutron targets according to claim 1, characterized in that the tritium separation system (3) comprises at least two tritium separation units;
the air outlet of the main pipeline (12) is communicated with the tritium-enriched cavity of one tritium separation unit;
the tritium-rich cavity of one tritium separation unit is communicated with the tritium-rich cavity of the other tritium separation unit through a pipeline, and the tritium-poor cavity of the other tritium separation unit is also communicated with the tritium-rich cavity of the one tritium separation unit through another pipeline;
the tritium-poor cavity of one tritium separation unit is also communicated with an exhaust gas collection tank (4);
the tritium-enriched chamber of the other tritium separation unit is also in communication with a tritium bed (5).
3. A tritium recovery and purification system for spent neutron targets according to claim 2,
and a one-way valve (9) is arranged on a pipeline for communicating the tritium-rich cavity of one tritium separation unit with the tritium-rich cavity of the other tritium separation unit.
4. The tritium recovery and purification system for the spent neutron target according to claim 3, characterized in that a valve (6) and a tritium pump (10) are arranged on a pipeline for communicating the tritium-rich chamber of one tritium separation unit with the tritium-rich chamber of the other tritium separation unit; and each tritium-rich cavity is provided with a tritium measuring instrument (11);
the hydrogen storage material particles are arranged in the tritium pump (10).
5. The tritium recovery and purification system for spent neutron targets according to any of claims 1 to 4, characterized in that the palladium alloy membrane (30) comprises a support material and an alloy coating coated on the support material;
the alloy coating is made of PdAuNiCuY alloy; wherein, the Ag content is 15-30 at.%, the Au content is 0-10 at.%, the Cu content is 0-10 at.%, the Ni content is 0-2 at.%, the Y content is 0-2 at.%, and the rest is Pd.
6. A tritium recovery and purification system for spent neutron targets according to claim 5, characterized in that the thickness of the alloy coating is 1-1000 μm.
7. The tritium recovery and purification system for spent neutron targets according to any one of claims 1 to 4, characterized in that the inner surfaces of the main pipeline (12), the pyrolysis chamber (1), the storage chamber (2), the vacuum pump set (7) and the tritium separation system (3) are coated with tritium-resistant coatings;
the tritium resistant coating comprises at least one of an oxide ceramic coating, a diamond-like coating, a SiC coating, and a nitride coating.
8. The tritium recovery and purification system for the spent neutron target according to any one of claims 1 to 4, wherein the particle size of the hydrogen storage material particles is 1 to 5mm; an isolation net with 100-500 meshes is fixed at the communication position of the storage cavity (2) and the main pipeline (12).
9. A tritium recovery and purification method for spent neutron targets using the system of any one of claims 1 to 8, comprising the steps of:
s1, closing an air inlet and an air outlet of the main pipeline (12); putting the waste neutron target into the pyrolysis cavity (1) for heating to obtain gas released by the waste neutron target;
s2, the gas enters the storage cavity (2) through the main pipeline (12), and the hydrogen storage material particles in the storage cavity (2) adsorb the hydrogen isotope mixed gas in the gas;
s3, starting the vacuum pump set (7), recovering impurity gas which is not adsorbed by the hydrogen storage material particles in the main pipeline (12) and the storage cavity (2), and closing the vacuum pump set (7) after recovery;
s4, heating the storage cavity (2) to 300-500 ℃, opening a gas outlet of the main pipeline (12), and preserving heat for 10-40 min to enable the hydrogen storage material particles to release the hydrogen isotope gas;
s5, the hydrogen isotope bonded gas enters a tritium separation unit of the tritium separation system (3) through the main pipeline (12) and is separated into tritium-rich gas and tritium-poor gas by the palladium alloy membrane (30); the tritium-rich gas flows out through the tritium-rich cavity and is recycled, and the tritium-poor gas flows out through the tritium-poor cavity and is recycled.
10. A tritium recovery and purification method for a spent neutron target according to claim 9, wherein in the step S1, the heating temperature of the spent neutron target is 300-600 ℃, the heating rate is 3-10 ℃/min, and the holding time is 40-80 min.
11. Tritium of spent neutron target according to claim 9The recovery and purification method is characterized in that in the step S3, the vacuum pump set (7) is started, and the vacuum degree in the system is less than 10 -6 Pa, continuously keeping the vacuum pump set (7) open for 30min, and then closing the vacuum pump set (7).
12. A tritium recovery and purification method for spent neutron targets according to claim 9,
after the step S5 is finished, opening an air inlet of the main pipeline (12), and closing an air outlet of the main pipeline (12); starting the vacuum pump group (7);
and introducing inert gas into the system for purging through the gas inlet of the main pipeline (12), so that the inert gas carries the residual tritium gas in the system to flow out through the vacuum pump set (7), and purifying the flowing gas.
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