CN115385302B - 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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- CN115385302B CN115385302B CN202210898869.7A CN202210898869A CN115385302B CN 115385302 B CN115385302 B CN 115385302B CN 202210898869 A CN202210898869 A CN 202210898869A CN 115385302 B CN115385302 B CN 115385302B
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- 229910052722 tritium Inorganic materials 0.000 title claims abstract description 350
- YZCKVEUIGOORGS-NJFSPNSNSA-N Tritium Chemical compound [3H] YZCKVEUIGOORGS-NJFSPNSNSA-N 0.000 title claims abstract description 346
- 238000011084 recovery Methods 0.000 title claims abstract description 35
- 239000002699 waste material Substances 0.000 title claims abstract description 30
- 238000000746 purification Methods 0.000 title claims abstract description 28
- 238000000034 method Methods 0.000 title claims abstract description 27
- 238000000926 separation method Methods 0.000 claims abstract description 151
- 239000007789 gas Substances 0.000 claims abstract description 99
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 claims abstract description 66
- 229910052739 hydrogen Inorganic materials 0.000 claims abstract description 66
- 239000001257 hydrogen Substances 0.000 claims abstract description 66
- 238000003860 storage Methods 0.000 claims abstract description 37
- 239000002245 particle Substances 0.000 claims abstract description 31
- 239000011232 storage material Substances 0.000 claims abstract description 31
- 229910001252 Pd alloy Inorganic materials 0.000 claims abstract description 30
- 239000012528 membrane Substances 0.000 claims abstract description 26
- 238000000197 pyrolysis Methods 0.000 claims abstract description 19
- 238000000576 coating method Methods 0.000 claims description 25
- 239000011248 coating agent Substances 0.000 claims description 22
- 229910045601 alloy Inorganic materials 0.000 claims description 19
- 239000000956 alloy Substances 0.000 claims description 19
- 238000010438 heat treatment Methods 0.000 claims description 16
- 239000000463 material Substances 0.000 claims description 14
- 238000004891 communication Methods 0.000 claims description 10
- 239000011261 inert gas Substances 0.000 claims description 6
- 239000012535 impurity Substances 0.000 claims description 5
- 238000010926 purge Methods 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
- 238000004321 preservation Methods 0.000 claims description 3
- 229910001069 Ti alloy Inorganic materials 0.000 claims description 2
- 229910052770 Uranium Inorganic materials 0.000 claims description 2
- 229910000756 V alloy Inorganic materials 0.000 claims description 2
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims description 2
- 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 7
- 230000009286 beneficial effect Effects 0.000 description 7
- 229910052805 deuterium Inorganic materials 0.000 description 7
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 6
- 239000010936 titanium Substances 0.000 description 6
- 229910052719 titanium Inorganic materials 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
- 239000010949 copper Substances 0.000 description 4
- 239000006096 absorbing agent Substances 0.000 description 3
- 239000000306 component Substances 0.000 description 3
- 238000001514 detection method Methods 0.000 description 3
- 230000000694 effects Effects 0.000 description 3
- 230000004927 fusion Effects 0.000 description 3
- 238000004817 gas chromatography Methods 0.000 description 3
- 229910052757 nitrogen Inorganic materials 0.000 description 3
- 239000012466 permeate Substances 0.000 description 3
- 239000002912 waste gas Substances 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
- 238000012544 monitoring process Methods 0.000 description 2
- 230000002285 radioactive effect Effects 0.000 description 2
- 238000004064 recycling Methods 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
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 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
- 238000009395 breeding Methods 0.000 description 1
- 230000001488 breeding effect Effects 0.000 description 1
- 201000011510 cancer Diseases 0.000 description 1
- 239000011247 coating layer Substances 0.000 description 1
- 238000009833 condensation Methods 0.000 description 1
- 230000005494 condensation Effects 0.000 description 1
- 238000001816 cooling Methods 0.000 description 1
- 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
- 239000010410 layer Substances 0.000 description 1
- 238000005259 measurement Methods 0.000 description 1
- 239000000203 mixture Substances 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
- 238000005371 permeation separation Methods 0.000 description 1
- 238000007747 plating Methods 0.000 description 1
- 239000000047 product Substances 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
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
Classifications
-
- 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
Landscapes
- 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 of a waste neutron target. The system comprises a main pipeline, wherein two ends of the main pipeline are respectively provided with an air inlet and an air outlet, and the two ends of the main pipeline are respectively communicated with a pyrolysis cavity and a storage cavity; the air 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; the storage cavity is internally provided with hydrogen storage material particles; 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 divides 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 the high-purity hydrogen isotope gas is directly obtained through a pyrolysis-adsorption cycle process; meanwhile, the hydrogen permeability and isotope effect of palladium alloy are adopted to realize the enrichment of tritium. The method greatly improves the separation coefficient and 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 of a waste neutron target.
Background
Neutron technology is widely applied to fields such as neutron photography, oilfield logging, cancer treatment, neutron security inspection, activation analysis, irradiation breeding, irradiation damage research, nuclear measurement and the like. Currently the main neutron sources are nuclear reactors, radioactive sources and accelerator type neutron sources. The accelerator type deuterium-tritium neutron source bombards a neutron target through an accelerated deuterium ion beam, and generates 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 for directly producing neutrons. Because titanium has higher tritium absorption density and stability, the most used neutron targets at home and abroad are obtained by plating a layer of titanium film on a substrate material and absorbing tritium through the titanium film.
Tritium is an important strategic resource, very scarce and expensive, and the price of 1 gram tritium is up to tens of millions. In the process of using the tritium-titanium target on the accelerator neutron source, only the position bombarded by the deuterium ion beam participates in the deuterium-tritium fusion reaction, so that the consumed tritium is far lower than the tritium stored 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 critical 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 absorber, a liquid nitrogen molecular sieve absorber and a liquid helium temperature molecular sieve absorber. Specifically, a low-temperature molecular sieve separation system is utilized, and gas heated and released by a waste neutron target passes through a low-temperature molecular sieve condensed by liquid nitrogen and is adsorbed by the molecular sieve due to low-temperature liquefaction of oxygen, nitrogen, water vapor and the like; the residual gas passes through the liquid helium condensed molecular sieve, the hydrogen isotope is condensed into liquid state and is absorbed into the molecular sieve again, and two low-temperature molecular sieves with cooling systems are needed for one purification. In order to obtain the high-purity hydrogen isotope gas, a plurality of low-temperature classifying screens are required to be connected in series to realize multistage purification, so that the device is large, the process is complex, and the separation coefficient is low. And the low-temperature freezing cost is high due to the low boiling point of the hydrogen isotope.
For isotope separation, a common solution in the prior art is low temperature 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 component is the rectification column. In order to improve efficiency and purity, the rectification column generally needs more than 2 rectification columns, each of which is 2-3m. Therefore, the whole equipment is huge and the operation is complex.
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 of a waste neutron target, which comprises a main pipeline, wherein two ends of the main pipeline are respectively provided with an air inlet and an air outlet, and the two ends of the main pipeline are respectively communicated with a pyrolysis cavity and a storage cavity; the air 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; the storage cavity is internally provided with hydrogen storage material particles; 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 divides 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 absorb hydrogen isotope gas, only a storage cavity with built-in hydrogen storage alloy is designed in the system, and the high-purity hydrogen isotope gas is directly obtained through a pyrolysis-adsorption cycle process; meanwhile, hydrogen permeability and isotope effect of palladium alloy are adopted, according to different difficulty of permeation of protium, deuterium and tritium through palladium alloy membrane, protium and deuterium of small molecules pass preferentially, tritium of larger molecules is difficult to permeate, and enrichment of tritium is realized. Meanwhile, the design of multistage circulating permeation separation is combined, so that the separation coefficient and 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-enriched cavity of one tritium separation unit; the tritium-enriched cavity of one tritium separation unit is communicated with the tritium-enriched cavity of the other tritium separation unit through a pipeline, and the tritium-depleted cavity of the other tritium separation unit is also communicated with the tritium-enriched cavity of the one tritium separation unit through another pipeline; the tritium-depleted cavity of one tritium separation unit is also communicated with the waste gas collection tank; the tritium-enriched cavity of the other tritium separation unit is also communicated with a tritium bed.
The beneficial effects of adopting the further scheme are as follows: the hydrogen isotope mixed gas can be circularly separated for a plurality of times between the two tritium separation units, so that the separation effect is better, and the purity of the recovered tritium gas is higher.
Further, a one-way valve is arranged on a pipeline which is communicated with the tritium-rich cavity of one tritium separation unit and the tritium-rich cavity of the other tritium separation unit.
The beneficial effects of adopting the further scheme are as follows: by arranging the one-way valve, tritium-containing gas can only flow from the previous-stage tritium-depleted chamber to the tritium-enriched chamber of the next-stage tritium separation unit and cannot flow reversely.
Further, a valve and a tritium pump are arranged on a pipeline which is communicated with the tritium-rich cavity of one tritium separation unit and the tritium-rich cavity of the other tritium separation unit; the tritium pump is internally provided with the hydrogen storage material particles; and each tritium-enriched cavity is provided with a tritium measuring instrument.
The beneficial effects of adopting the further scheme are as follows: the tritium measuring instrument can monitor the tritium content in the tritium-enriched cavity in real time, after the tritium content reaches a set limit, the tritium measuring instrument feeds back a measured signal to the control system, the control system opens a corresponding valve, and the tritium-containing gas in the tritium-enriched cavity of the separation unit is absorbed through the hydrogen storage material in the tritium pump; then, closing a valve, heating the tritium pump, and completely releasing the tritium adsorbed in the tritium pump to a tritium-enriched cavity of the upper-stage separation unit for further separation; the whole process can carry out full-automatic control on the monitoring signal feedback of the tritium content through a 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 PdArgAuNiCuY 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 effects of adopting the further scheme are as follows: the palladium alloy membrane with the components 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 group and the tritium separation system are coated with tritium-resisting 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 effects of adopting the further scheme are as follows: by applying a tritium-resistant coating, damage to the system caused by tritium in the gas can be prevented.
Further, the particle size of the hydrogen storage material particles is 1-5mm; and a 100-500-mesh isolation net is fixed at the communication part of the storage cavity and the main pipeline.
The beneficial effects of adopting the further scheme are as follows: by providing a separation net, particles of hydrogen storage material can be prevented from entering the main pipeline or other structure.
The invention provides a method for recycling 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; the waste neutron target is placed into the pyrolysis cavity to be heated, and gas released by the waste neutron target is obtained;
s2, enabling the gas to enter the storage cavity through the main pipeline, and enabling the hydrogen storage material particles in the storage cavity to adsorb hydrogen isotope mixed gas in the gas;
s3, starting the vacuum pump set, recovering impurity gas which is 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 an air outlet of the main pipeline, and preserving heat for 10-40min to enable the hydrogen storage material particles to release the hydrogen isotope combination gas;
s5, enabling the hydrogen isotope gas to enter one tritium separation unit of the tritium separation system through the main pipeline, and separating the hydrogen isotope gas into tritium-enriched gas and tritium-depleted gas by the palladium alloy membrane; the tritium-enriched gas flows out through the tritium-enriched cavity and is recovered, and the tritium-depleted gas flows out through the tritium-depleted cavity and is recovered.
The recovery and purification method can effectively solve the problem of recovery and reutilization of tritium in the waste neutron target, the recovery rate is more than 90%, and the purity of the recovered tritium gas is more than 99.9%.
In 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 40-80min.
Further, in the step S3, the vacuum pump set is turned on, and the vacuum degree in the system is less than 10 -6 Pa, continuously keeping the vacuum pump set on for 30min, and closing the vacuum pump set.
Further, after the step S5 is finished, opening the air inlet of the main pipeline, and closing the air outlet of the main pipeline; opening the vacuum pump group; and introducing inert gas into the system through an air inlet of the main pipeline for purging, so that the inert gas carries tritium gas remained in the system to flow out through the vacuum pump set, and purifying the flowing gas.
Drawings
FIG. 1 is a schematic diagram of a tritium recovery and purification system of a spent neutron target of the present invention;
FIG. 2 is a schematic diagram of a five-stage tritium separation system in one embodiment of the tritium recovery and purification system of the spent neutron target of the invention.
In the drawings, the list of components represented by the various numbers is as follows:
1. a pyrolysis chamber; 2. a storage chamber;
3. a tritium separation system; 30. a palladium alloy membrane; 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. an exhaust gas collection tank; 5. a tritium bed;
6. a valve;
7. a vacuum pump unit;
8. a purifying device;
9. a one-way valve;
10. a tritium pump;
11. a tritium measuring instrument;
12. and a main pipeline.
Detailed Description
The principles and features of the present invention are described below with reference to the drawings, the examples are illustrated for the purpose of illustrating the invention and are not to be construed as limiting the scope of the invention.
As shown in fig. 1, the pyrolysis device comprises a main pipeline 12, wherein two ends of the main pipeline 12 are respectively provided with an air inlet and an air outlet, and the two ends are 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; the storage cavity 2 is internally provided with hydrogen storage material particles; 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 internal cavity of the tritium separation unit into a tritium-rich cavity and a tritium-depleted cavity.
According to the tritium recovery and purification system, the waste neutron target is heated through the pyrolysis cavity 1, so that the waste neutron target releases mixed gas containing tritium; the mixed gas containing tritium 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 impurities which are not absorbed in the mixed gas can be separated from the system by the vacuum pump set 7 in a vacuumizing mode. Then, heating the storage cavity 2 to release the hydrogen isotope mixed gas from the hydrogen storage material particles; the hydrogen isotope mixed gas enters a tritium separation system 3 through an air outlet of a main pipeline 12 and is separated into tritium-enriched gas and tritium-depleted gas by a palladium alloy membrane 30 in a tritium separation unit; thus, 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, and can effectively solve the problem of recovery and reutilization of tritium in the waste neutron target, wherein the recovery rate is more than 90%, and the purity of the recovered tritium gas is more than 99.9%.
In the palladium alloy membrane 30 adopted by the invention, due to the isotope effect, lighter hydrogen atoms and deuterium atoms in the hydrogen isotope mixed gas are easier to permeate the palladium alloy membrane 30, and heavier tritium atoms are difficult to permeate the palladium alloy membrane 30, so that hydrogen and deuterium are enriched in a tritium-depleted cavity after permeating the palladium alloy membrane 30, and tritium is enriched in a tritium-enriched cavity.
The hydrogen storage material particles adopted by the invention are hydrogen storage alloy, and are specifically one or more than two of uranium, zirconium alloy, titanium alloy, laNi alloy, vanadium alloy and palladium alloy; these alloys are capable of efficiently absorbing hydrogen isotopes in the mixed gas.
Preferably, the pyrolysis chamber 1 is close to the air inlet of the main conduit 12 and the storage chamber 2 is close to the air outlet of the main conduit 12.
Preferably, the number of the storage cavities 2 can be multiple, and the multiple storage cavities 2 can be connected in parallel on the main pipeline 12; the hydrogen storage material particles in the plurality of storage cavities 2 can be made of different materials.
In the above scheme, preferably, the tritium separation system 3 comprises at least two tritium separation units.
Specifically, taking two tritium separation units as an example, the air outlet of the main pipeline 12 is communicated with the tritium-enriched cavity of the first tritium separation unit; the tritium-enriched cavity of the first tritium separation unit is also communicated with the tritium-enriched cavity and the tritium-depleted cavity of the second tritium separation unit through two pipelines respectively; the tritium-depleted cavity of the first tritium separation unit is also communicated with the waste gas collection tank 4; the tritium-rich chamber of the second tritium separation unit is also in communication with the tritium bed 5. The hydrogen isotope mixed gas entering the first tritium separation unit through the main pipeline 12 firstly enters a tritium-enriched cavity of the tritium separation unit, and the tritium-enriched gas enriched in the tritium-enriched cavity of the first tritium separation unit is circulated to the tritium-enriched cavity of the second tritium separation unit through the pipeline and is separated again by the palladium alloy membrane 30 in the second tritium separation unit; the enriched tritium-depleted gas in the tritium-depleted chamber of the second tritium separation unit is continuously circulated into the first tritium separation unit through the pipeline and is continuously separated by the palladium alloy membrane 30 in the first tritium separation unit.
Thus, the hydrogen isotope mixed gas can be circularly separated for a plurality of 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 air outlet of the main pipeline 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; the tritium pump 10 has hydrogen storage material particles disposed therein. 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 a tritium measuring instrument 11 is arranged on each tritium-rich cavity; the valves 6, the check valves 9, the tritium pumps 10 and the tritium measuring devices 11 are all connected with a control system circuit.
By arranging the one-way valve 9, tritium-containing gas can only flow from the previous-stage tritium-depleted chamber to the tritium-enriched chamber of the next-stage tritium separation unit and cannot flow reversely. Meanwhile, as 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 the gas cannot be split and cannot enter the previous separation unit.
The tritium measuring instrument 11 can monitor the tritium content in the tritium-enriched cavity in real time, after the tritium content reaches a set limit, the tritium measuring instrument 11 feeds back a measured signal to the control system, the control system opens the corresponding valve 6, and the tritium-containing gas in the tritium-enriched cavity of the separation unit is absorbed through the hydrogen storage material in the tritium pump 10; then, closing the valve 6, heating the tritium pump 10 to 200-450 ℃ to enable all tritium adsorbed in the tritium pump 10 to be released to a tritium-enriched cavity of the upper-stage separation unit for further separation; the whole process can be fully automatically controlled by monitoring signal feedback of tritium content through a tritium measuring instrument 11.
Preferably, the tritium content in the tritium-enriched cavity is set to be limited 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 secondary separation system; the higher the stage number of the separation system is, the better the recovery and purification effects of the system are, and the higher the purity of the recovered tritium gas is.
In theory, there may be a variety of ways of communication between the multiple tritium separation units of the present invention. The tritium separation units can basically realize that tritium-enriched gas and tritium-depleted gas can flow through a plurality of tritium separation units respectively and are separated for a plurality of times.
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 comprises 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-depleted cavity of the first tritium separation unit 31 is connected in series with the tritium-enriched cavity of the second tritium separation unit 32 through a pipeline, a one-way valve 9 is arranged on the pipeline connected in series, and the one-way valve 9 can control the gas in the pipeline to flow from the tritium-depleted cavity of the first tritium separation unit 31 into the tritium-enriched cavity of the second tritium separation unit 32 but not reversely. And so on until the tritium-depleted chamber of the fifth tritium separation unit 35 is in communication with the off-gas collection canister 4; in this way, the tritium-depleted gas separated by each tritium separation unit passes through the separation units in turn unidirectionally and is finally collected in the off-gas collection tank 4.
A tritium measuring instrument 11 is fixedly arranged on each tritium-enriched cavity. The tritium-enriched cavity of the second tritium separation unit 32 is communicated with the tritium-enriched 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 the tritium measuring instrument 11 monitors that the tritium content in the tritium-enriched cavity is reduced by 1% -5%, a monitored signal is transmitted to the control system, and the control system controls to open a valve 6 on a parallel pipeline between the first tritium separation unit 31 and the second tritium separation unit 32, and a hydrogen storage material in the tritium pump 10 on the pipeline absorbs the tritium-containing gas; thereafter, the valve 6 is closed, the tritium pump 10 is heated to 200-450 ℃, so that all tritium adsorbed in the tritium pump 10 is released into the tritium-enriched cavity of the first tritium separation unit 31, and the gas is further separated through the palladium alloy membrane 30 in the first tritium separation unit 31 and the pipeline connected in series with the second tritium separation unit 32.
The pipeline between the tritium-depleted cavity of the second tritium separation unit 32 and the tritium-enriched cavity of the third tritium separation unit 33 is communicated with the gas outlet of the main pipeline 12, and the 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. In this way, the tritium-enriched gas which is separated for the first time in the third tritium separating unit 33 passes through the second tritium separating unit 32 and the first tritium separating unit 31 in sequence to remove hydrogen or deuterium therefrom, and finally enters the tritium bed 5 to be recovered and fixed.
The tritium-enriched cavity of the fifth tritium separation unit 35 is communicated with the tritium-enriched cavity of the fourth tritium separation unit 34 and the tritium-enriched cavities of the fourth tritium separation unit 34 and the third tritium separation unit 33 in the above manner, so that the tritium in the tritium-depleted gas can be continuously recovered when passing through each separation unit and finally enter the tritium bed 5, thereby improving the recovery rate of the tritium.
The palladium alloy membrane 30 of the present invention includes a support material and an alloy coating layer coated on the support material; the alloy coating is made of PdArgAuNiCuY alloy; wherein, the Ag content is 15-30at%, the Au content is 0-10at%, the Cu content is 0-10at%, the Ni content is 0-2at%, the Y content is 0-2at%, 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 content of Pd is 65at.%, the content of Ag is 20at.%, the content of Au is 8at.%, the content of Ni is 1.5at.%, the content of Cu is 5at.%, and the content of Y is 0.5at.%.
Preferably, the thickness of the alloy coating is 1-1000 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-resisting 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 caused by tritium in the gas can be prevented.
Preferably, the particle size of the hydrogen storage material particles is 1-5mm; a 100-500 mesh isolation net is fixed at the communication part of the storage cavity 2 and the main pipeline 12; by providing a separation net, particles of hydrogen storage material may be prevented from entering the main pipeline 12 or other structure.
Preferably, the vacuum pump unit 7 is in communication with the purge device 8, and is configured to collect the impurity gas into the purge device 8 to be purged.
Preferably, in the system of the invention, the air inlet of the main pipe 12 is provided with a valve 6; the pipeline of the pyrolysis cavity 1 communicated with the main pipeline 12, the pipeline of the storage cavity 2 communicated with the main pipeline 12, the air outlet of the main pipeline 12, the air inlet of the waste gas collection tank 4 and the air inlet of the tritium bed 5 are all provided with a valve 6; by arranging the valve, each corresponding structure can be controlled, so that the system is more flexible and controllable to use.
Preferably, in the system of the present invention, all of the tubing and apparatus are 316 stainless steel.
Preferably, the leak rate of the system of the invention is better than 1X 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 removing the waste neutron target 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 released gas.
Preferably, the heating temperature is 300-600 ℃, the heating rate is 3-10 ℃/min, and the heat preservation time is 40-80min.
S2, enabling the gas to enter a storage cavity 2 through a main pipeline 12, and enabling hydrogen storage material particles in the storage cavity 2 to adsorb hydrogen isotopes in the gas;
s3, closing the communication between the pyrolysis cavity 1 and the main pipeline 12, and opening the vacuum pump set 7; recovering impurity gas which is not adsorbed by hydrogen storage material particles in the gas, and keeping the vacuum degree in the system to be less than 10 -6 Pa, continuously keeping the vacuum pump set 7 on for 30min, and closing the vacuum pump set 7 after recycling.
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, and enabling the hydrogen isotope mixed gas to enter a tritium separation unit of the tritium separation system 3 through the main pipeline 12 and be separated into tritium-enriched gas and tritium-depleted gas by the palladium alloy membrane 30; tritium-enriched gas flows out through the tritium-enriched cavity and is recovered, and tritium-depleted gas flows out through the tritium-depleted cavity and is recovered.
Preferably, after step S5 is completed, the air outlet of the main pipe 12 is closed. Introducing inert gas into the system through an air inlet of the main pipeline 12 for purging, so that the inert gas carries tritium gas remained in the system and is discharged out of the system; the gases exiting the system are collected and purified.
The invention is illustrated by the following specific examples:
example 1
The waste neutron target treated in this example was a neutron target with oxygen-free copper as the substrate material, the diameter of the neutron target was 100mm, and the thickness of the neutron target titanium film was 3. Mu.m.
The tritium recovery and purification method is adopted for treatment. In the system used in this example, tritium separation system 3 is a three-stage separation system, i.e., having three separation units.
In the method adopted in the embodiment, in the step S1, the pyrolysis cavity 1 is heated to 450+/-50 ℃, the heating rate is 3+/-1 ℃/min, and the temperature is kept for 40min.
After recovery and purification, the purity of the tritium gas finally obtained reaches 99.59% by gas chromatography detection, and the purity meets the use requirement.
Example 2
The waste neutron target treated in this example was a neutron target using CrZrCu alloy as a substrate material, the diameter was 260mm, and the thickness of the titanium film was 2. Mu.m.
The tritium recovery and purification method is adopted for treatment. In the system used in this example, the tritium separation system 3 is a two-stage separation system, i.e., having two separation units.
In the method adopted in the embodiment, in the step S1, the pyrolysis cavity 1 is heated to 350+/-50 ℃, the heating rate is 5+/-1 ℃/min, and the temperature is kept for 40min.
The purity of the tritium gas finally obtained is up to 99.2% by gas chromatography detection, and the purity meets the use requirement.
Example 3
The waste neutron target treated in this example was a neutron target using a molybdenum alloy as a substrate material, and had a diameter of 150mm and a zirconium film thickness of 3. Mu.m.
The tritium recovery and purification method is adopted for treatment. In the system employed in this example, tritium separation system 3 is a four-stage separation system, i.e., having four separation units.
In the method adopted in the embodiment, in the step S1, the pyrolysis cavity 1 is heated to 450+/-50 ℃, the heating rate is 5+/-1 ℃/min, and the temperature is kept for 40min.
The purity of the tritium gas finally obtained is up to 99.99% by gas chromatography detection, and the purity meets the use requirement.
As can be seen from the above examples, the system and method of the present invention have good tritium recovery efficiency and purification for waste neutron targets of different substrate materials. The purity of the tritium recovered in each example is more than 99%, and the recovery requirement of the tritium 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 defining "a first" or "a second" may explicitly or implicitly include at least one such feature. In the description of the present invention, the meaning of "plurality" means at least two, for example, two, three, etc., unless specifically defined otherwise.
In the description of the present specification, a description referring to terms "one embodiment," "some embodiments," "examples," "specific examples," or "some examples," etc., 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 present invention. In this specification, schematic representations of the above terms are not necessarily directed 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, the different embodiments or examples described in this specification and the features of the different embodiments or examples may be combined and combined by those skilled in the art without contradiction.
In the description of the present invention, it should be noted that, unless explicitly specified and limited otherwise, the terms "mounted," "connected," and "connected" are to be construed broadly, and may be either fixedly connected, detachably connected, or integrally connected, for example; can be mechanically or electrically connected; can be directly connected or indirectly connected through an intermediate medium, and can be communication between two elements. The specific meaning of the above terms in the present invention will be understood in specific cases by those of ordinary skill in the art.
The foregoing description of the preferred embodiments of the invention is not intended to limit the invention to the precise form disclosed, and any such modifications, equivalents, and alternatives falling within the spirit and scope of the invention are intended to be included within the scope of the invention.
Claims (7)
1. The tritium recovery and purification system of the waste neutron target is characterized by comprising a main pipeline (12), wherein two ends of the main pipeline (12) are respectively provided with an air inlet and an air outlet, and the two ends are respectively communicated with a pyrolysis cavity (1) and a storage cavity (2); the air outlet of the main pipeline (12) is communicated with a 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 hydrogen storage material particles are one or the combination of more than two of uranium, zirconium alloy, titanium alloy, laNi alloy, vanadium alloy and palladium alloy; the particle size of the hydrogen storage material particles is 1-5mm; an isolation net with 100-500 meshes is fixed at the communication part of the storage cavity (2) and the main pipeline (12);
the tritium separation system (3) comprises at least two tritium separation units, and a palladium alloy membrane (30) is fixedly arranged in each tritium separation unit; the palladium alloy membrane (30) divides the inner cavity of the tritium separation unit into a tritium-enriched cavity and a tritium-depleted cavity; the air outlet of the main pipeline (12) is communicated with a tritium-enriched cavity of one tritium separation unit; the tritium-enriched cavity of one tritium separation unit is communicated with the tritium-enriched cavity of the other tritium separation unit through a pipeline, and the tritium-depleted cavity of the other tritium separation unit is also communicated with the tritium-enriched cavity of the one tritium separation unit through another pipeline; the tritium-depleted cavity of one tritium separation unit is also communicated with an exhaust gas collection tank (4); the tritium-enriched cavity of the other tritium separation unit is also communicated with a tritium bed (5); a one-way valve (9), a valve (6) and a tritium pump (10) are arranged on a pipeline which is communicated with the tritium-rich cavity of one tritium separation unit and the tritium-rich cavity of the other tritium separation unit; and each tritium-enriched cavity is provided with a tritium measuring instrument (11); the tritium pump (10) is internally provided with the hydrogen storage material particles;
the inner surfaces of the main pipeline (12), the pyrolysis cavity (1), the storage cavity (2), the vacuum pump group (7) and the tritium separation system (3) are coated with tritium-resisting coatings;
the palladium alloy membrane (30) comprises a support material and an alloy coating coated on the support material;
the alloy coating is made of PdArgAuNiCuY alloy; wherein, the content of Pd is 65at%, the content of Ag is 20at%, the content of Au is 8at%, the content of Ni is 1.5at%, the content of Cu is 5at%, and the content of Y is 0.5at%.
2. The tritium recovery and purification system of a spent neutron target according to claim 1, wherein the thickness of the alloy coating is 1-1000 μm.
3. A tritium recovery and purification system for waste neutron targets according to any one of claim 1-2,
the tritium-resistant coating includes at least one of an oxide ceramic coating, a diamond-like coating, a SiC coating, and a nitride coating.
4. A method of tritium recovery and purification of waste neutron targets using the system of any one of claims 1-3, comprising the steps of:
s1, closing an air inlet and an air outlet of the main pipeline (12); the waste neutron target is placed into the pyrolysis cavity (1) to be heated, and gas released by the waste neutron target is obtained;
s2, enabling the gas to enter the storage cavity (2) through the main pipeline (12), and enabling the hydrogen storage material particles in the storage cavity (2) to absorb hydrogen isotope mixed gas in the gas;
s3, opening 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 an air outlet of the main pipeline (12), and preserving heat for 10-40min to enable the hydrogen storage material particles to release the hydrogen isotope mixed gas;
s5, enabling the hydrogen isotope mixed gas to enter at least two tritium separation units of the tritium separation system (3) through the main pipeline (12), and separating the hydrogen isotope mixed gas into tritium-enriched gas and tritium-depleted gas by the palladium alloy membrane (30); the tritium-enriched gas flows out through the tritium-enriched cavity and is recovered, and the tritium-depleted gas flows out through the tritium-depleted cavity and is recovered.
5. The method for recovering and purifying tritium from waste neutron target according to claim 4, wherein 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 40-80min.
6. The method according to claim 4, wherein in step S3, the vacuum pump set (7) is turned on until the vacuum degree in the system is less than 10 -6 Pa, continuing to keep the vacuum pump set (7) open for 30min, and closing the vacuum pump set (7).
7. The method for recovering and purifying tritium of a waste neutron target according to claim 4, wherein,
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); -opening the vacuum pump group (7);
and introducing inert gas into the system through an air inlet of the main pipeline (12) to purge, so that the inert gas carries tritium gas remained in the system to flow out through the vacuum pump set (7), and purifying the flowing gas.
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Citations (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN1715179A (en) * | 2005-06-07 | 2006-01-04 | 四川材料与工艺研究所 | Hydrogen isotope separating device and method |
| CN104122071A (en) * | 2014-08-07 | 2014-10-29 | 中国人民解放军陆军军官学院 | Thermal hydraulic characteristic testing device and method for tritium blowing helium of breeding blanket pebble bed of fusion reactor |
| CN105236354A (en) * | 2015-09-02 | 2016-01-13 | 中国工程物理研究院材料研究所 | Device of recycling tritium from tritiated water and implementation method of same |
| CN105923604A (en) * | 2016-04-20 | 2016-09-07 | 中国工程物理研究院核物理与化学研究所 | Device and method for quickly recovering deuterium and tritium in fusion reactor discharged ash gases |
| CN106629640A (en) * | 2017-01-20 | 2017-05-10 | 中国工程物理研究院材料研究所 | Low-release-ratio helium-3 gas purification device and purification method thereof |
| CN108328577A (en) * | 2017-03-28 | 2018-07-27 | 北京中联电科技术有限公司 | A kind of hydrogen purifier |
| CN112281004A (en) * | 2020-09-30 | 2021-01-29 | 武汉船用电力推进装置研究所（中国船舶重工集团公司第七一二研究所） | Preparation method of tubular palladium alloy membrane |
| CN112281016A (en) * | 2020-09-30 | 2021-01-29 | 武汉船用电力推进装置研究所（中国船舶重工集团公司第七一二研究所） | Palladium alloy for hydrogen permeation and preparation method thereof |
-
2022
- 2022-07-28 CN CN202210898869.7A patent/CN115385302B/en active Active
Patent Citations (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN1715179A (en) * | 2005-06-07 | 2006-01-04 | 四川材料与工艺研究所 | Hydrogen isotope separating device and method |
| CN104122071A (en) * | 2014-08-07 | 2014-10-29 | 中国人民解放军陆军军官学院 | Thermal hydraulic characteristic testing device and method for tritium blowing helium of breeding blanket pebble bed of fusion reactor |
| CN105236354A (en) * | 2015-09-02 | 2016-01-13 | 中国工程物理研究院材料研究所 | Device of recycling tritium from tritiated water and implementation method of same |
| CN105923604A (en) * | 2016-04-20 | 2016-09-07 | 中国工程物理研究院核物理与化学研究所 | Device and method for quickly recovering deuterium and tritium in fusion reactor discharged ash gases |
| CN106629640A (en) * | 2017-01-20 | 2017-05-10 | 中国工程物理研究院材料研究所 | Low-release-ratio helium-3 gas purification device and purification method thereof |
| CN108328577A (en) * | 2017-03-28 | 2018-07-27 | 北京中联电科技术有限公司 | A kind of hydrogen purifier |
| CN112281004A (en) * | 2020-09-30 | 2021-01-29 | 武汉船用电力推进装置研究所（中国船舶重工集团公司第七一二研究所） | Preparation method of tubular palladium alloy membrane |
| CN112281016A (en) * | 2020-09-30 | 2021-01-29 | 武汉船用电力推进装置研究所（中国船舶重工集团公司第七一二研究所） | Palladium alloy for hydrogen permeation and preparation method thereof |
Non-Patent Citations (2)
| Title |
|---|
| 山常起等.氚及防氚渗透材料.原子能出版社,2002,196. * |
| 沈建跃.大变革：甲醇能源时代.新华出版社,2018,36. * |
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