CN117809877A - Neutron source storage and transportation container - Google Patents
Neutron source storage and transportation container Download PDFInfo
- Publication number
- CN117809877A CN117809877A CN202311618022.XA CN202311618022A CN117809877A CN 117809877 A CN117809877 A CN 117809877A CN 202311618022 A CN202311618022 A CN 202311618022A CN 117809877 A CN117809877 A CN 117809877A
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- China
- Prior art keywords
- shielding layer
- neutron
- cladding
- neutron source
- source storage
- Prior art date
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- 238000003860 storage Methods 0.000 title claims abstract description 22
- 238000005253 cladding Methods 0.000 claims abstract description 67
- 229910052751 metal Inorganic materials 0.000 claims abstract description 17
- 239000002184 metal Substances 0.000 claims abstract description 17
- 210000004907 gland Anatomy 0.000 claims abstract description 12
- 239000000463 material Substances 0.000 claims description 15
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 claims description 9
- 229910052739 hydrogen Inorganic materials 0.000 claims description 9
- 239000001257 hydrogen Substances 0.000 claims description 9
- ZOXJGFHDIHLPTG-UHFFFAOYSA-N Boron Chemical compound [B] ZOXJGFHDIHLPTG-UHFFFAOYSA-N 0.000 claims description 8
- 229910052796 boron Inorganic materials 0.000 claims description 8
- 239000004698 Polyethylene Substances 0.000 claims description 6
- -1 polyethylene Polymers 0.000 claims description 6
- 229920000573 polyethylene Polymers 0.000 claims description 6
- 239000000941 radioactive substance Substances 0.000 abstract description 3
- 238000012546 transfer Methods 0.000 abstract description 3
- 230000000694 effects Effects 0.000 description 7
- 238000006243 chemical reaction Methods 0.000 description 6
- 230000017525 heat dissipation Effects 0.000 description 4
- 230000008901 benefit Effects 0.000 description 3
- 230000004992 fission Effects 0.000 description 3
- 238000004519 manufacturing process Methods 0.000 description 3
- 230000002285 radioactive effect Effects 0.000 description 3
- ZSLUVFAKFWKJRC-IGMARMGPSA-N 232Th Chemical compound [232Th] ZSLUVFAKFWKJRC-IGMARMGPSA-N 0.000 description 2
- 229910052778 Plutonium Inorganic materials 0.000 description 2
- 229910052776 Thorium Inorganic materials 0.000 description 2
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 239000012188 paraffin wax Substances 0.000 description 2
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 2
- 230000005855 radiation Effects 0.000 description 2
- 230000005258 radioactive decay Effects 0.000 description 2
- 238000011160 research Methods 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 2
- OYEHPCDNVJXUIW-FTXFMUIASA-N 239Pu Chemical compound [239Pu] OYEHPCDNVJXUIW-FTXFMUIASA-N 0.000 description 1
- 239000000956 alloy Substances 0.000 description 1
- 229910045601 alloy Inorganic materials 0.000 description 1
- 230000033228 biological regulation Effects 0.000 description 1
- 230000003749 cleanliness Effects 0.000 description 1
- 230000007547 defect Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000009826 distribution Methods 0.000 description 1
- 238000001125 extrusion Methods 0.000 description 1
- 230000002349 favourable effect Effects 0.000 description 1
- 230000004907 flux Effects 0.000 description 1
- 230000004927 fusion Effects 0.000 description 1
- 238000000034 method Methods 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 239000002245 particle Substances 0.000 description 1
- 230000000191 radiation effect Effects 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 238000002310 reflectometry Methods 0.000 description 1
- 238000004088 simulation Methods 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 238000006467 substitution reaction Methods 0.000 description 1
- 238000012360 testing method Methods 0.000 description 1
- JFALSRSLKYAFGM-OIOBTWANSA-N uranium-235 Chemical compound [235U] JFALSRSLKYAFGM-OIOBTWANSA-N 0.000 description 1
- 238000003466 welding Methods 0.000 description 1
Abstract
The invention belongs to the field of safe storage and transportation of radioactive substances, and particularly relates to a neutron source storage and transportation container. Including outer cladding, interior cladding, lug, support and lug are installed on the outer cladding, are equipped with the shielding layer between outer cladding and the interior cladding, and interior cladding is for holding neutron source part's two-layer metal tube and its overcoat metal tube, constitutes with the mode of flexonics between the metal tube, and interior cladding welds as a whole with the outer cladding, and the gland is installed to interior cladding tip, be provided with radiating fin on the outer cladding. According to the invention, the heat transfer performance is improved by arranging the radiating fins, so that the safety of the transport container can be ensured.
Description
Technical Field
The invention belongs to the field of safe storage and transportation of radioactive substances, and particularly relates to a neutron source storage and transportation container.
Background
Neutron sources are devices that release neutrons. There are a wide variety of neutron sources, from hand-held radioactive sources to the research stacks and fission sources of neutron research facilities. These devices have wide ranging uses in the physical, engineering, medical, nuclear weapons, oil exploration, biology, chemistry, nuclear power and other industries, depending on neutron energy, neutron flux, equipment size, cost, and government regulations.
A reactor neutron source refers to a neutron-producing substance or device that is an integral part of a nuclear reaction. It provides neutrons required for the nuclear reaction, thereby maintaining the normal progress of the nuclear reaction.
The natural radioactive substance refers to an element having reflectivity such as thorium, uranium, plutonium, etc. existing in nature. Neutrons are released during the radioactive decay of these elements. For example, neutrons are generated in the radioactive decay chain of plutonium, which can promote fission reactions with elements such as uranium and thorium, and thus generate more neutrons.
Artificial radioisotope refers to artificially produced reflective isotopes such as uranium-235, plutonium-239, etc. These isotopes emit neutrons in nuclear reactions and promote nuclear reactions with other substances, such as nuclear fission and nuclear fusion.
A neutron generating device is a device capable of generating neutrons and generally comprises a neutron generator and a neutron source. The neutron generator generates neutrons by collision of the particle beam with a target position, and the neutron source uses neutrons emitted by the radioactive isotope as a neutron source of the reactor.
It should be noted that the selection and rational utilization of the neutron source in the reactor is critical to the performance and stable operation of the reactor. Various neutron sources have different applications and performance performances in different reactors, and need to be selected and adjusted according to actual conditions. At the same time, certain expertise and safety precautions are also required for the use of the reflective isotopes to ensure safe operation of the reactor.
Chinese patent: CN202258393U discloses a storage and transportation container of a primary neutron source component for starting a nuclear reactor, the container is in a columnar structure, and comprises an outer cladding, an inner cladding, lifting lugs, a bracket, the bracket and the lifting lugs are arranged on the outer cladding, a neutron shielding layer of hydrogen-containing material and a gamma shielding layer formed by metal are arranged between the outer cladding and the inner cladding, the inner cladding is formed by two layers of metal tubes for accommodating the neutron source component and outer metal tubes thereof in a soft connection mode, the inner cladding and the outer cladding are welded into a whole, and a gland is arranged at the end part of the inner cladding. The patent provides a storage and transportation container which has a simple structure, good shielding property, impact resistance, extrusion resistance and ensured cleanliness of neutron source components and is specially suitable for primary neutron source components for starting nuclear reactors.
However, the neutron source decays during transportation, and decay heat can be dissipated to the atmosphere through the outer cladding, but there are cases where sufficient heat dissipation is not possible, and materials having excellent heat resistance are used for parts such as the outer cladding, which increases manufacturing costs.
In view of this, the present invention has been made.
Disclosure of Invention
In order to solve the technical problems in the prior art, the invention provides the neutron source storage and transportation container, and the heat transfer performance is improved by arranging the radiating fins, so that the safety of the transportation container can be ensured.
The invention comprises the following technical scheme:
the invention provides a neutron source storage and transportation container which comprises an outer cladding, an inner cladding, lifting lugs and a support, wherein the support and the lifting lugs are arranged on the outer cladding, a shielding layer is arranged between the outer cladding and the inner cladding, the inner cladding is a two-layer metal tube for accommodating neutron source components and an outer-layer metal tube thereof, the metal tubes are formed in a flexible connection mode, the inner cladding and the outer cladding are welded into a whole, a gland is arranged at the end part of the inner cladding, and radiating fins are arranged on the outer cladding.
Further, the radiating fins are long-strip-shaped, one ends of the radiating fins are fixed on the outer cladding, and the other ends of the radiating fins extend in a direction away from the outer cladding.
Further, the heat radiating fins are uniformly arranged on the outer cladding.
Further, the outer envelope is integrally formed with the heat radiating fins.
Further, the shielding layer is composed of a neutron shielding layer and a metal gamma shielding layer.
Further, the neutron shielding layer contains hydrogen materials, wherein the hydrogen materials are boron and polyethylene.
Further, the thickness of the neutron shielding layer is 100mm-150mm.
Further, the inner cladding is provided with the gamma shielding layer outside, and the first neutron shielding layer is arranged outside the gamma shielding layer; and the second neutron shielding layer is arranged outside the inner cladding and below the first neutron shielding layer and the second neutron shielding layer.
Further, the gland is arranged to be of a concave structure.
Further, the concave surface of the gland is a plane.
By adopting the technical scheme, the invention has the following advantages:
1. according to the invention, the heat transfer performance is improved by arranging the radiating fins, so that the safety of the transport container can be ensured.
2. The hydrogen-containing material is boron and polyethylene, so that the mechanical strength and shielding effect of the hydrogen-containing material can be improved; based on this, when designing neutron shielding layer, can make the thickness of neutron shielding layer littleer, the cost is reduced, sets up the volume of the corresponding outer cladding of neutron shielding layer littleer in addition, can reduce manufacturing cost.
3. The gland is arranged in a concave structure, and has the advantage of impact resistance; avoiding the influence on neutron source components in the transportation device when the transportation device falls.
Drawings
In order to more clearly illustrate the embodiments of the present invention or the technical solutions in the prior art, the drawings that are required in the embodiments or the description of the prior art will be briefly described, and it is obvious that the drawings in the following description are some embodiments of the present invention, and other drawings may be obtained according to these drawings without inventive effort for a person skilled in the art.
FIG. 1 is a schematic diagram of a neutron source storage and transportation container according to an embodiment of the present invention;
in the figure: 10-outer cladding, 20-inner cladding, 30-lifting lug, 40-bracket, 50-shielding layer, 51-neutron shielding layer, 511-first neutron shielding layer, 512-second neutron shielding layer, 52-gamma shielding layer, 60-neutron source component, 70-gland, 80-radiating fin, 90-simulation neutron source component.
Detailed Description
The following description provides many different embodiments, or examples, for implementing different features of the invention. The elements and arrangements described in the following specific examples are presented for purposes of brevity and are provided only as examples and are not intended to limit the invention.
For the purpose of making the objects, technical solutions and advantages of the embodiments of the present invention more apparent, the technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings in the embodiments of the present invention, and it is apparent that the described embodiments are some embodiments of the present invention, but not all embodiments of the present invention. All other embodiments, which can be made by those skilled in the art based on the embodiments of the invention without making any inventive effort, are intended to be within the scope of the invention.
The embodiment provides a neutron source storage and transportation container, which comprises an outer cladding 10, an inner cladding 20, lifting lugs 30 and a support 40, wherein the support 40 and the lifting lugs 30 are arranged on the outer cladding 10, a shielding layer 50 is arranged between the outer cladding 10 and the inner cladding 20, the inner cladding 20 is formed by two layers of metal pipes for accommodating a neutron source part 60 and outer metal pipes thereof in a flexible connection mode, the inner cladding 20 and the outer cladding 10 are welded into a whole, a gland 70 is arranged at the end part of the inner cladding 20, and radiating fins 80 are arranged on the outer cladding 10.
In some embodiments, the heat dissipating fin 80 is elongated, and one end of the heat dissipating fin 80 is fixed to the outer envelope 10, and the other end extends away from the outer envelope 10. It should be appreciated that the heat sink fins 80 are located on an extension of the diameter of the outer envelope 10 if the outer envelope 10 is cylindrical. The heat radiation fins 80 of this structure have a better heat radiation effect because of small interference with each other in heat radiation space. Of course, it should be noted that other structures of the heat dissipation fins 80 are also within the scope of the present invention.
In some embodiments, the heat sink fins 80 are uniformly disposed on the outer envelope 10. The structure has better heat dissipation effect.
When the outer envelope 10 and the radiating fins 80 are assembled, the radiating fins 80 can be welded to the outer envelope 10 by a welding machine, and thus, an improved operation can be performed on an existing transport container, so that new equipment investment is not required, and the manufacturing cost of the shield can be reduced.
In some embodiments, the outer envelope 10 is integrally formed with the heat sink fins 80. The integrated structure is more favorable to the heat dissipation, can avoid the weld defect to cause the poor problem of radiating effect.
In some embodiments, the shielding layer 50 is composed of a neutron shielding layer 51 and a metal gamma shielding layer 52.
In some embodiments, the neutron shielding layer 51 comprises a hydrogen-containing material that is boron and polyethylene. The shielding layer 51 adopts boron-containing polyethylene as a neutron shielding material, so that the shielding effect per unit volume is better than that of other conventional materials, and the shielding layer is organized into a solid, so that the problem of shielding performance reduction caused by boron sinking in the conventional paraffin material is avoided. Compared with paraffin materials, the anti-physical impact performance of the alloy is stronger, the boron distribution is more uniform, and the generation of shielding weak points is avoided.
In some embodiments, the neutron shielding layer 51 has a thickness of 100mm-150mm. Since the present invention uses the hydrogen-containing material of boron and polyethylene, the shielding effect and mechanical strength are improved, so that the thickness of the neutron shielding layer 51 can be reduced when the neutron shielding layer is provided, but if the thickness is too small, the shielding effect is poor; the thickness provided by the method can ensure the shielding effect, save materials and reduce cost.
In some embodiments, the inner cladding 20 includes a first neutron shielding layer 511 and a second neutron shielding layer 512, the gamma shielding layer 52 is disposed outside the inner cladding 20, and the first neutron shielding layer 511 is disposed outside the gamma shielding layer 52; a second neutron shielding layer 512 is disposed outside the inner cladding 20 below the first neutron shielding layer 511 and the second neutron shielding layer 512.
In some embodiments, the gland 70 is configured as a concave structure.
In some embodiments, the concave inward surface of the gland 70 is planar. Based on this, the neutron source member 60 is more advantageously fixed.
In some embodiments, the storage and transport vessel further includes a simulated neutron source component 90. The simulated neutron source element 90 is not radioactive and can be viewed at close range when used in a mode test.
Although the invention has been described in detail with reference to the foregoing embodiments, it will be understood by those of ordinary skill in the art that: the technical scheme described in the foregoing embodiments can be modified or some technical features thereof can be replaced by equivalents; such modifications and substitutions do not depart from the spirit and scope of the technical solutions of the embodiments of the present invention.
Claims (10)
1. The utility model provides a neutron source stores transport container, including outer cladding (10), interior cladding (20), lug (30), support (40) and lug (30) are installed on outer cladding (10), be equipped with shielding layer (50) between outer cladding (10) and interior cladding (20), interior cladding (20) are for holding the two-layer metal tube of neutron source part (60) and its overcoat metal tube, constitute with the mode of soft connection between the metal tube, interior cladding (20) weld with outer cladding (10) as a whole, gland (70) are installed to interior cladding (20) tip, a serial communication port, be provided with fin (80) on outer cladding (10).
2. The neutron source storage and transportation container according to claim 1, wherein the heat dissipating fins (80) are elongated, one end of each heat dissipating fin (80) is fixed to the outer cladding (10), and the other end extends away from the outer cladding (10).
3. A neutron source storage transport container as claimed in claim 1 or 2, wherein the heat dissipating fins (80) are uniformly arranged on the outer envelope (10).
4. The neutron source storage and transportation container according to claim 1, wherein the outer cladding (10) is integrally formed with the heat dissipating fins (80).
5. A neutron source storage transport container as defined in claim 1, wherein the shielding layer (50) is composed of a neutron shielding layer (51) and a metal gamma shielding layer (52).
6. The neutron source storage transport container of claim 5, wherein the neutron shielding layer (51) contains hydrogen material, the hydrogen material being boron and polyethylene.
7. The neutron source storage and transport container of claim 6, wherein the neutron shielding layer (51) has a thickness of 100mm-150mm.
8. The neutron source storage and transport container according to claim 5, wherein the container comprises a first neutron shielding layer (511) and a second neutron shielding layer (512), the gamma shielding layer (52) is arranged outside the inner cladding (20), and the first neutron shielding layer (511) is arranged outside the gamma shielding layer (52); and a second neutron shielding layer (512) is arranged outside the inner cladding (20) and is positioned below the first neutron shielding layer (511) and the second neutron shielding layer (512).
9. A neutron source storage and transportation container as claimed in claim 1, wherein the gland (70) is provided in a concave configuration.
10. The neutron source storage transport container of claim 9, wherein the concave surface of the gland (70) is planar.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202311618022.XA CN117809877A (en) | 2023-11-29 | 2023-11-29 | Neutron source storage and transportation container |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202311618022.XA CN117809877A (en) | 2023-11-29 | 2023-11-29 | Neutron source storage and transportation container |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CN117809877A true CN117809877A (en) | 2024-04-02 |
Family
ID=90425876
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202311618022.XA Pending CN117809877A (en) | 2023-11-29 | 2023-11-29 | Neutron source storage and transportation container |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN117809877A (en) |
-
2023
- 2023-11-29 CN CN202311618022.XA patent/CN117809877A/en active Pending
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