CN115394459A - Ultrahigh flux reactor core based on plate-shaped fuel assembly - Google Patents
Ultrahigh flux reactor core based on plate-shaped fuel assembly Download PDFInfo
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- 239000000446 fuel Substances 0.000 title claims abstract description 121
- 230000004907 flux Effects 0.000 title claims abstract description 41
- 230000000712 assembly Effects 0.000 claims abstract description 52
- 238000000429 assembly Methods 0.000 claims abstract description 52
- 239000002826 coolant Substances 0.000 claims description 27
- 238000005253 cladding Methods 0.000 claims description 15
- 239000006096 absorbing agent Substances 0.000 claims description 6
- 229910008894 U—Mo Inorganic materials 0.000 claims description 5
- 229910052726 zirconium Inorganic materials 0.000 claims description 4
- 238000004519 manufacturing process Methods 0.000 abstract description 7
- 239000000463 material Substances 0.000 abstract description 7
- 238000013461 design Methods 0.000 abstract description 6
- 238000011161 development Methods 0.000 abstract description 6
- 238000012360 testing method Methods 0.000 description 15
- 230000009286 beneficial effect Effects 0.000 description 5
- 239000007788 liquid Substances 0.000 description 5
- 230000009257 reactivity Effects 0.000 description 5
- 229910052797 bismuth Inorganic materials 0.000 description 3
- JCXGWMGPZLAOME-UHFFFAOYSA-N bismuth atom Chemical compound [Bi] JCXGWMGPZLAOME-UHFFFAOYSA-N 0.000 description 3
- 238000002844 melting Methods 0.000 description 3
- 230000008018 melting Effects 0.000 description 3
- 238000011160 research Methods 0.000 description 3
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 3
- XLYOFNOQVPJJNP-ZSJDYOACSA-N Heavy water Chemical compound [2H]O[2H] XLYOFNOQVPJJNP-ZSJDYOACSA-N 0.000 description 2
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 239000003054 catalyst Substances 0.000 description 2
- 238000009826 distribution Methods 0.000 description 2
- 230000002349 favourable effect Effects 0.000 description 2
- 230000002093 peripheral effect Effects 0.000 description 2
- 239000011148 porous material Substances 0.000 description 2
- 239000010935 stainless steel Substances 0.000 description 2
- 229910001220 stainless steel Inorganic materials 0.000 description 2
- 229910052580 B4C Inorganic materials 0.000 description 1
- 241000723353 Chrysanthemum Species 0.000 description 1
- 235000005633 Chrysanthemum balsamita Nutrition 0.000 description 1
- 229910045601 alloy Inorganic materials 0.000 description 1
- 239000000956 alloy Substances 0.000 description 1
- INAHAJYZKVIDIZ-UHFFFAOYSA-N boron carbide Chemical compound B12B3B4C32B41 INAHAJYZKVIDIZ-UHFFFAOYSA-N 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 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
- 238000005457 optimization Methods 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 230000002285 radioactive effect Effects 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C5/00—Moderator or core structure; Selection of materials for use as moderator
- G21C5/02—Details
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C11/00—Shielding structurally associated with the reactor
- G21C11/06—Reflecting shields, i.e. for minimising loss of neutrons
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/28—Selection of specific coolants ; Additions to the reactor coolants, e.g. against moderator corrosion
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/02—Fuel elements
- G21C3/04—Constructional details
- G21C3/06—Casings; Jackets
- G21C3/07—Casings; Jackets characterised by their material, e.g. alloys
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C5/00—Moderator or core structure; Selection of materials for use as moderator
- G21C5/12—Moderator or core structure; Selection of materials for use as moderator characterised by composition, e.g. the moderator containing additional substances which ensure improved heat resistance of the moderator
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C7/00—Control of nuclear reaction
- G21C7/06—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section
- G21C7/08—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods
- G21C7/10—Construction of control elements
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- 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
- Y02E30/30—Nuclear fission reactors
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- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
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- Chemical & Material Sciences (AREA)
- Chemical Kinetics & Catalysis (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
The invention discloses an ultrahigh flux reactor core based on a plate-shaped fuel assembly, which relates to the technical field of nuclear reactor design and comprises the plate-shaped fuel assembly, a control rod assembly and a reflecting layer assembly; the plate-shaped fuel assemblies and the control rod assemblies are arranged in the core active area compactly, and the control rod assemblies are distributed at the periphery of the core active area; the interior of the reflecting layer assembly is cooledThe core active region is located inside the reflecting layer assembly. By adopting the scheme, when the thermal power is not more than 200MW, the refueling period is not less than 90 full-power days, and the average component power density is not more than 1200MW/m 3 Under the condition that the maximum neutron flux in the reactor core exceeds 1 x 10 16 n/cm 2 And/s, the development of material irradiation examination is greatly improved, and the domestic important and scarce isotope production problem is solved.
Description
Technical Field
The invention relates to the technical field of nuclear reactor design, in particular to an ultrahigh flux reactor core based on a plate-shaped fuel assembly.
Background
Nuclear power engineering has evolved without departing from nuclear reactors, while nuclear reactors have evolved without departing from test reactors. The test reactor plays an important role in the development of various reactor types. The high neutron flux engineering test reactor is one of the important marks of national science and technology strength, and is essential infrastructure and an important tool for national independent and independent nuclear energy development. These all depend on the neutron flux level in the test reactor, and the higher the neutron flux, the better its irradiation and isotope production, etc.
The neutron flux of the advanced test reactor established internationally is 1.0 multiplied by 10 15 n/cm 2 In the order of/s, the flux exceeds 2.0 x 10 15 n/cm 2 The test piles per s are few. Typical advanced test piles are the Chinese advanced research pile (CARR pile) and the French JHR pile. The CARR stack adopts U3Si2-Al dispersed flat fuel and square box fuel components to form a square grid, and U-235 is enrichedThe degree is 20 percent, and the core uranium density is 4.0gU/cm 3 . Be is filled between the reactor core container and the fuel assembly, and a heavy water reflecting layer annular water tank is arranged outside the reactor core container. The JHR stack adopts a U3Si2-Al cylindrical fuel and daisy type grid arrangement mode, the enrichment degree of U-235 is 27 percent, and the core uranium density is 4.8gU/cm 3 . Be is selected as a reflecting layer at the periphery of the reactor core.
The new generation advanced test reactor design increasingly adopts the fourth generation reactor type, for example, the high flux reactor MBIR which is expected to be constructed in Russia belongs to the concept of sodium-cooled fast reactor, the thermal power is 150MW, and the maximum fast neutron flux level is 5.3 multiplied by 10 15 n/cm 2 And(s) in the presence of a catalyst. Currently, the atton national laboratory is working on developing a conceptual design of a radiation test stack called a multifunctional test stack (VTR). VTR belongs to the concept of sodium-cooled fast reactor, the thermal power of the reactor is 300MW, and the maximum fast neutron flux level is 4.0 multiplied by 10 15 n/cm 2 And(s) in the presence of a catalyst. The reflective layer design of these new pilot stacks typically uses depleted uranium or stainless steel materials.
However, the higher the flux and the greater the core power density, the higher the fuel core temperature and the cladding temperature increase, which requires that the coolant have sufficient capacity to carry away heat while ensuring that the fuel core maximum temperature and the cladding temperature are at a sufficient safe distance from the respective melting limits.
Disclosure of Invention
The invention aims to provide an ultrahigh flux reactor core based on a plate-shaped fuel assembly, by adopting the scheme, the average assembly power density is not more than 1200MW/m when the thermal power is not more than 200MW, the refueling period is not less than 90 full power days 3 Under the condition that the maximum neutron flux in the reactor core exceeds 1 x 10 16 n/cm 2 And/s, the development of material irradiation examination is greatly improved, and the domestic important and scarce isotope production problem is solved.
The invention is realized by the following technical scheme:
an ultra-high flux reactor core based on a plate-shaped fuel assembly, comprising a plate-shaped fuel assembly, a control rod assembly and a reflective layer assembly;
the plate-shaped fuel assemblies and the control rod assemblies are arranged in the core active area compactly, and the control rod assemblies are distributed at the periphery of the core active area;
the inside of the reflecting layer assembly is filled with a coolant, and the core active area is positioned inside the reflecting layer assembly.
Compared with the prior art, the problem that the neutron flux of nuclear facilities is lower is solved, the scheme provides a plate-shaped fuel assembly-based ultrahigh-flux reactor core, and in the concrete scheme, the core comprises a plurality of plate-shaped fuel assemblies, a plurality of control rod assemblies and a reflecting layer assembly located on the outer layer, wherein the plurality of plate-shaped fuel assemblies and the plurality of control rod assemblies are mutually and tightly arranged in the core active area, the plurality of control rod assemblies are divided into two groups, the compensation rod group and the safety rod group, the compensation rod group mainly compensates the reactivity loss caused by the fuel consumption, the safety rod group is mainly used for emergency shutdown, and the compensation rod group and the safety rod group respectively form a set of control system and can be used for independent shutdown. The control rod assemblies are mainly arranged at the periphery of the core active area, can control the power distribution of peripheral assemblies, improve the power density of the fuel assemblies in the central area and are beneficial to improving the maximum neutron flux density of the core; in addition, a reflecting layer is arranged outside the active region, is filled with a coolant and accommodates the core active region, so that the temperature can be reduced, and neutron leakage can be reduced;
the above arrangement is intended to achieve: when the thermal power is not more than 200MW, the refueling period is not less than 90 full-power days, and the average module power density is not more than 1200MW/m 3 Under the condition of (2), the maximum neutron flux in the reactor core exceeds 1 x 10 16 n/cm 2 And/s, thereby greatly improving the development of material irradiation tests and solving the domestic important and scarce isotope production problem. The indexes of the invention are far beyond the level of the current international test piles and the level of the international advanced test concept piles under research.
Further optimizing, the plate-shaped fuel assembly comprises a hexagonal structure with a hexagonal cross section, wherein the hexagonal structure is formed by mutually surrounding and splicing three sub-structures with parallelogram cross sections;
the substructure comprises a plurality of fuel plates which are parallel to each other and are arranged at equal intervals, each fuel plate comprises a fuel core and a fuel cladding for wrapping the fuel core, and both sides of each fuel plate are fixedly provided with structural plates;
and a coolant flow channel is reserved between two adjacent fuel plates.
Compared with the prior art, the higher the power density of the reactor core is, the higher the temperature of the fuel core and the temperature of the cladding are, so that the coolant is required to have enough capacity to take away heat, and meanwhile, the problem that the maximum temperature of the fuel core and the temperature of the cladding have enough safety distance from the corresponding melting limit value is solved; in the specific scheme, the fuel assembly comprises three sub-structures with cross sections in a parallelogram shape, and obtuse-angle end parts of the three sub-structures are mutually combined and spliced, so that a hexagonal structure with a cross section in a regular hexagon shape is formed, and the hexagonal structure is a plate-shaped fuel assembly; each substructure comprises a plurality of fuel plates and two structural plates, the plurality of fuel plates are parallel to each other and are arranged at equal intervals, and the two structural plates are respectively connected with two ends of each fuel plate so as to fix the fuel plates; at the moment, a coolant flow channel is formed between two adjacent fuel plates, and the wider strip-shaped coolant flow channel is beneficial to taking away the heat of the fuel plates, so that the temperature of the plate-shaped fuel assembly and the cladding is reduced, and the safety of the reactor core is improved.
Further optimizing, the thickness of the fuel core body is 0.5-0.9 mm, and the width of the coolant flow channel is 3-4 mm; for better carrying away the fuel plate heat.
Further optimizing, the fuel core adopts U-Zr, U-Mo or U-Pu-Zr; for increasing the maximum neutron flux density.
Preferably, the shell of the control rod assembly is an assembly box, and the cross section size of the assembly box is the same as that of the plate-shaped fuel assembly; the cross sections are all set to be hexagonal structures with the same size, so that compact splicing is facilitated.
Further optimization, a plurality of guide pipes are arranged inside the assembly box, and absorbers are arranged in the guide pipes; for absorbing neutrons.
Further optimizing, the control rod assemblies are divided into a plurality of compensation rod groups and a plurality of safety rod groups; the compensation rod group mainly compensates reactivity loss caused by burnup, the safety rod group is mainly used for emergency shutdown, and the compensation rod group and the safety rod group respectively form a set of control system and can be used for independent shutdown.
Further optimizing, a plurality of the plate-shaped fuel assemblies and the control rod assemblies are compactly arranged to form a combined structure with a hexagonal cross section, the safety rod groups are arranged at the center of the side edge of the combined structure, and the plurality of the compensation rod groups and the plurality of the safety rod groups are respectively arranged in a rotational symmetry manner; wherein, 6 boxes of safety rod groups are arranged according to 60-degree rotational symmetry, and the 6 boxes of safety rod groups are positioned at the center of 6 edges of a hexagon surrounded by the active area; and two circles of other compensating rod components also meet 60-degree rotational symmetry; the specific setting position is determined according to the actual splicing position.
Further optimizing, wherein the height of the core active area is between 40 and 60 cm; both axial ends of the reflecting layer assembly exceed the reactor core active area by 50-100 cm; and the neutron leakage can be reduced more favorably.
Further optimized, the coolant adopts liquid lead or liquid lead bismuth.
Further optimizing, the reactor core is applied to the reactor core with the thermal power not exceeding 200MW, the refueling period is not less than 90 full-power days, and the average component power density is not more than 1200MW/m 3 Under the conditions of (a).
Compared with the prior art, the invention has the following advantages and beneficial effects:
1. the reactor core has the advantages that the thermal power is not more than 200MW, the refueling period is not less than 90 full-power days, and the average component power density is not more than 1200MW/m 3 Under the condition of (2), the maximum neutron flux in the reactor core exceeds 1 x 10 16 n/cm 2 And/s, the maximum neutron flux of the reactor core provided by the invention is far higher than that of the currently built or planned reactor.
2. The plate-shaped fuel assembly is provided with the coolant flow channels on two sides, so that heat of the fuel element can be led out, and the temperature of the core and the cladding can be controlled within a proper temperature range.
3. The core reflecting layer can be provided with pore channels and the like, and can simultaneously realize tasks such as material irradiation, isotope production and the like.
Drawings
In order to more clearly illustrate the technical solutions of the exemplary embodiments of the present invention, the drawings that are required in the embodiments will be briefly described below, it should be understood that the following drawings only illustrate some embodiments of the present invention and therefore should not be considered as limiting the scope, and that those skilled in the art may also derive other related drawings based on these drawings without inventive effort. In the drawings:
FIG. 1 is a cross-sectional view of a plate-shaped fuel assembly according to an embodiment of the present invention;
FIG. 2 is a cross-sectional view of a control rod assembly according to one embodiment of the present invention;
FIG. 3 is a cross-sectional view of a core loading design according to an embodiment of the present invention.
Reference numbers and corresponding part names in the drawings:
1-structural plate, 2-fuel plate, 3-coolant flow channel, 4-component box, 5-absorber, 6-reflection layer component, 7-control rod component, 8-plate-shaped fuel component.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.
Examples
The present embodiment provides an ultra high flux reactor core based on plate-shaped fuel assemblies, as shown in fig. 2 and 3, comprising a plate-shaped fuel assembly 8, a control rod assembly 7 and a reflective layer assembly 6;
the plate-shaped fuel assemblies 8 and the control rod assemblies 7 are arranged in the core active area compactly, and the control rod assemblies 7 are distributed at the periphery of the core active area;
the inside of the reflective layer assembly 6 is filled with a coolant, and the core active region is located inside the reflective layer assembly 6.
Compared with the prior art, the problem that the neutron flux of nuclear facilities is lower is solved, the scheme provides a plate-shaped fuel assembly-based ultrahigh-flux reactor core, in the concrete scheme, the core comprises a plurality of plate-shaped fuel assemblies 8, a plurality of control rod assemblies 7 and a reflecting layer assembly 6 positioned on the outer layer, wherein the plurality of plate-shaped fuel assemblies 8 and the plurality of control rod assemblies 7 are mutually and tightly arranged in the core active area, the plurality of control rod assemblies 7 are divided into two groups, the compensating rod groups and the safety rod groups, the compensating rod groups mainly compensate the reactivity loss caused by the fuel consumption, the safety rod groups are mainly used for emergency shutdown, and the compensating rod groups and the safety rod groups respectively form a set of control system and can be used for independent shutdown. The control rod assemblies 7 are mainly arranged at the periphery of a core active area, can control the power distribution of peripheral assemblies, improve the power density of fuel assemblies in a central area and are beneficial to improving the maximum neutron flux density of the core; in addition, a reflecting layer is arranged outside the active region, is filled with a coolant and accommodates the core active region, so that the temperature can be reduced, and neutron leakage can be reduced; target assemblies are also dispersed within the core active area, which target assemblies can be used to place target channels for material irradiation or isotope production, which when not inserted, are filled with coolant.
In the above arrangement, the ultrahigh flux core preferably comprises 109 plate-shaped fuel assemblies 8, 18 control rod assemblies 7, the core outer diameter is 300cm and the core height is 150cm, as shown in FIG. 3. The fuel assembly center-to-center spacing was 7.4cm. The 18-box control rod assemblies 7 are divided into two groups: the compensating rod group and the safety rod group, wherein the compensating rod group has 12 boxes and mainly compensates reactivity loss caused by burnup, and the safety rod group has 6 boxes and is mainly used for scram. The compensation rod group and the safety rod group respectively form a set of control system, and can be independently stopped. The 6-box safety rod assemblies are arranged in 60-degree rotational symmetry, and the six-box safety rod assemblies are positioned in the center of the 6 sides of the approximate hexagon enclosed by the active area. The other compensating rod groups are provided with 60-degree rotational symmetry; as shown in the figure3, the thermal power of the reactor core formed by the 109-box plate-shaped fuel assemblies 8 is 200MW, the refueling period is 90 full-power days, and the maximum neutron flux in the refueling period is 1.03 multiplied by 10 16 n/cm 2 S, average module power density 1150MW/m 3 。
The above arrangement is intended to achieve: when the thermal power does not exceed 200MW, the refueling period is not less than 90 full-power days, and the average module power density does not exceed 1200MW/m 3 Under the condition that the maximum neutron flux in the reactor core exceeds 1 x 10 16 n/cm 2 And/s, thereby greatly improving the development of material irradiation examination and solving the domestic important and scarce isotope production problem. The indexes of the invention are far beyond the level of the current international test piles and the level of the international advanced test concept piles under research.
As shown in fig. 1, the present embodiment provides a plate-shaped fuel assembly 8, which includes a hexagonal structure with a hexagonal cross section formed by mutually enclosing and splicing three sub-structures with parallelogram cross sections;
the substructure comprises a plurality of fuel plates 2 which are parallel to each other and are arranged at equal intervals, the fuel plates 2 comprise fuel cores and fuel cladding used for wrapping the fuel cores, and the two sides of the plurality of fuel plates 2 are fixedly provided with structural plates 1;
a coolant flow channel 3 is left between two adjacent fuel plates 2.
Compared with the prior art, the higher the power density of the reactor core is, the higher the temperature of the fuel core and the temperature of the cladding are, so that the coolant has enough capacity to take away heat, and meanwhile, the problem that the maximum temperature of the fuel core and the temperature of the cladding have enough safety distance from the corresponding melting limit value is solved; in the specific scheme, the fuel assembly comprises three sub-structures with cross sections in a parallelogram shape, and obtuse-angle end parts of the three sub-structures are mutually combined and spliced, so that a hexagonal structure with a cross section in a regular hexagon shape is formed, and the hexagonal structure is the plate-shaped fuel assembly 8; each substructure comprises a plurality of fuel plates 2 and two structural plates 1, the plurality of fuel plates 2 are parallel to each other and are arranged at equal intervals, and the two structural plates 1 are respectively connected with two ends of each fuel plate 2 so as to fix the fuel plates 2; at the moment, the coolant flow channel 3 is formed between two adjacent fuel plates 2, and the wider strip-shaped coolant flow channel 3 is beneficial to taking away the heat of the fuel plates 2, so that the temperature of the plate-shaped fuel assembly 8 and the cladding is reduced, and the safety of the reactor core is improved.
As an embodiment more advantageous for taking away heat from the fuel plate 2, it is provided that: the thickness of the fuel core is 0.5-0.9 mm, and the width of the coolant flow channel 3 is 3-4 mm; in the scheme, as shown in fig. 1, one plate-shaped fuel assembly 8 consists of 24 fuel plates 2, each fuel plate 2 consists of a fuel core and a fuel cladding, the thickness of the fuel core is preferably 0.5mm, the width of a coolant flow channel 3 is 3-4 mm, and the wider the coolant flow channel 3 is, the more the heat of the fuel plates 2 can be taken away; wherein the fuel cladding is made of stainless steel and has good compatibility with lead bismuth or lead-based coolant.
As a specific implementation mode for improving the maximum neutron flux density, the following steps are set: the fuel core body adopts U-Zr, U-Mo or U-Pu-Zr; in the scheme, the fuel core can adopt metal fuels such as U-Zr, U-Mo or U-Pu-Zr, and the fuel containing Pu is favorable for improving the maximum neutron flux density, and the fuel core is made of U-Mo alloy.
Referring to FIG. 2, in the present embodiment, the control rod assembly 7 has a housing in the form of an assembly box 4, the assembly box 4 having a cross-sectional dimension that is the same as the cross-sectional dimension of the plate-shaped fuel assemblies 8; the cross sections are all set to be hexagonal structures with the same size, so that the splicing is convenient to compact.
Continuing with FIG. 2, as an embodiment for absorbing neutrons, the arrangement is; a plurality of guide pipes are arranged in the component box 4, and absorbers 5 are arranged in the guide pipes; in the scheme, the control rod assembly 7 comprises an assembly box 4 and seven control rod absorbers 5 positioned in the assembly box 4, wherein the absorbers 5 are made of boron carbide and placed in the guide tubes, and can absorb a large amount of neutrons so as to prevent the fission chain reaction from proceeding.
In the embodiment, the control rod assemblies 7 are divided into a plurality of compensating rod groups and a plurality of safety rod groups; the compensation rod set is mainly used for compensating reactivity loss caused by burnup, the safety rod set is mainly used for emergency shutdown, and the compensation rod set and the safety rod set respectively form a set of control system and can be used for independent shutdown.
Referring to fig. 3, in the present embodiment, a plurality of plate-shaped fuel assemblies 8 and control rod assemblies 7 are compactly arranged to form a combined structure with a hexagonal cross section, a safety rod set is disposed at a central position of a side of the combined structure, and a plurality of compensation rod sets and a plurality of safety rod sets are respectively disposed in rotational symmetry; wherein, 6 boxes of safety rod groups are arranged according to 60-degree rotational symmetry, and the 6 boxes of safety rod groups are positioned at the center of 6 edges of a hexagon surrounded by the active area; and two circles of other compensating rod components also meet 60-degree rotational symmetry; the specific setting position is determined according to the actual splicing position.
As a specific implementation mode which is more favorable for reducing neutron leakage, the method comprises the following steps: the height of the core active area is between 40 and 60 cm; both axial ends of the reflecting layer assembly 6 exceed the reactor core active area by 50-100 cm;
it can be understood that, in the present embodiment, the height of the ultrahigh flux core active region is 50cm, and both ends of the axial reflection layer respectively exceed the core active region by 50cm, the outer diameter of the reflection layer in the radial direction is not less than 200cm, the maximum cladding and core temperature is favorably reduced by the lower height of the active region, and the neutron leakage is favorably reduced by the thicker reflection layer; in addition, the thicker reflecting layer is filled with the coolant, so that ducts and loops for different purposes can be arranged on the reflecting layer, the thickness of the reflecting layer is between 50cm and 100cm, and the thicker reflecting layer not only can shield radioactive rays, but also can increase the space for arranging the loops.
In this embodiment, the coolant used is liquid lead or liquid lead bismuth; at the moment, the reflecting layer is equivalent to a liquid pool, so that the heat of the reactor core can be absorbed, and the safety of the reactor core is improved. Meanwhile, pore channels, loops and the like can be flexibly arranged in the reflecting layer.
In the embodiment, the reactor core is applied to the conditions that the thermal power does not exceed 200MW, the refueling period is not less than 90 full-power days, and the average component power density does not exceed 1200MW/m 3 Under the conditions of (a).
The above embodiments are provided to further explain the objects, technical solutions and advantages of the present invention in detail, it should be understood that the above embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalents, improvements, etc. made within the spirit and principle of the present invention should be included in the scope of the present invention.
Claims (10)
1. An ultra high flux reactor core based on plate shaped fuel assemblies, characterized by comprising a plate shaped fuel assembly (8), control rod assemblies (7) and a reflector assembly (6);
the plate-shaped fuel assemblies (8) and the control rod assemblies (7) are arranged in the core active area compactly, and the control rod assemblies (7) are distributed at the periphery of the core active area;
the inside of the reflecting layer assembly (6) is filled with a coolant, and the core active area is positioned inside the reflecting layer assembly (6).
2. The ultra high flux reactor core based on plate-shaped fuel assemblies according to claim 1, characterized in that the plate-shaped fuel assemblies (8) comprise a hexagonal structure with a hexagonal cross section formed by mutually surrounding and splicing three sub-structures with a parallelogram cross section;
the substructure comprises a plurality of fuel plates (2) which are parallel to each other and are arranged at equal intervals, the fuel plates (2) comprise fuel cores and fuel cladding used for wrapping the fuel cores, and structural plates (1) are fixedly arranged on two sides of the fuel plates (2);
a coolant flow channel (3) is reserved between two adjacent fuel plates (2).
3. The plate-shaped fuel assembly-based ultrahigh flux reactor core according to claim 2 wherein the thickness of the fuel core is 0.5 to 0.9mm and the width of the coolant flow channel (3) is 3 to 4mm.
4. The plate-shaped fuel assembly-based ultrahigh flux reactor core according to claim 2, wherein the fuel core is U-Zr, U-Mo or U-Pu-Zr.
5. The plate-shaped fuel assembly-based ultrahigh flux reactor core according to claim 1, characterized in that the outer shell of the control rod assembly (7) is an assembly box (4), the cross-sectional dimensions of the assembly box (4) and the plate-shaped fuel assembly (8) being the same.
6. The plate-shaped fuel assembly-based ultrahigh flux reactor core according to claim 5, characterized in that a plurality of guide tubes are arranged inside the assembly box (4), and an absorber (5) is arranged inside the guide tubes.
7. A plate-shaped fuel assembly based ultrahigh flux reactor core according to claim 1 characterized in that a number of the control rod assemblies (7) are divided into a number of compensation rod groups and a number of safety rod groups.
8. The plate-shaped fuel assembly-based ultrahigh-flux reactor core according to claim 7, wherein a plurality of the plate-shaped fuel assemblies (8) and the control rod assemblies (7) are compactly arranged to form a combined structure with a hexagonal cross section, the safety rod groups are arranged at the center positions of the sides of the combined structure, and a plurality of the compensation rod groups and a plurality of the safety rod groups are respectively arranged in a rotational symmetry manner.
9. The ultra high flux reactor core based on plate shaped fuel assemblies of claim 1, wherein the core active area height is between 40-60 cm; and both axial ends of the reflecting layer assembly (6) exceed the core active area by 50-100 cm.
10. The plate-shaped fuel assembly-based ultrahigh flux reactor core of claim 1 wherein the core is applied for a thermal power of no more than 200MW for a refueling cycleNo less than 90 full power days, and average module power density no more than 1200MW/m 3 Under the conditions of (a).
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Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109215811A (en) * | 2018-09-13 | 2019-01-15 | 中国核动力研究设计院 | Hexagon beryllium component and aluminium component nuclear design certificate authenticity reactor core and method of adjustment |
| CN114121308A (en) * | 2021-11-24 | 2022-03-01 | 西安交通大学 | Reactor core structure of lead bismuth cooling fast neutron research reactor with ultra-high flux |
| CN114446496A (en) * | 2022-02-17 | 2022-05-06 | 中国核动力研究设计院 | Ultra-high flux reactor core based on annular fuel element |
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Patent Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109215811A (en) * | 2018-09-13 | 2019-01-15 | 中国核动力研究设计院 | Hexagon beryllium component and aluminium component nuclear design certificate authenticity reactor core and method of adjustment |
| CN114121308A (en) * | 2021-11-24 | 2022-03-01 | 西安交通大学 | Reactor core structure of lead bismuth cooling fast neutron research reactor with ultra-high flux |
| CN114446496A (en) * | 2022-02-17 | 2022-05-06 | 中国核动力研究设计院 | Ultra-high flux reactor core based on annular fuel element |
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