CN108172318B - Molten salt reactor core, molten salt reactor system, fuel circulation system and fuel circulation method - Google Patents
Molten salt reactor core, molten salt reactor system, fuel circulation system and fuel circulation method Download PDFInfo
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- 150000003839 salts Chemical class 0.000 title claims abstract description 116
- 239000000446 fuel Substances 0.000 title claims abstract description 83
- 238000000034 method Methods 0.000 title claims abstract description 23
- 238000009377 nuclear transmutation Methods 0.000 claims abstract description 79
- 230000035755 proliferation Effects 0.000 claims abstract description 58
- 239000002915 spent fuel radioactive waste Substances 0.000 claims abstract description 54
- 150000004673 fluoride salts Chemical class 0.000 claims abstract description 40
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 claims abstract description 35
- 229910002804 graphite Inorganic materials 0.000 claims abstract description 35
- 239000010439 graphite Substances 0.000 claims abstract description 35
- 239000003758 nuclear fuel Substances 0.000 claims abstract description 26
- 238000001228 spectrum Methods 0.000 claims abstract description 23
- ZSLUVFAKFWKJRC-IGMARMGPSA-N 232Th Chemical compound [232Th] ZSLUVFAKFWKJRC-IGMARMGPSA-N 0.000 claims abstract description 18
- 229910052776 Thorium Inorganic materials 0.000 claims abstract description 18
- 206010020843 Hyperthermia Diseases 0.000 claims abstract description 11
- 230000036031 hyperthermia Effects 0.000 claims abstract description 11
- 150000003841 chloride salts Chemical class 0.000 claims abstract description 5
- 150000001804 chlorine Chemical class 0.000 claims description 37
- 239000000047 product Substances 0.000 claims description 28
- 239000012043 crude product Substances 0.000 claims description 21
- 230000004992 fission Effects 0.000 claims description 16
- 239000007788 liquid Substances 0.000 claims description 15
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 10
- 238000004519 manufacturing process Methods 0.000 claims description 9
- 239000004449 solid propellant Substances 0.000 claims description 9
- 230000008569 process Effects 0.000 claims description 6
- 238000000926 separation method Methods 0.000 claims description 6
- 238000000605 extraction Methods 0.000 claims description 5
- VEXZGXHMUGYJMC-UHFFFAOYSA-M Chloride anion Chemical compound [Cl-] VEXZGXHMUGYJMC-UHFFFAOYSA-M 0.000 claims description 3
- 238000007599 discharging Methods 0.000 claims description 3
- 239000000284 extract Substances 0.000 claims description 3
- 238000004334 fluoridation Methods 0.000 claims description 3
- 238000003682 fluorination reaction Methods 0.000 claims description 3
- 238000004064 recycling Methods 0.000 claims 4
- GFRMDONOCHESDE-UHFFFAOYSA-N [Th].[U] Chemical compound [Th].[U] GFRMDONOCHESDE-UHFFFAOYSA-N 0.000 abstract description 5
- 238000011161 development Methods 0.000 abstract description 5
- 238000009825 accumulation Methods 0.000 abstract description 4
- 239000002927 high level radioactive waste Substances 0.000 abstract description 3
- 239000000463 material Substances 0.000 description 6
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 6
- 229910052770 Uranium Inorganic materials 0.000 description 5
- 239000006052 feed supplement Substances 0.000 description 4
- KRHYYFGTRYWZRS-UHFFFAOYSA-M Fluoride anion Chemical compound [F-] KRHYYFGTRYWZRS-UHFFFAOYSA-M 0.000 description 2
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 description 2
- 238000007796 conventional method Methods 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 150000002221 fluorine Chemical class 0.000 description 2
- 238000012805 post-processing Methods 0.000 description 2
- 238000000746 purification Methods 0.000 description 2
- 230000002285 radioactive effect Effects 0.000 description 2
- 230000009257 reactivity Effects 0.000 description 2
- 230000001502 supplementing effect Effects 0.000 description 2
- 241000282414 Homo sapiens Species 0.000 description 1
- 230000008901 benefit Effects 0.000 description 1
- 230000005587 bubbling Effects 0.000 description 1
- 229910052799 carbon Inorganic materials 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 239000003153 chemical reaction reagent Substances 0.000 description 1
- 239000002826 coolant Substances 0.000 description 1
- 230000008878 coupling Effects 0.000 description 1
- 238000010168 coupling process Methods 0.000 description 1
- 238000005859 coupling reaction Methods 0.000 description 1
- 230000002950 deficient Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000004821 distillation Methods 0.000 description 1
- 238000005363 electrowinning Methods 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 239000012530 fluid Substances 0.000 description 1
- 230000007774 longterm Effects 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000001105 regulatory effect Effects 0.000 description 1
- 239000011780 sodium chloride Substances 0.000 description 1
- 230000004083 survival effect Effects 0.000 description 1
- 239000002699 waste material Substances 0.000 description 1
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
- G21F9/308—Processing by melting the waste
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C21/00—Apparatus or processes specially adapted to the manufacture of reactors or parts thereof
-
- 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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- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Manufacturing & Machinery (AREA)
- Plasma & Fusion (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
- Compounds Of Alkaline-Earth Elements, Aluminum Or Rare-Earth Metals (AREA)
Abstract
The invention discloses a molten salt reactor core, a molten salt reactor system, a fuel circulation system and a fuel circulation method. The active region of the molten salt reactor core is provided with a transmutation region and a proliferation region which are independent from each other, and the transmutation region is arranged in the proliferation region; the transmutation zone is used for fast spectrum transmutation of transuranic nuclides under the conditions of chloride salts and no moderator, and transmitting high-amount residual neutrons obtained by transmutation to the proliferation zone; the proliferation area is used for the hyperthermia proliferation of the nuclear fuel thorium under the condition that the fluoride salt and the moderator are graphite spheres. The molten salt reactor core, the molten salt reactor system, the fuel circulation system and the fuel circulation method have excellent thorium uranium proliferation capability and transmutation capability, can efficiently transmute TRUs in spent fuel, can fully utilize nuclear fuel thorium, and successfully solve the problems of nuclear fuel resource shortage and high-level waste accumulation existing in the current nuclear energy development.
Description
Technical Field
The invention relates to a molten salt reactor core, a molten salt reactor system, a fuel circulation system and a fuel circulation method.
Background
The nuclear power is used as a clean, low-carbon and high-energy-density energy source, and has the advantage incomparable with other energy sources. At present, a nuclear power unit in China mainly comprises a pressurized water reactor, a fissile material U-235 which exists only in nature is used as an ignition fuel, a once-through fuel circulation mode is adopted, and spent fuel obtained in the fuel circulation mode still contains un-fissile U-235, U-238, transuranic nuclides (TRUs) and high-radioactivity fission products, and the spent fuel is not subjected to post-treatment in the prior art, but is directly stored or buried. Therefore, the fuel circulation mode has the problems of high uranium ore demand, low nuclear fuel utilization rate and nuclear waste management caused by high radioactive spent fuel accumulation. The TRU has long-term radioactive hazard, and if the TRU is not properly disposed, the TRU can seriously affect the earth environment and biosphere, threaten the survival of human beings and other organisms, and seriously restrict the development of nuclear power industry in China.
In addition, the uranium reserves in China are deficient, the thorium reserves are far more abundant (3-4 times) than the uranium, the economy is higher than the uranium, and the pollution to the environment can be reduced. If thorium resources can be fully utilized, the development of the nuclear power industry in China can be favorably influenced.
However, in the prior art, a molten salt reactor core and a molten salt reactor system which can not only efficiently transmute TRUs in pressurized water reactor spent fuel, but also utilize nuclear fuel thorium are not yet available. Aiming at the national conditions of lean uranium and rich thorium in China, developing a set of molten salt reactor core and molten salt reactor system which can not only transmute TRUs in pressurized water reactor spent fuel, but also utilize nuclear fuel thorium is a technical problem which needs to be solved at present. In addition, a fuel circulation system for coupling TRU transmutation and thorium utilization is not yet available in the prior art, so developing a corresponding fuel circulation system is also a technical problem to be solved in the present day.
Disclosure of Invention
The invention aims to solve the technical problem that a set of molten salt reactor core and a molten salt reactor system which can not only transmute TRU in pressurized water reactor spent fuel but also utilize nuclear fuel thorium are not yet available in the prior art, and provides a novel molten salt reactor core, a molten salt reactor system, a fuel circulation system and a fuel circulation method. The molten salt reactor core, the molten salt reactor system, the fuel circulation system and the fuel circulation method have excellent thorium uranium proliferation capability and transmutation capability, can efficiently transmute TRUs in spent fuel, can fully utilize nuclear fuel thorium, and successfully solve the problems of nuclear fuel resource shortage and high-level waste accumulation existing in the current nuclear energy development.
The invention solves the technical problems by the following technical proposal:
the invention provides a molten salt reactor core, wherein an active region of the molten salt reactor core is provided with a transmutation region and a proliferation region which are independent from each other, and the transmutation region is arranged in the proliferation region;
the transmutation zone is used for fast spectrum transmutation of transuranic nuclides under the conditions of chloride salts and no moderator, and transmitting high-amount residual neutrons obtained by transmutation to the proliferation zone;
the proliferation area is used for the hyperthermia proliferation of the nuclear fuel thorium under the condition that the fluoride salt and the moderator are graphite spheres.
In the invention, a graphite reflecting layer is arranged outside the active region according to the conventional method in the field.
In the invention, the fast spectrum refers to a neutron energy spectrum with neutron energy greater than or equal to 0.5MeV.
In the invention, the hyperthermia spectrum refers to a neutron energy spectrum with neutron energy of 1eV-0.5 MeV.
In the present invention, preferably, said transmutation regions are disposed coaxially with said proliferation regions.
In the present invention, the transmutation zone is conventionally referred to in the art as the space in the molten salt reactor core for the transmutation.
In the invention, the transmutation area is provided with a feed supplement port according to the conventional in the field, and the feed supplement port is used for supplementing chlorine salts containing transuranic nuclides to the transmutation area.
In the present invention, the breeder region is conventionally referred to in the art as the space in the molten salt reactor core for the breeder.
In the present invention, preferably, the volume of molten salt in said transmutation zone is from 0.4 to 0.7 times the total volume of molten salt in said active zone. More preferably, the volume of molten salt in the transmutation region is 0.5 times the total volume of molten salt in the active region. Here, the total volume of the molten salt in the active region refers to the sum of the volume of the molten salt in the transmutation region and the volume of the molten salt in the breeder region.
In the present invention, preferably, the volume of the graphite nodules in the proliferation zone is 58% -62%, for example, 60% of the total volume of the proliferation zone.
In the present invention, preferably, a graphite ball inlet is provided at the top of the proliferation area, a graphite ball outlet is provided at the bottom of the proliferation area, the graphite ball inlet is used for providing graphite balls to the proliferation area during material changing, and the graphite ball outlet is used for discharging graphite balls in the proliferation area during material changing. According to the technical scheme, all the moderator graphite spheres can be replaced under the condition of no shutdown, and the method has great significance in industrial application.
The invention also provides a molten salt reactor system, which comprises the molten salt reactor core, wherein the transmutation area is provided with a chlorine salt circulation outer loop, and the transmutation area and the chlorine salt circulation outer loop form a chlorine salt circulation loop together; a first heat exchanger and a fission product post-treatment unit are sequentially arranged on the chlorine salt circulation outer loop along the chlorine salt flow direction, the first heat exchanger is used for removing heat in the chlorine salt, and the fission product post-treatment unit is used for removing fission products in the chlorine salt; the proliferation area is provided with a fluoride salt circulation external loop, and the proliferation area and the fluoride salt circulation external loop form a fluoride salt circulation loop together; and a second heat exchanger and an extractor are sequentially arranged on the fluoride salt circulation outer loop along the fluoride salt flowing direction, the second heat exchanger is used for removing heat in the fluoride salt, and the extractor is used for extracting and obtaining Pa-233 crude products in the fluoride salt.
In the present invention, sometimes, the fission product post-processing unit is connected in parallel with a first pipeline and the extractor is connected in parallel with a second pipeline because of limited processing capacity of the fission product post-processing unit and the extractor. In the technical scheme, the chlorine salt subjected to heat exchange is divided into two streams, one stream is purified and returned to the transmutation area, and the other stream is directly returned to the transmutation area without purification; the fluoride salt after heat exchange is divided into two streams, one stream returns to the proliferation area after extraction, and the other stream returns to the proliferation area directly without extraction.
In the invention, the first heat exchanger can be also connected with a steam generator and a generator set.
The invention also provides a fuel circulation system which comprises a spent fuel generation unit, a solid fuel post-treatment unit, the molten salt reactor system, a decay tank, a nuclear fuel manufacturing chamber, a small-sized modularized molten salt reactor and a liquid fuel post-treatment unit; wherein,,
the spent fuel generation unit is used for generating spent fuel and sending the spent fuel to the solid fuel post-treatment unit;
the solid fuel post-treatment unit is used for extracting transuranic nuclides from the spent fuel sent from the spent fuel generation unit and sending the extracted transuranic nuclides to the transmutation area;
the extractor is also used for sending the Pa-233 crude product to the decay tank;
the decay tank is used for decaying Pa-233 in the Pa-233 crude product into U-233, and carrying out fluorination separation on the obtained U-233 crude product to obtain a U-233 pure product;
the nuclear fuel manufacturing chamber is used for preparing the U-233 pure product into a starting fuel of the small-sized modularized molten salt reactor, and sending the starting fuel to a reactor core of the small-sized modularized molten salt reactor as reactor core fuel;
the small-sized modularized molten salt reactor is used for incinerating the reactor core fuel and sending spent fuel obtained by incineration to a liquid fuel post-treatment unit;
the liquid fuel post-treatment unit is used for extracting the transuranic nuclides from the spent fuel of the small-sized modular molten salt reactor and sending the extracted transuranic nuclides to a transmutation area of the molten salt reactor system.
In the present invention, preferably, the spent fuel generating unit is a pressurized water reactor.
The invention also provides a fuel circulation method, which adopts the fuel circulation system, and comprises the following steps:
(1) Extracting the transuranic nuclides in the spent fuel generated by the spent fuel generating unit to prepare chlorine salt fuel which is provided for the transmutation area;
(2) The transuranic nuclides undergo fast spectrum transmutation under the conditions of chloride and no moderator to obtain high-amount residual neutrons, and the high-amount residual neutrons are transmitted to the proliferation area; in the transmutation process, the chlorine salt always flows around the chlorine salt circulation loop;
(3) The nuclear fuel thorium in the proliferation zone absorbs the high residual neutrons under the condition that the fluoride salt and the moderator are graphite spheres and then carries out hyperthermia proliferation; during the proliferation process, the fluoride salt always flows around the fluoride salt circulation loop; when the fluoride salt flows through the extractor, pa-233 in the fluoride salt is extracted by the extractor to obtain Pa-233 crude product;
(4) The Pa-233 in the Pa-233 crude product decays in the decay tank to obtain a U-233 crude product, and the U-233 pure product is obtained through fluoridation separation;
(5) The U-233 pure product is manufactured into a start-up fuel of the small-sized modular molten salt reactor in the nuclear fuel manufacturing room, and the start-up fuel is sent to a reactor core of the small-sized modular molten salt reactor to serve as reactor core fuel;
(6) The reactor core fuel is burnt by the small-sized modularized molten salt reactor to obtain spent fuel, and the spent fuel is sent to the liquid fuel post-treatment unit;
(7) And the liquid fuel post-treatment unit extracts the transuranic nuclides from the spent fuel of the small-sized modular molten salt reactor and sends the extracted transuranic nuclides to a transmutation area of the molten salt reactor system.
According to the fuel circulation method, the full-closed circulation of thorium uranium fuel can be realized.
In step (2), the transmutation operating power may be 100MWt/m 3 The above.
In the step (3), the fast spectrum means neutron energy greater than or equal to 0.5MeV. Neutron energy spectrum of (a).
In the step (3), the hyperthermia spectrum refers to a neutron energy spectrum with neutron energy of 1eV-0.5 MeV.
In step (3), the extractant used for the extraction may be one conventionally used in the art, and may be Bi-Li, for example.
The above preferred conditions can be arbitrarily combined on the basis of not deviating from the common knowledge in the art, and thus, each preferred embodiment of the present invention can be obtained.
In the invention, the spent fuel is regulated by government; reagents and materials used, except spent fuel, are commercially available.
The invention has the positive progress effects that: the invention provides a molten salt reactor core, a molten salt reactor system, a fuel circulation system and a fuel circulation method. The molten salt reactor core, the molten salt reactor system, the fuel circulation system and the fuel circulation method have excellent thorium uranium proliferation capability and transmutation capability, can efficiently transmute TRUs in spent fuel, can fully utilize nuclear fuel thorium, and successfully solve the problems of nuclear fuel resource shortage and high-level waste accumulation existing in the current nuclear energy development.
Drawings
Fig. 1 is a top view of the molten salt reactor of example 1.
Fig. 2 is a schematic structural view of the molten salt reactor system of embodiment 1.
Fig. 3 is a schematic diagram of the fuel circulation system of example 1.
Reference numerals illustrate:
transmutation area 10
Proliferation zone 20 graphite sphere inlet 21 graphite sphere outlet 22
Graphite reflective layer 30
Chlorine salt circulation outer loop 40
First heat exchanger 50
Fission product aftertreatment unit 60
Fluoride salt circulation external loop 70
Second heat exchanger 80
Extractor 90
First pipeline 100
Second pipeline 110
Detailed Description
The invention is further illustrated by means of the following examples, which are not intended to limit the scope of the invention. The experimental methods, in which specific conditions are not noted in the following examples, were selected according to conventional methods and conditions, or according to the commercial specifications.
In the following examples, fast spectrum refers to neutron energy spectrum with neutron energy greater than or equal to 0.5 MeV; the hyperthermia spectrum refers to a neutron energy spectrum with neutron energies of 1eV-0.5 MeV.
In the following examples, a transmutation zone is conventionally referred to in the art as space for transmutation in a molten salt reactor core; the breeder zone is conventionally referred to in the art as space for breeder in a molten salt reactor core; the total volume of molten salt in the active zone refers to the sum of the volume of molten salt in the transmutation zone and the volume of molten salt in the breeder zone.
Example 1
1. Molten salt reactor core
The molten salt reactor shown in fig. 1, which has a molten salt reactor core, wherein an active region of the molten salt reactor core has a transmutation region 10 and a breeder region 20 which are independent from each other, and the transmutation region 10 is arranged in the breeder region 20; the transmutation zone 10 is used for fast spectrum transmutation of transuranic nuclides under the conditions of chloride salts and no moderator, and transmits high residual neutrons obtained by transmutation to the proliferation zone 20; the proliferation zone 20 is for the hyperthermia proliferation of nuclear fuel thorium in the presence of a fluoride salt and a moderator that is graphite nodules.
Wherein a graphite reflective layer 30 is provided outside the active region.
Wherein the transmutation zone 10 is coaxially disposed with the multiplication zone 20.
Wherein, the transmutation zone 10 is provided with a feed supplement port, and the feed supplement port is used for supplementing chlorine salts containing transuranic nuclides to the transmutation zone 10.
Wherein the volume of the molten salt in the transmutation zone 10 is 0.5 times the total volume of the molten salt in the active zone, the volume of the molten salt in the proliferation zone 20 is 0.5 times the total volume of the molten salt in the active zone, and the total volume of the molten salt in the active zone is 20m 3 The volume of graphite nodules in the multiplication region 20 is 60% of the total volume of the multiplication region 20.
Wherein, the top of the proliferation area 20 is provided with a graphite nodule inlet 21, the bottom of the proliferation area 20 is provided with a graphite nodule outlet 22, the graphite nodule inlet 21 is used for providing graphite nodules to the proliferation area 20 during material changing, and the graphite nodule outlet 22 is used for discharging the graphite nodules in the proliferation area 20 during material changing.
2. Molten salt reactor system
The molten salt reactor system shown in fig. 2 comprises a molten salt reactor with the molten salt reactor core, a chlorine salt circulation outer loop 40 is arranged in the transmutation area 10, and the transmutation area 10 and the chlorine salt circulation outer loop 40 together form a chlorine salt circulation loop; a first heat exchanger 50 and a fission product aftertreatment unit 60 are sequentially arranged on the chlorine salt circulation outer loop 40 along the chlorine salt flow direction, the first heat exchanger 50 is used for removing heat in the chlorine salt, and the fission product aftertreatment unit 60 is used for removing fission products in the chlorine salt; the proliferation zone 20 is provided with a fluoride salt circulation outer loop 70, and the proliferation zone 20 and the fluoride salt circulation outer loop 70 together form a fluoride salt circulation loop; the second heat exchanger 80 and the extractor 90 are sequentially arranged on the fluoride salt circulation outer loop 70 along the flow direction of the fluoride salt, the second heat exchanger 80 is used for removing heat in the fluoride salt, and the extractor 90 is used for extracting to obtain Pa-233 crude products in the fluoride salt. Wherein fission product aftertreatment unit 60 is in parallel with first conduit 100 and extractor 90 is in parallel with second conduit 110. That is, the chlorine salt after heat exchange is divided into two streams, one stream is purified and returned to the transmutation zone 10, and the other stream is directly returned to the transmutation zone 10 without purification; the heat exchanged fluoride is divided into two streams, one stream is extracted and returned to the proliferation zone 20, and the other stream is directly returned to the proliferation zone 20 without extraction.
Wherein the total thermal power of the molten salt reactor system is 2250MWth, the thermal conversion efficiency is 44.4 percent, the electric power is 1000MWe, the energy is mainly generated by the transmutation zone 10, and the power density of the transmutation zone 10 is about 200MW/m 3 。
3. Fuel circulation system
The fuel circulation system shown in fig. 3 includes a spent fuel generation unit, a solid fuel post-treatment unit, the above molten salt reactor system, a decay tank, a nuclear fuel manufacturing room, a small modular molten salt reactor, and a liquid fuel post-treatment unit. Wherein,,
the spent fuel generating unit is used for generating spent fuel and sending the spent fuel to the solid fuel post-treatment unit, and the spent fuel generating unit is a pressurized water reactor;
the solid fuel post-treatment unit is used for extracting the transuranic nuclides in the spent fuel sent from the spent fuel generation unit and sending the extracted transuranic nuclides to the transmutation area 10 through the chlorine salt fuel inlet;
the extractor 90 is also used for sending Pa-233 crude products to a decay tank;
the decay tank is used for decaying Pa-233 in the Pa-233 crude product into U-233, and carrying out fluorination separation on the obtained U-233 crude product to obtain a U-233 pure product;
the nuclear fuel manufacturing chamber is used for preparing a U-233 pure product into a starting fuel of the small-sized modular molten salt reactor, and sending the starting fuel to a reactor core of the small-sized modular molten salt reactor as reactor core fuel, wherein the annual yield TRU of the small-sized modular molten salt reactor is 0.06kg;
the small-sized modularized molten salt reactor is used for burning the reactor core fuel and sending spent fuel obtained by burning to the liquid fuel post-treatment unit;
the liquid fuel post-treatment unit is used for extracting the transuranic nuclides from the spent fuel of the small modular molten salt reactor and sending the extracted transuranic nuclides to the transmutation zone 10 of the molten salt reactor system.
4. Fuel circulation method
The fuel circulation method adopts the fuel circulation system, and comprises the following steps:
(1) The transuranic nuclides in the spent fuel generated by the spent fuel generating unit are extracted to prepare chlorine salt fuel which is provided for the transmutation area 10;
(2) The transuranic nuclides undergo fast spectrum transmutation under the conditions of chloride and no moderator to obtain high-volume residual neutrons, and the high-volume residual neutrons are transmitted to the proliferation area 20; chlorine salt always flows around a chlorine salt circulation loop in the transmutation process;
(3) The nuclear fuel thorium in the proliferation area 20 absorbs high residual neutrons under the condition that the fluoride salt and the moderator are graphite spheres and then carries out hyperthermia proliferation; during the proliferation process, the fluoride salt always flows around the fluoride salt circulation loop; when the fluorine salt flows through the extractor 90, pa-233 in the fluorine salt is extracted by the extractant Bi-Li in the extractor 90 to obtain Pa-233 crude product;
(4) The Pa-233 in the Pa-233 crude product decays in a decay tank to obtain a U-233 crude product, and the U-233 pure product is obtained through fluoridation separation;
(5) The U-233 pure product was prepared into 71.7mol% LiF-16mol% BeF in a nuclear fuel manufacturing room 2 -12mol％ThF 4 -0.3mol％U-233F 4 Is sent to the core of the small modular molten salt reactor as core fuel;
(6) The reactor core fuel is burnt by a small-sized modularized molten salt reactor to obtain spent fuel, and the spent fuel is sent to a liquid fuel post-treatment unit;
(7) The liquid fuel post-treatment unit extracts the transuranic nuclides from the spent fuel of the small modular molten salt reactor and sends the extracted transuranic nuclides to the transmutation zone 10 of the molten salt reactor system.
In step (1), the chlorine salt fuel is prepared from 45mol% HMCl 3 The molten salt consists of 55mol% NaCl, wherein HM is TRU (MA+Pu) obtained after post-treatment of spent fuel of a pressurized water reactor, the specific components are Np-237, pu-238, pu-239, pu-240, pu-241, pu-242, am-241, am-243, cm-244 and Cm-245, and the mole fractions of the nuclides are respectively as follows: 20.5%, 8.68%, 11.0%, 17.4%, 10.3%, 11.8%, 11.0%, 6.23%, 2.68%, 0.331%.
In the step (2), the high residual neutrons are at a transmutation operation power of 100MWt/m 3 The first heat exchanger 50 is also connected with a steam generator and a generator set, the cold fluid of the first heat exchanger 50 is 92mol% NaBeF 4 +8mol% NaF. Fission product post-treatment unit 60 removes fission products on line by means of He bubbling, reduced pressure distillation, electrowinning, etc., and adds TRU-containing chloride salt to transmutation zone 10 to compensate for reactivity loss according to reactor reactivity changes.
In step (3), the composition of the thorium-containing fluoride in the propagation zone 20 is as follows: 77.5mol% LiF and 22.5mol% ThF; the coolant in the second heat exchanger 80 is FNaBe molten salt.
Effect data:
1. the TRU transmutation rate is about 80%, and TRU annual transmutation can reach 750kg.
Since a 1 million kw-level pressurized water reactor discharges 25 tons per year, with TRU accounting for 1% and a mass of about 250kg, the molten salt reactor system can transmute 3 million kw-level pressurized water reactors each year.
2. The thorium uranium proliferation ratio of the molten salt reactor system is 1.06, 50kg U-233,6 years of the system can be proliferated each year, and a small modularized molten salt reactor with the thermal power of 500MW (the initial loading capacity of U-233 is 303 kg) can be started.
3. The annual output TRU of the small-sized modularized molten salt reactor is 0.06kg, so the molten salt reactor system can fully transmute TRUs generated by the small-sized modularized molten salt reactor.
Claims (9)
1. A molten salt reactor core, characterized in that an active zone of the molten salt reactor core has a transmutation zone and a breeder zone which are independent of each other, and the transmutation zone is arranged in the breeder zone; the transmutation zone and the proliferation zone are coaxially arranged;
the transmutation zone is used for fast spectrum transmutation of transuranic nuclides under the conditions of chloride salts and no moderator, and transmitting high-amount residual neutrons obtained by transmutation to the proliferation zone; the fast spectrum refers to a neutron energy spectrum with neutron energy greater than or equal to 0.5 MeV;
the proliferation area is used for the hyperthermia proliferation of the nuclear fuel thorium under the condition that the fluoride salt and the moderator are graphite spheres; the hyperthermia spectrum refers to a neutron energy spectrum with neutron energy of 1eV-0.5 MeV;
the volume of the molten salt in the transmutation area is 0.4-0.7 times of the total volume of the molten salt in the active area; the total volume of molten salt in the active zone refers to the sum of the volume of molten salt in the transmutation zone and the volume of molten salt in the breeder zone;
the volume of graphite spheres in the proliferation zone accounts for 58% -62% of the total volume of the proliferation zone;
and a graphite reflecting layer is arranged outside the active region.
2. The molten salt reactor core of claim 1, wherein the volume of molten salt in the transmutation region is 0.5 times the total volume of molten salt in the active region.
3. The molten salt reactor core of claim 1, wherein a graphite sphere inlet is provided at a top of the breeder region, a graphite sphere outlet is provided at a bottom of the breeder region, the graphite sphere inlet is used for supplying graphite spheres to the breeder region during a refueling operation, and the graphite sphere outlet is used for discharging graphite spheres from the breeder region during a refueling operation.
4. A molten salt reactor system comprising the molten salt reactor core of any one of claims 1-3, said transmutation regions being provided with chlorine salt circulation outer loops, and said transmutation regions and said chlorine salt circulation outer loops together forming a chlorine salt circulation loop; a first heat exchanger and a fission product post-treatment unit are sequentially arranged on the chlorine salt circulation outer loop along the chlorine salt flow direction, the first heat exchanger is used for removing heat in the chlorine salt, and the fission product post-treatment unit is used for removing fission products in the chlorine salt; the proliferation area is provided with a fluoride salt circulation external loop, and the proliferation area and the fluoride salt circulation external loop form a fluoride salt circulation loop together; and a second heat exchanger and an extractor are sequentially arranged on the fluoride salt circulation outer loop along the fluoride salt flowing direction, the second heat exchanger is used for removing heat in the fluoride salt, and the extractor is used for extracting and obtaining Pa-233 crude products in the fluoride salt.
5. A fuel recycling system comprising a spent fuel generation unit, a solid fuel post-treatment unit, the molten salt reactor system of claim 4, a decay tank, a nuclear fuel manufacturing room, a small modular molten salt reactor, and a liquid fuel post-treatment unit; wherein,,
the spent fuel generation unit is used for generating spent fuel and sending the spent fuel to the solid fuel post-treatment unit;
the solid fuel post-treatment unit is used for extracting transuranic nuclides from the spent fuel sent from the spent fuel generation unit and sending the extracted transuranic nuclides to the transmutation area;
the extractor is also used for sending the Pa-233 crude product to the decay tank;
the decay tank is used for decaying Pa-233 in the Pa-233 crude product into U-233, and carrying out fluorination separation on the obtained U-233 crude product to obtain a U-233 pure product;
the nuclear fuel manufacturing chamber is used for preparing the U-233 pure product into a starting fuel of the small-sized modularized molten salt reactor, and sending the starting fuel to a reactor core of the small-sized modularized molten salt reactor as reactor core fuel;
the small-sized modularized molten salt reactor is used for incinerating the reactor core fuel and sending spent fuel obtained by incineration to a liquid fuel post-treatment unit;
the liquid fuel post-treatment unit is used for extracting the transuranic nuclides from the spent fuel of the small-sized modular molten salt reactor and sending the extracted transuranic nuclides to a transmutation area of the molten salt reactor system.
6. The fuel recycling system of claim 5, wherein the spent fuel generating unit is a pressurized water reactor.
7. A fuel circulation method employing the fuel circulation system according to claim 5 or 6, comprising the steps of:
(1) Extracting the transuranic nuclides in the spent fuel generated by the spent fuel generating unit to prepare chlorine salt fuel which is provided for the transmutation area;
(2) The transuranic nuclides undergo fast spectrum transmutation under the conditions of chloride and no moderator to obtain high-amount residual neutrons, and the high-amount residual neutrons are transmitted to the proliferation area; in the transmutation process, the chlorine salt always flows around the chlorine salt circulation loop;
(3) The nuclear fuel thorium in the proliferation zone absorbs the high residual neutrons under the condition that the fluoride salt and the moderator are graphite spheres and then carries out hyperthermia proliferation; during the proliferation process, the fluoride salt always flows around the fluoride salt circulation loop; when the fluoride salt flows through the extractor, pa-233 in the fluoride salt is extracted by the extractor to obtain Pa-233 crude product;
(4) The Pa-233 in the Pa-233 crude product decays in the decay tank to obtain a U-233 crude product, and the U-233 pure product is obtained through fluoridation separation;
(5) The U-233 pure product is manufactured into a start-up fuel of the small-sized modular molten salt reactor in the nuclear fuel manufacturing room, and the start-up fuel is sent to a reactor core of the small-sized modular molten salt reactor to serve as reactor core fuel;
(6) The reactor core fuel is burnt by the small-sized modularized molten salt reactor to obtain spent fuel, and the spent fuel is sent to the liquid fuel post-treatment unit;
(7) And the liquid fuel post-treatment unit extracts the transuranic nuclides from the spent fuel of the small-sized modular molten salt reactor and sends the extracted transuranic nuclides to a transmutation area of the molten salt reactor system.
8. The fuel recycling method of claim 7, wherein in step (2), said transmuted operating power is 100MWt/m 3 The above.
9. The fuel recycling method according to claim 7, wherein in step (3), the extractant used for the extraction is Bi-Li.
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| CN109273121B (en) * | 2018-12-07 | 2020-11-27 | 中国科学院上海应用物理研究所 | A molten salt fuel multi-reactor system |
| CN110444311B (en) * | 2019-08-13 | 2021-05-18 | 中国科学院上海应用物理研究所 | Method for transmutation of transuranic elements in liquid molten salt reactor |
| EP4055623B1 (en) | 2019-11-08 | 2024-11-27 | Abilene Christian University | Identifying and quantifying components in a high-melting-point liquid |
| CN111627569B (en) * | 2020-05-14 | 2022-07-26 | 中国科学院上海应用物理研究所 | Transuranic fuel, preparation method thereof and transmutation method |
| CN111739665B (en) * | 2020-07-07 | 2022-10-04 | 中国科学院上海应用物理研究所 | A graphite ball moderated molten salt reactor |
| CN112530614B (en) * | 2020-10-14 | 2022-08-30 | 中国科学院上海应用物理研究所 | Liquid molten salt reactor nuclear diffusion protection system |
| WO2023049690A1 (en) | 2021-09-21 | 2023-03-30 | Abilene Christian University | Stabilizing face ring joint flange and assembly thereof |
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