DK181148B1 - A method of operating a nuclear reactor core of a molten salt nuclear reactor - Google Patents

A method of operating a nuclear reactor core of a molten salt nuclear reactor Download PDF

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Publication number
DK181148B1
DK181148B1 DKPA202170281A DKPA202170281A DK181148B1 DK 181148 B1 DK181148 B1 DK 181148B1 DK PA202170281 A DKPA202170281 A DK PA202170281A DK PA202170281 A DKPA202170281 A DK PA202170281A DK 181148 B1 DK181148 B1 DK 181148B1
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DK
Denmark
Prior art keywords
moderator
neutron reflector
salt
jacket
nuclear reactor
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DKPA202170281A
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Danish (da)
Inventor
Stubsgaard Aslak
Jam Pedersen Thomas
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Copenhagen Atomics As
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Priority to DKPA202170281A priority Critical patent/DK181148B1/en
Application filed by Copenhagen Atomics As filed Critical Copenhagen Atomics As
Priority to MX2023014035A priority patent/MX2023014035A/en
Priority to KR1020247000022A priority patent/KR20240032009A/en
Priority to JP2023573101A priority patent/JP2024521176A/en
Priority to PCT/DK2022/050109 priority patent/WO2022253392A1/en
Priority to BR122023025219-5A priority patent/BR122023025219A2/en
Priority to EP22815405.0A priority patent/EP4352750A4/en
Priority to AU2022284200A priority patent/AU2022284200A1/en
Priority to BR112023025005A priority patent/BR112023025005A2/en
Priority to CA3221998A priority patent/CA3221998A1/en
Publication of DK202170281A1 publication Critical patent/DK202170281A1/en
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Publication of DK181148B1 publication Critical patent/DK181148B1/en

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    • 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
    • G21C3/42—Selection of substances for use as reactor fuel
    • G21C3/44—Fluid or fluent reactor fuel
    • G21C3/54—Fused salt, oxide or hydroxide compositions
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactor types
    • G21C1/04—Thermal reactors ; Epithermal reactors
    • G21C1/06—Heterogeneous reactors, i.e. in which fuel and moderator are separated
    • G21C1/22—Heterogeneous reactors, i.e. in which fuel and moderator are separated using liquid or gaseous fuel
    • 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
    • G21C3/24—Fuel elements with fissile or breeder material in fluid form within a non-active casing
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C7/00—Control of nuclear reaction
    • 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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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Chemical Kinetics & Catalysis (AREA)
  • Structure Of Emergency Protection For Nuclear Reactors (AREA)

Abstract

A method of operating a nuclear reactor core (1) of a molten salt nuclear reactor (100) with a tubular cylindrical center moderator and neutron reflector vessel (10) for passage of a liquid moderator and neutron reflector (11) axially extending from a liquid moderator and neutron reflector inlet (12) to a liquid moderator and neutron reflector outlet (13), a cylindrical fuel salt jacket surrounding said center moderator vessel (10), said fuel salt jacket extending axially from a fuel salt inlet (22) to a fuel salt outlet (23) for passage of a molten fuel salt (21), and a cylindrical moderator and neutron reflector jacket surrounding said fuel salt jacket, said moderator and neutron reflector jacket extending axially from liquid moderator and neutron reflector inlet (32) to a moderator and neutron reflector outlet (33) for passage of a liquid moderator and neutron reflector (31), the method comprising: controlling the temperature of the fuel salt (21) in said fuel salt jacket to remain at a temperature between approximately 600 and 700°C, controlling the temperature of said moderator and liquid (11) in said center moderator and neutron reflector vessel (10) to remain at a temperature between approximately 20 and 90°C, and preferably controlling the temperature of said liquid moderator and neutron reflector (31) in said moderator and neutron reflector jacket (30) to remain at a temperature between approximately 20 and 90°C.

Description

DK 181148 B1 1
TECHNICAL FIELD
The disclosure relates to a method of operating a nuclear reactor of a molten salt nuclear reactor molten salt nuclear reactor, in particular to a method of operating a nuclear reactor core of a molten salt nuclear reactor.
BACKGROUND
A molten salt reactor (MSR) is a nuclear reactor where the nuclear reactor coolant and/or the nuclear fuel is a molten salt, typically a fluoride or chloride salt, with a melting point of around ~500 °C, operating temperature of around ~600 to 700 °C, and a boiling point of ~1000 °C above the melting point. One of the many advantages of this type of reactor is that molten salts can be used as the heat transfer media at very high temperatures while still operating at or close to atmospheric pressure. Heat can be extracted from such reactors by pumping the molten salt in a loop between the nuclear reactor core and a heat exchanger with the reactor power being directly proportional to the temperature drop across the heat exchanger and the flow rate. Due to the corrosive nature of molten fluoride and chloride salt, their operation requires an inert containment atmosphere, furthermore molten salt or molten salt vapors cannot be allowed to escape to the environment, putting strict requirements on molten salt reactor components to be completely leak tight. This poses a severe technical challenge, since the temperature, the aggressive nature of the molten salt combined with high radiation levels renders only very few suitable materials to
DK 181148 B1 2 work with. Thus, the materials that can be used to construct the core have to be carefully selected and combined in order to obtain a solution that provides a reliable and durable nuclear reactor core.
Molten salt reactors were built and operated at Oak Ridge
National Laboratory (ORNL) in the 1950s and 1960s with a research program lasting to the 1970s and other small programs around the world.
The nuclear reactor core has a special geometry designed to allow a nuclear chain reaction to take place, achieved by either 1. a large enough amount of fuel to make a critical assembly or by 2. combining enough moderator and fuel to make a critical assembly.
These are respectively called fast and thermal reactors because of the resulting neutron spectrum that each type exhibits.
There is a need for compact and mass manufacturable molten salt breeder reactors in order to achieve the goal of meeting the future global energy demand in a sustainable fashion.
This creates a significant challenge since the smaller one makes a reactor the harder it becomes for it to achieve breeding, since the probability of neutrons leaving the reactor, referred to as neutron leakage, roughly increases with the larger surface area to volume ratio of a small reactor.
DK 181148 B1 3
Fast reactors generally require a much larger fissile inventory to become critical when compared to thermal reactors and are thus not well suited to scale up rapidly to meet future energy demand because of limited availability of fissile material. Thus, a compact and mass manufacturable thermal molten salt breeder reactor is desired.
One of the challenges for thermal molten salt breeder reactors is the need for a moderator that effectively slows down neutrons while allowing for breeding, which means that only moderators based on carbon, beryllium, or deuterium can be used. The only practical moderators that will allow breeding in a thermal spectrum molten salt reactor are: solid carbon, solid beryllium, molten enriched lithium 7 deuteroxide salt (7LiOD), or liquid heavy water (D20). Of these, carbon is the only one that can withstand direct contact with the fuel salt, while the others need to have a structural material separating them from the fuel salt. All these moderators have their pros and cons and have been proposed and studied in the past for use as moderators in a molten salt reactor.
Another challenge is the choice of materials for the vessels that contain the fuel salt or moderator material if the moderator is separated from the fuel salt. The materials need to resist degradation under extremely high temperatures, intense radiation exposure, and must have suitable neutron absorption properties as well as resistance to the corrosive effect of molten salt for the vessels containing molten salt.
Various materials have been proposed and studied in the past for use in the construction of the components of a molten salt reactor. However, each of these materials has practical
DK 181148 B1 4 limitations in relation to the shapes that are possible to produce in commercial manufacturing.
The only known molten salt compatible materials with low neutron absorption are carbon and silicon carbide based materials. The most promising of these are the ceramic composites of silicon carbide and carbon, e.g. carbon fiber- carbon matrix (C/C) composite, silicon carbide fiber-silicon carbide matrix (SiC/SiC) composite, and silicon carbide fiber-carbon matrix (SiC/C) composite. The construction of complex shaped composite materials is much more challenging than from metals since these complex shaped composite materials cannot be bent and or welded in their final form but rather have to be molded to shape and treated in multiple steps, e.g. furnace firing.
CN112259263A discloses a nuclear reactor core according to the preamble of claim 1 with a construction with straight cylindrical vessels, that can only be used as a fast reactor, which requires operating with high temperatures for the liquids used during its operation.
SUMMARY
It is an object to provide a nuclear reactor core for a nuclear reactor that overcomes or at least reduces the problem above.
The foregoing and other objects are achieved by the features of the independent claims. Further implementation forms are apparent from the dependent claims, the description and the figures.
DK 181148 B1
According to a first aspect, there is provided nuclear reactor core for a molten salt nuclear reactor, the nuclear reactor core comprising: 5 - a tubular cylindrical center moderator and neutron reflector vessel for passage of a first liquid moderator and neutron reflector axially extending from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet, - a plurality of tubular fuel salt vessels, each tubular fuel salt vessel extending axially from a fuel salt inlet to a fuel salt outlet for passage of a molten fuel salt, the plurality of tubular fuel salt vessels being assembled to form a cylindrical fuel salt jacket surrounding the center moderator and neutron reflector vessel, - a plurality of tubular moderator and neutron reflector vessels, each tubular moderator neutron reflector vessel extending axially from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet for passage of a liquid moderator and neutron reflector, the plurality of tubular moderator and neutron reflector vessels being assembled to form a cylindrical moderator and neutron reflector jacket surrounding the cylindrical fuel salt jacket.
By constructing the nuclear reactor core with a fuel salt jacket that is formed by a plurality of tubular vessels that, when assembled, form the fuel salt jacket and by constructing the nuclear reactor core with a neutron reflector jacket that is formed by a plurality of tubular vessels than when assembled form the neutron reflector jacket, it becomes
DK 181148 B1 6 possible to construct a nuclear reactor core with a design of concentric cylinders that can have a varying radius to allow a larger volume in the respective concentric spaces between the respective cylinders in which the moderator and reflector liquid and the molten fuel salt is received for achieving a critical assembly and reducing neutron leakage, i.e. an efficient reactor geometry, while still being able to manufacture from materials that are suitable for use in a molten salt nuclear reactor. The components are made from materials that typically need to be made in a mold which in turn poses a range of restrictions in the design for these components. The core construction according to the present invention allows these components to be designed in such a way that they can be manufactured from suitable materials, such as ceramics composite and metal alloy, preferably zirconium alloy. The construction of the nuclear reactor core according to the first aspect allows for an optimized core design with an efficient reactor geometry where only a small amount of fuel salt is needed to create a critical reactor assembly while still being practical to assemble and also allowing for a sufficient fuel salt circulation rate to extract the produced heat.
In a possible implementation of the first aspect, the ceramic composite material is a ceramic matrix composite.
In a possible implementation of the first aspect, the ceramic composite material is a ceramic composite of silicon carbide and carbon, preferably a carbon fiber-carbon matrix (C/C) composite, silicon carbide fiber-silicon carbide matrix
DK 181148 B1 7 (siC/siC) composite, and/or silicon carbide fiber-carbon matrix (SiC/C) composite.
In a possible implementation form of the first aspect, the cylindrical center moderator and neutron reflector vessel has a largest inner cross-sectional area medially between the liquid moderator inlet and the liquid moderator and neutron reflector outlet, with the inner cross-sectional area of the center moderator and neutron reflector vessel preferably increasing gradually towards the medial position between the liquid moderator and neutron reflector inlet and the liquid moderator and neutron reflector outlet. The inner cross- sectional area preferably varying continuously between the moderator and neutron reflector inlet and the moderator and neutron reflector outlet.
In a possible implementation form of the first aspect, the cylindrical center moderator and neutron reflector vessel and/or the liquid moderator and neutron reflector vessels have a wall of a metal alloy, preferably iron alloy, preferably stainless steel and most preferably zirconium alloy or ceramic composite material, and wherein the fuel salt vessels preferably having a wall of ceramic composite material, a metal alloy, an iron alloy, preferably stainless steel, or nickel alloy, preferably Inconel® or Hastelloy®.
In a possible implementation form of the first aspect, each fuel salt vessel in the fuel salt jacket is substantially identical to the other fuel salt vessels in the fuel salt jacket.
DK 181148 B1 8
In a possible implementation form of the first aspect, each moderator and neutron reflector vessel in the moderator and neutron reflector jacket is substantially identical to the other moderator and neutron reflector vessels in the moderator and neutron reflector jacket.
In a possible implementation form of the first aspect, a first cylindrical insulation jacket is provided between the center moderator vessel and the fuel salt jacket, the first insulation jacket preferably being formed by one or more first insulation members, the first insulation members preferably being shaped like sectors of the first cylindrical insulation jacket or shaped as a strip that is spirally wound around the center moderator vessel, and the first insulation members preferably made being of a graphite felt or insulation aerogel impregnated graphite felt. The insulation allows a much lower heat loss to the moderator and neutron reflector liquid and for the moderator and neutron reflector liquid to be consistently kept operating at a much lower temperature than the fuel salt, which is in particularly a significant advantage if a hydroxide or deuteroxide moderator is used, due to the reduced corrosive effect at lower temperatures.
In a possible implementation form of the first aspect, the second cylindrical insulation jacket is provided between the fuel salt jacket and the moderator and neutron reflector jacket, the second insulation jacket preferably being formed by one or more second insulation members, the second insulation members preferably being shaped like sectors of the second cylindrical insulation jacket or shaped as a strip that is spirally wound around the fuel salt jacket, and the
DK 181148 B1 9 second insulation members preferably being made of a graphite felt or insulation aerogel impregnated graphite felt.
In a possible implementation form of the first aspect, the nuclear reactor core comprises a plurality of tubular blanket salt vessels, each tubular blanket salt vessel extending axially from an inlet to an outlet for passage of a molten blanket salt, the plurality of tubular blanket salt vessels being assembled to form a cylindrical blanket salt jacket surrounding the moderator and neutron reflector jacket.
In a possible implementation form of the first aspect, a third cylindrical insulation jacket is provided between the neutron reflector vessel and the blanket salt jacket, the third insulation jacket preferably being formed by one or more third insulation members, the third insulation members preferably being shaped like sectors of the third cylindrical insulation jacket or shaped as a strip that is spirally wound around the moderator and neutron reflector jacket, and the third insulation members preferably being made of a graphite felt or insulation aerogel impregnated graphite felt.
In a possible implementation form of the first aspect, the center moderator and neutron reflector vessel, the fuel salt vessels, the moderator and neutron reflector vessels, and/or the blanket salt vessels are each assembled from two parts that are separated by a symmetry plane M that extends radially from the medial position.
In a possible implementation form of the first aspect, the center moderator and neutron reflector vessel has a variable
DK 181148 B1 10 inner and outer radii R11 and R12, the inner and outer Radii
R11 and R12 preferably being largest at or near the medial position and gradually decreasing from the medial position.
In a possible implementation form of the first aspect, the fuel salt jacket is shaped as a hollow circular cylinder with variable inner and outer radii R21 and R22, and wherein the fuel salt vessels are preferably shaped as a circular cylindrical sector with variable inner and outer radii R21 and R22, respectively, and a given sector angle A2, the inner and outer Radii R21 and R22 preferably being largest at or near the medial position and gradually decreasing from the medial position.
In a possible implementation form of the first aspect, the moderator and neutron reflector jacket is shaped like a hollow circular cylinder with variable inner and outer radii R31 and
R32, and wherein the moderator neutron reflector vessel are preferably shaped like a circular cylindrical sector with variable inner and outer radii R31 and R32, respectively, and a given sector angle A3, the inner and outer Radii R31 and
R32 preferably being largest at or near the medial position and gradually decreasing from the medial position.
In a possible implementation form of the first aspect, the blanket salt jacket is shaped as a hollow circular cylinder with variable inner and outer radii R41 and R42, and wherein the blanket salt vessels is shaped like a circular cylindrical sector with variable inner and outer radii R41 and R42, respectively, and a given sector angle A4, the inner and outer
Radii R41 and R42 preferably being largest at or near the
DK 181148 B1 11 medial position and gradually decreasing from the medial position.
In a possible implementation form of the first aspect, the fuel salt jacket, the neutron reflector jacket, and/or the blanket jacket are circumferentially potentially continuous jackets, preferably formed by the matching sides of the tubular vessels forming these jackets touching one another along a substantial portion of their axial extent.
In a possible implementation form of the first aspect, the center moderator vessel is segmented and formed by a plurality of tubular moderator vessels, each moderator vessel extending axlally from a moderator inlet to a moderator outlet for passage of a moderator, the plurality of tubular moderator vessels being assembled to form the center moderator vessel.
In a possible implementation form of the first aspect, the center moderator vessel has a circumferential outer wall and a circumferential inner wall, the circumferential inner wall preferably creating a lumen for receiving a control rod.
In a possible implementation form of the first aspect, the liquid moderator is heavy water.
In a possible implementation form of the first aspect, the fuel salt comprises fissile components, preferably comprising enriched lithium 7 fluoride, thorium tetrafluoride, uranium tetrafluoride, uranium trifluoride and/or plutonium trifluoride (7LiF)- (ThF4)- (UF4)- (UF3)- (PuF3) salt.
DK 181148 B1 12
In a possible implementation form of the first aspect, the blanket salt is a molten salt comprising fertile components, preferably comprising enriched lithium 7 fluoride and/or thorium tetrafluoride (7LiF)- (ThF4) salt.
In a possible implementation form of the first aspect, the nuclear reactor core is provided with movable neutron absorbing control elements between vessels of the jackets, preferably between vessels making up the fuel salt jacket and/or between vessels forming a jacket that surrounds the fuel jacket or is surrounded by the fuel jacket. The neutron absorbing control elements being similar to control rods, except that the neutron absorbing control elements are not rod shaped but are shaped like a plate or wedge to match the adjacent vessels in the jacket concerned, and are arranged such that they can be inserted into and retracted from the nuclear reactor core, for controlling nuclear reactivity of the nuclear reactor core.
In a possible implementation form of the first aspect a second fuel salt jacket surrounds the moderator and neutron reflector jacket, and a second moderator and neutron reflector jacket surrounds the second fuel salt jacket. In this implementation, the blanket jacket, if present, surrounds the second neutron reflector jacket. In this implementation, the second fuel salt jacket is formed by a plurality of tubular fuel salt vessels, each tubular fuel salt vessel extending axially from a fuel salt inlet to a fuel salt outlet for passage of a molten fuel salt, the plurality of tubular fuel salt vessels being assembled to form a second cylindrical fuel salt jacket surrounding the moderator and neutron reflector jacket. The
DK 181148 B1 13 second neutron reflector jacket is formed by a plurality of tubular moderator and neutron reflector vessels. Each tubular moderator neutron reflector vessel extending axially from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet for passage of a liquid moderator and neutron reflector.
According to a second aspect, there is provided a method of operating a nuclear reactor core of a molten salt nuclear reactor, the nuclear reactor core comprising: - a tubular cylindrical center moderator and neutron reflector vessel for passage of a liquid moderator and neutron reflector axially extending from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet, - a cylindrical fuel salt jacket surrounding the center moderator vessel, the fuel salt jacket extending axially from a fuel salt inlet to a fuel salt outlet for passage of a molten fuel salt, and a cylindrical moderator and neutron reflector jacket surrounding the fuel salt jacket, the moderator and neutron reflector jacket extending axially from liquid moderator and neutron reflector inlet to a moderator and neutron reflector outlet for passage of a liquid moderator and neutron reflector, the method comprising: -controlling temperature of the fuel salt in the fuel salt jacket to remain at a temperature between approximately 600 and 700°C, -controlling the temperature of the moderator and neutron reflector liquid in the central moderator and neutron
DK 181148 B1 14 reflector vessel to remain at a temperature between approximately 20 and 90°C, and - controlling the temperature of the moderator and neutron reflector liquid in the moderator and neutron reflector jacket to remain at a temperature between approximately 20 and 90°C.
By controlling the temperature of the moderator between approximately 20 and the person 90°C, the moderator and neutron reflector, typically heavy water, does not need to be pressurized to prevent it from boiling. Thus, the vessels that contain the moderator and neutron reflector will not need to act as a pressure vessel and can thus be constructed in a much lighter and less solid way. This is a significant advantage, since the suitable materials that have the required properties are difficult for use in a construction that needs to withstand pressure (pressure vessel). Thus, the construction of the nuclear core is significantly facilitated.
In a possible implementation form of the second aspect, the method comprises: -adjusting the liquid level of the first liquid moderator and neutron reflector in the center moderator and neutron reflector vessel, and/or - adjusting the liquid level of the second liquid moderator and neutron reflector in the moderator and neutron reflector jacket, and/or -adjusting the liquid level of the molten blanket salt in the blanket salt jacket, and/or -adjusting temperature of fuel salt in the fuel jacket, and/or
DK 181148 B1 15 -adjusting temperature of the first liquid moderator and neutron reflector in the center moderator and neutron reflector vessel, and/or -adjusting temperature of the second liquid moderator and neutron reflector in the moderator and neutron reflector jacket, and/or - adjusting the chemical composition of the first or second liquid moderator and neutron reflector, and/or -adjusting the chemical composition of the fuel salt, and/or -adjusting flow rate of the first liquid moderator and neutron reflector in the center moderator and neutron reflector vessel, and/or - adjusting the position of a control rod that is at least partially inserted in a lumen in the center moderator and neutron reflector vessel, and/or - adjusting the position of a control rod that is at least partially inserted between fuel salt vessels, and/or - adjusting the position of a control rod that is at least partially inserted between moderator and neutron absorption vessels.
In a possible implementation form of the second aspect, the control rod is of a neutron absorbing material such as boron or hafnium.
In a possible implementation form of the second aspect, the liquid moderator is heavy water.
In a possible implementation form of the second aspect, the fuel salt comprises fissile components, preferably comprising enriched lithium 7 fluoride, thorium tetrafluoride, uranium
DK 181148 B1 16 tetrafluoride, uranium trifluoride and/or plutonium trifluoride (7LiF)- (ThF4)- (UF4)- (UF3)- (PuF3) salt.
In a possible implementation form of the second aspect, the blanket salt is a molten salt comprising fertile components, preferably comprising enriched lithium 7 fluoride and/or thorium tetrafluoride (7LiF)- (ThF4) salt.
According to a third aspect there is provided a molten salt nuclear reactor comprising a controller configured to perform the method according to the second aspect or any possible implementations of the second aspect.
According to a fourth aspect, there is provided nuclear reactor core of a molten salt nuclear reactor, the nuclear reactor core comprising: - a tubular cylindrical center moderator and neutron reflector vessel for passage of a first liquid moderator and neutron reflector axially extending from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet, the center moderator and neutron reflector vessel having a wall of metal alloy, preferably zirconium alloy or ceramic composite material, - a cylindrical fuel salt jacket surrounding the center moderator vessel, the fuel salt jacket extending axially from a fuel salt inlet to a fuel salt outlet for passage of a molten fuel salt, the fuel salt jacket having walls of ceramic composite material, and a cylindrical moderator and neutron reflector jacket surrounding the fuel salt jacket, the moderator and neutron reflector jacket extending axially from liquid moderator and
DK 181148 B1 17 neutron reflector inlet to a moderator and neutron reflector outlet for passage of a second liquid moderator and neutron reflector, the moderator and neutron reflector jacket having walls of metal alloy, preferably zirconium alloy or ceramic composite material.
The choice of materials for the vessels, and the fourth aspect ensures relatively low neutron absorption by the vessels while still providing good tolerance towards the molten salts and moderator liquids.
In a possible implementation form of the fourth aspect, the nuclear reactor core comprises a first cylindrical insulation jacket between the center moderator vessel and the fuel salt jacket, the first insulation jacket comprising graphite felt or insulation aerogel impregnated graphite felt, or other high temperature insulation material, such as ceramic fiber insulation.
In a possible implementation form of the fourth aspect, the nuclear reactor core comprises a second cylindrical insulation jacket between the fuel salt jacket and the moderator neutron reflector jacket, the second insulation jacket comprising graphite felt or insulation aerogel impregnated graphite felt.
In a possible implementation form of the fourth aspect, the first and/or second liquid moderator and neutron reflector is heavy water or a molten hydroxide, preferably molten enriched lithium 7 deuteroxide salt (7LiOD).
DK 181148 B1 18
In a possible implementation form of the fourth aspect, the control rod comprises a neutron absorbing material, preferably boron or hafnium.
In a possible implementation form of the fourth aspect, the fuel salt comprises fissile components, preferably comprising enriched lithium 7 fluoride, thorium tetrafluoride, uranium tetrafluoride, uranium trifluoride and/or plutonium trifluoride (7LiF)- (ThF4)- (UF4)- (UF3)- (PuF3) salt.
In a possible implementation form of the fourth aspect, the blanket salt is a molten salt comprising fertile components, preferably comprising enriched lithium 7 fluoride and/or thorium tetrafluoride (7LiF)- (ThF4) salt.
According to a fifth aspect there is provided a molten salt nuclear reactor comprising a nuclear reactor core according to any the fourth aspect or any possible implementations of the fourth aspect.
These and other aspects will be apparent from and the embodiments described below.
BRIEF DESCRIPTION OF THE DRAWINGS
In the following detailed portion of the present disclosure, the aspects, embodiments, and implementations will Lbe explained in more detail with reference to the example embodiments shown in the drawings, in which:
DK 181148 B1 19
Fig. 1 is a diagrammatic representation of a molten salt reactor with a nuclear reactor core according to an embodiment;
Fig. 2 is an elevated view of a nuclear reactor core according to an embodiment,
Fig. 3 is a sectional view of the nuclear reactor core of
Fig. 2 including an enlarged detail,
Fig. 4 is a cross-sectional view of the nuclear reactor core of Fig. 2 including an enlarged detail,
Fig. 5 is an elevated and exploded view of a portion of the nuclear reactor core of Fig. 2, and
Fig. 6 is an elevated and exploded view of a portion of the nuclear reactor core of Fig. 2 together with the remaining non-exploded portions of the nuclear reactor core.
DETAILED DESCRIPTION
Fig. 1 illustrates an embodiment of a molten salt nuclear reactor 100 using an embodiment of the nuclear reactor core 1. The molten salt nuclear corrector 100 uses a plurality of flow machines 1 (pumps) for circulating the molten fuel salt.
Fach flow machine 2 is connected to a heat exchanger 3 via a fuel salt loop 4, a reactor coolant loop 5 and a secondary coolant loop 6 for heat exchanging with the nuclear reactor core 1. The fuel salt loop 4 provides fuel salt for driving and controlling the nuclear reaction. Further, the heat exchangers 3 provide the reactor coolant and secondary coolants via the fuel salt loop 4 and the reactor coolant loop 5 and secondary coolant loop 6. In order to circulate and drive the flow of molten salt, flow machines 1 (pumps) are used.
DK 181148 B1 20
With reference to Figs. 2 to 6, an embodiment of the nuclear reactor core 1 is shown with an inlet and/or outlet area 7 at the upper end and an inlet and/or outlet area 8 at the lower end.
The nuclear reactor core 1 has an axis X and extends along the axis X extends between the inlet area 7 and the outlet area 8.
In the shown embodiment, the nuclear reactor core 1 is cylindrical and almost spherical. The nuclear reactor core 1 has a cylindrical and concentric construction. A tubular cylindrical center moderator and neutron reflector vessel 10 extends along the axis X. The cylindrical center moderator and neutron reflector vessel 10 provides for passage of a liquid moderator and neutron reflector 11 between a liquid moderator and neutron reflector inlet 12 and a liquid moderator and neutron reflector outlet 13. The liquid moderator and neutron reflector can be heavy water or a molten hydroxide, such as molten enriched lithium 7 deuteroxide salt (7LiOD).
The cylindrical center moderator and neutron reflector vessel 10 has a largest inner cross-sectional area medially between the liquid moderator inlet 12 and the liquid moderator and neutron reflector outlet 13. The inner cross-sectional area of the center moderator and neutron reflector vessel 10 preferably increasing gradually towards a medial position between the liquid moderator and neutron reflector inlet 12 and the liquid moderator and neutron reflector outlet 13 the inner cross-sectional area of the center moderator and neutron
DK 181148 B1 21 reflector vessel 10 preferably varying continuously between the moderator and neutron reflector inlet 12 and the moderator and neutron reflector outlet 13. Thus, the center moderator and neutron reflector vessel 10 has a varying inner radius.
The inner radius is varied in a way such as to achieve a semi- spherical shape of the cylindrical center moderator and neutron reflector vessel 10. The center moderator and neutron reflector vessel 10 has variable inner and outer radii RII and R12, the inner and outer Radii R11 and R12 preferably being largest at or near the medial position and gradually decreasing from the medial position. During operation of the nuclear reactor core 1, the center moderator and neutron reflector vessel 10 is at least partially filled or completely with a liquid moderator neutron reflector 11, and the liquid moderator and neutron reflector 11 is exchanged at a controlled rate by flow through the moderator and neutron reflector vessel 10.
The cylindrical center moderator and neutron reflector vessel 10 is in an embodiment a wall of a metal alloy, preferably iron alloy (when the moderator and neutron absorption liquid is heavy water), preferably stainless steel, and most preferably zirconium alloy or ceramic composite material. The ceramic composite material is in an embodiment ceramic matrix composite. The ceramic matrix composite comprises ceramic fibers embedded in a ceramic matrix. The fibers and the matrix both can consist of any ceramic material, whereby carbon and carbon fibers can also be regarded as a ceramic material. The ceramic composite material is in an embodiment a composite of silicon carbide and carbon, e.g. carbon fiber-carbon matrix (C/C) composite, silicon carbide fiber-silicon carbide matrix
DK 181148 B1 22 (siC/siC) composite, and or silicon carbide fiber-carbon matrix (SiC/C) composite.
A cylindrical fuel salt jacket extends along the axis X between the inlet and outlet area 7,8 and surrounds the center moderator and neutron reflector vessel 10. The fuel salt jacket is formed by assembling a plurality of tubular fuel salt vessels 20. Each tubular fuel salt vessel 20 extends axlally from a fuel salt inlet 22 to a fuel salt outlet 23 for passage of a molten fuel salt 21. Fach fuel salt vessel 20 in the fuel salt jacket is substantially identical to the other fuel salt vessels 20 in the fuel salt jacket. During operation of the nuclear reactor core 1, the fuel salt jacket 20 is at least partially or completely filled with fuel salt 21, and the fuel salt 21 is exchanged at a controlled rate by flow through the fuel salt vessels 20.
A first cylindrical insulation jacket is provided between the center moderator vessel 10 and the fuel salt jacket. The first insulation jacket is formed by one or more first insulation members 15. The first insulation members 15 can, as shown, be shaped like sectors of the first cylindrical insulation jacket. Alternatively, the first insulation members 15 can be shaped as a strip that is spirally wound around the center moderator vessel 10. The first insulation members 15 are in an embodiment made of a graphite felt or insulation aerogel impregnated graphite felt. The first insulation members 15 are in an embodiment made of a graphite felt or insulation aerogel impregnated graphite felt. The first insulation jacket 15 allows a much lower heat loss to the moderator and neutron reflector liquid 11 and for the moderator and neutron
DK 181148 B1 23 reflector liquid 11 to be consistently kept operating at a much lower temperature than the fuel salt.
The cylindrical fuel salt jacket is shaped to match the shape of the cylindrical center moderator and neutron reflector vessel 10 (with the first insulation jacket therebetween), i.e. complementary therewith. Hence, the inner radius of the fuel salt jacket varies in a similar way to the outer radius of the cylindrical center moderator and neutron reflector vessel 10, such that the first insulation jacket 15 fits in between.
The fuel salt jacket is shaped as a hollow circular cylinder with variable inner and outer radii R21 and R22. The fuel salt vessels 20 are preferably shaped as a circular cylindrical sector with variable inner and outer radii R21 and R22, respectively, and a given sector angle A2, the inner and outer Radii R21 and R22 preferably being largest at or near the medial position and gradually decreasing from the medial position. In the present embodiment, the sector angle is the same for all elements of the nuclear reactor core 1 and hence the sector angle has been indicated in Fig. 4 as “An”. However, it should be understood that the sector angle “An” does not need to be the same for the various jackets and not even for the various vessels that make up a single jacket.
The fuel salt jacket is a circumferentially substantially continuous jacket that is formed by the matching sides of the tubular fuel salt vessels 20 touching one another along a substantial portion of their axial extent.
DK 181148 B1 24
A cylindrical moderator and neutron reflector jacket surrounds the cylindrical fuel salt jacket. The cylindrical moderator and neutron reflector jacket is formed by assembling a plurality of tubular moderator and neutron reflector vessels 30. Each tubular moderator neutron reflector vessel 30 extends axially from a liquid moderator and neutron reflector inlet 32 to a liquid moderator and neutron reflector outlet 33 for passage of a liquid moderator and neutron reflector 31. The elements that make up the tubular moderator neutron reflector vessel 30 have walls of ceramic composite material or metal alloy, preferably zirconium alloy. The ceramic composite material is in an embodiment ceramic matrix composite. The ceramic matrix composite comprises ceramic fibers embedded in a ceramic matrix. The fibers and the matrix both can consist of any ceramic material, whereby carbon and carbon fibers can also be regarded as a ceramic material. The ceramic composite material is in an embodiment a composite of silicon carbide and carbon, e.g. carbon fiber-carbon matrix (C/C) composite, silicon carbide fiber-silicon carbide matrix (SiC/SiC) composite, and or silicon carbide fiber-carbon matrix (SiC/C) composite. In an embodiment each moderator neutron reflector vessel 30 in the moderator neutron reflector jacket 30 substantially identical to the other moderator neutron reflector vessels 30 in the fuel salt jacket.
The moderator and neutron reflector jacket is shaped as a hollow circular cylinder with variable inner and outer radii
R31 and R32. The moderator neutron reflector vessels 30 are shaped as a circular cylindrical sector with variable inner and outer radii R31 and R32, respectively, and a given sector angle A3. The inner and outer Radii R31 and R32 are largest
DK 181148 B1 25 at or near the medial position and gradually decrease from the medial position.
During operation of the nuclear reactor core 1, the moderator and neutron reflector jacket is at least partially filled or completely with a liquid moderator neutron reflector 31, and the liquid moderator and neutron reflector 31 is exchanged at a controlled rate by flow through the moderator and neutron reflector vessels 30.
A second cylindrical insulation jacket is provided between the fuel salt jacket and the moderator and neutron absorption jacket. The second insulation jacket is formed by one or more second insulation members 25. The second insulation members 25 can, as shown, be shaped like sectors of the second cylindrical insulation jacket. Alternatively, the second insulation members 25 can be shaped as a strip that is spirally wound around the fuel salt jacket. The second insulation members 25 are in an embodiment made of a graphite felt or insulation aerogel impregnated graphite felt. The second insulation jacket allows a much lower heat loss to the moderator and neutron reflector liquid 31 and for the moderator and neutron reflector liquid 31 to be consistently kept operating at a much lower temperature than the fuel salt 21.
An optional blanket salt jacket surrounds the moderator and neutron absorption jacket, preferably with a third insulation jacket there between. The blanket salt jacket comprises a plurality of tubular blanket salt vessels 40, each tubular blanket salt vessel 40 extending axially from an inlet to an
DK 181148 B1 26 outlet for passage or be constructed without blanket salt inlet and outlet holding of a molten blanket salt 41 in a stationary fashion. The plurality of tubular blanket salt vessels 40 are assembled to form a cylindrical blanket salt jacket surrounding the moderator and neutron reflector jacket.
The blanket salt jacket is shaped as a hollow circular cylinder with variable inner and outer radii R41 and R42. The blanket salt vessels 40 are shaped as a circular cylindrical sector with variable inner and outer radii R41 and R42, respectively, and a given sector angle A4. The inner and outer
Radii R41 and R42 are largest at or near the medial position and gradually decrease from the medial position. The blanket salt jacket can be constructed without blanket salt inlets and outlets, in which the blanket salt 41 is not continuously exchanged and could be either maintained molten, near its melting point to prevent radiolysis of the blanket salt, or frozen.
A third cylindrical insulation jacket is provided between the neutron reflector vessel and the blanket salt jacket. The third insulation jacket is preferably formed by one or more third insulation members 35. The third insulation members 35 can be shaped like sectors of the third cylindrical insulation jacket or shaped as a strip that is spirally wound around the moderator and neutron reflector jacket. The third insulation members 35 are preferably made of a graphite felt or insulation aerogel impregnated graphite felt.
DK 181148 B1 27
The fuel salt vessels 20, the moderator and neutron reflector vessels 30, and/or the blanket salt vessels 40 are, in an embodiment, each assembled from two parts that are separated by a symmetry plane M that extends radially from the medial position (radially relative to axis A). The inlet and outlet sections of these vessels 20, 30, and 40 can, as shown, curve radially outward and be provided with a flange or the like for connection to piping. However, it should be understood that the inlet and outlet sections can also extend axially and do not need to be provided with a flange for connection to piping since other solutions for connecting the vessels 20, 30, and 40 to piping are known in the art. The inlet and outlet 12,13 of the moderator and neutron reflector vessel 10 are in an embodiment provided with a flange for connection to pimping, but it is understood that the inlet and outlet 12,13 can a be formed without a flange and be connected to piping by another form of assembly.
The vessels 20, 30, and 40 and possibly their inlet and outlet sections are held together by a structure that is not shown in the figures for reasons of simplicity. Such support structures are thought well known in the art.
The design of the nuclear reactor core 1 is an attempt to obtain a nuclear reactor core that is as spherical as achievable, but still constructible from materials that can withstand the harsh conditions in the nuclear reactor core 1 while allowing for a sufficient fuel salt circulation rate to extract the produced heat. The result is a design with a structure that is both cylindrical and spherical, as well as both layered and segmented, i.e. resembling a hybrid of the
DK 181148 B1 28 layered structure and shape of onions and the segmented structure and shape of citrus fruit.
Traditional ceramics are relatively brittle, have relatively low thermal shock resistance, and relatively low fracture toughness. Ceramic composite materials or ceramic matrix composite materials are made from short, continuous, or braided ceramic fiber material, usually embedded in a ceramic matrix, providing reinforcement of the matrix ceramic.
Ceramic composites subdue or at least reduce the drawbacks of traditional ceramics.
The fibers and the matrix are made from a variety of ceramic materials and in a variety of different processes resulting in differences in mechanical properties, such as strength and porosity, and impurity content. Most common commercially available ceramic composite materials are made from carbon and/or silicon carbide fibers and carbon or silicon carbide matrix, abbreviated (C/C), (C/sicC), (SiC/C), and (SiC/SicC).
Both naturally occurring carbon, mainly containing the carbon 12 isotope, and silicon, mainly containing silicon 28 isotope, have low neutron capture probability in a thermal neutron energy spectrum. Furthermore, both carbon and silicon carbide have good corrosion resistance to the molten salt used in the molten salt reactor, i.e. both are corrosive resistant to molten salt and have low neutron absorption in the thermal spectrum. However, both are too brittle to be used as a construction material for vessels and jackets in a molten salt nuclear reactor core 1. Yet, their ceramic composites are interesting candidates for use as construction material in molten salt reactors because of their relatively high
DK 181148 B1 29 fracture toughness and high thermal shock resistance combined with their low neutron absorption in thermal spectrum and exceptional resistance to radiation damage.
The construction of the vessels (moderator and neutron reflector vessel 10, and/or moderator and neutron reflector vessels 30 and/or fuel salt vessels 20 and/or blanket salt vessels 40) from ceramic composites comprises lay-up and fixation, where ceramic fiber or resin impregnated ceramic fibers are wound or placed in or around a mold/core, giving the part the shape of the final ceramic composites part, then polymer is infiltrated into the fibers, then the part is cured. After curing the part is demolded (removed from mold/core) and the polymer is pyrolyzed in a furnace at high temperatures, usually above 800°C in an inert atmosphere to form the matrix composite (in an embodiment the mold/core is a one use mold/core that is burned, melted, crushed or destroyed after first use. This leaves the part porous and the part is re-infiltrated (resin impregnated) and the paralyzing step is repeated until the desired porosity is reached.
Several minor variations of the above process can be used, however, all these processes depend on the first step of shaping the fibers and polymer in a mold and/or on a core.
This limits the geometry of the jackets, vessels and other parts that can be manufactured, since the jackets, vessels and other parts of the nuclear reactor core 1 have to be able to release from the mold or fixture (core) with the vessel in its final form, allowing for some intermediate and or final machining to tolerances. The process does not allow for
DK 181148 B1 30 creation of complex shapes compared to traditional metal manufacturing, such as casting, forming (e.g. forging, rolling, extruding, die forming, indenting, stretching, deep drawing, stamping and bending). Further, machining (e.g. turning, drilling, boring, milling and cutting) but more technically challenging compared with machining conventional metal parts. However, the disclosed construction and geometry of the present nuclear reactor core allows each of the vessels and jackets to be manufactured by the commercially available processes for manufacturing parts in composite ceramic material.
In an embodiment (not shown in the Figs.), the center moderator vessel 10 is segmented and formed by a plurality of tubular moderator vessels, each moderator vessel extending axlally from a moderator inlet to a moderator outlet for passage of a moderator, the plurality of tubular moderator vessels being assembled to form the center moderator vessel 10. In a variation of this embodiment, the center moderator vessel 10 has a circumferential outer wall and a circumferential inner wall, the circumferential inner wall creating a lumen for receiving a control rod. The control rod is of a neutron absorbing material such as boron or hafnium.
In an embodiment (not shown in the Figs.) a second fuel salt jacket surrounds the moderator and neutron reflector jacket, and a second moderator and neutron reflector jacket surrounds the second fuel salt jacket. In this embodiment, the blanket jacket, 1f present, surrounds the second neutron reflector jacket. In this embodiment, the second fuel salt jacket is formed by a plurality of tubular fuel salt vessels, each
DK 181148 B1 31 tubular fuel salt vessel extending axially from a fuel salt inlet to a fuel salt outlet for passage of a molten fuel salt 21, the plurality of tubular vessels being assembled to form a second cylindrical fuel salt jacket surrounding the moderator and neutron reflector jacket. The second neutron reflector jacket is formed by a plurality of tubular moderator and neutron reflector vessels. Each tubular moderator neutron reflector vessel extending axially from a liquid moderator and neutron reflector inlet to a liquid moderator and neutron reflector outlet for passage of a liquid moderator and neutron reflector.
In an embodiment (not shown in the Figs.) the nuclear reactor core 1 1s provided with movable neutron absorbing control elements between vessels of the jackets, preferably between vessels making up the fuel salt jacket and/or between vessels forming a jacket that surrounds the fuel jacket or is surrounded by the fuel jacket. The neutron absorbing control elements are similar to control rods, except that the neutron absorbing control elements are not rod shaped but are shaped like a plate or wedge to match the adjacent vessels in the jacket concerned, and are arranged such that they can be inserted into and retracted from the nuclear reactor core 1, for controlling nuclear reactivity of the nuclear reactor core.
In an embodiment, the reactivity of the nuclear chain reaction of the nuclear reactor core 1 is controlled by controlling the fuel salt temperature in the fuel salt jacket to remain at a temperature between approximately 600 and 700°C, and controlling the moderator and neutron reflector liquid in the
DK 181148 B1 32 central moderator and neutron reflector vessel 10 to remain at a temperature between approximately 20 and 90°C. Preferably also the temperature of the moderator and neutron reflector liquid 21 in the moderator and neutron reflector jacket 20 is controlled to remain at a temperature between approximately 20 and 90°C.
The method of controlling the reactivity of the nuclear chain reaction and temperature household of the nuclear reactor core 1 may further comprise: -adjusting the liquid level of the liquid moderator and neutron reflector 11 in the central moderator and neutron reflector vessel 10, and/or -adjusting the liquid level of the liquid moderator and neutron reflector 31 in the moderator and neutron reflector jacket 30, and/or -adjusting the liquid level of the molten blanket salt 41 in the blanket salt jacket, and/or -adjusting temperature of fuel salt 21 through the fuel jacket, and/or -adjusting temperature of the liquid moderator and neutron reflector 11 in the center moderator and neutron reflector vessel 10, and/or -adjusting temperature of the liquid moderator and neutron reflector 31 in the moderator and neutron reflector jacket, and/or - adjusting the chemical composition of the fuel salt 21, and/or - adjusting the chemical composition of the liquid moderator and neutron reflector 11, 31, and/or
DK 181148 B1 33 - adjusting the position of a control rod that is at least partially inserted in a lumen in the center moderator and neutron reflector vessel 10, and/or - adjusting the position of a control rod that is at least partially inserted between fuel salt vessels 20, and/or - adjusting the position of a control rod that is at least partially inserted between moderator and neutron absorption vessels 30.
Increasing the liquid level of the liquid moderator and neutron reflector in the central moderator and neutron reflector vessel 10, increases the reactivity of the nuclear chain reaction by providing more moderation and or reflection, and vice versa.
Increasing the liquid level of the liquid moderator and neutron reflector in the moderator and neutron reflector jacket 30, increases the reactivity of the nuclear chain reaction by providing more moderation and or reflection and or less neutron leakage, and vice versa.
Increasing the liquid level of the molten blanket salt 41 in the blanket salt jacket 40, if present, increases the reactivity of the nuclear chain reaction by providing more reflection and or less neutron leakage, and vice versa.
Increasing the temperature of the fuel salt in the fuel salt jacket 20, decreases the reactivity of the nuclear chain reaction by providing a lower average density in the nuclear reactor core and thus presence of less fissile fuel, and vice versa.
Increasing the temperature of the liquid moderator and neutron reflector in the center moderator and neutron reflector vessel 10, decreases the reactivity of the nuclear chain reaction by
DK 181148 B1 34 providing a lower average density in the nuclear reactor core and thus more moderation and or reflection, and vice versa.
Increasing the temperature of the liquid moderator and neutron reflector in the moderator and neutron reflector jacket, decreases the reactivity of the nuclear chain reaction by providing a lower average density in the nuclear reactor core and thus more moderation and or reflection and or more neutron leakage, and vice versa.
Adding fissile material to the fuel salt 21 increases the reactivity of the nuclear chain reaction, and vice versa.
Adding fertile material, e.g. thorium, or neutron absorbing material to the fuel salt 21 decreases the reactivity of the nuclear chain reaction, and vice versa.
Adding fertile material, e.g. thorium, or neutron absorbing material to the moderator and reflector liquid 11, 31 decreases the reactivity of the nuclear chain reaction, and vice versa.
Temperatures of the respective moderator liquid 11,31 or fuel salt 21 are changed by e.g.: - increasing the flow rate resulting a smaller temperature drop across the nuclear reactor core 1 and thus a higher or lower density, and/or - adjusting the cooling rate of the moderator liquids 11,31 or fuel salt 21 outside the nuclear reactor core 1 resulting in shifting the average temperature over the nuclear reactor core 1 and thus higher or lower densities.
Liquid levels of the respective moderator liquid 11,31 or fuel salt 21 or molten blanket salt 41 is changed by e.g.
DK 181148 B1 35 adjusting the rate at which liquid is pumped in the respective vessel or jacket from the top and the rate at which the liquid is passively allowed to drained from the bottom.
In an embodiment (not shown), a controller, e.g. an electronic control unit is coupled to the elements of the molten salt nuclear reactor 100 and configured to control the reactivity of the nuclear reactor core 1 in accordance with the measures described above.
The various aspects and implementations have been described in conjunction with various embodiments herein. However, other variations to the disclosed embodiments can be understood and effected by those skilled in the art in practicing the claimed subject-matter, from a study of the drawings, the disclosure, and the appended claims. In the claims, the word “comprising” does not exclude other elements or steps, and the indefinite article “a” or “an” does not exclude a plurality. The mere fact that certain measures are recited in mutually different dependent claims does not indicate that a combination of these measures cannot be used to advantage.
The reference signs used in the claims shall not be construed as limiting the scope. Unless otherwise indicated, the drawings are intended to be read (e.g., cross-hatching, arrangement of parts, proportion, degree, etc.) together with the specification, and are to be considered a portion of the entire written description of this disclosure. As used in the description, the terms “horizontal”, “vertical”, “left”, “right”, “up” and “down”, as well as adjectival and adverbial
DK 181148 B1 36 derivatives thereof (e.g., “horizontally”, “rightwardly”, “upwardly”, etc.), simply refer to the orientation of the illustrated structure as the particular drawing figure faces the reader.
Similarly, the terms “inwardly” and “outwardly” generally refer to the orientation of a surface relative to its axis of elongation, or axis of rotation, as appropriate.

Claims (7)

DK 181148 B1 37 PATENTKRAV:DK 181148 B1 37 PATENT REQUIREMENT: 1. Fremgangsmåde til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100), hvilken atomreaktorkerne omfatter: - en rørformet, cylindrisk centermoderator- og neutronreflektorbeholder (10) til passage af en første væskemoderator og neutronreflektor (11), der strækker sig aksialt fra et væskemoderator- og neutronreflektorindløb (12) for centermoderator- og neutronreflektorbeholderen (10) til et væskemoderator- og neutronreflektorudløb (13) fra centermoderator- og neutronreflektorbeholderen (10), - en cylindrisk brændselssaltkappe, der omgiver centermoderatorbeholderen (10), hvilken brændselssaltkappe strækker sig aksialt fra et brændselssaltindløb (22) for den cylindriske brændselssaltkappe til et brændselssaltudløb (23) fra den cylindriske brændselssaltkappe til passage af et smeltet brændselssalt (21), og en cylindrisk moderator- og neutronreflektorkappe, der omgiver brændselssaltkappen, hvilken moderator- og neutronreflektorkappe strækker sig aksialt fra den cylindriske moderator- og neutronreflektorkappes væskemoderator- og neutronreflektorindløb (32) til et moderator- og neutronreflektorudløb (33) fra den cylindriske moderator- og neutronreflektorkappe til passage af en anden væskemoderator og neutronreflektor (31), kendetegnet ved, at fremgangsmåden omfatter: -styring af temperaturen på brændselssaltet (21) i brændselssaltkappen til at forblive på en temperatur mellem omtrent 600 og 700 °C,1. Method for operating a nuclear reactor core (1) for a molten salt nuclear reactor (100), which nuclear reactor core comprises: - a tubular, cylindrical central moderator and neutron reflector container (10) for the passage of a first liquid moderator and neutron reflector (11) extending axially from a liquid moderator and neutron reflector inlet (12) for the center moderator and neutron reflector container (10) to a liquid moderator and neutron reflector outlet (13) from the center moderator and neutron reflector container (10), - a cylindrical fuel salt jacket surrounding the center moderator container (10), which fuel salt jacket extends axially from a fuel salt inlet (22) for the cylindrical fuel salt jacket to a fuel salt outlet (23) from the cylindrical fuel salt jacket for the passage of a molten fuel salt (21), and a cylindrical moderator and neutron reflector jacket surrounding the fuel salt jacket, which moderator and neutron reflector jacket extends say axially from the cylindrical one moderator and neutron reflector jacket liquid moderator and neutron reflector inlet (32) to a moderator and neutron reflector outlet (33) from the cylindrical moderator and neutron reflector jacket for the passage of another liquid moderator and neutron reflector (31), characterized in that the method comprises: -controlling the temperature on the fuel salt (21) in the fuel salt jacket to remain at a temperature between approximately 600 and 700 °C, DK 181148 B1 38 - styring af temperaturen på den første væskemoderator og neutronreflektor (11) i centermoderator- og neutronreflektorbeholderen (10) til at forblive på en temperatur mellem omtrent 20 og 90 °C, og - styring af temperaturen på den anden væskemoderator og neutronreflektor (31) i moderator- og neutronreflektorkappen (30) til at forblive på en temperatur mellem omtrent 20 og 90DK 181148 B1 38 - controlling the temperature of the first liquid moderator and neutron reflector (11) in the center moderator and neutron reflector vessel (10) to remain at a temperature between approximately 20 and 90 °C, and - controlling the temperature of the second liquid moderator and neutron reflector (31) in the moderator and neutron reflector jacket (30) to remain at a temperature between about 20 and 90 °C.°C. 2. Fremgangsmade til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100) ifølge krav 1, hvilken fremgangsmåde omfatter styring af atomkædereaktionens reaktivitet i atomreaktorkernen (1) ved: - justering af væskemoderatorens og neutronreflektorens (11) væskeniveau 1 centermoderator- og neutronreflektorbeholderen (10), og/eller - justering af væskemoderatorens og neutronreflektorens (31) væskeniveau i moderator- og neutronreflektorkappen, og/eller - justering af væskeniveauet af et smeltet dæksalt (41) i en dæksaltkappe til atomreaktorkernen (1), og/eller - justering af temperaturen på brændselssaltet (21) 1 brændselskappen, og/eller - justering af den første væskemoderators og neutronreflektors (11) temperatur i centermoderator- og neutronreflektorbeholderen (10), og/eller - justering af den anden væskemoderators og neutronreflektors (31) temperatur i moderator- og neutronreflektorkappen (30), og/eller - justering af den første og/eller anden væskemoderators og neutronreflektors (11, 31) kemiske sammensætning, og/eller2. Method for operating a nuclear reactor core (1) for a molten salt nuclear reactor (100) according to claim 1, which method comprises controlling the reactivity of the nuclear chain reaction in the nuclear reactor core (1) by: - adjusting the liquid level of the liquid moderator and neutron reflector (11) 1 the center moderator and neutron reflector container (10), and/or - adjusting the liquid level of the liquid moderator and neutron reflector (31) in the moderator and neutron reflector jacket, and/or - adjusting the liquid level of a molten covering salt (41) in a covering salt jacket for the nuclear reactor core (1), and/or - adjusting the temperature of the fuel salt (21) 1 fuel tank, and/or - adjusting the temperature of the first liquid moderator and neutron reflector (11) in the center moderator and neutron reflector container (10), and/or - adjusting the temperature of the second liquid moderator and neutron reflector (31) in the moderator and neutron reflector cover (30), and/or - adjustment of the first and/or second liquid moderator and neutronr the chemical composition of eflector (11, 31), and/or DK 181148 B1 39 -justering af brændselssaltets (21) kemiske sammensætning, og/eller - justering af den første væskemoderators og neutronreflektors (11) strømningsrate 1 centermoderator- og neutronreflektorbeholderen (10), og/eller - justering af positionen for en styrestang, der mindst delvist indføres i et lumen i centermoderator- og neutronreflektorbeholderen (10), og/eller - justering af positionen for en styrestang, der mindst delvist indføres mellem brændselssaltbeholdere (20), og/eller - justering af positionen for en styrestang, der mindst delvist indføres mellem moderator- og neutronabsorptionsbeholdere (30).DK 181148 B1 39 - adjusting the chemical composition of the fuel salt (21), and/or - adjusting the flow rate of the first liquid moderator and neutron reflector (11) 1 the center moderator and neutron reflector container (10), and/or - adjusting the position of a control rod that at least partially introduced into a lumen in the center moderator and neutron reflector container (10), and/or - adjusting the position of a control rod that is at least partially introduced between fuel salt containers (20), and/or - adjusting the position of a control rod that at least partially is introduced between moderator and neutron absorption containers (30). 3. Fremgangsmåde til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100) ifølge krav 1 eller 2, hvor den første og/eller anden væskemoderator og neutronreflektor (11, 31) er tungt vand eller smeltet hydroxid, fortrinsvis smeltet, beriget lithium-7-deuteroxidsalt (7LiOD).3. Method for operating a nuclear reactor core (1) for a molten salt nuclear reactor (100) according to claim 1 or 2, wherein the first and/or second liquid moderator and neutron reflector (11, 31) is heavy water or molten hydroxide, preferably molten, enriched lithium -7-deuteroxide salt (7LiOD). 4. Fremgangsmåde til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100) ifølge krav 2 eller krav 3, når det afhænger af krav 2, hvor styrestangen, der mindst delvist indføres mellem brændselssaltbeholdere (20), omfatter et neutronabsorberende materiale, fortrinsvis bor eller hafnium.4. Method for operating a nuclear reactor core (1) for a molten salt nuclear reactor (100) according to claim 2 or claim 3, when it depends on claim 2, wherein the control rod, which is at least partially introduced between fuel salt containers (20), comprises a neutron absorbing material, preferably boron or hafnium. 5. Fremgangsmåde til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100) ifølge et hvilket som helst af kravene 1 til 4, hvor brændselssaltet (21) omfatter fissile komponenter, og fortrinsvis omfatter beriget lithium-7- fluorid, thorium-tetrafluorid, uran-tetrafluorid, uran-Method for operating a nuclear reactor core (1) for a molten salt nuclear reactor (100) according to any one of claims 1 to 4, wherein the fuel salt (21) comprises fissile components, and preferably comprises enriched lithium-7-fluoride, thorium-tetrafluoride , uranium tetrafluoride, uranium- DK 181148 B1 40 trifluorid og/eller plutonium-trifluorid 7LiF-ThF4-UF4-UF3- PuF3-salt.DK 181148 B1 40 trifluoride and/or plutonium trifluoride 7LiF-ThF4-UF4-UF3- PuF3 salt. 6. Fremgangsmåde til drift af en atomreaktorkerne (1) til en smeltetsaltatomreaktor (100) ifølge krav 2 eller et hvilket som helst af kravene 3 til 5, når det afhænger af krav 2, hvor dæksaltet (41) er et smeltet salt, der omfatter fertile komponenter, og fortrinsvis omfatter beriget lithium-7- fluorid og/eller thorium-tetrafluorid (7LiF-ThF4)-salt.A method of operating a nuclear reactor core (1) for a molten salt nuclear reactor (100) according to claim 2 or any one of claims 3 to 5 when dependent on claim 2, wherein the covering salt (41) is a molten salt comprising fertile components, and preferably comprises enriched lithium 7-fluoride and/or thorium tetrafluoride (7LiF-ThF4) salt. 7. Smeltetsaltatomreaktor (100), der omfatter en styreenhed, der er konfigureret til udføre fremgangsmåden ifølge et hvilket som helst af kravene 1 til 6.A molten salt atomic reactor (100) comprising a control unit configured to perform the method according to any one of claims 1 to 6.
DKPA202170281A 2021-05-31 2021-05-31 A method of operating a nuclear reactor core of a molten salt nuclear reactor DK181148B1 (en)

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DKPA202170281A DK181148B1 (en) 2021-05-31 2021-05-31 A method of operating a nuclear reactor core of a molten salt nuclear reactor
KR1020247000022A KR20240032009A (en) 2021-05-31 2022-05-25 molten salt reactor core
JP2023573101A JP2024521176A (en) 2021-05-31 2022-05-25 Molten Salt Reactor Core
PCT/DK2022/050109 WO2022253392A1 (en) 2021-05-31 2022-05-25 Molten salt nuclear reactor core
MX2023014035A MX2023014035A (en) 2021-05-31 2022-05-25 Molten salt nuclear reactor core.
BR122023025219-5A BR122023025219A2 (en) 2021-05-31 2022-05-25 MELTED SALT NUCLEAR REACTOR CORE
EP22815405.0A EP4352750A4 (en) 2021-05-31 2022-05-25 Molten salt nuclear reactor core
AU2022284200A AU2022284200A1 (en) 2021-05-31 2022-05-25 Molten salt nuclear reactor core
BR112023025005A BR112023025005A2 (en) 2021-05-31 2022-05-25 MELTED SALT NUCLEAR REACTOR CORE
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