KR101681793B1 - A nuclear fission reactor having flow control assembly - Google Patents

Abstract

A fission cracking reaction, a flow control assembly, a method thereof, and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module configured to have at least a portion of the progressive combustion wave at a location relative to the nuclear fission module. The flow control assembly includes a flow regulator subassembly configured to operate in accordance with operating parameters associated with the nuclear breakdown module. The flow regulator subassembly can also be reconfigured according to a predetermined input to the flow regulator subassembly. The flow control assembly also includes a carriage subassembly coupled to the flow regulator subassembly for adjusting the induction regulator subassembly to change fluid flow to the nuclear division module.

Description

A NUCLEAR FISSION REACTOR HAVING FLOW CONTROL ASSEMBLY WITH A FLOW CONTROL ASSEMBLY

Cross-reference of related application

The present application claims the benefit of priority based on and related to the earliest applicable filing date (s) from the following patent application (s) ("Related Application") (for example, (S) of the relevant application (s), any and all applications, grandparent applications (originals of originals), Great-Grant parent applications The origin of the source of origin (eg, the origin of the source) of 35 USC §119 (e). All claimed applications, and all claims, grand-parent applications, Great-Grand-parent applications, etc., are hereby incorporated by reference to the extent not inconsistent with the claims set forth herein.

Related Application:

For USPTO special-statutory requirements, the present application is directed to Charles E. Ahlfeld, Roderick A, et al., Entitled " A NUCLEAR FISSION REACTOR, FLOW CONTROL ASSEMBLY, METHODS THEREFOR AND A FLOW CONTROL ASSEMBLY SYSTEM " Hyde, Muriel Y. Ishikawa, David G. McAlees, Jon D. McWhirter, Nathan P. Myhrvold, Ashok Odedra, Clarence T. Tegreene, Thomas Allan Weaver, Charles Whitmer, Victoria YH Wood, Lowell L. Wood, Jr., And George B. Zimmerman, which constitute a continuation-in-part application of U.S. Patent Application 12 / 386,495, filed April 16, 2009, the application of which is currently pending, or which is currently pending, Application.

For USPTO special-statutory requirements, the present application is directed to Charles E. Ahlfeld, Roderick A. Hyde, and Muriel Y., "Inventors of the invention," A NUCLEAR FISSION REACTOR, FLOW CONTROL ASSEMBLY, METHODS THEREFOR AND A FLOW CONTROL ASSEMBLY SYSTEM. Ishikawa, David G. McAlees, Jon D. McWhirter, Nathan P. Myhrvold, Ashok Odedra, Clarence T. Tegreene, Thomas Allan Weaver, Charles Whitmer, Victoria YH Wood, Lowell L. Wood, Jr., and George B. Zimmerman , Filed on July 13, 2009, which is a pending application filed on or prior to which the pending application is currently pending.

For USPTO special-statutory requirements, the present application is directed to Charles E. Ahlfeld, Roderick A. Hyde, and Muriel Y., "Inventors of the invention," A NUCLEAR FISSION REACTOR, FLOW CONTROL ASSEMBLY, METHODS THEREFOR AND A FLOW CONTROL ASSEMBLY SYSTEM. Ishikawa, David G. McAlees, Jon D. McWhirter, Nathan P. Myhrvold, Ashok Odedra, Clarence T. Tegreene, Thomas Allan Weaver, Charles Whitmer, Victoria YH Wood, Lowell L. Wood, Jr., and George B. Zimmerman , Filed on July 13, 2009, which application is an application pending or currently pending based on the priority claim.

For USPTO special-statutory requirements, the present application is directed to Charles E. Ahlfeld, Roderick A. Hyde, and Muriel Y., "Inventors of the invention," A NUCLEAR FISSION REACTOR, FLOW CONTROL ASSEMBLY, METHODS THEREFOR AND A FLOW CONTROL ASSEMBLY SYSTEM. Ishikawa, David G. McAlees, Jon D. McWhirter, Nathan P. Myhrvold, Ashok Odedra, Clarence T. Tegreene, Thomas Allan Weaver, Charles Whitmer, Victoria YH Wood, and Lowell L. Wood, Jr. Which constitutes a continuation-in-part application of U.S. Patent Application No. 12 / 460,159, filed on July 13, 2009, the application being pending or the pending application being the basis of the priority claim.

The USPTO announces the entry into force of the USPTO's computer program that requires the applicant to state the serial number and to indicate whether the application is an ongoing application or a part-time continuation application . This is confirmed in Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette March 18, 2003, http://www.uspto.gov/web/offices/com/sol/og/2003/weekll/patbene.htm It is possible. Applicant (hereinafter "the applicant") has previously described specific details of the application on which the priority claim is based, as set forth in the Act. The applicant understands that the specific description of the act is clear and does not require a characterization or serial number such as "continue" or "part-continuation" to claim priority to a US patent application. Notwithstanding the foregoing, the applicant hereby acknowledges that the computer program of the United States Patent and Trademark Office requires a special data entry requirement whereby the applicant has hereby described the application as a part-time application of the parent application as described above, In addition to the contents of this application, does not imply any type of reference and / or acknowledgment as to whether or not new content is included.

Technical field

In general, the present invention relates to a process comprising an induced nuclear reaction and to a structure for carrying out such a process, said structure comprising an orifice or fluid control means, outlet or coolant channel located at the inlet, To a nuclear fission reaction, a flow control assembly, an operating method thereof, and a flow control assembly system.

It is known that in the operation of the nuclear fission reactor, the neurtron of known energy is absorbed by the atomic nuclide. The resulting compound nuclei are separated into a cleavage product and a decay product containing a cleavage fragment of two smaller atomic weights. Nuclides known to undergo such a split by all energy neutrons include the fragmented nuclear species uranium-233, uranium-235 and plutonium-239. For example, thermal neutrons with a kinetic energy of 0.0253 eV (electron volts) could be used to break up the U-235 nuclide. The thorium-232 and uranium-238, the fertile nuclides, do not induce induction cleavage, except when fast neutrons with kinetic energy of at least 1 MeV (million electron volts) are used. The total kinetic energy emitted from each cleavage is about 200 MeV. This kinetic energy eventually converts to heat.

In the nuclear reactor furnace, the above-described fission and / or raw materials are typically housed in a plurality of closely packed fuel assemblies, which form a nuclear reaction core. It has been observed that due to heat accumulation, the fuel assemblies packed together densely and other reactor components will thermally expand differently from one another, which can lead to misalignment of reactor core components. Heat accumulation can also contribute to fuel rod creep, which can increase the risk of fuel rod expansion and craddage failure during reactor operation. This can increase the risk of fuel pellet cracking and / or fuel rod bending. Fuel pellet cracking can be a precursor to cracking failure mechanisms, such as pellet-clad mechanical interaction, and can lead to fission gas release. Fission gas release can result in higher radioactivity levels than the normal radioactive levels in the reactor core. The fuel rod bow can cause clogging (failure) of the coolant flow channel.

Efforts have been made to provide adequate coolant flow to the fuel assembly as a nuclear reaction. U.S. Patent 4,505,877, entitled " Device for Regulating the Flow of a Fluid ", issued Mar. 19, 1985 to Jacky Rion, discloses a device comprising a series of grids that are perpendicular to the fluid flow and change the direction of the fluid flow . According to Rion's patent, this device is for use in directing cooling fluid circulating in the base of the assembly with a liquid metal-cooled nuclear reaction. Such a device relates to a given pressure drop for a given downstream pressure and a given nominal flow rate without forming cavitation.

Another attempt to provide an appropriate coolant flow to the fuel assembly as a nuclear reactor is described in U.S. Patent 5,066,453, entitled " Nuclear Fuel Assembly Coolant Control, " issued Nov. 19, 1991 to Neil G. Heppenstall et al. This patent discloses an apparatus for controlling coolant flow through a nuclear fuel assembly, the apparatus comprising a variable flow restrictor, which may be located within the fuel assembly, at a location within the fuel assembly in a manner that induces neutron induced growth of the response means Means for responding to neutron radiation, and means for connecting neutron radiation response means to the variable flow restrictor to control the flow of coolant through the fuel assembly. The variable flow restriction device includes a plurality of longitudinally aligned conduits and plugging means having an array of plugging members that can be positioned within a portion of the conduits, the length of the plugging means by the connecting means The plugging members have different lengths so that the directional displacement gradually opens or closes a portion of the conduits.

Another attempt to provide an appropriate coolant flow to the fuel assembly as a liquid reaction is disclosed in U.S. Patent 5,198,185 entitled " Nuclear Reactor Flow Control Method and Apparatus " issued to John P. Church on March 30, 1993 . This patent describes a coolant flow distribution that can improve flow during accidental conditions, without degrading the flow during nominal conditions. According to this patent, a universal sleeve housing surrounds the fuel element. The universal sleeve housing has a plurality of holes to allow coolant to pass therethrough. To vary the size and number of holes in the sleeve housing from one sleeve to the other, in order to increase the amount of coolant flowing into the fuel at the center of the core and, to a lesser extent, to reduce flow to the circumferential fuel. Also, according to this patent, the number of holes changed and the size of the holes meet a certain power shape across the core.

It is an object of the present disclosure to provide a process comprising an induced nuclear reaction and a structure for carrying out such a process.

According to one aspect of the present disclosure there is provided a nuclear fission reactor, wherein the reactor is a nuclear fission module configured to have at least a portion of a traveling burn wave at a location relative to the nuclear fission module; And a flow control assembly configured to couple to the nuclear fission module and configured to change the flow of fluid in response to the progressive firing at a location relative to the nuclear fission module.

According to another aspect of the present disclosure, there is provided a nuclear fission reactor, wherein the nuclear fission reactor comprises a heat-generating nuclear fuel assembly configured to have at least a portion of the progressive combustion wave at a location relative to the nuclear fuel assembly; And a flow control assembly configured to be coupled to the nuclear fission fuel assembly and capable of altering the flow of the fluid stream in response to progressive combustion waves at a location relative to the nuclear fission fuel assembly.

According to another aspect of the present disclosure, there is provided a flow control assembly comprising a flow regulator subassembly for use in a traveling wave nuclear fission reactor.

According to another aspect of the present disclosure there is provided a flow control assembly comprising a flow regulator subassembly for use in a traveling wave nuclear fission reactor, the flow regulator subassembly comprising: a first sleeve having a first hole; A second sleeve configured to be inserted into the first sleeve, wherein the second sleeve has a second hole that can be aligned with the first hole, and wherein the first sleeve rotates to align the first hole with the second hole, A second sleeve configured to receive the first sleeve; And a carriage subassembly configured to be coupled to the flow regulator subassembly.

According to another aspect of the present disclosure there is provided a flow control assembly configured for connection to a fuel assembly for use in a traveling wave nuclear fission reactor, the flow control assembly comprising a flow regulator sub- Assembly.

According to another aspect of the present disclosure, there is provided a flow control assembly configured for connection to a fuel assembly for use in a traveling wave nuclear fission reactor, the flow control assembly being configured to be disposed within a fluid stream, 12. An assembly, wherein the adjustable flow regulator subassembly includes a first sleeve having a first hole, the flow regulator subassembly; And a second sleeve configured to be inserted into the first sleeve, wherein the second sleeve has a second hole, the first hole can be progressively aligned with the second hole, As the fluid is progressively aligned with the second hole, a variable amount of the fluid stream flows through the first and second holes and the first sleeve is forced into the second sleeve to align the second hole with the first hole To relatively translate in the axial direction.

According to a further aspect of the present disclosure there is provided a flow control assembly configured for connection to a fuel assembly for use in a traveling wave nuclear fission reactor, the flow control assembly comprising: an adjustable flow regulator subassembly; And a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly.

According to another aspect of the present disclosure there is provided a flow control assembly for use in a traveling wave nuclear fission reactor, the flow control assembly being capable of coupling to a nuclear fission fuel assembly selected from a plurality of fission fuel assemblies arranged to be disposed within a nuclear fission reactor Wherein the flow control assembly is an adjustable flow regulator subassembly for modifying the flow of a fluid stream flowing through a nuclear fuel assembly selected from a plurality of nuclear fuel assemblies, An adjustable flow regulator subassembly comprising an outer sleeve having a first hole; An inner sleeve inserted into the outer sleeve, the inner sleeve having a plurality of second holes, the first holes being progressively aligned with the second holes to form a variable flow area, And an inner sleeve through which a variable amount of the fluid stream flows through the first and second holes when the second hole is progressively aligned to form a variable flow area. And a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly.

According to a further aspect of the present disclosure there is provided a method of operating with a nuclear fission reaction, the method comprising: generating at least a portion of a progressive combustion wave at a location relative to the nuclear fission module; And actuating a flow control assembly coupled to the nucleation module to change the flow of fluid in response to the position relative to the nucleation module.

According to another aspect of the present disclosure, there is provided a method of assembling a flow control assembly for use in a traveling wave nuclear fission reactor, the method comprising the step of receiving a flow regulator subassembly.

According to another aspect of the present disclosure, there is provided a method of assembling a flow control assembly for use in a traveling wave nuclear fission reactor, the method comprising the step of receiving a carriage subassembly.

According to another aspect of the present disclosure, there is provided a method of assembling a flow control assembly for use in a traveling wave nuclear fission reactor, the method comprising: receiving a first sleeve having a first hole; Inserting a second sleeve into the first sleeve, wherein the second sleeve has a second hole that can be aligned with the first hole, the first sleeve translating the first hole axially, Inserting a second sleeve into the first sleeve, the second sleeve configured to rotate to align with the second hole; And coupling the carriage subassembly to the flow regulator subassembly.

According to a further aspect of the present disclosure, there is provided a flow control assembly system for use in a traveling wave nuclear fission reactor, the system including a flow regulator subassembly.

According to another aspect of the present disclosure, there is provided a flow control assembly system for use in a nuclear fission reactor, the system comprising: a flow regulator subassembly comprising a first sleeve having a first hole; A second sleeve configured to be inserted into the first sleeve, the second sleeve having a second hole that can be aligned with the first hole, the first sleeve translating the first hole axially, A second sleeve configured to rotate to align with the hole; And a carriage subassembly configured to be coupled to the flow regulator subassembly.

According to another aspect of the present disclosure there is provided a flow control assembly system for use in a nuclear fission reactor, the system being configured to couple to a nuclear fission fuel assembly, the adjustable flow regulator sub- Assembly.

According to another aspect of the disclosure there is provided a flow control assembly system for use in a nuclear fission reactor, the system being capable of coupling to a nuclear fission fuel assembly selected from a plurality of nuclear fuel fission assemblies disposed in the nuclear fission reactor , Such a system is an adjustable flow regulator subassembly for controlling the flow of a fluid stream flowing through a nuclear fission fuel assembly selected from among the nuclear fission fuel assemblies, including an outer sleeve having a plurality of first holes A possible flow regulator subassembly; An inner sleeve inserted into the outer sleeve, the inner sleeve having a plurality of second holes, the first holes being progressively aligned with the second holes to form a variable flow area, An inner sleeve through which a variable volume fluid stream flows through the first and second holes as the first hole progressively aligns with the second hole to form a variable flow area; And a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly.

A feature of the present disclosure is to provide a flow control assembly that is capable of controlling the flow of fluid in response to the position of the combustion wave.

Another aspect of the present disclosure is to provide a flow control assembly comprising a flow regulator subassembly comprising an outer sleeve and an inner sleeve, the outer sleeve having a first hole and the inner sleeve having a first So that when a second hole is aligned with the first hole, a predetermined amount of the fluid stream flows through the first hole and the second hole.

It is a further feature of the present disclosure to provide a carriage subassembly configured to be coupled to a flow regulator subassembly for transporting and configuring the flow regulator subassembly.

In addition to the foregoing, various other methods and / or embodiments of the apparatus are described and illustrated in the written description of the specification (e.g., claims and / or detailed description) and / or drawings.

The foregoing is a summary and may, accordingly, include simplification, generalizations, inclusions, and / or omissions of details; As a result, those skilled in the art will appreciate that the summary is merely exemplary and should not be construed in any way as limiting. Additional aspects, embodiments and features, in addition to the aspects, embodiments, and features described above, may be best understood from the accompanying drawings and the following detailed description.

BRIEF DESCRIPTION OF THE DRAWINGS While the specification concludes with claims particularly pointing out and distinctly claiming the subject matter of this disclosure, the disclosure will be more clearly understood from the following detailed description when read in conjunction with the accompanying drawings. Also, in various figures, the same symbol typically represents a similar or identical item.
1 is a view showing a nuclear fission reactor.
IA is a cross-sectional view of a nuclear fission module or nuclear fuel assembly belonging to a nuclear fission reactor.
FIG. 1B is a perspective view showing a part of a nuclear fuel rod belonging to the nuclear fission module. FIG.
Figure 2 is a cross-sectional view of a hexagonal nucleation reactor core having a plurality of hexagonal nucleation modules disposed therein.
Figure 3 is a cross-sectional view of a cylindrical reactor core in which a plurality of hexagonal nucleation modules are disposed.
Figure 4 is a cross-sectional view of a parallelepiped reactor core in which a plurality of hexagonal nucleation modules are disposed, wherein the reactor core comprises at least a portion of a progressive softwave having a width "x " Is a cross-sectional view of the reactor core.
5 is a cross-sectional view showing a plurality of adjacent hexagonal nuclear cracking modules, in which a plurality of longitudinally movable control rods are disposed in addition to the fuel rods therein.
FIG. 5A is a cross-sectional view showing a plurality of adjacent hexagonal nuclear fission modules, in addition to the fuel rods, a nuclear fission module in which a plurality of raw material breeding rods are disposed.
FIG. 5B is a cross-sectional view showing a plurality of adjacent hexagonal nuclear cracking modules, in addition to the fuel rods, in which a plurality of neutron reflectors are disposed. FIG.
Figure 5C is a cross-sectional view of a reactor core having a proximal blanket fuel assembly disposed about an inner perimeter as a plurality of adjacent parallelepiped reactor core.
6 is a cross-sectional view taken along line 6-6 of Fig.
Figure 7 is a partial cross-sectional view illustrating a plurality of flow regulator subassemblies and a plurality of adjacent nucleation modules pertaining to a flow control assembly and coupled to each one of the nuclear mitigation modules.
Figure 8 is an exploded view of a flow regulator subassembly.
8A is an exploded view showing a partial cross-section of a flow regulator subassembly.
8B is a partial cross-sectional view illustrating the flow regulator subassembly in an open configuration to permit full fluid flow.
8C is a partial cross-sectional view illustrating the flow regulator subassembly in a closed configuration for completely shutting off fluid flow.
8D is a view taken along line 8D-8D of FIG. 8B, which shows in a horizontal section the anti-rotation configuration belonging to the lower portion of the flow regulator subassembly.
FIG. 8E is a vertical cross-sectional view showing a nipple that can be freely rotated, showing a lower portion of the flow regulator subassembly with some portions omitted for clarity. FIG.
9 is a view of a flow regulator subassembly coupled to a nuclear fission module, in a fully open position that allows fluid to flow into the nuclear fission module.
Figure 10 is a view of a flow regulator subassembly coupled to a nuclear fission module, in a fully closed position to prevent fluid from flowing into the nuclear fission module.
11 is a vertical cross-sectional view illustrating a plurality of flow regulator subassemblies coupled to one of a plurality of adjacent nucleation modules and nucleation modules.
Figure 12 is a vertical cross-sectional view of a plurality of flow regulator subassemblies coupled to one of a plurality of adjacent nucleation modules and nucleation modules, wherein the flow regulator subassembly includes a full An open position, a partially closed or open position, and a fully closed position.
Figure 13 is a perspective view illustrating a carriage subassembly belonging to a flow control assembly, partially shown for clarity.
14 is a diagram showing a plurality of sensors disposed in each one of a plurality of adjacent nucleation modules and nucleation modules.
Figure 15 illustrates a plurality of flow regulator subassemblies, with some portions omitted for clarity, wherein a selected one of the plurality of flow regulator subassemblies is connected to a lead screw arrangement And a plurality of socket wrenches driven in the rotational direction and axially driven by the gear alignment body.
16 is a perspective view of a gear alignment body for driving a socket wrench selected from a plurality of socket wrenches;
Figure 17 illustrates a plurality of flow regulator subassemblies coupled by a socket wrench selected from among a plurality of socket wrenches, partially omitted for clarity, wherein the socket wrench is electrically coupled to a control or control unit Lt; RTI ID = 0.0 > hermetically < / RTI >
Figure 18 illustrates a plurality of flow regulator subassemblies coupled by a socket wrench selected from a plurality of socket wrenches, partially omitted for clarity, wherein the socket wrench is capable of transmitting radio frequency signals And at least partially controlled by a hermetically sealed electric motor alignment body in response to a radio transmitter-receiver arrangement belonging to a control or control unit.
19 shows a plurality of flow regulator subassemblies coupled by a socket wrench selected from a plurality of socket wrenches, wherein the socket wrench is connected to a fiber optic transmitter-receiver arrangement body belonging to a control unit capable of delivering a signal by a light beam Lt; RTI ID = 0.0 > at least < / RTI >
20A to 20C are flowcharts showing a method of operating the nuclear fission reactor.
Figures 21A-21H are flow diagrams illustrating a flow control assembly assembly method.

In the following description, reference is made to the accompanying drawings, which form a part hereof. In the accompanying drawings, where there is no other content, similar symbols typically denote similar elements. The illustrative embodiments set forth in the description, drawings, and claims are not intended to be limiting. Other embodiments may be utilized and other changes may be made without departing from the scope or spirit of the claimed subject matter.

In addition, the present application uses formal suffixes for clear display. It will be appreciated, however, that such a descriptor is for indication and that various types of claims are described throughout the disclosure (e.g., device (s) / structure (s) (S) may be described below and / or the process (s) / operation (s) may reside in the structure (s) / process (s) and be described below). Accordingly, formal endorsements should never be construed as limiting.

In addition, the claims set forth herein may also describe various ingredients that are included or linked to several other ingredients. Such described architectures are to be understood as merely illustrative, and in fact, many other architectures that achieve the same functionality may be implemented. From a conceptual standpoint, any arrangement of components to achieve the same function is effectively "associated" to achieve the desired function. Accordingly, any two components combined to achieve a particular function may be seen as "associated" with each other, to achieve the desired function, regardless of the architecture or mediator component. Likewise, any two components so associated may be seen as being "operably connected" or "operatively coupled" to one another for the sake of achieving the desired function, The components may also be viewed as "operably coupled" to one another to achieve the desired function. Specific examples that may be operatively coupled include physically mateable and / or physically interacting components, and / or components that can interact wirelessly and / or wirelessly, and / / RTI > and / or < / RTI > logically interacting and / or logically interacting components.

In some instances, one or more of the components may be "configured to", "can be", "operative to", "adaptable , "Possible "," conformable / matched ", and the like. Those skilled in the art will recognize that the term " comprising, "" operable / actuable," " adaptable / adaptable ", & It will be appreciated that, unless otherwise stated, it may generally comprise an active-state component and / or an inactive-state component and / or an atmospheric-state component.

In many cases, and in connection with the present disclosure, as noted above, in all neutrons absorbed into the fissioning species, one or more neutrons are liberated until the fissioning nucleus is depleted. This phenomenon is used in a commercial nuclear reactor to generate heat continuously, and the heat is used again to produce electricity.

However, due to the "peak" temperature (i.e., the high temperature channeling factor) that arises due to the uneven neutron flux distribution in the reactor core, thermal damage to the reactor furnace structure material may occur. As is well known in the art, the neutron flux is defined as the population of neutrons passing through a unit area per unit time. This peak temperature is again due to the heterogeneous control rod / fuel rod distribution. If the peak temperature exceeds the material limit, thermal damage may occur. In addition, reactor reactors operating in the fast neutron spectrum may be designed to have a fertile "proliferation blanket" material present around the core. Such reactors will tend to propagate fuel through proton absorption to proliferative branket materials. This results in increasing the power output at the periphery of the reactor as the reactor approaches the end of the fuel cycle. At the beginning of the reactor furnace fuel cycle, the coolant flow through the periphery assembly will be able to maintain a safe operating temperature and account for the power increase due to burn-up increases during the fuel cycle.

&Quot; Reactivity "(i. E., Change in reactor power) is obtained due to fuel" burnup ". Generations are typically defined as the amount of energy generated per unit mass of fuel and are generally expressed in megawatt-days (MWd / MTHM) per metric tonne of heavy metals or in gigawatts of metric tons of heavy metals - day (GWd / MTHM). More specifically, the change in reactivity relates to the relative ability of the reactor to produce more or less neutrons than is accurate to maintain a critical chain reaction. The responsiveness of the reactor is typically characterized as the time derivative of the response change in which the reactor exponentially increases or decreases the power.

In this regard, a control rod, typically made of a neutron absorbing material, is used to adjust and control the varying degrees of reactivity. Such control rods are reciprocated in and out of the reactor core to variably control neutron absorption and thereby variably control neutron flux levels and reactivity within the reactor core. The neutron flux level will be suppressed around the control rod and higher than far away from the control rod. As a result, the neutron flux becomes nonuniform throughout the reactor core. This results in higher starvation in such areas where the neutron flux is higher. It will also be understood by those skilled in the art of nuclear power generation that neutron flux and power density fluctuations are caused by many factors. Proximity to the control rod may or may not be the primary factor. For example, the neutron flux typically drops considerably at the core boundary without close control rods. This effect will again cause an overheating or peak temperature in the high region of the neutron flux. Such a peak temperature will alter the mechanical properties of the structure and undesirably reduce the operating life of the structure exposed to such peak temperatures. In addition, power density is limited by the ability of the core structure material to withstand such peak temperatures, as a reaction proportional to neutron generation and fission fuel concentration.

Accordingly, referring to FIG. 1, there is shown a nuclear fission reactor designated generally by the numeral 10 and solving the above-mentioned problems, by way of example only and not as a limitation. As will be described in more detail below, the reactor 10 may be a progressive wave nuclear decomposition reaction. The nuclear fission reactor 10 produces electricity delivered to an electrical user through a plurality of transmission lines (not shown). Instead, the reactor 10 may be used to conduct tests such as testing for the effect of temperature on the material as a reaction.

Referring to Figures 1, 1a, 1b and 2, the reactor 10 comprises a nuclear fission reactor core, generally designated by the numeral ' 20 ', such cores comprising a plurality of nuclear fission fuel assemblies, And a nuclear fission module 30, as shown in FIG. The core 20 in the nuclear fission reaction is encapsulated within the reactor core enclosure 35. By way of example only and not as a limitation, each nuclear fission module 30 forms a hexagonal-shaped structure as shown in cross-section, and thus has a nuclear sacrificial module of the most different shape, such as a cylindrical or circular shape 30), it may be packed more densely with other nuclear division modules 30. Each nuclear fission module 30 includes a plurality of fuel rods 40 for heat generation due to the nuclear fission chain reaction process described above. The fuel rods 40 may be surrounded by a fuel rod canister 43 to add structural rigidity to the nuclear fission module 30 and to isolate the nuclear fission module 30 from each other, . Isolating the nucleation modules 30 from each other prevents lateral coolant crossflow between adjacent nucleation modules 30. Prevention of transverse coolant crossflow prevents lateral oscillation of the nucleation module 30. Such transverse vibration will increase the risk of damage to the fuel rods 40. In addition, isolating the nucleation modules 30 from each other may enable coolant flow control in the form of individual module-by-module (per module), as will be described in greater detail below. Controlling the coolant flow into the individual, pre-selected nuclear fission modules 30 allows effective management of the coolant flow in the reactor core 20 and can be achieved, for example, substantially at a non-uniform temperature Allowing the coolant flow to be directed in accordance with the distribution. The canister 43 may include an annular shoulder portion 46 (see FIG. 7) for allowing the fuel rods 40 to be bundled together. The coolant has an average nominal flow rate of about 5.5 m ³ / s (ie, about 194 cubic feet per second (ft ³ / s)) and an average nominal flow rate of about 2.3 m / s About 7.55 ft / sec). The fuel rods 40 form a coolant flow channel 47 therebetween (see FIG. 7), which are adjacent to each other and allow the coolant to flow along the outside of the fuel rods 40. The fuel rods 40 are bundled together to form the hexagonal nucleation module 30 described above. Although the fuel rods 40 are adjacent to each other and are nevertheless non-conforming to techniques known to those skilled in the art of nuclear power reactors design, the fuel rods 40 are made of a wire wrapper wire that spirally extends along the length of each fuel rod 40 wrapper 50 (see FIG. 7).

1B, each fuel rod 40 has a plurality of nuclear fuel pellets 60 laminated in an end-to-end manner in the interior thereof, and such nuclear fuel pellets 60 are provided with a plurality of nuclear fuel pellets 60, (70). ≪ / RTI > The nuclear fuel pellets 60 include the above-described fission nuclides such as uranium-235, uranium-233 or plutonium-239. Alternatively, the nuclear fuel pellets 60 may include a raw radionuclide such as thorium-232 and / or uranium-238, which is changed to the fissile radionuclide immediately above during the fission process. Another alternative is that the nuclear fuel pellets 60 may comprise a predetermined (hereinafter, 'predetermined') blend of a fission radionuclide and a raw radionuclide. As a more specific, an example only and non-(which is also referred to as also thorium oxide) as the limiting, the nuclear fuel pellet 60 is uranium monoxide (UO), uranium dioxide (UO _2), thorium dioxide (ThO _2), uranium trioxide (UO _3), uranium oxide-plutonium oxide (UO-PuO), will be made from the tree uranium oktok side _{(octoxide) (U 3 O 8} ) , and an oxide selected from the group consisting essentially of a mixture thereof . Alternatively, the nuclear fuel pellets 60 may comprise substantially unalloyed uranium, alloyed with other metals such as, but not limited to, zirconium or thorium metal. As yet another alternative, the nuclear fuel pellets 60 may substantially comprise carbides of uranium (UC _x ) or thorium carbides (ThC _x ). For example, the nuclear fuel pellets 60 may be made from a mixture of uranium monocarbide (UC), uranium dicarbide (UC ₂ ), uranium sesqui carbide (U ₂ C ₃ ), thorium dicarbide (ThC ₂ ) Lt; / RTI > (ThC), and mixtures thereof. Limiting example, the nuclear fuel pellet 60 is uranium nitride (nitride) (U ₃ N _2), uranium nitride - - Other non-zirconium _{(U 3 N 2 Zr 3 N} 4), uranium-flow thorium nitride (U-Pu) N, thorium nitride (ThN), uranium-zirconium alloy (UZr), and mixtures thereof. The fuel rod crating material 70 sealingly surrounding the laminate of nuclear fuel pellets 60 may be a suitable zirconium alloy such as ZIRCOLOY ™ (registered trademark of Westinghouse Electric Corporation), and such alloys may be used for corrosion and cracking It has a known tolerance. The cladding 70 may also be made of other materials such as fritic martensite steel.

1, reactor core 20 is disposed within reaction vessel 80 to prevent radioactive particles, gases, or liquid from leaking from reactor core 20 into the surrounding biosphere . The pressure vessel 80 may be made of steel, concrete, or other material having a suitable thickness and size to reduce the risk of radiation leakage and to support the required pressure loads. A containment vessel (not shown) sealingly surrounds portions of reactor core 20 to further prevent leakage of radioactive particles, gas, or liquid from reactor core 20 into the surrounding biosphere, May be provided.

Referring to Figure 1, a premixed loop coolant pipe 90 for allowing a suitable coolant to flow through the reactor core 20 to cool the reactor core 20, 20). The primary loop coolant pipe 90 can be made from any suitable material, such as stainless steel. It will be appreciated that if desired, the primary loop refrigerant pipe 90 may not be made of an iron-based alloy and may also be made of a non-ferrous alloy, a zirconium-based alloy or other structural material or composite. The coolant transferred by the primary loop coolant pipe 90 may be a noble gas or a mixture of rare gases. Alternatively, the coolant may be "hard water" (H ₂ O) or other fluid such as gas or supercritical carbon dioxide (CO ₂ ). As another example, the coolant may be a liquid metal. Such liquid metal may be a lead (Pb) alloy such as lead-bismuth (Pb-Bi). In addition, the coolant may be an organic coolant, such as polyphenyl or fluorocarbon. In the exemplary embodiment described herein, the coolant may suitably be a liquid sodium (Na) metal or a sodium metal mixture, for example sodium-potassium (Na-K). By way of example and depending on the core design in a particular reaction and depending on the operating history, the normal operating temperature of the core in the sodium-cooled reaction can be relatively high. For example, in the case of a 500-1,500 MWe sodium-cooled reactor furnished with a mixed uranium-plutonium oxide fuel, the reactor outlet temperature during normal operation is about 510 ° C (i.e., 950 ° F) to about 550 ° C I.e., 1,020 < 0 > F). On the other hand, in the case of a Loss of Coolant Accident (LOCA) or a Loss of Flow Transient Accident (LOFTA), the peak fuel cladding temperature is about 600 ° C (ie, 1,110 ° F) or more . ≪ / RTI > Also, decay heat accumulation during LOCA or post-LOFTA scenarios and during suspension of reactor operation may result in unacceptable heat accumulation. Accordingly, in some cases, it may be appropriate to control the coolant flow to reactor core 20 during both normal operation and post-accident scenarios.

In addition, the temperature profile in reactor core 20 is varied as a function of position. In this regard, the temperature distribution within the reactor core 20 will closely follow the power density spatial distribution within the reactor core 20. In the absence of a suitable neutron reflector or neutron propagation "blanket" around the perimeter of the reactor core 20, the power density close to the center of the reactor core 20 is generally close to the periphery of the reactor core 20 More generally. The coolant flow parameter for the reactor core perturbation module 30 at the beginning of the core life of the reactor core 20 is determined by the coolant flow parameter for the nucleation module 30 close to the reactor core 20 center, . Thus, in this case, it would be unnecessary to provide the same or a uniform mass flow rate for each nuclear fission module 30. [ Depending on the location of the nuclear fission module 30 in the reactor core 20 and the desired reaction furnace operation results, as will be described in greater detail below, Technology is provided.

1, for reasons presented here, heat-bearing coolant, which is generated by the reactor core 20, flows along the coolant flow path 95 to the intermediate heat exchanger 100, do. Coolant flowing along the coolant flow path 95 flows through the intermediate heat exchanger 100 and into the plemum volume 105 associated with the intermediate heat exchanger 100. After flowing into the plenum skin 105, the coolant continues through the primary loop pipe 90, as shown by the plurality of arrows 107. It will be appreciated that the coolant leaving the plenum skin 105 is cooled due to the heat transfer occurring in the intermediate heat exchanger 100. The first pump 110 is coupled to the primary loop pipe 90 and is in fluid communication with the reactor coolant carried by the primary loop pipe 90 so that the reactor coolant flows into the primary loop pipe 90, Through the reactor core 20, along the coolant flow path 95, into the intermediate heat exchanger 100, and into the plenum skin 105.

Referring again to FIG. 1, a secondary loop pipe 120 is provided to remove heat from the intermediate heat exchanger 100. The secondary loop pipe 120 includes a secondary " hot "leg pipe segment 130 and a secondary " cold" leg pipe segment 140. The secondary low temperature leg pipe segment 140 is formed integrally with the secondary high temperature leg pipe segment 130 to form a closed loop forming the secondary loop pipe 120, as shown. The secondary loop pipe 120 formed by the hot leg pipe segment 130 and the cold leg pipe segment 140 includes a fluid, which may suitably be a liquid sodium or liquid sodium mixture. The secondary hot leg pipe segments 130 extend from the intermediate heat exchanger 100 to a steam generator and a superheater combination 143 (hereinafter referred to as "steam generator 143") . The coolant that flows through the secondary loop pipe 120 and exits the steam generator 143 after passing through the steam generator 143 is discharged to the steam generator 143 due to the heat transfer occurring in the steam generator 143 143). ≪ / RTI > After passing through the steam generator 143, the coolant is pumped along the "cold" leg pipe segment 140, for example by the second pump 145, which is terminated in the intermediate heat exchanger 100. The manner in which the steam generator 143 produces steam is schematically described below.

Referring again to FIG. 1, a body of water 150 maintained at a predetermined temperature and pressure is disposed in the steam generator 143. The fluid flowing through the secondary hot leg pipe segment 130 conveys the heat to the body of water 15 and the body of water is lower in temperature than the fluid flowing through the secondary hot leg pipe segment 130. The fluid flowing through the secondary hot leg pipe segment 130 provides heat to the body of water 15 so that a portion of the body 15 of water is heated to the steam 160 Evaporates. The steam 160 will travel through the steam line 170 and one end of the steam line is in vapor communication with the steam 160 and the other end is in liquid communication with the body 15 of water. The rotatable turbine 180 is coupled to the steam line 170 and thus the turbine 180 is rotated as the steam 160 passes therethrough. For example, a generator 190 connected to a turbine 180 by a rotatable turbine shaft 195 produces electricity when the turbine 180 is rotated. The condenser 200 is also connected to the steam line 170 and receives steam passing through the turbine 180. Condenser 200 condenses the vapor to liquid water and transfers the wastewater to a heat sink, such as cooling tower 210, associated with reactor 10. The liquid water condensed by the condenser 200 is condensed by the third pump 220 disposed between the condenser 200 and the steam generator 143 from the condenser 200 to the steam generator 143 Lt; / RTI >

Referring now to Figures 2, 3 and 4, an exemplary configuration for reactor core 20 is shown in cross-section. In this regard, for the reactor core 20, the nucleation module 30 is aligned to form a hexagonal-shaped structure (generally indicated at 230). Alternatively, for the reactor core 20, the nucleation module 30 may be arranged to form a cylindrical-shaped structure (generally indicated at 240). Alternatively, for the reactor core 20, the nucleation module 30 is aligned to form a parallelepiped-shaped structure (generally indicated at 250). In this regard, for reasons described below, reactor core 250 includes a first end 252 and a second end 254.

Referring to Figure 5, regardless of the configuration selected for the reactor core 20, a plurality of spaced, longitudinally extending, longitudinally displaceable control rods 260 are provided in the control rod guide tube or cladding To extend the length of the predetermined number of nucleation modules 30. The control rod 260, shown as being disposed within a predetermined number of hexagonal-shaped nuclear fission modules 30, controls the neutron fission reactions occurring within the nuclear fission module 30. The control rod 260 comprises a suitable neutron absorber material having a high neutron absorption cross section that can be tolerated. In this regard, the sorbent material may be a metal or a metalloid selected from the group consisting essentially of lithium, silver, indium, cadmium, boron, cobalt, hafnium, dysprosium, gadolinium, samarium, erbium, europium, have. Alternatively, the sorbent material may be a compound or alloy selected from the group consisting essentially of silver-indium-cadmium, boron carbide, zirconium diboride, titanium diboride, hafnium diboride, gadolinium titanate, dysprosium titanate, . The control rod 260 will controllably supply the negative reactivity reactivity to the reactor core 20. Accordingly, control rod 260 provides responsiveness management capability for reactor core 20. In other words, the control rod 260 can be configured to control or control the neutron flux profile across the reactor core 20, thereby affecting the temperature profile across the reactor core 20 .

Referring to FIGS. 5A and 5B, an alternate embodiment of a nuclear fission module 30 is shown. It will be appreciated that nuclear division module 30 need not be neutronically active. In other words, the nucleation module 30 need not contain any fissile material. In this case, the nuclear fission module 30 could be a pure relective assembly or a pure ferrial assembly or a combination of both. In this regard, the nuclear fission module 30 may be a proliferative nucleation module comprising a nuclear proliferation material or a reflective nuclear proliferation module comprising a reflective material. Alternatively, in one embodiment, the nucleation module 30 may include a fuel rod 40 in combination with a nuclear proliferation rod or a reflective rod. For example, in FIG. 5A, a plurality of raw nuclear proliferating rods 270 are disposed in the nuclear fission module 30 in combination with the fuel rods 40. A control rod 260 may also be present. The raw nuclear growth material in the nuclear proliferation rod 270 may be thorium-232 and / or uranium-238, as described above. In this manner, the nuclear fission module 30 forms the raw nuclear proliferation assembly. In Fig. 5B, a plurality of neutron reflecting rods 274 are arranged in the nuclear fission module 30 in combination with the fuel rods 40. Fig. A control rod 260 may also be present. The reflective material may be a material selected from the group consisting essentially of beryllium (Be), tungsten (W), vanadium (V), depleted uranium (U), thorium (Th) will be. In addition, the reflective portion rods 274 may be selected from a variety of steel alloys. In this manner, the nucleation module 30 forms a neutron reflector assembly. In addition, those skilled in the art of nuclear in-core fuel management will appreciate that the nuclear fission module 30 can be used in conjunction with the nuclear fuel rods 40, the control rods 260, the proliferation rods 270, Combinations thereof. ≪ RTI ID = 0.0 >

FIG. 5C illustrates another embodiment of reactor core 20 described above. In FIG. 5c, the raw material is placed around the inner periphery of the parallelepiped reactor core 250 with a multiplication blanket containing a plurality of proliferation nucleation modules 276 that receive. The proliferation blanket proliferates the raw material internally.

Referring to Figure 4, regardless of the configuration selected for the core 20 in the nuclear fission reactor, the core 20 in the nuclear fission reactor may be a core in a traveling wave nuclear fission reactor, for example, an exemplary reactor core 250, Lt; / RTI > In this regard, a relatively small and removable nuclear fission igniter, including, but not limited to, a moderate radioisotope enrichment such as U-233, U-235 or Pu-239. 280 are properly positioned within reactor core 250. By way of example only and not limitation, the igniter 280 may be located proximate the first end 252 opposite the second end of the reactor core 250. The neutrons are emitted by the igniter 280. The neutrons emitted by the igniter 280 are captured by the raw material and / or the raw material in the nuclear fission module 30 to initiate a fission chain reaction. If the fission chain reaction is self-sustaining, the igniter 280 may be removed, if necessary.

Referring again to FIG. 4, the igniter 280 discloses a three-dimensional deflagration wave or "soft wave" 290 having a width ("x"). The combustion wave 290 is directed from the igniter 280 proximate the first end 252 to the second end 254 of the reactor core 250 while the igniter 280 emits neutrons to induce & To form a soft-wave wave 290 propagating along the wave-propagating direction. In other words, as the softwaves 290 propagate through the reactor core 250, each nucleation module 30 may receive at least a portion of the progressive softwaves 290. The speed of the moving soft wave 290 may be constant or may not be constant. Therefore, the speed at which the soft-magnetic wave 290 propagates can be controlled. For example, the longitudinal movement (see FIG. 5) of the control rod 260 described above, in a predetermined or programmed manner, may reduce or reduce the neutron response of the fuel rod 40 disposed within the nuclear fission module 30 . In this way, the neutron reactivity of the fuel rod 40 currently burned at the location of the combustion wave 290 is relatively reduced compared to the neutron response of the fuel rod 40 that is "ahead of the burner & Or reduced. This results in providing the direction of the radio wave propagation indicated by the arrow 295.

The basic principle of such a traveling wave nuclear cracking reactor is the United States of America named as Roderick A. Hyde et al., Entitled " Automated Nuclear Power Reactor For Long-Term Operation " Which is more fully described in patent application 11 / 605,943, the entire contents of which are incorporated herein by reference.

Referring to Figures 6 and 7, there is shown an upright and adjacent hexagonal-shaped nucleation module 30. It will be appreciated that only three adjacent nuclear fission modules 30 are shown and that a greater number of nuclear fission modules 30 may be present in the reactor core 20. In addition, each nuclear fission module 30 will include a plurality of the above-described fuel rods 40. Each nucleation module 30 is mounted to the reactor core lower support plate 360 extending in the horizontal direction. The reactor core lower support plate 360 extends across all nucleation modules 30. The reactor core lower support plate 360 has a corresponding bore 370 therethrough for reasons to be described below. The corresponding bore 370 has an open end 380 for allowing the flow of coolant. A reactor core upper support plate 400 that caps each nuclear fission module 30 extends horizontally across the upper or outlet portion of each nuclear fission module 30 and is removably connected do. The reactor core upper support plate 400 also defines a plurality of flow slots 410 for allowing passage of the coolant flow.

As described above, it is important to control the temperature of the reactor core 20 and the nuclear fission module 30 therein, regardless of the configuration selected for the reactor core 20. Proper temperature control is important for several reasons. For example, if the peak temperature exceeds the substrate temperature, thermal damage may occur in the reactor core structure material. Such a peak temperature can undesirably reduce the working life of the structure exposed to such peak temperature by changing the mechanical properties of the structure, in particular by changing the properties associated with heat creep. In addition, the reactor furnace power density will be limited by the ability of the core structure material to withstand such high temperatures without damage. Alternatively, the reactor 10 may be used in test runs, such as testing to determine the effect of temperature on the reactor furnace.

Controlling the reactor core temperature is important in the successful implementation of such tests. Also, in the absence of a neutron reflector or neutron propagation blanket surrounding the perimeter of the reactor core 20, the nucleation module 30 located at or near the center of the reactor core 20 is a reactor Will generate more heat than the nucleation module 30 located at or adjacent to the periphery of the core 20. Accordingly, it may not be sufficient to provide a uniform coolant mass flow across the reactor core 20 because the higher temperature nucleation module 30, near the center of the reactor core 20, 20 will have a greater coolant mass flow rate than the nuclear fission module 30 in the vicinity of the perimeter thereof. The present disclosure provides techniques for solving such problems.

Referring to Figures 1, 6, and 7, the first pump 110 and the primary loop pipe 90 communicate the reactor coolant through a fluid stream or coolant flow path, indicated by arrow 420, 30). The primary coolant then continues through the coolant flow path 420 and through the open end 380 formed in the lower support plate 360. As will be described in more detail below, the reactor coolant can be used to remove or cool the heat from those selected from the nucleation modules 30 at the location of the progressive softwaves 290. As will be described in more detail below, the kerosene module 30 is configured to at least partially determine whether the kerosene 290 is located, or detected, or within or near the nuclear division module 30 As shown in FIG.

Referring again to Figures 1, 6 and 7, an adjustable flow regulator subassembly 430 is connected to the nucleation module (not shown) to achieve the desired goal of cooling the nuclear mitigation module selected from among the nucleation modules 30 30). The flow regulator subassembly 430 includes a plurality of flow regulator subassemblies 430 that are responsive to the location of the softwaves 290 relative to the nucleation module 30 (see FIG. 4) and also to the flow of coolant . In other words, the flow regulator subassembly 430 may be configured to allow the nuclear mitigation module 30 to generate a smaller amount of the combustion waves 290 (i.e., a lower intensity of the combustion waves 290) And is configured to supply or supply a relatively small amount of coolant. On the other hand, the flow regulator subassembly 430 is relatively less susceptible to nucleation module 30 when there is a greater amount of combustion waves 290 (i.e., a higher intensity of combustion wave 290) To supply or supply a large amount of coolant. The presence and intensity (intensity) of the softwave 290 may be identified by the heat release rate, the neutron flux level, the power level, or other appropriate operating characteristics associated with the nuclear division module 30. [

Referring to Figures 7, 8, 8a, 8b, 8c, and 8d, a suitable flow regulator subassembly 430 may be used to control the flow of fluid stream 420 into nucleation module 30, 370). Those skilled in the art will understand that to regulate the flow of fluid stream 420, flow regulator subassembly 430 provides a controllable flow resistance. The flow regulator subassembly 430 includes a generally cylindrical first or outer sleeve 450 having a plurality of first ligaments 460 which includes a plurality of radially disposed outer sleeves 450 Each of the axially spaced first holes or first controllable flow openings 470 is formed. The outer sleeve 450 further includes a first nipple 480 that may have a hexagonal-shaped cross-section for reasons described herein. The first nipple 480 forms a threaded inner cavity 500 for the reasons described below.

Referring again to Figures 7, 8, 8a, 8b, 8c, and 8d, the flow regulator subassembly 430 includes a generally cylindrical < RTI ID = 0.0 > And further includes a second or inner sleeve 530. The inner sleeve 530 may be integrally formed with the nuclear fission module 30 during manufacture of the nuclear fission module 30 such that the inner sleeve 530 is permanently attached to the nuclear fission module 30, It becomes a permanent part. The inner sleeve 530 can be detachably connected to the nucleation module 30 such that the inner sleeve 530 can be easily detached from the nucleation module 30 and accordingly, It may not be a permanent part of the mitigation module 30. [ In both embodiments, the inner sleeve 530 includes a plurality of second coupling portions 540, which includes a plurality of axially spaced apart second holes or radially disposed second sleeves 530 around the inner sleeve 530, To form each one of the controllable flow openings 550. The inner sleeve 530 includes an outer threaded second nipple 560 sized to be threadably received within the threaded inner cavity 500 of the lower portion 490 that belongs to the outer sleeve 450. The upper portion 570 of the inner sleeve 530 includes a cap 580 that may or may not be permanently formed with the nucleation module 30 as described above. An inner bore 590 extends through upper portion 570 and through cap 580 so that coolant can pass therethrough. The cap 580 and the fuel rods 40 are connected to the inner and outer bores 590 of the canister 43 to allow coolant to pass from the inner bore 590 into the canister 43 where the fuel rods 40 are located. A frusto-conical funnel portion 600 having an inner surface 605 communicating with the frusto-conical funnel portion 600 can be coupled. As described above, the nuclear cleavage module 30 can be configured to have or have a temperature dependent reactivity change. Accordingly, the flow control regulator subassembly 430 is at least partially configured to control the temperature in the nucleation module 30 by controlling the flow of coolant into the nucleation module 30 to affect temperature dependent reactivity changes .

8A and 8D, to prevent relative rotation of outer sleeve 450 relative to inner sleeve 530, the lower portion 490 of outer sleeve 450 is generally formed of a < RTI ID = 0.0 > - < / RTI > In this regard, the outer sleeve 450 includes a plurality of grooves 607a and 607b, such as grooves 607a and 607b, for accommodating each one of the plurality of tabs 608a and 608b integrally formed with the inner sleeve 530 Respectively. Accordingly, as the outer sleeve 450 is rotated, the inner sleeve 530 is prevented from rotating relative to the outer sleeve 450 because the tabs 608a and 608b are coupled within the grooves 607a and 607b, respectively Because.

The first nipple 480 is rotated relative to the outer sleeve 450, as best seen in Figure 8E. In this regard, the first nipple 480 includes an annular flange 608c that is slidably received within the cold formed annular slot 608d of the outer sleeve 450. In this manner, the first nipple 480 can be freely and slidably rotated relative to the outer sleeve 450. The first nipple 480 can be freely and slidably rotated along the direction indicated by the curved arrows 608e and 608f. Also, since the first nipple 480 can be freely and slidably rotated along one direction, such as the arrow 608e, the threaded inner cavity 500 can be screwed with the external threads of the second nipple 560, Lt; / RTI > It will be appreciated that since the threads of the inner cavity 500 are threadedly coupled to the outer threads of the second nipple 560, the first nipple 480 will contact the first sleeve 450, such as surface 608g will be. Because the first nipple 480 abuts the first sleeve 450, the first sleeve 450 will translate upward or downward along the longitudinal axis in the direction indicated by the vertical arrow 608h. The first sleeve 450 will only translate upward or downward in the direction of arrow 608h due to the presence of the anti-rotation feature 606. [ The first hole 470 is progressively closed, covered, and blocked by the second coupling portion 540 of the inner sleeve 530 as the first sleeve 450 is upwardly translated or lowered by a predetermined amount , And will be blocked. The second hole 550 is progressively closed and covered by the first binding portion 460 of the outer sleeve 450 as the first sleeve 450 is upwardly translated or lowered by a predetermined amount, It will be blocked, and it will be blocked. The progressive closure, covering, blocking, and other clogging of the first hole 470 and the second hole 550 in this manner variably reduces the flow of coolant through the first hole 470 and the second hole 550 . The rotation of the first nipple 480 along the opposite direction, such as the direction of the curved arrow 608f, causes the first hole 470 and the second hole 550 to progressively open, uncover, release, It is understood that the flow of coolant through the first hole 470 and the second hole 550 can be variably increased.

Thus, referring to Figures 7, 8, 8a, 8b, 8c, 8d, 8e, 9 and 10, it can be seen that the flow control in the nucleation module 30 is, at least in part, By an outer sleeve 450 and an inner sleeve 530 as shown in FIG. As described above, when the nucleation module 30 is first fabricated, the inner sleeve 530 may be formed integrally with the nucleation module 30. However, if necessary, instead of being formed integrally with the nuclear fission module 30 when the nuclear fission module 30 is first manufactured, the inner sleeve may be formed separately from the nuclear fission module 30, There will be. The inner sleeve 530 forms a plurality of second holes 550 to allow passage of the coolant through the nucleation module 30. The outer sleeve 450 is slid on top of the inner sleeve 530 and has a corresponding plurality of first holes 470. The outer sleeve 450 and the inner sleeve 530 are concentric and the holes 470/550 are always aligned to match along the radial or rotational direction axis. The coolant flow is controlled by the relative positions of the inner sleeve 530 and the outer sleeve 450 along the axial or vertical direction. In this regard, FIG. 8B shows a fully regulated flow regulator subassembly 430 that allows fluid to flow into the breakdown module 30, and FIG. 8C shows a flow regulator subassembly Flow regulator subassembly 430 in a fully-shut-off configuration that prevents the flow of fluid from the fluid reservoir to the fluid reservoir. Coupling the tabs 608a and 608b into the respective grooves 607a and 607b limits the rotation of the outer sleeve 450 relative to the inner sleeve 530, as described above. This feature allows the outer sleeve 450 to slide axially on the inner sleeve 530 but does not allow relative rotation between the outer sleeve 450 and the inner sleeve 530. Fine adjustment of the coolant flow is achieved by progressive axial sliding of the outer sleeve 450 relative to the inner sleeve 530. [ The rotation of the first nipple 480 along the direction 608e progressively opens the flow regulator subassembly 430 and the rotation of the first nipple 480 along the direction 608f causes the flow regulator subassembly 430, (Not shown) to gradually close the holes 430 to achieve fine adjustment of the holes 470/550 and thus fine adjustment of the coolant flow.

As best shown in FIG. 11, there may be a plurality of smaller flow regulator subassemblies, such as flow regulator subassemblies 609a and 609b assigned to a single nucleation module 30. Assigning a plurality of small flow regulator subassemblies 609a and 609b to a single nuclear fission module 30 provides another configuration for providing a coolant flow to the nuclear fission module 30. In addition, assigning a plurality of small flow regulator subassemblies 609a and 609b to an individual or single nuclear fission module 30 substantially reduces the temperature distribution within the discrete portions of the individual or single nuclear fission fuel module 30 It provides the possibility to do. This is possible because the fluid flow through each small flow regulator subassembly 609a and 609b can be individually controlled.

Referring to Figures 12, 13, 14, 15, and 16, there is shown a flow regulator subassembly 430 in operative condition for adjusting or regulating the coolant fluid flow to nucleation module 30. The flow regulator subassembly 430 and carriage subassembly 610 together form a flow control assembly, generally designated 615, as will be described in greater detail below. In other words, the flow control assembly 615 includes a flow assembly 615 and a carriage subassembly 610. In this regard, the carriage subassembly 610 is disposed beneath the reactor core 20, such as underneath the core lower support plate 360, and is positioned below the flow regulator subassembly 430 to adjust the flow regulator subassembly 430. [ May be coupled to, or coupled to, the assembly 430. Adjustment of the flow regulator subassembly 430 variably controls the flow of coolant to the nucleation module 30, as described above. Also, if necessary, carriage subassembly 610 may transfer outer sleeve 450 to nucleation module 30.

13, 14, 15, and 16, the configuration of the carriage subassembly 610 will be described. The carriage subassembly 610 includes a elongated bridge 620 spanning the reactor core 20 to support a plurality of vertically movable socket wrenches 630 thereon. Each socket wrench 630 has a shaft 700 and is movably disposed within a socket well 635 for reasons to be described below. The first bridge mover 640a and the second bridge mover 640b are each connected to opposite ends of the bridge 620. The bridge mover 640a and 640b may be operated by a gear device (not shown) driven by a motor (not shown). Such a motor would be located outside reactor core 20 to avoid the heat and corrosive effects of coolant such as liquid sodium circulating through reactor core 20. Each of bridge mover 640a and 640b includes one or more wheels 650a and 650b, respectively, such that bridge mover 640a and 640b are spaced at the same time along each of the laterally spaced, parallel tracks 660a and 660b To move. Bridge mover 640a and 640b may be configured to move or move bridge 620 along tracks 660a and 660b in either of the directions indicated by arrow 663. Track support portions 665a and 665b for supporting the tracks 660a and 660b may be connected to the respective tracks 660a and 660b, respectively.

Referring to Figures 13, 14, 15, 16, 17, 18 and 19, a socket wrench 630 is provided in the socket well 635 to engage and disengage the first nipple 480 of the outer sleeve 450, (635). ≪ / RTI > In one embodiment of the carriage subassembly 610, rows of socket wrenches 630 are configured to be driven by a leadscrew device, generally designated 670. The leadscrew device 670 includes a leadscrew 680 configured to be threadably engaged with an outer threaded portion 690 that surrounds the shaft 700 that belongs to each socket wrench 630. The lead screw 680 may be driven by a mechanical drive system 705 including a mechanical linkage 707 coupled to the lead screw 680. [ Due to the threaded engagement between the leadscrew 680 and the outer threaded portion 690 surrounding the shaft 700 when the mechanical linkage 707 drives the leadscrew 680, 700). ≪ / RTI > As shown, when the hexagonal-shaped recess 700a in the upper portion of the shaft 700 is engaged with the hexagon-shaped first nipple 480, the rotating or rotating shaft 700 is engaged with the first nipple 480, (480).

Referring to Figs. 15 and 16, the manner in which each shaft 700 is selectively raised and lowered will be described. In this regard, an outer threaded elongate mechanical linkage extension 708 may include a first gear wheel 709 and a second gear wheel 709 to rotate the first gear wheel 709 in the direction indicated by the curved arrows 709a and 709b, . For example, as the mechanical linkage extension 708 translates along one of the directions indicated by the double-headed arrow 709c, the first gear wheel 709 is rotated in the first direction, such as in the direction of arrow 709a, Direction. On the other hand, as the mechanical linkage extension 708 translates in the direction opposite to the direction indicated by the double-headed arrow 709c, the first gear wheel 709 rotates along a second direction, such as the direction of arrow 709b will be. As the first gear wheel 709 is rotated in the direction of arrow 709a, for example, the first rod 709d in the center of the outer threaded portion will rotate by a similar amount, (Not shown) that is formed through the center of the first gear wheel 709. The first gear wheel 709 is provided with an outer screw thread (not shown) The second gear wheel 709e has an internal threaded portion (not shown) formed through the center thereof to be threadably engaged with the external threaded portion of the first rod 709d. Accordingly, as the first rod 709d is rotated by the first gear wheel 709, due to the threaded engagement of the first rod 709d and the second gear wheel 709e, the second gear wheel 709e Will translationally move along the first rod 709d. The second gear wheel 709e will translate along the first rod 709d until a predetermined one of the shafts 700 is reached. In order to prevent the interference with the pitch of the outer thread portion surrounding the shaft 700 so that the translation of the second gear wheel 709e along the first rod 709d progresses unimpeded, It will be understood that the pitch of the outer threaded portion or the gear tooth portion of the gear 709e can be determined. Also, for the reasons described herein, a third gear wheel 709f is provided. In this regard, the third gear wheel 709f is connected to the elongated second rod 709g and to the third rod 709d, which is disposed on either side of the first rod 709d closest to the center, 0.0 > 709h. ≪ / RTI > The third gear wheel 709f is connected to the aforementioned mechanical linkage extension 708 which can be moved from the first engagement position with the first gear wheel 709 to the second engagement position with the third gear wheel 709f . As the third gear wheel 709f rotates, the second rod 709g and the third rod 709h are rotated about the longitudinal axis of the first rod 709d to rotate in the longitudinal direction of the first rod 709d And will rotate the second gear wheel 709e around the axis. As the second gear wheel 709e rotates, the outer threaded portion of the second gear wheel 709e is threadedly engaged with the outer threaded portion of the shaft 700 to cause the shaft 700 to translate vertically. In this manner, the socket wrench 630 is translated upwardly or downwardly. It will be appreciated that the mechanical linkage extension 708 may be replaced by a fourth gear wheel (not shown) or by a pulley belt assembly (not shown).

17, 18, and 19, in another embodiment of the carriage assembly 610, the socket wrench 630 includes a plurality of hermetically sealed, reversible, Can be individually rotated by each one of the plurality of actuators 710 and can be translated in the axial direction. The first electric motor 710 may be hermetically sealed and gas cooled to protect the first electric motor 710 from corrosion effects and heat of the coolant, such coolant may be a liquid sodium or liquid sodium mixture . The first electric motor 710 is configured to selectively move the shaft 700 vertically. In order to move the shaft 700 upwards or downwards, respectively, the motor 710 is reversible in that the rotor of the motor 710 can be operated in a first direction or in a second direction opposite to the first direction . The operation of the mechanical drive system 705 or the motor 710 is suitably controlled by a control unit or control unit 720 coupled thereto. Each motor 710 may be a custom designed DC servomotor such as is available from ARC System, Incorporated, New Haven, New York. The control unit 720 may be a motor control unit custom-designed such as is available from the Bodine Electric Company of Chicago, Ill. According to another embodiment, the socket wrenches 630 may include a plurality of hermetically sealed, gas-cooled, reversible second (not shown) receivers that can be individually actuated by receipt of a radio frequency signal transmitted by the wireless transmitter 740 And can be moved individually by a wireless transmitter-receiver device including an electric motor. To protect the second electric motor 730 from the thermal and corrosive effects of the sodium refrigerant, the second electric motor 730 may be hermetically sealed and gas cooled. The power supply of the second electric motor 730 may be a battery or other power supply (not shown). A second electric motor 730 configured to receive a radio signal, a wireless transmitter 740 may be a custom designed motor and transmitter available from Myostat Motion Control, Incorporated of Ontario, Canada . According to another embodiment, the socket wrench 630 has a plurality of fiber optic cables 745 for actuating the reversible motor arrangement by light transmission, and individually by means of a fiber optic transmitter-receiver arrangement, indicated generally by the reference numeral 742 Can be moved.

14, the flow control assembly 615, and hence the flow regulator subassembly 430, may be operated in response to or in response to the operating parameters associated with the nucleation module 30. In this regard, one or more sensors 750 may be disposed within the nuclear fission module 30 to sense the status of operational parameters. The operating parameter sensed by the sensor 750 may be the current temperature in the nucleation module 30. The operating parameter sensed by the sensor 750 may instead be the previous temperature in the nucleation module 30. To sense the temperature, the sensor 750 may be a temperature sensor or thermocouple device available from Thermocoax, Incorporated of Alpharetta, GA, USA. Alternatively, the operating parameter sensed by the sensor 750 may be a neutron flux in the nucleation module 30. To sense the neutron flux, the sensor 750 may be a "PN9EB20 / 25" neutron flux proportional counter detector, such as is available from Centronic House, Surrey, England or the like. As another example, the operating parameter sensed by the sensor 750 may be a unique isotope within the nuclear cleavage module 30. Specific isotopes may be cleavage products, activated isotopes, modified products produced by proliferation, or other specific isotopes. In another example, the operating parameter sensed by the sensor 750 may be neutron fluence in the nucleation module 30. As is well known in the art, neutron fluences represent the number of neutrons per unit area that is defined as an integral neutron flux over a period of time and passed during that time. As another example, the operating parameter sensed by the sensor 750 may be the divider module pressure, which is about 138 bar (i.e., about 2000 bar for the "hard water" psi) or about 10 bar (i.e., about 145 psi) in the case of an exemplary sodium-cooled reactor. Alternatively, the fission module pressure sensed by the sensor 750 may be a static fluid pressure or a fission product pressure. In order to sense the dynamic or static dissociation module pressure, the sensor 750 may be a custom designed pressure detector available from Kaman Measuring Systems, Incorporated of Colorado Springs, Colo. As an alternative, the sensor 750 may be a suitable flow meter such as "BLANCETT 1100 TURBINE FLOW METER" available from Instrumart, Incorporated of Williston, Vermont, USA. Also, the operating parameters sensed by the sensor 750 may be determined by a suitable computer-based algorithm. Including known algorithms that generate a signal representing a pressure or temperature from a direct or indirect measurement of an ideal gas law, i. E. An algorithm such as PV = nRT, or other properties such as flow, temperature, Various algorithms can be implemented. According to another example, the operational parameter may be an operator initiated action. That is, the flow regulator subassembly 430 may be altered in response to any appropriate operating parameters determined by the operator. In addition, the flow regulator subassembly 430 may be modified in response to operational parameters determined by appropriate feedback control. In addition, the flow regulator subassembly 430 may be modified in response to operating parameters determined by the automatic control system. In addition, the flow regulator subassembly 430 can be altered in response to a change in decay heat. In this regard, the decay heat is reduced at the "tail" of the soft wave 290 (see FIG. 4). To account for this decrease in the decay heat found at the end of the combustion wave 290, the detection of the presence of the end of the combustion wave 290 is used to reduce the coolant flow rate over time. This is especially true if the nuclear cleavage module 30 is behind the flare 290. In this case, the flow regulator subassembly 430 illustrates the change in the decay heat output of the nuclear fission module 30 as the distance from the combustion wave 290 of the nuclear fission module 30 is changed. The monitoring status of such operating parameters may facilitate appropriate control and modification of the flow control assembly 615 operation and thereby facilitate proper control and modification of the temperature within reactor core 20. [

Referring to Figures 14, 15, 17, 18 and 19, it can be seen from the above description that control 720 and 740 combined with the flow regulator subassembly 430, to control 720 and 740, It will be appreciated that the flow regulator subassembly 430 may be reconfigured according to a predetermined input. That is, the predetermined input to the control units 720 and 740 is the signal generated by the sensor 750 described above. For example, a predetermined input to the control units 720 and 740 may be a signal generated by the thermocouple or temperature sensor described above. Instead, a predetermined input to the control units 720 and 740 may be a signal generated by the flow meter described above. As another example, a predetermined input to the control units 720 and 740 may be a signal generated by the neutron detector described above. As another example, signals received by the control units 720 and 740 may be processed by a reactor control system (not shown). For example, a signal generated by the control system in such a reaction may originate from a meter or detector and may be processed by a computer or reaction operator in a control room and output to a carriage subassembly 610, The bridge 620 and the socket wrench 630 are moved to operate the subassembly 430. [

4, 10 and 14, it will be appreciated by those skilled in the art, based on the description herein, that when the progressive softwaves 290 arrive at the nucleation module 30 and / or from the nucleation module 30 It will be appreciated that the flow control assembly 615 may control or change the flow of coolant as it exits. The flow control assembly 615 may also control or change the flow of coolant by the flow control assembly 615 as the progressive softwaves 290 approach or proximate the nucleation module 30. The flow control assembly 615 may also control or change the flow of coolant in accordance with the width ("x") of the above described softwave 290. The arrival and departure of the softwave 290 is detected by sensing any of the above mentioned operating parameters as the softwave 290 travels through the nucleation module 30. For example, the flow control assembly 615 may control or change the flow of coolant in accordance with the sensed heat release rate within the nucleation module 30. Those skilled in the art will appreciate that, in some cases, the input signal alone can control the flow control assembly 615 changes and the associated fluid flow in the nucleation module 30 associated therewith.

Referring to FIGS. 14 and 15, as described above, the flow control assembly 615 is actuated to provide a variable fluid flow to a nuclear fission module selected from the nuclear fission modules 30. The nucleation module 30 is configured to determine the nucleation temperature of the nucleation module 30 based on the desired value for the operating parameters (e.g., temperature) in the nucleation module 30 compared to the actual values of the operating parameters sensed in the nucleation module 30 Is selected. As will be described in more detail, the fluid flow to the nucleation module 30 is adjusted so that the actual value for the operating parameter is substantially consistent with the desired value for the operating parameter. To achieve this result, the bridge 620 belonging to the carriage subassembly 630 is moved along the tracks 660a and 660b by simultaneously operating bridgemovers 640a and 640b. As the bridge 620 moves along the tracks 660a and 660b, the bridge 620 will move under the core lower support plate 360. As a result, the bridge 620 can be configured to determine the operating parameters of the operating parameters detected by the sensors 750 in the nucleation module 30 compared to the desired values of the operating parameters for the nucleation module 30, And stops moving at a predetermined position below the core lower support plate 360 based on the actual value. The activation and movement range of the bridge mover 640a and 640b may be controlled by an appropriate control unit such as the control unit 720 or 740. In this regard, the control unit 720 or 740 may stop the movement of the bridge 620 based on the location of the nuclear fission module selected from among the plurality of nuclear fission modules 30. The tuning nucleation module 30 determines whether there is a substantial match between the actual value of the operating parameter sensed by the sensor 750 and the desired operating parameter value for the nucleation module 30 As shown in FIG. Next, a socket wrench selected from a plurality of hexagonal socket wrenches 630 is vertically moved upward to be engaged with the hexagonal first nipple 480 in a matching manner. After the socket wrench 630 is engaged with the first nipple 480, the shaft 700 is rotated to rotate the socket wrench 630. The shaft 700 may be rotated by the lead screw device 670, the first electric motor 710, or the second motor 730, which is coupled to the control unit 720 or 740.

Referring to Figures 7, 8, 8a, 8b, 8c, 8d, 8e, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18 and 19, after engagement with the first nipple 480 , The rotation of the socket wrench 630 along the first direction rotates the first or outer sleeve 450 in the same direction. As the outer sleeve 450 rotates, the outer sleeve 450 will slide axially along the outer side of the inner sleeve 530, which will be referred to as the first nipple 480, which belongs to the outer sleeve 450, Due to the threaded engagement of the second nipple 560 belonging to the inner sleeve 530. [ As the outer sleeve 450 slides upwards along the inner sleeve 530 the first binding portion 460 of the outer sleeve 450 progressively closes the second hole 550 of the inner sleeve 530 And the second binding portion 540 of the inner sleeve 530 simultaneously closes, covers, blocks, and closes the first hole 470 of the outer sleeve 450, And will stop. Gradually closing, covering, blocking, and blocking the first hole 470 and the second hole 550 allows the flow of coolant through the first hole 470 and the second hole 550 to be variably reduced . In this case, the second hole 550 and the first hole 470 may be pre-aligned to allow full through flow of the coolant. Instead, the second hole 550 and the first hole 470 may be partially pre-aligned to allow for partial through flow of the coolant.

Referring again to Figures 7, 8, 8a, 8b, 8c, 8d, 8e, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18 and 19, the coupling with the first nipple 480 Thereafter, when the socket wrench 630 rotates along the second direction opposite to the first direction, the first or outer sleeve 450 is rotated in the second direction. As the outer sleeve 450 rotates, the outer sleeve 450 will slide down axially along the outer side of the inner sleeve 530, which will move the first nipple 480, which belongs to the outer sleeve 450, Due to the threaded engagement of the second nipple 560 belonging to the inner sleeve 530. [ As the outer sleeve 450 slides downward along the inner sleeve 530 the first binding portion 460 of the outer sleeve 450 progressively opens the second hole 550 of the inner sleeve 530 Uncovering, releasing, and otherwise opening the first sleeve 470 of the inner sleeve 530 and the second binding portion 540 of the inner sleeve 530 simultaneously opens and uncovers the first hole 470 of the outer sleeve 450, Release and open it differently. Gradually opening, uncovering, releasing, and otherwise opening the first and second holes 470 and 550 may be accomplished by opening the first hole 470 and the second hole 550 through the first hole 470 and the second hole 550, Variably increasing the flow. In this case, the second hole 550 and the first hole 470 may be misaligned in advance to block or block the flow of coolant through. Alternatively, the second hole 550 and the first hole 470 may be partially misaligned in advance to partially block or partially block the flow of coolant through.

Accordingly, the use of the flow control assembly 615, including the flow regulator subassembly 430 and the carriage subassembly 630, allows the use of a module-by-module (i.e., per fuel assembly) Lt; RTI ID = 0.0 > a < / RTI > This allows the coolant flow to be variable across the reactor core 20 depending on the non-uniform temperature distribution in the reactor core 20 or the location of the softwaves 290.

An exemplary method

In the following, an exemplary method in connection with an exemplary embodiment of a nuclear fission reactor and flow control assembly is described.

Referring to Figures 20a-20s, an exemplary method for operating a nuclear fission reactor is presented.

Referring to FIG. 20A, an exemplary method 760 for operating the nuclear fission reactor begins at block 770. FIG. At block 780, the method includes generating at least a portion of the progressive wave at a location relative to the nucleation module. At block 790, the flow control assembly is actuated to change the flow of fluid in response to the position relative to the nucleation module. The method ends at block 800.

In FIG. 20B, an exemplary method 810 for operating the nuclear fission reactor begins at block 820. FIG. At block 830, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 840, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. The method ends at block 860.

In FIG. 20C, another exemplary method 870 for operating the nuclear fission reactor begins at block 880. FIG. At block 890, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 900, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. The flow regulator subassembly is actuated in block 910. At block 920, the flow regulator subassembly operates in accordance with the operating parameters associated with the nucleation module. This method ends at block 930. [

In Figure 20d, another exemplary method 940 for operating the nuclear fission reactor begins at block 950. [ At block 960, at least a portion of the progressive wave is generated at a location relative to the nucleation module. At block 970, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. The flow regulator subassembly is actuated in block 980. At block 990, the flow regulator subassembly is changed in accordance with the operating parameters associated with the nucleation module. This method ends at block 1000.

In FIG. 20E, another exemplary method 1010 for operating the nuclear fission reactor begins at block 1020. FIG. At block 1030, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1040, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. The flow regulator subassembly is operated in block 1050. [ At block 1060, the flow regulator subassembly is reconfigured according to a predetermined input to the flow regulator subassembly. This method ends at block 1070.

In Figure 20f, another exemplary method 1080 for operating the nuclear fission reactor begins at block 1090. [ At block 1100, at least a portion of the progressive wave is generated at a location relative to the nucleation module. At block 1110, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to a position relative to the nucleation module. The flow regulator subassembly is actuated in block 1120. At block 1130, a controllable flow resistance is obtained. This method ends at block 1140. [

In FIG. 20G, another exemplary method 1150 for operating the nuclear fission reactor begins at block 1160. FIG. At block 1170, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1180, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. The flow regulator subassembly is operated in block 1190. [ In block 1200, a second sleeve is inserted into the first sleeve, with the first sleeve having a first hole and the second sleeve having a second hole that can be aligned with the first hole. This method ends at block 1210.

In Figure 20h, another exemplary method 1220 for operating the nuclear fission reactor begins at block 1230. [ At block 1240, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1250, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to a position relative to the nucleation module. The flow regulator subassembly is operated in block 1260. [ At block 1270, a carriage subassembly coupled to the flow regulator subassembly is actuated. This method ends at block 1280. [

In Figure 20i, another exemplary method 1290 for operating the nuclear fission reactor begins at block 1300. [ At block 1310, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1320, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position relative to the nucleation module. A flow regulator subassembly is operated in block 1330. [ At block 1340, a temperature sensor is coupled to the nuclear fission module and the flow regulator subassembly. This method ends at block 1350. [

In Figure 20j, a further exemplary method 1360 for operating the nuclear fission reactor begins at block 1370. [ At block 1380, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1390, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to a position relative to the nucleation module. At block 1400, the flow of fluid is controlled in response to a position relative to the position of the nucleation module by actuating the flow control assembly according to the time at which the softwave reaches a position relative to the location of the nucleation module. This method ends at block 1410. [

In FIG. 20K, another exemplary method 1420 for operating the nuclear fission reactor begins at block 1430. At block 1440, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1450, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to a position relative to the nucleation module. At block 1460, the flow of fluid is controlled in response to the position with respect to the nucleation module by actuating the flow control assembly in accordance with the timing at which the soft wave deviates (originates) from the position relative to the nucleation module. This method ends at block 1470.

In Figure 20I, another exemplary method 1480 for operating the nuclear fission reactor begins at block 1490. [ At block 1500, at least a portion of the progressive wave is generated at a location relative to the nucleation module. At block 1510, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to the position with respect to the nucleation module. At block 1520, the flow of fluid is controlled in response to a position relative to the nucleation module by activating the flow control assembly in accordance with when the burning wave approaches the location relative to the nucleation module. This method ends at block 1530. [

In Figure 20m, an exemplary method 1540 for operating the nuclear fission reactor begins at block 1550. [ At block 1560, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1570, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the relative position with respect to the nucleation module. At block 1580, the flow of fluid is controlled according to the width of the combustion wave. This method ends at block 1590. [

In Figure 20n, another exemplary method 1600 for operating the nuclear fission reactor begins at block 1610. At block 1620, at least a portion of the progressive wave is generated at a location relative to the nucleation module. At block 1630, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to location relative to the nucleation module. At block 1640, the flow of fluid is controlled by operating the flow control assembly in accordance with the rate of heat generation in the nuclear breakdown module. This method ends at block 1650.

In Figure 20O, another exemplary method 1660 for operating the nuclear fission reactor begins at block 1670. At block 1680, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1690, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to the position with respect to the nucleation module. At block 1700, the flow of fluid is controlled by operating the flow control assembly in accordance with the temperature in the nucleation module. This method ends at block 1710. [

In FIG. 20P, another exemplary method 1720 for operating the nuclear fission reactor begins at block 1730. At block 1740, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1750, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to the position with respect to the nucleation module. At block 1760, the flow of fluid is controlled by operating the flow control assembly along with the neutron flux in the nucleation module. This method ends at block 1770.

In Figure 20q, another exemplary method 1780 for operating the nuclear fission reactor begins at block 1790. [ At block 1800, at least a portion of the progressive wave is generated at a location relative to the nucleation module. At block 1810, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to the position with respect to the nucleation module. At block 1820, at least a portion of the progressive combustion wave is generated at a location relative to the nuclear fuel assembly. This method ends at block 1830.

In Figure 20r, another exemplary method 1840 for operating the nuclear fission reactor begins at block 1850. [ At block 1860, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1870, a flow control assembly coupled to the nucleation module is activated to change the flow of fluid in response to the position with respect to the nucleation module. At block 1880, at least a portion of the progressive softwaves are generated at a location relative to the source nuclear proliferation assembly. This method ends at block 1890.

In Figure 20s, another exemplary method 1900 for operating the nuclear fission reactor begins at block 1910. At block 1920, at least a portion of the progressive softwave is generated at a location relative to the nucleation module. At block 1930, a flow control assembly coupled to the nucleation module is actuated to change the flow of fluid in response to position with respect to the nucleation module. At block 1940, at least a portion of the progressive combustion wave is generated at a position relative to the neutron reflector assembly. This method ends at block 1950.

21A-21H, exemplary methods for assembling a flow control assembly for use in a nuclear fission reactor are provided.

Referring to FIG. 21A, an exemplary method 1960 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 1970. At block 1980, a flow regulator subassembly is received. The method ends at block 1990.

In Fig. 21B, an exemplary method 2000 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2010. At block 2020, a carriage subassembly is received. The method ends at block 2030.

In FIG. 21C, another exemplary method 2040 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2050. A flow regulator subassembly is received in block 2060. A first sleeve having a first hole is received in block 2070. At block 2080, a second sleeve is inserted into the first sleeve, the second sleeve has a second hole that can be aligned with the first sleeve, and the first sleeve is configured to rotate, And can be rotated in alignment with the second hole. At block 2090, the carriage subassembly is coupled to the flow regulator subassembly. The method ends at block 2100.

In Figure 21d, another exemplary method 2110 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2120. [ A flow regulator subassembly is received in block 2130. A first sleeve having a first hole is received in block 2140. At block 2150, a second sleeve is inserted into the first sleeve and the second sleeve has a second hole that can be aligned with the first sleeve. At block 2160, the carriage subassembly is coupled to the flow regulator subassembly. At block 2170, the carriage subassembly is coupled to the flow regulator subassembly, such that the carriage subassembly transfers the flow regulator subassembly to the fuel assembly. The method ends at block 2180.

In Figure 21E, another exemplary method 2190 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2200. [ A flow regulator subassembly is received in block 2210. A first sleeve having a first hole is received in block 2220. At block 2230, a second sleeve is inserted into the first sleeve and the second sleeve has a second hole that can be aligned with the first sleeve. At block 2240, the carriage subassembly is coupled to the flow regulator subassembly. At block 2250, the carriage subassembly is coupled to the flow regulator subassembly, such that the carriage subassembly is driven by the leadscrew device. The method ends at block 2260.

In Figure 21F, another exemplary method 2270 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2280. [ A flow regulator subassembly is received in block 2290. A first sleeve having a first hole is received in block 2300. In block 2310, a second sleeve is inserted into the first sleeve, the second sleeve has a second hole that can be aligned with the first sleeve, and the first sleeve is configured to rotate, Can be rotated in alignment with the second hole. At block 2320, the carriage subassembly is coupled to the flow regulator subassembly. At block 2330, the carriage subassembly is coupled such that the carriage subassembly is driven by a reversible motor arrangement. The method ends at block 2340.

In FIG. 21G, an exemplary method 2350 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2360. A flow regulator subassembly is received in block 2370. A first sleeve having a first hole is received in block 2380. At block 2390, a second sleeve is inserted into the first sleeve, the second sleeve has a second hole that can be aligned with the first hole, and the first sleeve is configured to rotate, Can be rotated in alignment with the second hole. At block 2400, the carriage subassembly is coupled to the flow regulator subassembly. At block 2410, the carriage subassembly is coupled such that the carriage subassembly is at least partially controlled by a wireless transmitter-receiver device that operates a reversible motor arrangement. The method ends at block 2415.

In Figure 21h, another exemplary method 2420 for assembling a flow control assembly for use in a nuclear fission reactor begins at block 2430. [ A flow regulator subassembly is received in block 2440. A first sleeve having a first hole is received in block 2450. At block 2460, a second sleeve is inserted into the first sleeve, the second sleeve has a second hole that can be aligned with the first hole, and the first sleeve is configured to rotate, Can be rotated in alignment with the second hole. At block 2470, the carriage subassembly is coupled to the flow regulator subassembly. At block 2480, the carriage subassembly is coupled such that the carriage subassembly is at least partially controlled by a fiber optic transmitter-receiver device that operates a reversible motor arrangement. The method ends at block 2490.

Those skilled in the art will appreciate that the components (e.g., operations), devices, objects, and the accompanying description described herein are used as examples for conceptual clarity and that various structural variations are possible will be. As a result, as used herein, the specific examples mentioned and the accompanying description are intended to represent the more general categories. In general, the use of any particular specimen herein is intended to indicate its scope, and the non-inclusion of such specific components (e.g., operations), apparatus, and objects herein should not be treated as a limitation something to do.

The particular illustrative processes and / or apparatus and / or techniques described above may be embodied in other specific forms without departing from the spirit or essential characteristics thereof, such as the scope of the appended claims and / or elsewhere in this application A person skilled in the art will understand.

It will be apparent to those skilled in the art that, although the claims of certain embodiments are illustrated and described herein, it will be apparent to those skilled in the art, on the basis of the present teachings, that variations and modifications may be made without departing from the scope of the claims and the broader aspects set forth herein, , It is intended that the appended claims cover all such modifications and variations as fall within the true spirit and scope of the claims set forth herein. It is also to be understood that the present disclosure is defined by the appended claims. In general, terms used herein and particularly in the appended claims (e.g., the text of the appended claims) are intended to be inclusive in a manner similar to the term "Quot;, "having ", the word " having ", and the term " including" Also, those skilled in the art will appreciate that if a particular number of the claim citation introductory is intended, such intent is expressly recited in the claim, and such citation is not intended to be so. For example, to facilitate understanding, subsequent claims may include the use of the phrases "at least one" and "more than one" to introduce a claim citation. It should be understood, however, that the use of such phrases is intended to cover such phrases even if the same claim includes singular expressions (indefinite articles such as "a" or "an") and introduction phrases such as "one or more" Quot; a "or" an ") should not be construed to mean limiting the entry of such a claim citation to a claim that includes only such quotation (e.g., quot; a "and / or" an "are usually to be interpreted to mean" at least one " The same is true of the case where the definite article is used in the lead-in citation of the claim. Also, those skilled in the art will recognize that, even if a particular number of a claim citation entry is explicitly quoted, it should normally be understood that such quotation should at least be construed as referring to the number quoted (e.g., Quot; two quoted "means usually at least two citations or two or more citations). Also, in the example where idioms similar to "at least one of A, B, and C" are used, this interpretation is generally intended in such a way that those skilled in the art understand the idiom (e.g., B, and C includes a system having only A, only B, only C, A and B, A and C, B and C, and / or A, B, But is not limited thereto). In an example where idioms similar to "at least one of A, B, or C" are used, this interpretation is generally intended in such a way that those skilled in the art understand the idiom (eg, Or C includes, but is not limited to, systems having only A, only B, only C, A and B, A and C, B and C, and / or A, B, . Furthermore, in the description, the claims, or the drawings, it is to be understood that substantially any disjunctive words and / or phrases of two or more alternative terms may include the term, one of the terms, As will be understood by those skilled in the art. For example, it will be understood that the phrase "A or B" includes the possibility of "A" or "B" or "A and B"

In the context of the claims, those skilled in the art will understand that the described operations may generally be performed in any order. In addition, although optional flows are presented sequentially, it will be understood that various operations may be performed in a different order than that described, or may be performed simultaneously. Examples of such other sequences may include overlapping, intervening, interrupting, rearranging, increasing, preparing, supplementing, synchronizing, reversing, or other altering orders, unless otherwise stated. Also, unless stated otherwise, "responding", "related", or other past adjectives generally should not be construed as excluding such modifications.

Thereby, in a nuclear fission reaction, a flow control assembly, methods thereof, and a flow control assembly system are provided.

While various aspects and embodiments have been disclosed, other aspects and embodiments will be apparent to those skilled in the art. For example, a horizontally disposed orifice plate may replace the flow regulator subassembly, and such orifice plate has a plurality of through orifices. A plurality of individually operable shutters may be associated with each of the orifices and the shutters may progressively close and open the orifices to adjust or change the flow of coolant to the cracking module.

In addition, from the teachings herein, the flow control assembly and system of the present disclosure, unlike the devices described in the prior art patents cited herein, dynamically change the amount of fluid flow, Avoids relying on different and precisely configured neutron-induced growth properties of the material and, if necessary, can be changed dynamically during operation as a reaction.

In addition, the various aspects and embodiments disclosed herein are for the purpose of illustration and not of limitation, the true scope and spirit being determined by the claims.

Claims (14)

A flow control assembly system for use in a nuclear fission reactor configured to be connected to a nuclear fission fuel assembly,
An adjustable flow regulator subassembly configured to be disposed within the fluid stream,
The adjustable flow regulator subassembly
(a) a first sleeve having a first hole; And
(b) a second sleeve configured to be inserted into the first sleeve,
Wherein the first hole is alignable with the second hole such that when the first hole is aligned with the second hole a predetermined amount of fluid stream flows through the first and second holes,
The first sleeve is configured to translate axially relative to the second sleeve to align the first hole with the second hole,
The adjustable flow regulator subassembly includes a rotation-preventing arrangement for engaging the first sleeve and the second sleeve, wherein the rotation-preventing arrangement is configured such that the first sleeve rotates relative to the second sleeve And permits axial translation of the first sleeve relative to the second sleeve. delete delete The flow control assembly system of claim 1, wherein the flow control assembly system further comprises a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly. 5. The flow control assembly system of claim 4, wherein the carriage subassembly is driven by a leadscrew device. 5. The flow control assembly system of claim 4, wherein the carriage subassembly is driven by a reversible motor arrangement. 7. The flow control assembly system of claim 6, wherein the carriage subassembly is at least partially controlled by a wireless transmitter-receiver device that operates the reversible motor arrangement. 7. The flow control assembly system of claim 6, wherein the carriage subassembly is at least partially controlled by a fiber optic transmitter-receiver arrangement that operates the reversible motor arrangement. delete delete delete delete delete delete

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