Patent Application
20230112687
The present subject matter relates to examples of nuclear reactor systems and nuclear reactors for power production and propulsion, e.g., in remote regions, such as outer space. The present subject matter also encompasses a nuclear reactor that includes an in-vessel neutron shield and a method for fabricating the in-vessel neutron shield.
Nuclear fission reactors include thermal or fast type reactors. Currently, almost all operating nuclear fission reactors are thermal. Nuclear fission reactors include nuclear fuel inside a nuclear reactor core and a moderator to slow down fast neutrons so that nuclear fission can continue. Typically, the nuclear fuel is formed in cylindrical shaped fuel compacts or pellets. The fuel compacts are loaded into fuel pins or rods, cladded, and stacked inside the numerous columns of fuel elements in the nuclear reactor core.
The nuclear reactor core increases neutron fluence or neutron flux, which is the distance free neutrons move in a volume per time. Neutron fluence is necessary in nuclear reactors to generate heat and energy, but neutron fluence, particularly fast neutron fluence, or the neutron fluence of neutrons moving at a speed of 14,000 km/s or higher, is very dangerous to both matter and human life. Therefore, the high fast neutron fluence is decreased to safe levels at distances further from the nuclear reactor core, ideally to safe values just beyond the nuclear reactor itself. Some components of a nuclear reactor, such as neutron reflectors, perform this action: reflectors direct free neutrons back toward the nuclear reactor core, thereby increasing neutron fluence inside the nuclear reactor, improving energy efficiency, and decreasing neutron fluence outside the nuclear reactor, making the surrounding area safer.
When building a conventional nuclear reactor system for conventional terrestrial land applications, e.g., a nuclear power plant, the size (e.g., space or volume) and mass of the nuclear reactor is not a major concern. Typically, the actual thickness of the neutron reflector to reflect neutrons back to the nuclear reactor core for energy efficiency is much greater than the required thickness needed to appropriately reduce fast neutron fluence to safe levels beyond the nuclear reactor. Therefore, the conventional nuclear reactor tends to have a very thick neutron reflector in order to reduce neutron fluence external to the nuclear reactor. Typically, the neutron reflector is much thicker than required for optimal neutronic performance needed to reflect the free neutrons back to the nuclear reactor core. Essentially, the neutron reflector additionally acts as a neutron shield.
However, the size and mass of a nuclear reactor system are important factors for nuclear thermal propulsion (NTP) and providing nuclear power (e.g., thermal and/or electrical power) in remote region applications including outer space, celestial bodies, planetary bodies, and remotes regions on Earth. For example, the mass of the nuclear reactor system directly affects performance, such as power per mass, in the NTP and space applications; and may add unnecessary manufacturing cost. Moreover, in the remote region applications, a smaller (more compact) form factor reduces construction costs and increases transportability while allowing operation of the nuclear reactor at high power densities. Accordingly, improvements to the conventional implementation of a neutron reflector as a space-inefficient neutron shield are needed.
The various examples disclosed herein relate to nuclear technologies for nuclear reactor systems both for space or terrestrial land applications. The nuclear reactor system 100 advantageously reduces the size and mass of the nuclear reactor 107 by implementing integrated in-vessel shield 105 to separate the role of neutron reflector and the neutron shield and still obtain optimal neutronic performance. The integrated in-vessel shield 105 brings radiation shielding into the pressure vessel 160 of the nuclear reactor 107, which allows for reduction of mass of the nuclear reactor system 100 and achieves a space and mass efficient neutron shield.
Integrated in-vessel shield 105 has the following benefits. First, in-vessel shield 105 reduces the required distance between the pressure vessel 160 and the active nuclear reactor core 101 and thus the total size of the pressure vessel 160. For example, in-vessel shield 105 enables a thinner reflector 140 and smaller overall pressure vessel 160. Prior neutron reflectors unnecessarily increase diameter of the pressure vessel 160, leading to a larger diameter of the pressure vessel 160 that is more expensive to manufacture and difficult to transport and field. Second, in-vessel shield 105 extends the life of the pressure vessel 160 by reducing fast flux to the pressure vessel 160. Third, in-vessel shield 105 reduces activation outside of the nuclear reactor core 101 and therefore makes the nuclear reactor 107 easier to install. Fourth, in-vessel shield 105 can be incorporated (e.g., retrofitted) into an already built nuclear reactor 107 or the design of a new nuclear reactor 107.
To achieve breakthrough performance, in-vessel shield 105 includes a near black neutron absorbing material and a gray neutron absorbing material to survive the radiation environment and thermal environment that is found inside of the pressure vessel 160 of the nuclear reactor 107. For fabrication of the in-vessel shield 105, advanced 3D printing and spark plasma sintering methods can be employed. In-vessel shield 105 is shaped as a plurality of in-vessel shield tiles 131A-N that shape the near black neutron absorbing material and the gray neutron absorbing material into an interlocking geometry pattern well-suited for placement on the interior wall 133 inside of the pressure vessel 160 and removes neutron streaming paths. In-vessel shield 105 solves the problem of limiting fast fluence to the pressure vessel 160 in a volume-constrained nuclear reactor 107 and decreases fast flux to the pressure vessel 160 with the outside. In-vessel shield 105 reduces the radiation field around the nuclear reactor 107, which in turn decreases the activation in the area around the nuclear reactor 107 and facilitates installation.
An example nuclear reactor system 100 includes a pressure vessel 160 including an interior wall 133 and a nuclear reactor core 101 located within the interior wall 133 of the pressure vessel 160. Nuclear reactor core 101 includes a plurality of fuel elements 104A-N and at least one moderator element 103. Nuclear reactor system 100 includes a reflector 140 located inside the pressure vessel that includes a plurality of reflector blocks 141A-N laterally surrounding the plurality of fuel elements 104A-N and the at least one moderator element 103. Nuclear reactor system 100 includes an in-vessel shield 105 located on the interior wall 133 of the pressure vessel 160 to surround the reflector blocks 141A-N. In-vessel shield 105 is formed of two or more neutron absorbing materials. The two more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material.
An example method includes selecting two or more neutron absorbing materials to form an in-vessel shield 105 (block 610). The two or more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material. The method further includes eutectic sintering the near black neutron absorbing material to fabricate a ceramic absorbing powder (block 615). The method further include kinetically mixing the gray neutron absorbing material and the ceramic absorbing powder to create an in-vessel shield mixture (block 620). The method further includes cold press sintering the in-vessel shield mixture into an in-vessel shield 105 (block 625).
Additional objects, advantages and novel features of the examples will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following and the accompanying drawings or may be learned by production or operation of the examples. The objects and advantages of the present subject matter may be realized and attained by means of the methodologies, instrumentalities and combinations particularly pointed out in the appended claims.
The drawing figures depict one or more implementations, by way of example only, not by way of limitations. In the figures, like reference numerals refer to the same or similar elements.
100
101
101A-B
102A-N
103A-N
104A-N
105
105
107
107A-K
112
113
114
115A-N
116
117
118A-N
131A-N
132
133
140
141A-N
160
200
205A-K
206A-K
210A-K
211A-K
300
301A-B
305
305A-B
310
400
401
402
403
405
406
411
412
413
500
600
601
602
In the following detailed description, numerous specific details are set forth by way of examples in order to provide a thorough understanding of the relevant teachings. However, it should be apparent to those skilled in the art that the present teachings may be practiced without such details. In other instances, well known methods, procedures, components, and/or circuitry have been described at a relatively high-level, without detail, in order to avoid unnecessarily obscuring aspects of the present teachings.
The term “coupled” as used herein refers to any logical or physical connection. Unless described otherwise, coupled elements or devices are not necessarily directly connected to one another and may be separated by intermediate components, elements, etc.
Unless otherwise stated, any and all measurements, values, ratings, positions, magnitudes, sizes, angles, and other specifications that are set forth in this specification, including in the claims that follow, are approximate, not exact. Such amounts are intended to have a reasonable range that is consistent with the functions to which they relate and with what is customary in the art to which they pertain. For example, unless expressly stated otherwise, a parameter value or the like may vary by as much as ± 5% or as much as ± 10% from the stated amount. The term “approximately” or “substantially” means that the parameter value or the like varies up to ± 10% from the stated amount.
The orientations of the nuclear reactor core 101, nuclear reactor 107, associated components, and/or any nuclear reactor system 100 incorporating the in-vessel shield 105, such as shown in any of the drawings, are given by way of example only, for illustration and discussion purposes. In operation for a particular nuclear reactor system 100, the nuclear reactor 107 may be oriented in any other direction suitable to the particular application of the nuclear reactor 107, for example upright, sideways, or any other orientation. Also, to the extent used herein, any directional term, such as lateral, longitudinal, up, down, upper, lower, top, bottom, and side, are used by way of example only, and are not limiting as to direction or orientation of any nuclear reactor 107 or component of the nuclear reactor 107 constructed as otherwise described herein. Reference now is made in detail to the examples illustrated in the accompanying drawings and discussed below.
In a third example, the nuclear reactor core 101 can be implemented like the nuclear reactor core 101 described in
As further shown in
In-vessel shield 105 is formed of two or more neutron absorbing materials. The two more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material. Near black neutron absorbing material includes a composited ceramic material. Gray neutron absorbing material includes a heavy metal material. The composited ceramic material and the heavy metal material form the in-vessel shield. Composited ceramic material includes boron carbide (B4C), hafnium carbide (HfC), or gadolinium oxide (Gd2O3). Composited ceramic material can further include aluminum oxide (Al2O3) or silicon carbide (SiC). Heavy metal material includes tungsten (W), iron (Fe), Nickel (Ni), or copper (Cu).
More specifically, composited ceramic material can include: boron-10 carbide (10B4C), a boron-10 carbide and aluminum oxide composite (10B4C—Al2O3) of 50% boron-10 carbide by weight percent, a boron-10 carbide and silicon carbide composite (10B4C—SiC) of 50% boron-10 carbide by weight percent, or a borated stainless steel alloy of 5% boron-10 by weight percent. The heavy metal material includes atomized tungsten heavy metal with a tungsten content greater than or equal to 90% by weight percent. Composited ceramic material can include a plurality of composited ceramic particles with an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns. The plurality of composited ceramic particles can be embedded inside a heavy metal matrix of the heavy metal material to form the in-vessel shield 105.
In-vessel shield 105 can be formed integrally (e.g., as one component, part, or piece) as an in-vessel shield liner 132 that is a suitable shape to line the interior wall 133, e.g., a tube or pipe shape. Alternatively, the in-vessel shield 105 is formed as several components formed separately and then connected together. Hence, as depicted in
In-vessel shield tiles 131A-N are tolerant to swelling and reduce neutron streaming paths. In-vessel shield tiles 131A-N have a composition and geometry that can be designed particularly for the nuclear reactor system 100, for example, the plurality of in-vessel shield tiles 131A-N include a base shape with an interlocking geometry pattern. In
Interior wall 133 can be formed of a continuous surface, e.g., rounded, aspherical, or spherical surfaces to form a cylinder or other conical surfaces to form a quadric surface, such as a hyperboloid, cone, ellipsoid, paraboloid, etc. Alternatively or additionally, the interior wall 133 can be formed of a plurality of discontinuous surfaces (e.g., to form a cuboid or other polyhedron). As used herein, “discontinuous” means that the surfaces in aggregate do not form a continuous round (e.g., circular or oval) perimeter of the interior wall 133. In
Returning now to
In the depicted example, the nuclear reactor system 100 with the nuclear reactor core 101 is utilized in a space environment, such as in a nuclear thermal propulsion (NTP) system. An example NTP system that the integrated in-vessel shield 105 can be implemented in is described in
In such an NTP system (e.g., compact space nuclear reactor), a generated thrust propels a vehicle that houses, is formed integrally with, connects, or attaches to the nuclear reactor core 101, such as a rocket, drone, unmanned air vehicle (UAV), aircraft, spacecraft, missile, etc. Typically, this is done by heating a propellant, typically low molecular weight hydrogen, to over 2,600° Kelvin by harnessing thermal energy from the nuclear reactor core 101. In addition, the NTP nuclear reactor system 100 can be used in the propulsion of submarines or ships.
As noted above, the nuclear reactor system 100 can also be a nuclear power plant in a terrestrial land application, e.g., for providing nuclear power (e.g., thermal and/or electrical power) for remote region applications, including outer space, celestial bodies, planetary bodies, and remotes regions on Earth. An example terrestrial land nuclear reactor system that the in-vessel shield 105 can be implemented in is described in
Nuclear reactor system 100 can also be a terrestrial power system, such as a nuclear electric propulsion (NEP) system for fission surface power (FSP) system. NEP powers electric thrusters such as a Hall-effect thruster for robotic and human spacecraft. FSP provides power for planetary bodies such as the moon and Mars. In the NEP and FSP power applications, the nuclear reactor system 100 heats a working fluid (e.g., He, HeXe, Ne, CO2) through a power conversion system (e.g., Brayton) to produce electricity. Moreover, in the NEP and FSP power applications, the nuclear reactor system 100 does not include a propellant, but rather includes a working fluid that passes through a reactor inlet when producing power. In the NEP and FSP power applications, the moderator elements 103A-N can be cooled via the reactor inlet working fluid (e.g., the flow coming out of a recuperator) before the working fluid passes through the fuel elements 104A-N.
Each of the fuel elements 104A-N, shown as nuclear fuel tiles 104A-N, includes a nuclear fuel. The nuclear fuel includes a fuel compact comprised of coated fuel particles, such as tristructural-isotropic (TRISO) fuel particles embedded inside a high-temperature matrix. In some implementations, the nuclear fuel includes a fuel compact comprised of bistructuralisotropic (BISO) fuel particles embedded inside the high-temperature matrix. The high-temperature matrix includes silicon carbide, zirconium carbide, titanium carbide, niobium carbide, tungsten, molybdenum, or a combination thereof. Each of the TRISO fuel particles can include a fuel kernel surrounded by a porous carbon buffer layer, an inner pyrolytic carbon layer, a binary carbide layer (e.g., ceramic layer of SiC or a refractory metal carbide layer), and an outer pyrolytic carbon layer. The refractory metal carbide layer of the TRISO fuel particles can include at least one of titanium carbide (TiC), zirconium carbide (ZrC), niobium carbide (NbC), tantalum carbide, hafnium carbide, ZrC—ZrB2 composite, ZrC—ZrB2—SiC composite, or a combination thereof. The high-temperature matrix can be formed of the same material as the binary carbide layer of the TRISO fuel particles.
A description of TRISO fuel particles dispersed in a silicon carbide matrix to form a cylindrical shaped nuclear fuel compact is provided in the following patents and publications of Ultra Safe Nuclear Corporation of Seattle, Washington: U.S. Pat. No. 9,299,464, issued Mar. 29, 2016, titled “Fully Ceramic Nuclear fuel and Related Methods”; U.S. Pat. No. 10,032,528, issued Jul. 24, 2018, titled “Fully Ceramic Micro-encapsulated (FCM) fuel for CANDUs and Other Reactors”; U.S. Pat. No. 10,109,378, issued Oct. 23, 2018, titled “Method for Fabrication of Fully Ceramic Microencapsulation Nuclear Fuel”; U.S. Pat. Nos. US 9,620,248, issued Apr. 11, 2017 and 10,475,543, issued Nov. 12, 2019, titled “Dispersion Ceramic Micro-encapsulated (DCM) Nuclear Fuel and Related Methods”; U.S. Pat. Pub. No. 2020/0027587, published Jan. 23, 2020, titled “Composite Moderator for Nuclear Reactor Systems”; and U.S. Pat. No. 10,573,416, issued Feb. 25, 2020, titled “Nuclear Fuel Particle Having a Pressure Vessel Comprising Layers of Pyrolytic Graphite and Silicon Carbide,” the entireties of which are incorporated by reference herein. As described in those Ultra Safe Nuclear Corporation patents, the nuclear fuel can include a cylindrical fuel compact or pellet comprised of TRISO fuel particles embedded inside a silicon carbide matrix to create a cylindrical shaped nuclear fuel compact.
As shown, nuclear reactor core 101 includes an insulator element array 112 of insulator elements 102A-N and a moderator element array 113 of moderator elements 103A-N. Insulator elements 102A-N are formed of a high-temperature thermal insulator material with low thermal conductivity. The high-temperature thermal insulator material can include low density carbides, metal-carbides, metal-oxides, or a combination thereof. More specifically, the high-temperature thermal insulator material includes low density SiC, stabilized zirconium oxide, aluminum oxide, low density ZrC, low density carbon, or a combination thereof. Moderator elements 103A-N are formed of a low-temperature solid-phase moderator. The low-temperature solid-phase moderator includes MgHx, YHx, ZrHx, CaHx, ZrOx, CaOx, BeOx, BeCx, Be, enriched boron carbide, 11B4C, CeHx, LiHx, or a combination thereof.
In an NTP, NEP, or FSP nuclear reactor system 100, the nuclear reactor 107 can include a plurality of control drums 115A-N and a reflector 140. The control drums 115A-N may laterally surround the insulator element array 112 of insulator elements 102A-N, the moderator element array 113 of moderator elements 103A-N, and nuclear fuel tile array 114 of nuclear fuel tiles 104A-N to change reactivity of the nuclear reactor core 101 by rotating the control drums 115A-N. As depicted, the control drums 115A-N reside on the perimeter or periphery of a pressure vessel 160 and are positioned circumferentially around the insulator elements 102A-N, moderator elements 103A-N, and nuclear fuel tiles 104A-N of the nuclear reactor core 101. Control drums 115A-N may be located in an area of the reflector 140, e.g., an outer reflector region formed of reflector blocks 141A-N immediately surrounding the nuclear reactor core 101, to selectively regulate the neutron population and nuclear reactor power level during operation. For example, the control drums 115A-N can be a cylindrical shape and formed of both a reflector material 116 (e.g., beryllium (Be), beryllium oxide (BeO), BeSiC, BeMgO, Al2O3, etc.) on a first outer surface and an absorber material 117 on a second outer surface.
The reflector material 116 and the absorber material 117 can be on opposing sides of the cylindrical shape, e.g., portions of an outer circumference, of the control drums 115A-N. The reflector material 116 can include a reflector substrate shaped as a cylinder or a truncated portion thereof. The absorber material 117 can include an absorber plate or an absorber coating. The absorber plate or the absorber coating are disposed on the reflector substrate to form the cylindrical shape of each of the control drums 115A-N. For example, the absorber plate or the absorber coating covers the reflector substrate formed of the reflector material to form the control drums 115A-N.
Rotating the depicted cylindrical-shaped control drums 115A-N changes proximity of the absorber material 117 (e.g., boron carbide, B4C) of the control drums 115A-N to the nuclear reactor core 101 to alter the amount of neutron reflection. When the reflector material 116 is inwards facing towards the nuclear reactor core 101 and the absorber material 117 is outwards facing, neutrons are scattered back (reflected) into the nuclear reactor core 101 to cause more fissions and increase reactivity of the nuclear reactor core 101. When the absorber material 117 is inwards facing towards the nuclear reactor core 101 and the reflector material 116 is outwards facing, neutrons are absorbed and further fissions are stopped to decrease reactivity of the nuclear reactor core 101. In a terrestrial land application, the nuclear reactor core 101 may include control rods 118A-N (see
Neutron reflector 140, e.g., shown as the outer reflector region, can be filler elements disposed between outermost nuclear fuel tiles 104A-N and the control drums 115A-N as well as around the control drums 115A-N. Reflector 140 can be formed of a moderator that is disposed between the outermost nuclear fuel tiles 104A-N and an optional barrel (e.g., formed of beryllium). The reflector 140 can include hexagonal or partially hexagonal shaped filler elements and can be formed of a neutron moderator (e.g., beryllium oxide, BeO). Although not required, nuclear reactor 107 can include the optional barrel (not shown) to surround the bundled collection that includes the insulator element array 112, moderator element array 113, nuclear fuel tile array 114 of the nuclear reactor core 101, as well as the reflector 140. As depicted, the control drums 115A-N reside on the perimeter of the pressure vessel 160 and can be interspersed or disposed within the reflector 140, e.g., surround a subset of the filler elements (e.g., reflector blocks 141A-N) forming the reflector 140.
Pressure vessel 160 can be formed of aluminum alloy, carbon-composite, titanium alloy, a radiation resilient SiC composite, nickel based alloys (e.g., Inconel™ or Haynes™), or a combination thereof. Pressure vessel 160 and nuclear reactor system 100 can be comprised of other components, including cylinders, piping, and storage tanks that transfer a moderator coolant that flows through moderator coolant passages 121A-N; and a separate nuclear fuel coolant, such as a propellant (e.g., hydrogen gas or liquid) that flows through the fuel coolant passages 141A-N. The moderator coolant and the nuclear fuel coolant can be a gas or a liquid, e.g., that transitions from a liquid to a gas state during a burn cycle of the nuclear reactor core 101 for thrust generation in an NTP nuclear reactor system 100. Hydrogen is for an NTP nuclear reactor system 100. In NEP or FSP applications, the nuclear reactor system 100 circulates a working fluid, such as He, neon, HeXe, CO2, instead.
In the example of
In an alternative implementation, a coolant that is shared between the moderator elements 103A-N and the nuclear fuel tiles 104A-N may be flowed through both the moderator coolant passages 121A-N and the fuel coolant passages 141A-N. In the alternative implementation, the coolant that flows through the plurality of fuel elements 104A-N can include helium, FLiBe molten salt formed of lithium fluoride (LiF), beryllium fluoride (BeF2), sodium, He, HeXe, CO2, neon, or HeN. The shared coolant flows through the moderator coolant passages 121A-N before the shared coolant is heated in the nuclear fuel tiles 104A-N. This keeps the moderator elements 103A-N cool.
Reflector 140 may be an integrally formed body (e.g. a tube or pipe), to surround the nuclear reactor core 101, or it may be several components or parts, such as a reflector region made of reflector blocks 141A-N, which surround the nuclear reactor core 101. Reflector 140 elastically scatters neutrons attempting to escape the nuclear reactor core 101, and redirects these neutrons back toward the nuclear reactor core 101 in order to create more fission events and therefore more energy. Reflector 140 also has a secondary purpose: by redirecting neutrons back toward the nuclear reactor core 101, the reflector 140 necessarily reduces the amount of neutrons impacting the pressure vessel 160 and areas beyond. Reflector 140 reduces the fast neutron fluence or fast neutron flux, the total length travelled by all free fast neutrons per time and volume outside the nuclear reactor core 101. Efficacy of the reflector 140 in both the electron scattering and the fast neutron fluence reduction is logarithmically improved by increasing the thickness of the reflector 140.
In a conventional nuclear reactor, the thickness of the reflector 140 is increased until the fast neutron fluence outside the reflector 140 is acceptably low: this thickness is often materially beyond the thickness required to efficiently reflect neutrons back toward the nuclear reactor core 101. In contrast, the nuclear reactor system 100 described herein additionally includes in-vessel shield 105, which can behave as a neutron poison, and is designed to absorb free neutrons, rather than reflect free neutrons back toward the nuclear reactor core 101. In-vessel shield 105 can absorb free neutrons to stop fast neutron fluence through the neutron poison. In-vessel shield 105, while being significantly thinner than the reflector 140, can nevertheless reduce the fast neutron fluence in an area by the same amount as the reflector 140. Although the in-vessel shield 105 can have the effect of neutron poisoning, the in-vessel shield 105 does not behave as a pure neutron poison because the neutron poison is composited within a moderator to form the in-vessel shield 105. By using a combination of materials with large neutron absorption cross-sections as in-vessel shield 105 between the pressure vessel 160 and the reflector 140 of the nuclear reactor 107, the fast neutron fluence can be reduced to acceptable levels.
Pressure vessel 160 includes the reflector 140 disposed therein, for the primary purpose of redirecting neutrons back to the nuclear reactor core 101, and secondarily for the purpose of reducing fast neutron fluence to an acceptable level at and beyond the pressure vessel 160. Pressure vessel 160 is then lined with in-vessel shield 105, with the primary purpose of reducing fast neutron fluence to an acceptable level at and beyond the pressure vessel 160. The in-vessel shield 105 is disposed between the pressure vessel 160 and the reflector 140. Together, the in-vessel shield 105 and the reflector 140 are able to reduce fast neutron fluence to an acceptable level at and beyond the pressure vessel 160, while the combined thickness of the in-vessel shield 105 and the reflector 140 is overall thinner than an equivalent reflector 140 without a paired in-vessel shield 105 reducing fast neutron fluence to the same acceptable or equivalent level. Overall, this allows for a combined thinner reflector 140 and in-vessel shield 105 than a reflector 140 alone, and a smaller pressure vessel 160 and nuclear reactor 107 with a reduced volume and mass.
Because of fast neutron fluence concerns, any size nuclear reactor system 100 (large or small) can benefit from the in-vessel shield 105. In a large nuclear reactor system 100, the fast neutron fluence to the pressure vessel 160 is limited by the distance of the active nuclear reactor core 101 from the pressure vessel 160. For a volume-constrained small nuclear reactor system 100, the use of in-vessel shield 105 can be very advantageous. In-vessel shield 105 reduces the radiation field around the nuclear reactor system 100, which in turn decreases the activation in the area outside (e.g., around) the nuclear reactor 107, decreases the footprint, and aids in ease of install.
During operation of the nuclear reactor system 100, the materials disposed within the pressure vessel 160 become activation products, made radioactive by the free neutrons. Nuclear reactor system 100 includes an inner density of activation product within the nuclear reactor core 101 and an outer density of activation product outside the nuclear reactor core 101. The outer density of activation product is lower than the inner density of activation product during operation of the nuclear reactor core 101. This is because the purpose of the reflector 140 is to redirect free neutrons back into the nuclear reactor core 101, increasing core activation. The in-vessel shield 105 is made of materials that are selected because the materials absorb free neutrons while becoming minimally radioactive: this further reduces activation outside the nuclear reactor core 101, in comparison to the activation within the nuclear reactor core 101.
In-vessel shield 105 is formed out of two or more neutron absorbing materials, at least one of which is a near black neutron absorbing material, and another is a gray neutron absorbing material. Near black neutron absorbing material is more efficient at absorbing free neutrons and reducing neutron fluence than a gray neutron absorbing material in similar quantities. Near black neutron absorbing material is a composited ceramic material, that is, a composited high-temperature neutron absorbing ceramic. For example, the composited ceramic material includes mixtures of neutron absorbing ceramics, such as B4C, HfC, and Gd2O3 with radiation tolerant materials, such as Al2O3 and SiC, to minimize radiation induced swelling. Adding radiation tolerant materials, such as Al2O3 and SiC, to form the composted ceramic material increases temperature and radiation endurance of the near black neutron absorbing material.
The gray neutron absorbing material includes a heavy metal. Together, the near black neutron material and the gray neutron absorbing material form an in-vessel shield 105. In-vessel shield 105 is fabricated through advanced manufacturing of the composited materials specifically engineered for shielding. This contrasts with the current paradigm of shield selection based on selection of the generic class of engineering alloys and structural ceramics, perhaps with some degree of isotopic enrichment. In-vessel shield 105 may be primarily made of composited ceramics, and therefore could also be described in such examples as a composited ceramic shield.
In the overall shielding thickness graph 200, white bars represent the in-vessel shield thickness 210A-K added by the in-vessel shield 105, while the black bars represent the reflector thickness 211A-K added by the reflector 140. Total thickness 206A-K is the sum of the in-vessel shield thickness 210A-K and the reflector thickness 211A-K. As shown, in a first implementation of nuclear reactor 107A, nuclear reactor 107A includes a total thickness 206A of 30 centimeters (cm), where the reflector thickness 211A is 30 cm and the in-vessel shield thickness 210K is 0 cm. In the eleventh implementation of nuclear reactor 107K, nuclear reactor 107K includes a total thickness 206K of 10 cm, where the reflector thickness 211K is 0 cm and the in-vessel shield thickness 210K is 10 cm.
Based on the total thicknesses 206A-K of the overall shielding thickness graph 200, in-vessel shield 105 has a shield diffusion length three times shorter than a reflector diffusion length of the reflector 140. Diffusion length is in part a factor of the diffusion coefficient of the material, as well as the macroscopic absorption cross section of that material. This diffusion length difference of three between the reflector 140 and in-vessel shield 105 is based on differences in the materials forming the in-vessel shield 105 and the reflector 140. Diffusion length difference may vary, e.g., based on materials selected, differences in the in-vessel shield thickness 210A-K, reflector thickness 211A-K, and overall geometry of the nuclear reactor core 101.
In
A second implementation of nuclear reactor 107B has 1 cm of in-vessel shield thickness 210B, which reduces total thickness 206B by 2 cm compared to the first implementation of the nuclear reactor 107A where no in-vessel shield 105 is present. Every 1 cm of in-vessel shield thickness 210 reduces the need for 3 cm of reflector thickness 211. Based on the overall shielding thickness graph 200, an optimal balance can be met: if only 15 cm reflector thickness 211F is needed to efficiently reflect free neutrons back toward the nuclear reactor core 101, then the sixth implementation of nuclear reactor 107F shows that only 5 cm in-vessel shield thickness 210F is needed to efficiently reduce fast neutron fluence to an acceptable level at and beyond the pressure vessel 160.
In a conventional nuclear reactor without the in-vessel shield 105, the pressure vessel 160 needs 30 cm of additional radius to properly reflect free neutrons back toward the nuclear reactor core 101 and to efficiently reduce fast neutron fluence to an acceptable level at and beyond the pressure vessel 160. Advantageously, the nuclear reactor system 100 with the in-vessel shield 105 only requires 20 cm of additional radius to properly reflect free neutrons back toward the nuclear reactor core 101 and to efficiently reduce fast neutron fluence to an acceptable level at and beyond the pressure vessel 160. An even further improvement in reduction in size and mass can be achieved if the nuclear reactor system 100 includes control rods 118A-N in a drum form factor to alternatively face reflector blocks 141A-N of the reflector 140 and moderator elements 103A-N of nuclear reactor core 101. In such a nuclear reactor system 100 with control rods 118A-N, the pressure vessel 160 does not necessarily require a fully efficient reflector 140 to reflect free neutrons back toward the nuclear reactor core 101.
Accordingly, overall shielding thickness graph 200 of
As explained in
Three Pareto fronts 411-413 of
Second, B4C—Al2O3 composite or B4C—SiC composites can be on the order of 50% B4C by weight and better accommodate the swelling of B4C. B4C—Al2O3 composite or B4C—SiC composites are less effective in shielding fast neutrons than pure B4C. Thus, B4C—Al2O3 composite or B4C—SiC composites are on the line of the medium Pareto front 412.
Third, borated stainless steel is less than 5% boron by weight, but is very dimensionally-stable under irradiation. Borated stainless steel is on the line of the long Pareto front 413. Based on the in-vessel shield material optimization chart 400, the two or more neutron absorbing materials to form the in-vessel shield 105 are selected, including a near black neutron absorbing material and a gray neutron absorbing material.
In block 510, the in-vessel shield material selection method 500 includes determining the lifetime 305 of the nuclear reactor system 100. Moving to block 520, the in-vessel shield material selection method 500 further includes determining a reflector thickness 211 that is efficient based on: (i) the nuclear reactor core 101; (ii) a reflector material utilized to form the reflector 140; (iii) geometry of the pressure vessel 160; or (iv) a combination thereof. Ultimately, a reflector thickness 211 is determined that most efficiently redirects neutrons back to the nuclear reactor core 101.
Continuing to block 530, the in-vessel shield material selection method 500 includes determining the reduction of in vessel fast neutron fluence at and beyond the pressure vessel 160 caused by the reflector 140. Any amount of fast neutron fluence reduction by the reflector 140 reduces the amount of fast neutron fluence reduction required by the in-vessel shield 105 to preserve the targeted reduction in pressure vessel 160 fast neutron fluence for the nuclear reactor system 100. In block 540, the in-vessel shield material selection method 500 includes determining a maximum pressure vessel radius of the pressure vessel 160 based on: (i) installation; (ii) transportation; and/or (iii) footprint requirements of the nuclear reactor system 100, e.g., a nuclear reactor core radius of the nuclear reactor core 101.
In block 550, the in-vessel shield material selection method 500 includes selecting the in-vessel shield material of the in-vessel shield. Referring to the in-vessel shield material optimization chart 400 of
With the reduction in the fast neutron fluence requirement of the pressure vessel 160 and the maximum thickness of the in-vessel shield 105, an in-vessel shield material on the proper Pareto front 411-413 is found with a diffusion length able to reduce fast neutron fluence sufficiently to meet the reduction in fast neutron fluence requirement of the pressure vessel 160 while no more than the maximum thickness of the in-vessel shield 105. The list of factors considered in the selection of the in-vessel shield material can further include: average and peak temperature of the nuclear reactor core 101; material and volume of the coolant; geometry of the nuclear reactor system 100; and the nuclear, thermal, and mechanical limits of the various in-vessel shield material to form the in-vessel shield 105.
Selecting the in-vessel shield material can include determining the effectiveness of in-vessel shield 105 based on a computational multi-physics computational model that combines thermal, mechanical, and time-dependent irradiation effects. The computational multi-physics model may be a tool with time-dependent neutron fluxes. The computational multi-physics model may also be used to inform the lifetime 305 and the geometric form of the in-vessel shield 105. Finishing in block 560, the in-vessel shield material selection method 500 includes accepting the in-vessel shield material cost. Once an in-vessel shield material from the in-vessel shield material optimization chart 400 is selected that satisfies the reduction in fast neutron fluence of the pressure vessel 160 on the appropriate Pareto fronts 411-413, an in-vessel shield material with the lowest cost 405 on the appropriate Pareto front 411-413 with sufficient reduction in fast neutron fluence 406 of the pressure vessel 160 is accepted for the nuclear reactor system 100.
Beginning in block 605, the in-vessel shield method 600 includes selecting a nuclear reactor system 100, which includes unique parameters of the nuclear reactor system 100, such as a fast neutron fluence to the pressure vessel 160, radius of the pressure vessel 160, reflector thickness 211, and other parameters relevant to design of the in-vessel shield 105. Nuclear reactor system 100 may include a small nuclear reactor 107, which is designed to produce up to approximately 500 thermal megawatts. A smaller nuclear reactor 107 may benefit more greatly from the space savings achieved by the in-vessel shield 105.
Moving to block 610, the in-vessel shield method 600 includes selecting two or more neutron absorbing materials. In addition to the material selection method 500 of
The two or more neutron absorbing materials include at least one near black neutron absorbing material and at least one gray neutron absorbing material. The selection process of block 610 is like that described in blocks 510, 520, 530, 540, 550, and 560 of the in-vessel shield material selection method 500 of
Composited ceramic material can include boron carbide (B4C), hafnium carbide (HfC), or gadolinium oxide (Gd2O3); each acting as strong neutron poisons, providing the near black neutron absorbing material potency in reducing neutron fluence. Composited ceramic material can further include aluminum oxide (Al2O3) or silicon carbide (SiC); these structural materials are hardy, inert, and have small neutron capture-cross sections: they physically support the strong neutron poisons while not being strong neutron poisons themselves, and may be categorized as neutron moderators. The near black neutron absorbing material selected may be an isotopically-tailored near black neutron absorbing material. Particular combinations of strong neutron poisons and structural materials for forming the composited ceramic material include: boron-10 carbide (10B4C), a boron-10 carbide and aluminum oxide composite (10B4C—Al2O3) of 50% boron-10 carbide by weight percent, a boron-10 carbide and silicon carbide composite (10B4C—SiC) of 50% boron-10 carbide by weight percent, or a borated stainless steel alloy of 5% boron-10 by weight percent. Boron-10 carbide and borated stainless steel are not composited ceramics alone, but must be paired with other ceramics in order to form a composited ceramic.
Heavy metal material of the gray neutron absorbing material includes tungsten (W), iron (Fe), Nickel (Ni), or copper (Cu), for example, in the form of atomized tungsten heavy metal. The heavy metal material, when in the form of atomized tungsten heavy metal can include atomized tungsten heavy metal with a tungsten content greater than or equal to 90% by weight percent.
Continuing to block 615, the in-vessel shield method 600 includes eutectic sintering base powders of the near black neutron absorbing material to fabricate a ceramic absorbing powder. Eutectic sintering the near black neutron absorbing material to fabricate a ceramic absorbing powder is performed with spark plasma sintering. For example, base powders to form the ceramic absorbing powder include an average particle size in the range of approximately 80 nm to 50 microns. The base powders are spark plasma sintered to form the ceramic absorbing powder. The ceramic absorbing powder includes a plurality of composited ceramic particles with an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns. Sintering in blocks 615 and 625 can include high-vacuum direct-current sintering. Sintering operations in some examples may be substituted for or include an advanced 3D printing process.
In block 620, the in-vessel shield method 600 further includes kinetically mixing the gray neutron absorbing material with the ceramic absorbing powder fabricated in block 615, creating an in-vessel shield mixture. This ensures thorough, impurity free dispersion.
Moving to block 625, the in-vessel shield method 600 includes cold press sintering the in-vessel shield mixture into an in-vessel shield 105. The cold press sintering the in-vessel shield mixture into the in-vessel shield 105 includes embedding the plurality of composited ceramic particles inside a heavy metal matrix of the heavy metal material to form the in-vessel shield 105. As noted above, composited ceramic particles include an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns.
The cold press sintering the in-vessel shield mixture into the in-vessel shield 105 further includes forming the in-vessel shield 105 as a plurality of in-vessel shield tiles 131A-N. All or a subset of the plurality of in-vessel shield tiles 131A-N are a curved polyhedron shape or a truncated portion thereof. The plurality of in-vessel shield tiles 131A-N can include a base shape with an interlocking geometry pattern. The base shape of the in-vessel shield tiles 131A-N can be selected based in part on: (i) tolerance towards swelling; (ii) reduction of neutron streaming paths; or (iii) a combination thereof.
Finishing in block 630, the in-vessel shield method 600 includes mounting the in-vessel shield 105 on an interior wall 133 of a pressure vessel 160, for example, by disposing the plurality of in-vessel shield tiles 131A-N on the interior wall 133 of the pressure vessel 160. Interior wall 133 is formed of a continuous surface or a plurality of discontinuous surfaces. Hence, block 630 can include covering the continuous or discontinuous surfaces of the interior wall 133 with the plurality of in-vessel shield tiles 131A-N by joining the interlocking geometry pattern.
The scope of protection is limited solely by the claims that now follow. That scope is intended and should be interpreted to be as broad as is consistent with the ordinary meaning of the language that is used in the claims when interpreted in light of this specification and the prosecution history that follows and to encompass all structural and functional equivalents. Notwithstanding, none of the claims are intended to embrace subject matter that fails to satisfy the requirement of Sections 101, 102, or 103 of the Patent Act, nor should they be interpreted in such a way. Any unintended embracement of such subject matter is hereby disclaimed.
It will be understood that the terms and expressions used herein have the ordinary meaning as is accorded to such terms and expressions with respect to their corresponding respective areas of inquiry and study except where specific meanings have otherwise been set forth herein. Relational terms such as first and second and the like may be used solely to distinguish one entity or action from another without necessarily requiring or implying any actual such relationship or order between such entities or actions. The terms “comprises,” “comprising,” “includes,” “including,” “has,” “having,” “with,” “formed of,” or any other variation thereof, are intended to cover a non-exclusive inclusion, such that a process, method, article, or apparatus that comprises or includes a list of elements or steps does not include only those elements or steps but may include other elements or steps not expressly listed or inherent to such process, method, article, or apparatus. An element preceded by “a” or “an” does not, without further constraints, preclude the existence of additional identical elements in the process, method, article, or apparatus that comprises the element.
In addition, in the foregoing Detailed Description, it can be seen that various features are grouped together in various examples for the purpose of streamlining the disclosure. This method of disclosure is not to be interpreted as reflecting an intention that the claimed examples require more features than are expressly recited in each claim. Rather, as the following claims reflect, the subject matter to be protected lies in less than all features of any single disclosed example. Thus, the following claims are hereby incorporated into the Detailed Description, with each claim standing on its own as a separately claimed subject matter.
While the foregoing has described what are considered to be the best mode and/or other examples, it is understood that various modifications may be made therein and that the subject matter disclosed herein may be implemented in various forms and examples, and that they may be applied in numerous applications, only some of which have been described herein. It is intended by the following claims to claim any and all modifications and variations that fall within the true scope of the present concepts.
This application claims priority to U.S. Provisional Pat. Application No. 62/910,561, filed on Oct. 4, 2019, titled “Nuclear System for Power Production in Space,” the entirety of which is incorporated by reference herein. This application relates to International Application No. PCT/US2020/XXXXXX, filed on Oct. 4, 2020, titled “Nuclear Reactor Core Architecture with Enhanced Heat Transfer and Safety,” the entirety of which is incorporated by reference herein. This application also relates to International Application No. PCT/US2020/XXXXXX, filed on Oct. 4, 2020, titled “Automatic Shutdown Controller for Nuclear Reactor System with Control Drums,” the entirety of which is incorporated by reference herein.
| Filing Document | Filing Date | Country | Kind |
|---|---|---|---|
| PCT/US2020/054188 | 10/4/2020 | WO |
| Number | Date | Country | |
|---|---|---|---|
| 62910561 | Oct 2019 | US |
