Patent Description:
Radioisotope power systems (RPS) are devices that produce power by harnessing heat from decay of a concentrated mass of radioactive material. Radioisotope heater units (RHU) use this heat directly. Radioisotope thermoelectric generators (RTGs) convert heat produced from radioactive decay into electricity. RTGs are desirable as a compact, always-on, no maintenance source of power without need for refueling over periods of months to decades. The utility of an RPS depends on the availability and characteristics of radioisotopes used to fuel it. In general, the characteristics of plutonium-<NUM> (<NUM>Pu) are favorable for use as a radioisotope for RPS devices. However, plutonium-<NUM> has a very limited supply, making plutonium-<NUM> impractical for most applications. Alternatives radioisotopes for RPS devices include beta emitting radioisotopes (e.g., strontium-<NUM> (<NUM>Sr)), which have been used for RPSs in the past, but which are not optimal because of the need for heavy radiation shielding. RTGs are established devices used since the <NUM> in applications where their specialized capabilities warranted their cost premium, such as for military purposes, space probes, etc. Previous versions of RTGs were used in scenarios (e.g., space probes) where minimal or no maintenance or human intervention was needed. These RTGs generally produced a few hundred watts or less of power, operated for relatively short durations of time, and/or were designs customized for a specific use. The fuel form and configuration of previous versions of RTGs were determined based on the radioisotope to be used and the application of the RTG. Many past RTG designs consisted of fuel capsules containing large diameter discs of Strontium Titanate (SrTiO<NUM>) surrounded by large amounts of dense metals (e.g., lead (Pb)) or concrete.

Strontium-<NUM> decays by beta emission into yttrium-<NUM> (<NUM>Y), which itself is a beta radiation source. The energy of the two primary beta particles produced by the <NUM>Sr decay chain are <NUM> keV and <NUM>,<NUM> keV. The initial strontium-<NUM> half-life is <NUM> years with the yttrium-<NUM> decay being only <NUM> hours. Yttrium-<NUM> decays into stable zirconium-<NUM> (<NUM>Z). Beta particles (e.g., electrons or positrons), when slowed by interaction with matter through which they are moving, can produce bremsstrahlung radiation (x-rays). Bremsstrahlung radiation is a process whereby a charged particle, such as an electron, is decelerated, causing electromagnetic radiation (i.e., a photon) to be produced. To satisfy the law of conservation of energy, the energy of the photon will equal the kinetic energy of the charged particle prior to being decelerated minus the charged particle's energy after being decelerated (e.g., Eγ = E(i)e- - E(f)e-). Interaction with higher atomic number (i.e., Z) materials (e.g., incident with) will cause a larger deceleration, which will increase the average energy of the x-rays. This indicates that lower Z materials are preferable for shielding these particles due to the decreased x-ray production. Like beta particles, the resulting bremsstrahlung x-rays technically have a range of energies up to a maximum energy equal to the maximum energy of the beta particle (assuming that the beta particle is completely stopped in the material). The x-ray energy and intensity can be high enough to require a considerable radiation shield layer to protect humans and the local environment, including nearby equipment.

Historical designs of <NUM>Sr RTGs performed well for their given tasks. However, the heavy shielding of these RTGs limited their practical applications. Furthermore, previous fabrication processes focused on hot pressing the strontium-based material (or other isotope) alone. Additionally, previous fabrication techniques included obtaining the fuel after it had been separated and purified into larger chunks. Fuel pellets were then created by powdering the asymmetric chunks and hot pressing the powder. While uniformity in the powder was desired, it was not required and the fabrication process could be optimized by creating the fuel pellet from the powder and utilizing any remaining non-uniformity. Still further, previous fabrication processes were not concerned with using graphite in the creation of the final fuel material. These and/or other drawbacks exist. <CIT> describes a high pressure resistant compact housing structure which is particularly useful in connection with thermoelectric radioisotope generator constructions. The housing is particularly constructed for use at high external pressures as at great depths in the ocean and is resistant to galvanic corrosion and high pressure. An outer housing shell transfers high pressures to interior structure which absorbs substantially the entire pressure load. Preferably a radioisotope heat source is enclosed in the housing. Heat at both ends of the source is utilized to obtain maximized efficiency of the source in thermoelectric conversion. A novel mounting structure for the heat source and associated thermoelectric conversion means comprises utility mounting plates and associated radiation shielding preferably incorporating a novel double seal arrangement in accordance with this invention. The construction permits low power-weight ratios in high pressure resistant radioisotope thermoelectric generators.

The following is a non-exhaustive listing of some aspects of the present techniques. These and other aspects are described in the following disclosure.

Some aspects include a method of fabricating a heat source for a radioisotope thermoelectric generator (RTG), comprising: reducing a particle size in a strontium compound by powdering and sieving the strontium compound and/or dissolving the strontium compound into an aqueous solution; mixing the strontium compound with graphite to obtain a strontium-graphite mixture; performing a press to the strontium-graphite mixture; and encapsulating the pressed strontium-graphite mixture into an x-ray shielding to obtain the heat source.

The above-mentioned aspects and other aspects of the present techniques will be better understood when the present application is read in view of the following figures in which like numbers indicate similar or identical elements:.

While the present techniques are susceptible to various modifications and alternative forms, specific embodiments thereof are shown by way of example in the drawings and will herein be described in detail. The drawings may not be to scale. It should be understood, however, that the drawings and detailed description thereto are not intended to limit the present techniques to the particular form disclosed, but to the contrary, the intention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the present techniques as defined by the appended claims.

To mitigate the problems described herein, the inventors had to both invent solutions and, in some cases just as importantly, recognize problems overlooked (or not foreseen) by others in the field of radioisotope thermoelectric generators (RTGs) and fabrication techniques for manufacturing radioactive source materials for RTGs. Indeed, the inventors wish to emphasize the difficulty of recognizing those problems. Further, because multiple problems are addressed, it should be understood that some embodiments are problem-specific, and not all embodiments address every problem with traditional systems described herein or provide every benefit described herein. That said, improvements that solve one or more permutations of these problems and/or other problem(s) are described below.

Power sources including high-energy beta emitting sources may be used to create radioisotope thermoelectric generators (RTGs). RTGs convert heat to electricity, which may then be output from the RTG to power various devices. Such devices may include, for example, batteries, satellites, unmanned facilities, solar panels, communications devices, lighting, motors, etc. The heat may be produced by the decay of a radioactive source material or combinations of different radioactive source materials. The decay process, for example, may include a radioactive element decaying to another element while outputting one or more particles. A radioactive element, as described herein, refers to an element that includes an unstable nuclei, where a number of protons and a number of neutrons in the nucleus is unbalanced. One type of radioactive decay is a beta decay, where an initially unstable atomic element decays to another element (stable or unstable) while outputting an electron or a position. A half-life, τ, of a radioactive element indicates an amount of time for the unstable atomic element to decay to half of its initial value. Each radioactive isotope may have a different half-life, and these half-lives are generally well-known in scientific communities.

In some embodiments, an RTG may be formed that includes strontium-<NUM> as its radioactive element. Strontium-<NUM> includes <NUM> protons and <NUM> neutrons (e.g., <NUM> + <NUM> = <NUM>). Strontium-<NUM> is an unstable strontium (Sr) isotope, having a half-life of <NUM> years. Strontium-<NUM> decays to yttrium-<NUM> via a beta decay, emitting a <NUM> keV electron. Yttrium-<NUM> includes <NUM> protons and <NUM> neutrons (e.g., <NUM> + <NUM> = <NUM>). Yttrium-<NUM> is an unstable yttrium (Y) isotope, having a half-life of <NUM> hours. Yttrium-<NUM> decays to zirconium-<NUM> via a beta decay, emitting a <NUM>,<NUM> keV (<NUM> MeV) electron. While the foregoing description relates generally to radioactive source materials including strontium-<NUM>, alternative RTGs may be formed including other radioactive elements that produce high-energy beta radiation (e.g., beta radiation greater than <NUM> MeV). For example, strontium-<NUM>, plutonium-<NUM>, polonium-<NUM>, curium-<NUM>, or americium-<NUM> are alternative radioisotopes that could be used instead of strontium-<NUM>.

RTGs, due to the radioactivity of the materials used, require some form of shielding for safety. Generally speaking, the goal of a shielding design, regardless of use-case, is to reduce the escape of radiation by interposing a material that interacts with that type of radiation to reduce the radiation's energy. One issue with high-energy beta decay is that high energy x-rays result from the initial interaction(s) with shielding materials, and these x-rays can require considerable shielding in turn. Historical designs for shielding resulted in bremsstrahlung being produced before the beta radiation could escape the fuel (e.g., radioactive material) itself. However, fuel designs that reduce/eliminate bremsstrahlung radiation generated within the fuel itself have yet to be fabricated, particularly for larger activity energy sources. The historical fuel designs where bremsstrahlung radiation is produced within the fuel itself have only been possible by limiting the fuel to small amounts of strontium-<NUM> and, even so, surrounding them with unwieldy shields. For instance, current strontium-<NUM> radioactive power source designs use large-radius fuel designs even though it has been known since at least <NUM> where virtually all bremsstrahlung x-rays are generated inside the fuel. Therefore, a shielding design that reduced bremsstrahlung radiation generation for large activity beta radiation sources has not been pursued because the x-ray generation had already occurred in the fuel. Consequently, to shield high-energy beta radiation sources and their associated bremsstrahlung x-ray generation required using large masses of heavy metals or concrete that severely constrained the applications (e.g., use cases) of such beta emitting RTGs.

Due to the need for large masses of metals or concrete for shielding, lightweight shielding of high-energy beta radiation sources was not practical for portable, terrestrial, or space use. This has limited the use of beta source RTGs to applications where either higher allowable radiation levels permit use of less shielding or where the large shield mass, volume, and cost are acceptable. A shielding design intended for portable use-cases outside of a remote environment requires the inherent safety mechanisms afforded by large masses of metals or concrete, while also being lightweight enough to be portable. Now, a confluence of factors warrants reassessing use of strontium-<NUM> RTGs. Namely, a lack of alternative radioisotopes where strontium-<NUM> would be in plentiful supply if nuclear fuel is reprocessed and the need for small amounts of unattended, reliable, long-term power to highly energy efficient devices being deployed where humans cannot go. One of the main driving forces in propelling the widespread use of RTGs is reducing the x-ray shield cost and weight.

Additionally, manufacturing the fuel material for RTGs, such as those described above, requires development of unique set fabrication steps to ensure that the fuel material will have the appropriate density, size, and uniformity. In particular, the inclusion of graphite in the fuel material requires identifying optimized pressing parameters, as the graphite drives the pressing process, not the fuel as was the case in previous fabrication processes. Furthermore, the parameters for pressing the powdered material are optimized around the fuel design and bremsstrahlung generation to ensure adequate protection from radiation and appropriate power output levels.

Previous techniques for fabricating fuel sources had no need to include graphite, so the fabrication process was not concerned with how including graphite could or would alter the fabrication steps and parameters. For instance, if additional elements were included that increased the volume of the fuel, the volume of heavy metal shielding would subsequently be increased. On the other hand, the inclusion of graphite serves an important function - to minimize the bremsstrahlung radiation. Additionally, previous fuel fabrication processes did not impose the uniformity requirements of the powdered radioactive materials that are required for the fuel sources described herein. For instance, the hot pressing for the fuel sources described herein may use a die that is the shielding itself. Furthermore, the fabrication process described herein utilizes a hot cell because, during processing, the fuel may be exposed to or placed near metal that is not graphite until it is in its final form, which can cause very high bremsstrahlung radiation to be generated and, subsequently, produces high radiation exposure during processing.

<FIG> illustrates an example of a geometry <NUM> of a historical fuel design for a radioisotope thermoelectric generator (RTG) and a geometry <NUM> of an optimized fuel design for an RTG, in accordance with various embodiments. In some embodiments, geometry <NUM> for an optimized fuel design of an RTG may be crafted such that bremsstrahlung radiation generation may be reduced as compared to geometry <NUM> for the historical fuel design. For instance, geometry <NUM> of the optimized fuel design may reduce bremsstrahlung x-ray generation by allowing most high-energy beta radiation to escape from a fuel source. The fuel source of the optimized fuel design, for example, may be a high-energy beta emitter, such as strontium-<NUM>.

As seen in <FIG>, geometry <NUM> of the historical fuel design may have a radius that is greater than a mean free path <NUM> of the fuel source's beta radiation. In this example, the electron <NUM> produced via the beta decay of the radioactive source material is converted to bremsstrahlung radiation <NUM> (e.g., an x-ray) within the fuel source <NUM> (e.g., the fuel source of the historical fuel design). However, in some embodiments, geometry <NUM> of the optimized fuel design may have a radius that is less than the mean free path of the beta radiation, thereby allowing the beta radiation to escape from the fuel source <NUM> (e.g., the fuel source of optimized fuel design) without first being converted to bremsstrahlung x-rays. After escaping the fuel source (e.g., the fuel source of optimized fuel design), the beta radiation may be decelerated within a material, such as a portion of a shielding, which may inhibit bremsstrahlung radiation generation. The material may be selected based on atomic number (i.e., Z), or effective atomic number, if a compound material. In some cases, the material may be a material with a lowdensity.

<FIG> illustrates an example of a two-phase shielding <NUM> for an RTG <NUM>, in accordance with various embodiments. In some embodiments, RTG <NUM> may include a fuel <NUM>, also referred to herein interchangeably as fuel source <NUM>, a fuel container <NUM>, and a two-phase shielding system <NUM>, also referred to herein interchangeably as two-phase shielding <NUM>, shielding <NUM>, and/or system <NUM>. Two-phase shielding system <NUM> may include a first shield <NUM> and a second shield <NUM>. In some embodiments, first shield <NUM> may be referred to herein interchangeably as primary shield <NUM>, and second shield <NUM> may be referred to herein interchangeably as secondary shield <NUM>. Additionally, in some embodiments, RTG <NUM> may not include fuel container <NUM>.

In some embodiments, two-phase shielding <NUM> may be designed to reduce bremsstrahlung radiation generation, minimize radiation outside RTG <NUM>, and also provide structural stability for RTG <NUM>. In some embodiments, first shield <NUM> may be formed of a material having a low effective atomic number. First shield <NUM> may be configured to attenuate at least some of the bremsstrahlung radiation that escapes fuel <NUM> and fuel container <NUM> (e.g., as seen in geometry <NUM> of <FIG>). In some embodiments, second shield <NUM> may be formed of a material having a high effective atomic number. Second shield <NUM> may be configured to absorb bremsstrahlung radiation that escapes first shield <NUM> and/or is generated within first shield <NUM>. Second shield <NUM> may also be configured to provide structural stability and ruggedness for two-phase shielding <NUM>, thus also providing stability and ruggedness for RTG <NUM>. In some embodiments, RTG <NUM> may be configured to operate at very high temperatures to maximize thermal efficiency. For example, RTG <NUM> may be configured to operate at, or above, approximately <NUM> degrees Celsius. The temperature ranges for which RTG <NUM> may operate may range from approximately <NUM> degrees Celsius to the degradation temperature of the fuel (e.g., approximately <NUM> degrees Celsius for SrF<NUM>, <NUM> degrees Celsius for SrTiO<NUM>). In some embodiments, RTG <NUM> may be configured to operate at low temperatures to simplify the thermal design and handling. Regardless, the geometry <NUM> can support a wide range of possible operating temperatures to best suit the application. In some embodiments, first shield <NUM> may serve to attenuate as much bremsstrahlung radiation as possible that escaped from fuel <NUM>, or fuel <NUM> and fuel container <NUM>. Additionally, first shield <NUM> may be designed to produce a minimal amount of bremsstrahlung radiation. In some embodiments, first shield <NUM> must be configured to act as a heat transfer medium to transfer heat between a heat source and a heat sink. First shield <NUM> may be formed of one or more materials. For example, first shield <NUM> may be formed of graphite, lithium hydride, hydrogenous oils or resins, molten salts, and the like. The one or more materials used to form first shield <NUM> may include materials having a low density, a low atomic number, high thermal conductivity, and/or high material degradation temperature, and may additionally or alternatively be a material that is compatible with a material (e.g., strontium titanate) including the radiation source to be used (e.g., strontium-<NUM>).

In some embodiments, graphite may be selected as a material to be used for first shield <NUM> because graphite has a good thermal conductivity and a low effective atomic number of <NUM>. Additionally, graphite has high chemical compatibility with other materials and has been used as a material compatibility buffer in previous RTG designs. The melting temperature of graphite is higher (e.g., <NUM>,<NUM> degrees Celsius) than a melting temperature of strontium titanate (e.g., <NUM>,<NUM> degrees Celsius), making it thermally stable. Graphite is also not toxic, relatively inexpensive, and is used for many existing nuclear applications.

In some embodiments, lithium hydride or hydrogenous oils may be selected as a material to be used for first shield <NUM> because of their low densities, acceptable thermal conductivity, and low effective atomic numbers. For example, lithium hydride has an effective atomic number of <NUM>, while hydrogenous oil will be generally higher and dependent on the specific oil used. Both substances may be compatible with the other materials in the fuel design, though lithium hydride is toxic and chemically reactive, which may increase manufacturing cost and increase risks in the event of a loss of containment. While hydrogenous oils are relatively inexpensive, lithium hydride is not. In some cases, the maximum temperature before degradation or boiling of both lithium hydride and hydrogenous oils can limit the operating temperature of the fuel design. For example, lithium hydride and hydrogenous oils can cause the operating temperature of RTG <NUM> to be below <NUM> degrees Celsius.

In some embodiments, a molten salt option, such as FLiBe (Li<NUM>BeF<NUM>), may be selected as a material to be used for first shield <NUM>. The effective atomic number of the molten salt option is <NUM>, representing a bremsstrahlung intensity reduction of approximately <NUM>% compared to past strontium-<NUM> designs. The density of the molten salt option is approximately <NUM>/cm<NUM>. The molten salt option has an acceptable thermal conductivity and can be made relatively inexpensively. Molten salts can be used as high temperature heat transfer medium for certain applications, such as, for example, central solar power stations, and has been planned for use in nuclear reactors.

In some embodiments, a refractory compound such as boron nitride (BN) or boron carbide (B<NUM>C) may be selected as a material to be used for first shield <NUM>. The effective atomic number of the refractory compounds is similar to graphite, representing in a bremsstrahlung intensity reduction similar to that of using graphite. The density of the refractory compound option is approximately <NUM> - <NUM>/cm<NUM>. The refractory compound option has an acceptable thermal conductivity and can be made relatively inexpensively. Refractory compounds have high thermal and chemical stability and high hardness to resist damage from impacts.

In some embodiments, a thickness and material for second shield <NUM> are chosen to reduce remaining beta rays that escape first shield <NUM> to low enough values to allow handling and use. Second shield <NUM> may be configured to stop or absorb most bremsstrahlung radiation that may have been produced by fuel <NUM>, fuel container <NUM>, and/or primary shield <NUM> that was not attenuated by fuel container <NUM> or first shield <NUM>. Second shield <NUM> may be formed of one or more materials, which may differ from the one or more materials of first shield <NUM>. For instance, the one or more materials used to form second shield <NUM> may include materials having a high density, a high effective atomic number, high thermal conductivity, high material degradation temperature, as well as being a material that is compatible with other materials in RTG <NUM>. As an example, second shield <NUM> may be formed of tungsten, lead, depleted uranium, a combination thereof, or other materials. In some embodiments, combinations of tungsten, lead, and/or depleted uranium may be used to control heat transfer.

In some embodiments, an electron <NUM> produced as a result of a radioactive decay (e.g., a beta decay) of a radioactive substance (e.g., strontium-<NUM>) within fuel <NUM> may escape fuel <NUM> and be attenuated in fuel container <NUM>. For instance, a radius of fuel <NUM> may be less than a mean free path of electron <NUM>, allowing electron <NUM> to escape fuel <NUM> without being converted into bremsstrahlung radiation. In some embodiments, an electron <NUM> produced as a result of a radioactive decay (e.g., a beta decay) of a radioactive substance (e.g., strontium-<NUM>) within fuel <NUM> may escape fuel <NUM>, pass through fuel container <NUM>, and may be attenuated in first shield <NUM>. In some embodiments, an electron <NUM> produced as a result of a radioactive decay (e.g., a beta decay) of a radioactive substance (e.g., strontium-<NUM>) within fuel <NUM> may not escape fuel <NUM>, and may be converted into bremsstrahlung radiation (e.g., an x-ray) <NUM>. Bremsstrahlung radiation <NUM> may escape fuel <NUM>, pass through fuel container <NUM> and first shield <NUM>, and may be attenuated by second shield <NUM>. In some embodiments, an electron <NUM> produced as a result of a radioactive decay (e.g., a beta decay) of a radioactive substance (e.g., strontium-<NUM>) within fuel <NUM> may escape fuel <NUM>, pass through fuel container <NUM> and first shield <NUM>, and may be attenuated by second shield <NUM>. Alternatively, an electron produced as a result of a radioactive decay (e.g., a beta decay) of a radioactive substance (e.g., strontium-<NUM>) within fuel <NUM> may escape fuel <NUM>, and be converted to bremsstrahlung radiation within fuel container <NUM>, first shield <NUM>, or second shield <NUM>. Regardless of whether the bremsstrahlung radiation was produced within fuel container <NUM>, first shield <NUM>, or second shield <NUM>, the bremsstrahlung radiation may be attenuated prior to arriving at, or by, second shield <NUM> so as to not escape RTG <NUM>.

Many nuclear fuel designs may be contained in multiple levels of containment structures as an inherent safety element. However, some prior designs for fuel capsules included a single layer of containment. Some embodiments include a graduated containment structure to increase safety of the fuel design by reducing radiation exposure and loss of fuel containment if the shield is punctured or the containment is lost in some other way. Additionally, the graduated containment structure may provide lightweight attenuation for any bremsstrahlung radiation that is generated. In some embodiments, the graduated shielding may include geometric combinations making use of self-shielding, sectioning off of fuel into separate compartments, containing fuel such that the spread of fuel is limited if the shielding is impacted, and geometric designs that reduce radiation exposure under off-normal conditions.

In some embodiments, fuel designs are described herein for a high-energy beta radiation emitter, such as strontium-<NUM>. As an example, the range (e.g., a mean-free path) of electrons in strontium titanate (SrTiO<NUM>) may be approximately <NUM>. An electron produced within a strontium titanate source therefore could be stopped in approximately <NUM> of strontium titanate source material.

<FIG> illustrates an example of a dispersed fuel design <NUM> for an RTG, in accordance with various embodiments. In some embodiments, fuel design <NUM> may include dispersed spherical structures <NUM> formed of the high-energy beta radiation emitter. The spherical structures may be (i) of a substantially same size and (ii) substantially spherical. For instance, a volume V of each sphere may be equal to <NUM>/3πr<NUM>, where "r" corresponds to a radius of a given spherical structure. Each spherical structure's volume may be within a predetermined tolerance level. For example, each spherical structure may have a volume V that within N standard deviations σ of an average spherical structure volume, where, for example, N can be any number such as <NUM>, <NUM>, etc. Spherical structures having a volume greater than V + Nσ or less than V - Nσ may be excluded from fuel design <NUM>. Furthermore, each spherical structure may have a radius, r, that is substantially constant throughout its volume.

In some embodiments, each spherical structure may have a radius that is equal to or less than <NUM>, corresponding to a range of electrons produced by a beta decay of strontium-<NUM>. In some embodiments, each spherical structure may be formed of strontium titanate and may be coated with graphite <NUM>. In some embodiments, the spherical structures may be dispersed in the primary shield (e.g., first shield <NUM>) and may also be sintered or hot pressed together and encased in the secondary shield (e.g., second shield <NUM>). In other embodiments, the dispersed structures may be other shapes such as polyhedral or irregular, but having a characteristic size, surface area, or volume commensurate with the exemplary spherical structures as described herein.

<FIG> illustrates an example of a concentric fuel design <NUM> for an RTG, in accordance with various embodiments. In some embodiments, concentric fuel design <NUM> may include concentric cylinders of a high-energy beta radiation emitter <NUM> having a thickness less than the mean free path of the high-energy beta radiation emitter. For example, concentric fuel design <NUM> may include concentric cylinders of strontium titanate having a thickness of <NUM>. In some embodiments, concentric fuel design <NUM> may further include concentric cylinders of a first shielding material <NUM>, such as graphite. For example, concentric fuel design <NUM> may include a first cylinder of strontium titanate having a first radius, encased by a cylinder of graphite having a thickness, encased by another cylinder of strontium titanate, encased by another cylinder of graphite, and so on. In some embodiments, the first radius may be less than <NUM> (e.g., the mean free path of electrons emitted by strontium titanate). In some embodiments, the thickness of the cylinder of graphite may be <NUM> or less. In some embodiments, an average distance between particles of strontium-<NUM> may be approximately <NUM>. The number of cylinders of the radioactive source material and the number of cylinders of the first shielding material may vary and may depend on the power output a given fuel design is to produce. For example, the number of cylinders may be <NUM> or more (e.g., <NUM> cylinders of strontium titanate and <NUM> cylinders of graphite), <NUM> or more (e.g., <NUM> cylinders of strontium titanate and <NUM> cylinders of graphite), etc. In some embodiments, the concentric cylinders of the high-energy beta radiation emitter (e.g., strontium titanate) and the first shielding material (e.g., graphite) may be encased in a second shielding material <NUM>, such as tungsten. The thickness may be selected such that radiation exposure from the RTG will be less than a threshold amount of radiation exposure (e.g., less than <NUM> mrem/hr at <NUM> meter away).

<FIG> illustrates another example of a concentric fuel design <NUM> for an RTG, in accordance with various embodiments. <FIG> illustrate a side view and a top view, respectively, of an example of a concentric fuel design for RTG <NUM>, in accordance with various embodiments. In some embodiments, concentric fuel design <NUM> may be similar to concentric fuel design <NUM> of <FIG>, and the previous description may apply. Concentric fuel design <NUM> may include first concentric cylinders <NUM> of a high-energy beta emitter, such a strontium-<NUM>. For example, first concentric cylinders <NUM> may be formed of strontium titanate. First concentric cylinders <NUM> may be interposed with second concentric cylinders <NUM>. Second concentric cylinders <NUM> may, in some embodiments, form a first, or primary, shield. For example, second concentric cylinders <NUM> may be similar to first shield <NUM> of <FIG>. In some embodiments, second concentric cylinders <NUM> may be formed of one or more materials having a low density, a low atomic number, high thermal conductivity, high material degradation temperature, as well as being a material that is compatible with the radiation source to be used (e.g., strontium-<NUM>). For example, second concentric cylinders <NUM> may be formed of graphite.

In some embodiments, a central cylinder of first concentric cylinders <NUM> may be surrounded by a first instance of second concentric cylinders <NUM>. The first instance of second concentric cylinder <NUM> may be sandwiched between a first instance of first concentric cylinder <NUM>. This alternating pattern of first concentric cylinders <NUM> and second concentric cylinders <NUM> may be repeated a number of times depending on the power requirements of the RTG. However, an outer cylinder of fuel design <NUM> will be formed of second concentric cylinder <NUM>. In some embodiments, a thickness of each concentric cylinder <NUM> and <NUM> may be substantially similar. For example, concentric cylinders <NUM> and <NUM> may both have a thickness that is equal to or less than <NUM>, <NUM>, <NUM>, <NUM>, <NUM>, etc. As another example, concentric cylinders <NUM> and <NUM> may both have a same thickness in a range of <NUM>-<NUM>. In some embodiments, a thickness of each concentric cylinders <NUM> may be substantially similar to one another, and a thickness of each concentric cylinder <NUM> may be substantially similar to one another but different than that of concentric cylinders <NUM>. For example, concentric cylinders <NUM> and <NUM> may each have a thickness that is equal to or less than one of <NUM>, <NUM>, <NUM>, <NUM>, <NUM>, etc. As another example, concentric cylinders <NUM> and <NUM> may each have a thickness in a range of <NUM>-<NUM>. In some embodiments, a radius of each concentric cylinder <NUM> and each concentric cylinder <NUM> may increase as the number of cylinders increases (e.g., an inner cylinder may have a smaller radius than an outer cylinder). For example, a first concentric cylinder <NUM> may have a first radius, and first concentric cylinder <NUM> may be surrounded by a first concentric cylinder <NUM> having a second radius that is greater than the first radius. First concentric cylinder <NUM> may be surrounded by a second concentric cylinder <NUM> having a third radius that is greater than the second radius, and so on. The radius may gradually increase, for example, starting at <NUM>. The second radius may be one selected in a range of <NUM>-<NUM>, <NUM>-<NUM>, <NUM>-<NUM>, etc., and the third radius may be selected in a range of <NUM>-<NUM>, <NUM>-<NUM>, <NUM>-<NUM>, etc..

In some embodiments, third cylinder <NUM> may surround concentric cylinders <NUM> and <NUM>. Third cylinder <NUM> may be substantially similar to second shield <NUM> of <FIG>, and the previous description may apply. In some embodiments, third cylinder <NUM> may be formed of one or more materials with high density, high effective atomic number, high thermal conductivity, high material degradation temperature, as well as being a material that is compatible with other materials in the RTG. For example, third cylinder <NUM> may be formed of tungsten. In some embodiments, a thickness of third cylinder <NUM> may be dependent on design requirements of the RTG (e.g., power requirements, number of instances of first shield, material composition, etc.). In some embodiments, third cylinder <NUM> may be encompassed by a layer <NUM> of air, or may be exposed to the natural environment.

<FIG> illustrates another example of a dispersed fuel design for an RTG <NUM>, in accordance with various embodiments. <FIG> illustrate a side view and a top view, respectively, of an example of a dispersed fuel design for RTG <NUM>, in accordance with various embodiments RTG <NUM> may include dispersed spherical structures <NUM> formed of the high-energy beta radiation emitter. The spherical structures may be (i) of a substantially similar size and (ii) substantially spherical. For example, the spherical structures may be the same or similar to the spherical structures described above with references to <FIG>. In some embodiments, each spherical structure may have a radius that is equal to or less than <NUM>, corresponding to a range of electrons produced by a beta decay of strontium-<NUM>. In some embodiments, each spherical structure may be formed of strontium titanate and may be coated with graphite. In some embodiments, the spherical structures may be dispersed in a first shield <NUM>. For example, first shield <NUM> may be formed of one or more materials having a low density, a low atomic number, high thermal conductivity, and/or high material degradation temperature, and may additionally or alternatively be a material that is compatible with a material (e.g., strontium titanate) including the radiation source to be used (e.g., strontium-<NUM>). As an example, first shield <NUM> may be formed of graphite. In some embodiments, first shield <NUM> may include a buffer layer between a portion including spherical structures <NUM>, and second shields <NUM> and <NUM>. The purpose of this buffer layer is to prevent any spherical structures <NUM> from contacting the second shields <NUM> and <NUM> and generating high-intensity bremsstrahlung.

Second shields <NUM> and <NUM> may each be formed of one or more materials having a high density, a high effective atomic number, high thermal conductivity, high material degradation temperature, as well as being a material that is compatible with other materials in RTG <NUM>. As an example, second shields <NUM> and <NUM> may be formed of tungsten. In some embodiment, second shield <NUM> may have a first thickness and second shield <NUM> may have a second thickness. For example, the thickness of second shield <NUM> may be <NUM> and the thickness of second shield <NUM> may be <NUM>. Each of second shields <NUM> and <NUM> may have a thickness selected from within a range of thickness. The thickness is selected such that radiation exposure from the RTG will be less than a threshold amount of radiation exposure. For example, the thickness of the second shield may be selected such that the radiation exposure is less than <NUM> mrem/hr at <NUM> meter away. In some embodiments, second shield <NUM> may be substantially constant in thickness. For example, second shield <NUM> may be a sidewall of RTG <NUM> having a substantially constant thickness (e.g. selected as being a value within a range of <NUM>-<NUM>). In some embodiments, second shield <NUM> may have a varying thickness, whereby the thickness proximate to a perimeter of RTG <NUM> is less than a thickness along a central axis of RTG <NUM> (e.g., selected as being a value within a range of <NUM>-<NUM>).

<FIG> illustrates a perspective view of the dispersed fuel design for an RTG of <FIG>, in accordance with various embodiments. As seen from <FIG>, RTG <NUM> - described in detail in <FIG> above, is shown. In <FIG>, a perspective view of RTG <NUM> is depicted.

<FIG> illustrates another example of a cross-sectional view of a fuel design for an RTG, in accordance with various embodiments. As seen in <FIG>, the RTG includes heat pipes <NUM>, thermoelectric <NUM>, insulation <NUM>, fuel (e.g., strontium titanate) and graphite matrix <NUM>, x-ray shielding <NUM>, and a pyrolytic graphite radiative fin <NUM>. In some embodiments, the fuel design of <FIG> depicts an example of how the heat source can be incorporated into an RTG designed for space applications.

<FIG> illustrates an example of a zoomed-in region of an interface of a radioactive source material and shielding of a fuel design for an RTG, in accordance with various embodiments. As seen in <FIG>, strontium titanate particles are encapsulated in graphite, and the graphite and particle combination are dispersed within the fuel. At an edge of the fuel, in some embodiments, may be a graphite beta shield. The graphite beta shield may be sandwiched between the graphite/particle combination and a tungsten layer of shielding. The graphite beta shield can prevent a strontium titanate particle from being directly on the high atomic number, tungsten shield which can result in high-intensity bremsstrahlung generation.

The fabrication process for creating a radioactive source material, such as the radioactive source material described above with reference to <FIG>, is now described. In some embodiments, the fuel fabrication process relates to a strontium-<NUM> source material, however alternative radioactive isotopes may be used.

Pressure sensitivity studies on a strontium fluoride-graphite mixture of <NUM> micrometer diameter powders indicate that an appropriate pressure to press the strontium fluoride-graphite mixture to achieve a highest density is, approximately, <NUM> kilopounds of force per square inch (<NUM> MPa). Using this pressure, densities of approximately <NUM>-<NUM>% of the theoretical maximum density of the strontium fluoride-graphite mixture were obtained. In some embodiments, the mixture ratio is approximately <NUM>: <NUM> (e.g., graphite to strontium fluoride). By mass, the density of strontium fluoride is approximately <NUM>/cc, and the density of graphite is <NUM>/cc. The density of the strontium fluoride-graphite mixture is approximately <NUM>/cc. To perform the pressure sensitivity studies, samples of strontium fluoride-graphite mixture were cold isostatically pressed (CIP).

It should be noted that, as described herein, the use of strontium fluoride as a radioactive source material may, alternatively, be replaced with strontium titanate, or other radioactive source materials, and the present disclosure is not intended to be limited to only strontium fluoride. In other embodiments, any titanate-based source disclosure herein may be replaced with fluoride.

Temperature sensitivity studies were also performed on the strontium fluoride-graphite mixture to identify an appropriate temperature to press the strontium fluoride-graphite mixture to achieve a maximum density. The strontium fluoride-graphite mixture, in powder form, was hot isostatically pressed (HIP). The resulting material's density was not analyzed for appropriate temperature due to asymmetric dimensions caused by removing the samples from the can used to perform the HIP. By visual inspection, the appropriate temperature to achieve the best resulting material was determined to be approximately <NUM> degrees Celsius.

HIP has also been used to press simulated fuel pellets for an electrically heated RTG prototype. First, CIP was performed to the strontium fluoride-graphite mixtures to improve a resulting net shape. Next, HIP was performed to the resulting CIP-ed mixtures to further increase density. However, this technique was determined to produce samples that were not large enough, and additional prototypes are to be made and tested.

CIP and HIP of the strontium fluoride-graphite mixture has been determined to produce suitable densities, however the resulting shape is not of adequate uniformity. Furthermore, the samples required post processing that resulted in a large amount of dust being created, which is not desirable. Hot pressing, instead of HIP, of pellets, or hot pressing directly into the x-ray shielding using the shield as a die, has been determined to be more suitable.

The dimensional discrepancy previously detected may result from a process known as cold sintering. Putting powdered material (having a theoretical density of approximately <NUM>%) under pressure compacts the material and reduces the porosity of the material to achieve densities that are limited to about <NUM>% of theoretical density. Sintering the powdered material induces material changes that can eliminate porosity and potentially reach greater than <NUM>% of theoretical density. However, the sintering temperature of graphite is greater than the melting temperature of SrTiO<NUM> and SrF<NUM>. For example, the melting temperature of strontium fluoride is approximately <NUM> degrees Celsius, and the melting temperature of strontium titanate is approximate <NUM> degrees Celsius. Cold sintering suggests that the sintering temperature of a material can be reduced at high pressures, thereby allowing the materials to density at lower temperatures. This may explain the dimensional discrepancies if the material reached higher densities than anticipated during the initial pressing steps.

<FIG> illustrates an example of a process <NUM> for manufacturing a radioactive source material used for an RTG, in accordance with various embodiments. In some embodiments, process <NUM> may be conducted using a hot cell or a shielded glovebox. Minor alterations may be made to either the hot cell or the shielded glovebox to limit exposure doses to occupational workers. It should be noted that, although the foregoing description relates to a strontium-<NUM> radioisotope as a radioactive source material, alternative radioactive source materials may be used and are not limited to the few potential radioisotopes discussed previously in this disclosure.

In some embodiments, process <NUM> may begin at step <NUM>. At step <NUM>, a strontium compound (e.g., strontium-<NUM>) may be obtained. In some embodiments, the strontium compound may be obtained after impurities have been removed. For example, for stronitium-<NUM>, the removed impurities may include the decay product zirconium-<NUM>. Removal of the impurities from the source material may function to maximize a specific power of the fuel for the RTG. In some embodiments, the strontium compound (e.g., <NUM>Sr) may be converted to strontium titanate (e.g., SrTiO<NUM>) or strontium fluoride (e.g., SrF<NUM>). If waste or contaminated materials remain from the strontium conversion process, they may be disposed of appropriately.

At step <NUM>, the strontium compound (e.g., <NUM>SrTiO<NUM> or <NUM>SrF<NUM>) may be reduced in particle size. In some embodiments, the strontium compound may be powdered and sieved. In some embodiments, the strontium compound may be powdered using a ball mill. After being powdered, the heterogenous powder may be transferred to a system of sieves to obtain a uniform powder. In some embodiments, the strontium compound (e.g., <NUM>SrTiO<NUM> or <NUM>SrF<NUM>) may be powdered using grinding media in the top of the sieve to grind and sieve at the same time. The powder may be sieved for uniform particle size as well. The size of a radius of each granule of the powders used may be between <NUM>-<NUM> microns, <NUM>-<NUM> microns, <NUM>-<NUM> microns, <NUM>-<NUM> microns, or other ranges. For example, a size of a radius of the granules in the powders used may be <NUM> micrometers. As another example, a size of a radius of a granule of the powders may be less than <NUM> microns. Persons of ordinary skill in the art will recognize that the aforementioned dimensions are approximates and some granules may be slightly larger or smaller than others.

As an alternative (or additional) method of particle size reduction, the strontium compound can also be dissolved in aqueous solution. In some embodiments, the strontium compound may be dissolved in an acid, solvent, or other liquid media (e.g., acetone, trichloroethylene, acetic acid, nitric acid, methanol, isopropanol, deionized water, tap water, salt water, or ammonium hydroxide) to decrease powder contamination. The size of a radius of each granule of the particles used may be between <NUM>-<NUM> microns, <NUM>-<NUM> microns, <NUM>-<NUM> microns, <NUM>-<NUM> microns, or other ranges. For example, a size of a radius of the granules in the particles used may be <NUM> micrometers. As another example, a size of a radius of a granule of the particles may be less than <NUM> microns. In other embodiments, strontium titanate may be placed in aqueous solution or solvent and mixed to reduce the particle size. For example, <NUM>SrTiO<NUM> may be placed in ammonium hydroxide (NH<NUM>OH) and allow <NUM>-<NUM> minutes to size reduce with ultra-sonication.

At step <NUM>, powdered graphite may be mixed with the strontium compound (made for example by either/both methods described above in step <NUM>) to suspend the <NUM>SrTiO<NUM> particles in a strontium-graphite mixture (which may be a slurry, if the strontium compound is in aqueous solution). The strontium-graphite mixture may then be heated (e.g., to between <NUM> and <NUM> C, optionally to approximately 80C) and spun (e.g., between <NUM> and <NUM> RPM, optionally at approximately <NUM> RPM) to remove any ammonium hydroxide and return the mixture to a dry powder before pressing. In some embodiments, the strontium-graphite mixture may be mixed with graphite powder of a same or similar granule size as the strontium. The mixing may be performed using a powder mixture device, such as a V-shape mixer, in accordance with a predefined ratio. In some embodiments, the ratio of graphite to strontium may be <NUM>:<NUM> (e.g., graphite:SrF<NUM>), however other ratios (e.g., <NUM>:<NUM>, <NUM>:<NUM>, etc.) may be used depending on power outputs desired for the RTG, chemical form of the strontium used, or other factors. The ratio used can vary depending on a chemical form of the radioactive source material and a desired power output of the RTG.

At step <NUM>, the strontium-graphite mixture is pressed. In some embodiments, the strontium-graphite mixture may be pressed into a "fuel pellet" using a hot press, and subsequently placed into x-ray shielding. The pressing process may include various types of pressing, including, but not limited to, (which is not to imply that other lists are limiting), HIP, CIP then HIP, CIP then hot press, cold press, cold press and sinter, or other forms of pressing. In embodiments where the strontium-graphite mixture/slurry is transferred to a heat loaded die, one example of heating before pressing can include being at <NUM> degrees C for <NUM> minutes, <NUM> degrees C for <NUM> minutes, <NUM> degrees C for <NUM> minutes, and <NUM> degrees C for <NUM> minutes. In some embodiments, when HIP is used to press the strontium-graphite mixture, a pressure of approximately <NUM> ksi and a temperature of <NUM> degrees C can be used. However, the pressure and temperature may differ from these values. For example, the pressure may be selected from a range of <NUM>-<NUM> ksi (<NUM>-<NUM> MPa), <NUM>-<NUM> ksi (<NUM>-<NUM> MPa), <NUM>-<NUM> ksi, approximately <NUM> ksi (<NUM> MPa), or other pressure values. As another example, the temperature may be selected from a range of <NUM>-<NUM> degrees C, <NUM>-<NUM> degrees C, <NUM>-<NUM> degrees C, <NUM>-<NUM> degrees C, or other temperatures. In some embodiments, the pressing temperature may be <NUM> degrees C. Persons of ordinary skill in the art will recognize that additional temperature ranges and pressure ranges may be used. For instance, the temperature and pressure ranges may depend on the chemical form of the radioactive source material (e.g., strontium titanate), and other materials (e.g., graphite), that are used. As an alternative example, boron nitride or boron carbide, which has a similar effective atomic number as graphite, may have different characteristic temperatures and pressures. Additionally, an additive or binder (e.g. sugar, stearic acid, polycaprolactone, or propylene carbonate) may be added to the mixture to increase pellet density after decomposition during pellet pressing.

At step <NUM>, the pressed mixture may be encapsulated. In some embodiments, the pressed mixture may be encapsulated into the x-ray shielding and secured to a leak tight fit.

At step <NUM>, one or more tests may be performed to the encapsulated mixture. In some embodiments, the test may include measuring a radiation dose rate at various distances from the heat source. For example, the radiation dose rate at the surface of the encapsulated mixture, at one meter away, at <NUM> meters away, etc., may be measured. In some embodiments, the measured radiation dose rate may be compared to a threshold radiation dose rate for each distance to determine whether the measured radiation dose rate is greater than the threshold radiation dose rate. For example, to satisfy safety requirements, the radiation dose rate at the surface of the heat source should be less than <NUM> mrem/hr (<NUM>µSv/hr) and less than <NUM> mrem/hr (<NUM>µSv/hr) at <NUM> meter from the surface. The surface, as described herein, refers to a portion of the fuel source that could be contacted or touched. If so, then a new heat source may be needed and the one tested may not be used. In some embodiments, the heat output of the heat source may be measured. The heat output may be compared to a threshold heat output to determine whether the measured heat output is greater than the threshold heat output. If so, then a new heat source may be needed and the one tested may not be used. In some embodiments, a leak testing may be conducted to confirm whether the heat source is sealed.

At step <NUM>, any process losses may be disposed of in accordance with known safety protocols. For example, the process losses may be disposed of as LLW using standard procedures.

In some embodiments, the fuel designs described herein may be used for a variety of applications including, but not limited to, heater units for space environments, heater units for cold environments, thermal power sources for energy conversion, medical radioisotope storage shielding, spent nuclear fuel transportation shielding, medical radioisotope transportation shielding, spent nuclear fuel shielding, emergency radiation containment, biological radiation protection for space environments, and biological radiation protection for emergency radiation environments.

The reader should appreciate that the present application describes several independently useful techniques. Rather than separating those techniques into multiple isolated patent applications, applicants have grouped these techniques into a single document because their related subject matter lends itself to economies in the application process. But the distinct advantages and aspects of such techniques should not be conflated. In some cases, embodiments address all of the deficiencies noted herein, but it should be understood that the techniques are independently useful, and some embodiments address only a subset of such problems or offer other, unmentioned benefits that will be apparent to those of skill in the art reviewing the present disclosure.

Claim 1:
A method of fabricating a heat source for a radioisotope thermoelectric generator (RTG), comprising:
reducing (<NUM>) a particle size in a radioisotope compound that produces high-energy beta radiation by powdering and sieving the radioisotope compound and/or dissolving the radioisotope compound into an aqueous solution;
mixing (<NUM>) the radioisotope compound with graphite to obtain a radioisotope-graphite mixture;
performing (<NUM>) a press to the radioisotope-graphite mixture; and
encapsulating (<NUM>) the pressed radioisotope-graphite mixture into an x-ray shielding to obtain the heat source.