Patent Application
20160329113
Small modular nuclear reactors (SMRs) offer specific attributes that make them attractive for remote and isolated communities with limited/seasonable access to fossil fuel supplies, to countries with small electric grid capacity, and island nations with no or limited access to an electric grid. SMRs are also an attractive option to electric utilities operating in regions with modest growth in electricity demand and/or modest financial resources.
All or most components of SMRs, depending on the installed electrical power of the plant, are built in the factory. They are small enough to be shipped to the site by rail, truck or on a barge. During nominal operation and after shutdown, SMRs can be designed to operate fully or partially passive with no or a few circulation pumps and redundant means for removing the decay heat generated in the reactor core after a routine or an emergency shutdown. These reactors can also be designed with long operation lives of 5-10 years, or even longer, without refueling.
Small modular nuclear reactors can provide for up to 300 MWe of electricity. In highly populated and industrial areas, the cost of electricity generation using SMRs may not be comparable to, or lower than that of medium (>300 and <1000 MWe) and large (>1000 MWe) nuclear power plants.
A typical large nuclear plant provides 1,000-1,500 MWe at a capital cost of $6 B-$10 B and takes 5-6 years to build. A SMR plant with an installed capacity of ˜300 MWe at cost <$2 B takes ˜18 months to build. A 10 MWe SMR plant may take only 6 months or less to bring on line at a cost of ˜$80 M. In addition to design simplicity and partial or fully passive operation and safety features, an SMR vessel typically has few penetrations, and some have an in-vessel compact steam generator or heat exchanger. Large nuclear plants typically require emergency planning zones of up to 10 miles in radius, while those for SMRs are <0.5 miles.
SMRs are constructed and brought on line incrementally in future years, commensurate with the increase in electricity demand. This avoids the financial obligations associated with building a large plant, the need to develop a consortium of multiple utilities in more than one state, the interest on a large loan ($6 B-10 B) for ˜5-6 years of the construction time, and the lack of revenues during these years. SMRs offer simple designs with partial or fully passive operation and safety features. During nominal operation and after shutdown, SMRs are safely cooled by natural convection or using a few circulation pumps. They also have redundant and passive means for removing decay heat generated in the reactor core after a routine or an emergency shutdown. They are designed to survive a loss of both onsite and off-site power for weeks, without compromising safety. SMRs may offer independent and passive means for generating auxiliary power to support vital plant functions, in the unlikely event of a Fukushima-Daiichi type accident.
Many SMRs have thermal neutron spectra, but only a few are being developed with fast neutron spectra. Those with thermal spectra have an operation cycle length between refueling of 1-5 years. By contrast, fast neutron spectrum SMRs are smaller in size and may operate for decades without refueling and effectively minor actinides. The light water moderated and cooled SMRs with thermal spectra capitalize on the existing technology base and long and successful experience operating large Light Water Reactors (LWRs) and use low fuel enrichment <5%.
These SMR designs offer a number of passive operation and safety features. Because of their thermal energy spectra, they minimally reduce minor actinides in used fuel, which could effective in fast spectra reactors.
High temperature gas cooled HTG-SMRs with epithermal or fast neutron spectra for burning minor actinides operating at exit temperatures up to 1000-1200 K are considered or actively being developed for a near term deployment. In addition to the high thermal efficiency (>40%) for electricity generation, the HTG-SMR plants with a hybrid operation provide for thermochemical production of hydrogen fuel and process heat for a multitude of industrial uses. The HTG-SMR designs, which effectively burn actinides and utilize used fuel from LWR plants, are well known.
The sodium cooled SMRs with fast neutron spectra, are not only the smallest but also effective in destroying minor actinide during reactor operation. The ratio of fission to capture cross sections is higher than in reactors with thermal or epithermal energy spectra. Owing to the low vapor pressure of sodium coolant, SMRs operate at relatively high exit temperatures (<850 K), plant thermal efficiency close to 40%, and below atmospheric pressure. By contrast, light water SMRs operating at high pressures of 5-15 MPa, require a massive reactor vessel and operate at a lower plant thermal efficiency of 30 to 33%. The potential of other liquid metals (molten lead or lead-bismuth) cooled and molten salt cooled SMRs are also being investigated.
In one embodiment, the present invention is directed toward a sodium-cooled, small modular reactor capable of generating tens of MWth and has a long operation life without refueling. The reactor of the present invention includes an in-vessel coiled tube Na/Na heat exchanger (HEX) and two redundant control/shutdown systems. It may be fabricated, assembled and sealed at the factory and shipped to the site by rail, a heavy truck or on a barge. It is installed below ground to alleviate a missile or a spacecraft impact and mounted onto seismic insulators to withstand Earthquakes.
In another embodiment, the present invention uses natural circulation of liquid sodium with the aid of in-vessel helically coiled tubes, Na/Na HEX, and a chimney to cool the fast neutron spectrum of the reactor during nominal operation and after shutdown. The design uses a fuel rod cladding, core structure and fuel materials, operates below 820 K and has two redundant control and emergency shutdown systems. It operates fully passive, except for the control drives, uses rods of UN fuel with enrichment less than 18% in the driver core, and has radial and axial blankets of depleted UN (DUN). Other embodiments may use a BeO reflector blanket to reduce or lower fuel enrichment. The high heavy metal atom ratio of UN increases the operation cycle length without refueling. The UN has good retention of fission gasses and its high thermal conductivity typically keeps the temperatures of the UN fuel in the reactor core below 1400 K. These inherent characteristics minimize or eliminate concerns of fuel swelling and fission gas release and enhance safety and fuel performance for a long operation life of the reactor.
In yet another embodiment, the reactor core has two redundant control systems with separate drives, one for safety shutdown and the other controls the reactor operation. Each system is capable of shutting down the reactor in case of an emergency. Both safety shutdown systems may consist of enriched B4C control rods, with BeO followers.
In the drawings, which are not necessarily drawn to scale, like numerals may describe substantially similar components throughout the several views. Like numerals having different letter suffixes may represent different instances of substantially similar components. The drawings illustrate generally, by way of example, but not by way of limitation, a detailed description of certain embodiments discussed in the present document.
Detailed embodiments of the present invention are disclosed herein; however, it is to be understood that the disclosed embodiments are merely exemplary of the invention, which may be embodied in various forms. Therefore, specific structural and functional details disclosed herein are not to be interpreted as limiting, but merely as a representative basis for teaching one skilled in the art to variously employ the present invention in virtually any appropriately detailed method, structure or system. Further, the terms and phrases used herein are not intended to be limiting, but rather to provide an understandable description of the invention.
The reactor design of the present invention takes advantage of the high heavy metal atom ratio, high melting point and high thermal conductivity of UN fuel and uses HT-9 steel for cladding, reactor vessel and core structure. It may be fabricated, assembled and sealed offsite such as in a factory, shipped to the construction site by rail, heavy truck or barge and installed below ground on seismic insulators. Natural circulation of liquid sodium cools the reactor core during nominal operation and after shutdown using in-vessel helically coiled tubes such as a Na/Na heat exchanger (HEX). In other embodiments, other known heat exchangers may be used as well. It also estimates the operation life of the reactor in full power mode (100 MWth of six years and as much as 34 years at 20 MWth.
The reactor of the present invention operates fully passive, except for the control drives, uses UN fuel with enrichment <18% in the driver core with radial and axial blankets of either depleted uranium nitride (DUN) or BeO. The high thermal conductivity of UN decreases fuel temperatures during nominal operation below 1400 K, depending on the reactor thermal power. At such temperatures, fuel swelling and fission gas release are non-issues. In addition to its compatibility with the HT-9 and liquid sodium, UN has a higher volumetric heat capacity than UO2 and U-metal fuels for enhanced safety.
In addition to the high heavy metal ratio for long operational life of the reactor, the UN high thermal conductivity, ˜10 times that of UO2, decreases the maximum UN fuel temperature during nominal reactor operation. The UN fuel is compatible with the HT-9 cladding and with sodium coolant at the operation temperatures of the reactor (<820 K).
As shown in
As further shown in
Fuel assemblies 1-36 may be retained by hexagonal corner assemblies 120-125 which may be made of steel or BeO rods. A hexagonal cladding may also be used. Also surrounding and retaining the fuel assemblies are blanket assemblies 130-147 which may be comprised of rods containing depleted (DUN) fuel or BeO pellets. As further shown in
As shown in
The fuel rods have an upper gas plenum 210 section for accommodating released fission gasses, sections of axial blankets of depleted (DUN) or BeO pellets 212 and 216, and fuel section 214. As shown in
As shown in
As described above, using hexagon assemblies allows for the creation of rings of assemblies that may be interlocked together. In a preferred embodiment, each assembly uses the same hexagonal frame or housing which allows for the assemblies to be mated together into hexagonal rings or units. For example, assemblies 1-6 form a first hexagonally-shaped ring around assembly 190. Assemblies 7-18, in turn, form a second hexagonally-shaped ring around assemblies 1-6. Assemblies 19-36, in turn, form a third hexagonally-shaped ring around assemblies 7-18. Assemblies 120-125 and 130-147, in turn, form a fourth hexagonally-shaped ring around assemblies 19-36. Lastly, a retaining wall 108 having wedges 170 locks the assemblies together.
As shown in
Also shown is the placement of Na/Na HEX 422 and the natural circulation paths of heated liquid sodium. The paths are indicated by arrows 440-443 and cooled liquid sodium is indicated by arrows 450-455.
The internal pressure in the reactor vessel is kept slightly below atmospheric, owing to the low vapor pressure of the sodium. Thus, the steel reactor vessel wall is only a few inches thick. In a preferred embodiment, the pressure of the argon cover gas is slightly below atmospheric and the total pressure at the bottom of the reactor vessel is less than 2.0 MPa, depending on the chimney height.
In-vessel natural circulation of liquid sodium cools the reactor core during nominal operation and after shutdown, with the aid of chimney 420 and in-vessel HEX 422. The difference between the static pressure head of sodium in downcomer 466 and the sum of those in the core and the chimney drives the natural circulation of liquid sodium through the core. The circulating liquid sodium removes and transports the fission or the decay heat generated in the core to HEX 422 near the top of downcomer 466. The cooled sodium flows downward to lower plenum 429 before entering the reactor core to repeat the cycle.
In another embodiment using the same core design and having an inlet temperature of liquid sodium at the core of about 610 K, increasing the chimney height increases the nominal thermal power of the reactor and the temperature of liquid sodium exiting the core, which remains <820 K. At these temperatures, corrosion of HT-9 for the fuel rods cladding and core structure by liquid sodium is negligible. In another embodiment, the sodium vapor pressure at its exit temperature from the core during nominal reactor operation is very low at about 1.6 kPa. This provides a large safety and operation margin to boiling incipience.
Leak and radiation detectors are also placed in the gap between the inner and outer vessels for monitoring reactor operation. Each vessel is capable of containing the reactor core and the internal structure and components.
As indicated earlier, Na/Na HEX 422 removes the reactor thermal power during nominal operation and the decay heat after shutdown. The heat removed from the circulating liquid sodium in the reactor vessel is transported to one or more steam generators (not shown) in a superheated steam Rankine cycle that produces electricity at a plant thermal efficiency of up to 40% or even higher. Nominal electrical power and the thermal efficiency of the reactor plant depend on the operating reactor thermal power and the temperature of the liquid sodium coolant exiting the reactor core as shown in
The solution assumes a constant sodium inlet temperature into the core of 610 K, which may be varied to change the temperature rise in the core. The pressure losses in the momentum balance equation include those in the core, chimney, Na/Na HEX and the rest of the downcomer. They also account for the changes in the flow area in the various regions. The highest-pressure losses are those calculated in the reactor core followed by those in the helically coiled tube Na/Na HEX and then the chimney. For a reactor power of 10 MWth the temperature of liquid sodium exiting the core decreases from ˜667 to only 655 K as the chimney height increases from 2 to 8 m. For a reactor power of 100 MWth a chimney that is at least 6 m tall keeps the temperature of the sodium exiting the core at or below 820 K. This temperature decreases to 809 K when the chimney height increases to 8 m. At these temperatures, sodium corrosion of the HT-9 cladding, core structure and reactor vessel is negligible.
As described above, during nominal reactor operation and after shutdown, liquid sodium continues to flow by natural circulation through the core and the chimney and downward in the downcomer separating the core barrel from the reactor vessel. The low thermal conductivity of the inert argon gas that fills gap or chamber 430 of the embodiment shown in
To further reduce heat losses, the reactor may include an above ground vent on the outside surface of the guard vessel which remains shut. The vent may be configured to open only intermittently when the air heats up and expands against the weight of the vent cover.
The variable conductance, liquid metal heat pipes 130A-130D in the reactor vessel wall may be configured to capture some of the side heat losses, both during reactor operation and after shutdown, and transport it to an offsite location such as containment building for other uses. There, it may be partially converted to electricity using static modules comprised of segmented TE elements. The modules may be made of a number of parallel strings of TE elements for redundancy and avoidance of a single point failure. The TI: modules generate kilowatts of electrical power at relatively high voltage of ˜200-400 VDC. They serve as an auxiliary power source for operating critical functions of the plant, particularly in the unlikely event of a loss of both onsite and offsite power. The waste heat removed from the cold side of the TE modules by natural circulation of ambient air. This passive design feature not only enhances reactor and power plant safety, but also increases the total thermal power utilization for the plant.
In yet another preferred embodiment, the reactor is modular and fabricated, fully inspected and assembled and sealed at an offsite location such as a factory, enhancing the quality assurance and reliability of the reactor. It may be shipped to the site by a truck, rail or on a barge, with the coolant frozen prior to shipping so as to prevent coolant from moving during transport. Th reactor may be installed below grade and brought on line within a short time <12 months, depending on the nominal reactor thermal power. At the end of its operational life, which may be several years to decades, the reactor units may be replaced with new ones, in a relatively short period of time, which may be as little as days. The used units may then be shipped back to the vender for fuel reprocessing in a safe and secure facility, thus eliminating proliferation concerns.
The reactor design of the present invention takes advantages of the high heavy metal atom ratio, high melting point and high thermal conductivity of UN fuel and uses stainless steel for the fuel rod cladding and the reactor core structure and a steel vessel. As a result of the modularity, with the same reactor core design, materials and dimensions, the nominal thermal power of the reactor increases simply by increasing the height of the chimney up to 8 m, while keeping the temperature of liquid sodium coolant entering the core the same temperature of about 610 K. The increased chimney height increases the circulation rate of liquid sodium through the core as well as its exit temperature. The exit temperature generally remains below 820 K to maintain a negligible corrosion rate of the HT-9 steel components such as the cladding of the UN fuel rods and core structure materials. The low reactor thermal power density increases its operation life and together with the high thermal conductivity of UN fuel, decreases its temperature during nominal reactor operation. Such low temperature results in negligible fuel swelling and fission gas release, consistent with the long operation life of the reactor without refueling.
In additional embodiments of the present invention, the reactor core has a negative temperature reactivity feedback that helps passive shut down in case of a large temperature rise. The reactor operation is monitored using two redundant systems, each with separate drivers; the safety shutdown (190 in
Reactor control systems 180 and 181 include enriched B4C rods with BeO followers which may be located in the fuel assemblies described above. In a preferred embodiment, the rods are inserted at the center of fuel rod assemblies which may be in place of rod 50 that is shown in
After a reactor shutdown, the coiled tube Na/Na heat exchanger continues to remove the decay heat generated in the reactor core, and some of the sensible heat of the liquid sodium in the reactor vessel, decreasing its temperature with time after shutdown. The large mass of liquid sodium in the reactor vessel serves as a huge heat sink, enhancing safety by limiting the initial increase in its temperature shortly after reactor shutdown. In a preferred embodiment, the liquid sodium mass increases with the height of the chimney, from 18 MT to 38 MT as the chimney height increases from 2 m to 8 m as shown in
In the unlikely event of a malfunction, the two redundant control systems may be used to shut down the reactor using either system or both systems. In addition, the reactor core negative temperature reactivity feedback limits potential power increase and help reactor shutdown. The decay heat generated in the reactor core after shutdown is removed safely and passively by a backup system of natural circulation of ambient air on the outside of the guard vessel wall. Replacing the argon gas in the small gap between the reactor vessel and the guard vessel with liquid sodium increases the gap conductance, thus enhancing the decay heat removal from the reactor core. Initially, the decay heat is partially stored in the large sodium mass within the reactor vessel. It is also transferred by conduction from the reactor vessel to the guard vessel wall, which has longitudinal metal tins, where removed by natural circulation of ambient air.
In addition, within a few days after shutdown, depending on the reactor nominal power before shutdown, natural circulation of air alone is capable of removing all the decay heat generated in the core and some of the sensible heat stored earlier by the liquid sodium in the vessel, gradually lowering its temperature. The maximum temperature of liquid sodium in the vessel after reactor shutdown remains below its boiling point. The decay heat removal from the reactor guard vessel is aided by the array of variable conductance heat pipes 103A-103D located in the reactor vessel wall as shown in
Owing to the negative temperature reactivity feedback of the reactor core, the plant is inherently load following. Thus, it accommodates a change in the load demand, as a percentage of the nominal reactor power without active control. In addition to the enhanced reliability, this design feature increases the plant's availability or capacity factor. Table I below lists the primary features of some embodiments of the reactor of the present invention and power plant performance.
The rate of decay heat generation in core at shutdown may vary from ˜8% to 10% of the reactor's nominal power before shutdown. After reactor shutdown, the rate of heat removal by natural circulation of ambient air from the outer surface of the guard vessel wall would initially be lower than that of the decay heat generation in the core. The difference would be partially stored in the large sodium mass in the primary vessel (Table I) and partially removed by heat pipes to TE energy conversion modules of the auxiliary power system. Thus, the peak temperature of liquid sodium in the primary vessel after reactor shutdown would remain well below its boiling point (˜1156 K) by an acceptable margin.
Another feature of some embodiments of the present invention is that the hot air riser on the outside of the guard vessel wall has a long chimney to enhance its circulation by natural convection. The hot air riser is insulated from the intake duct to minimize heat losses to incoming cold air and maximize the driving pressure for natural circulation. Within a few days after shutdown, natural circulation of ambient air along the surface of the guard vessel would remove not only the decay heat generated in the core, but also some of the sensible heat stored earlier in the liquid sodium in the primary vessel. The temperatures of the in-vessel liquid sodium and the reactor's primary and guard vessel walls would then decrease gradually with time. This fully passive backup system for removing the decay heat from the core enhances reactor safety.
Neutronic analyses of the reactor was undertaken using the Monte Carlo neutron transport code MCNP5. The performed criticality calculations without reaction rate tallies used 20,000 source particles per history and 50 skipped and 9,00 active histories. Those performed with reaction rate tallies used 20,000 source particles per history and 50 skipped and 5,000 active histories. The calculations of the cold-clean reactivity assume uniform temperature of 400 K throughout the reactor core. Those of the hot-clean reactivity are at the calculated temperatures in the various regions of the reactor core such as the UN fuel pellets, HT-9 cladding, liquid Na and the core BeO and HT-9 structures. The temperatures vary with the nominal power of the reactor (10-100 MWth).
The performed neutronic analyses investigated the effects of using different materials for the follower rods in the two control systems of the reactor on the cold- and hot-clean reactivity and the reactivity shutdown margin. Varied is the material (BeO, HT-9 and DU) of the followers for the B4C rods in the central assembly for emergency shutdown and in the fifteen B4C rods in the reactor control system. Also investigated were the effects of using HT-9 and DUN corner assemblies in the radial blanket, which was described above as the 4th ring in the core both on the clean reactivity and the shutdown margin as well as the reactivity worth of the thickness of the HT-9 core barrel wall up to 20 cm.
To avoid distorting the neutrons flux profile in the core and remove the tally dependence on keff, the calculations of the neutron reaction rate and flux tallies were performed with the reactor core critical and at hot condition (keff=1). The neutron energy spectrum in the critical hot-clean core was also determined. In the calculations of the neutron energy spectrum, the fuel composition and power production in the UN fuel bundles in each of the three rings of the driver core were tracked independently using MCNPX version 2.7.0 code. This code also tracked the fuel depletion in the core throughout the reactor's operation life, when fully depleting the hot-clean excess reactivity. The calculations were repeated at reactor nominal thermal powers of 10-100 MWth and critical (keff=1.0) hot conditions.
The base design parameters for the reactor included a chimney height of 8 m. UN fuel enrichment of 17.65%, HT-9 core barrel wall-thickness of 10 cm, and DUN radial blanket with HT-9 steel corner assemblies. In addition, the B4C rods in the central assembly for safety shutdown and in the fuel assemblies in the 1st and 2nd rings of the driver core have BeO followers as discussed above. The neutronic calculations of reactor varied one base design parameter at a time.
As indicated earlier, for the natural circulation cooled reactor at a given nominal thermal power and sodium inlet temperature to the core of 610 K, increasing the in-vessel chimney height increases the liquid sodium flow rate through the core and decreases its exit temperature as shown in
The temperatures, calculated separately, are used to adjust densities of the UN fuel and core structure materials, core dimensions and neuron cross-sections, including resonance broadening, in the neutronic and fuel depletion analyses and lifetime estimates. The height of the reactor core increases when hot and as well as the diameters of the fuel rods and the dimensions of the core and the BeO and HT-9 shrouds for the core assemblies. In addition, the density of the liquid Na in the reactor vessel decreases and its total volume increases. Increasing the UN fuel enrichment from 17.25% to 17.65% increases the cold-clean excess reactivity but reduces the reactivity shutdown margin. This margin varies from −$3 to −$2.5 at a fuel enrichment of 17.25% to −$1.75 to −$1.6 at slightly higher enrichment of 17.65% in the base design of the reactor core. These reactivity margins are more than sufficient to safely shutdown the reactor using the B4C rods with BeO followers of either the emergency shutdown assembly or the control rods described above.
For the embodiment concerning a base design of the reactor, which provides HT-9 steel corner assemblies 120-125 in the radial blanket.
The HT-9 core barrel helps reflect leaking neutrons out of the core and could affect the reactivity of the reactor.
The results shown in
Owing to the slight tail of the neutron energy spectra in the epithermal range, the effect of resonance broadening of the neutron cross-sections on reactivity at hot condition would be small. Other factors that affect the hot-clean reactivity are: (a) the decreases in densities of UN fuel, HT-9 cladding and core structure and other core materials and (b) the changes in the dimensions of active core and UN fuel. For example, the axial expansion of the UN fuel not only decreases the fission rate but also increases the active core height and hence neutron leakage, partially decreasing the hot-clean reactivity.
As indicated in
The neutron energy spectra in
The hard neutron energy spectrum of the SLIMM reactor is advantageous in reducing the inventory minor actinides generated in the UN fuel during reactor operation. The effectiveness in burning minor actinides stems from the fact that the ratio of fission to the capture neutron cross sections is higher in a fast spectrum than in a thermal spectrum, such as of LWRs.
In another embodiment, the present invention provides a modular nuclear reactor comprising a reactor pressure vessel having a lower section having a first wall and a second wall and an upper section having a first wall and a second wall. The walls of the lower section and the upper section are adapted to be connectable. The second wall of the upper section defining a chimney having a first open end and an opposingly located second open end. The chimney maybe disposed in the upper section and defines a first passageway. Also provided is a second passageway disposed between the chimney and the first wall of said upper section. The upper section further includes an upper plenum in communication with the first open end of the chimney and connecting the first and second passageways of the upper section. A heat exchanger is disposed in the second passageway of the upper section. The first wall lower of said lower section is opposingly located from the second wall of the lower section to form a first passageway that is communication with the first passageway of the upper section. The second wall of the lower section defines a second passageway that is in communication with the second passageway of the upper section. A lower plenum connects the first and second passageways of said lower section. The reactor core is located within the second passageway of the lower section and is adapted to heat a heat transfer fluid by having one or more passageways in communication with the second passageway of the lower section. The first and second passageways create a circulation loop wherein a heated heat transfer fluid circulates up from the reactor core, through the chimney, through the upper plenum and downwardly past the heat exchanger, into the lower plenum and back into the core.
In other embodiments, the nuclear core includes a hexagonal shutdown drive assembly comprised of a plurality of B4C rods with BeO followers, a first hexagonal ring of six hexagonal UN fuel assemblies surrounding the shutdown drive, a second hexagonal ring of twelve hexagonal UN fuel assemblies surrounding the first ring, a third hexagonal ring of eighteen hexagonal UN fuel assemblies surrounding the second ring, a fourth ring hexagonal ring comprised of six one-half hexagonal BeO of DUN blanket assemblies at the vertices and eighteen BeO or DUN hexagonal blanket assemblies surrounding the third ring.
The reactor may further including a plurality of heat pipes in the vessel wall to enhance air-cooling by natural circulation. An auxiliary power source driven by heat extracted from the wall of the vessel may also be used. The power source may be adapted to generate electric power. Sensors between the inner and outer vessel walls for monitoring reactor operation and to provide early warning may also be used.
While the foregoing written description enables one of ordinary skill to make and use what is considered presently to be the best mode thereof, those of ordinary skill will understand and appreciate the existence of variations, combinations, and equivalents of the specific embodiment, method, and examples herein. The disclosure should therefore not be limited by the above described embodiments, methods, and examples, but by all embodiments and methods within the scope and spirit of the disclosure.
This application claims priority to and the benefit of the filing date of U.S. provisional application Ser. No. 61/913,097 filed Dec. 6, 2013 and incorporated herein for all purposes.
| Filing Document | Filing Date | Country | Kind |
|---|---|---|---|
| PCT/US2014/068910 | 12/5/2014 | WO | 00 |
| Number | Date | Country | |
|---|---|---|---|
| 61913097 | Dec 2013 | US |
