Patent Description:
<CIT> discloses high pressure gas cooled liquid moderated nuclear reactors comprising a moderator containment tank pierced by a plurality of tubes containing fuel elements cooled by the flow of pressurised gas in a coolant circuit including the tubes. <CIT> discloses a nuclear fuel bundle including a first end face and a second end face axially spaced apart from the first end face and a plurality of elongate nuclear fuel elements supported by at least one spacer. <CIT> discloses a fuel bundle design for a heavy water moderated and cooled reactor. <CIT> discloses a method for operating a closed loop regenerative thermodynamic power generation cycle system. <CIT> discloses improvements in nuclear reactors in which the coolant is gaseous and the moderator is heavy water. <CIT> discloses a pressure-tube nuclear reactor including an outer shell having an interior to contain a moderator at a first pressure and a coolant plenum to receive the coolant fluid at a second pressure. <CIT> discloses nuclear power station with carbon dioxide cooling system for producing electrical energy.

The subject matter for which protection is sought is defined by the independent claims.

Various aspects of at least one example are discussed below with reference to the accompanying figures, which are not intended to be drawn to scale. The figures are included to provide an illustration and a further understanding of the various aspects and examples and are incorporated in and constitute a part of this specification, but are not intended as a definition of the limits of a particular example. The drawings, together with the remainder of the specification, serve to explain principles and operations of the described and claimed aspects and examples.

This disclosure describes various configurations and components of a gas-cooled pressure tube nuclear reactor (GPTR). For the purposes of this application, embodiments of a GPTR that use a uranium fuel will be described. However, it will be understood that any type of nuclear fuel, now known or later developed, may be used and that the technologies described herein may be equally applicable regardless of the type of fuel used. Note that the minimum and maximum operational temperatures of fuel within a reactor may vary depending on the fuel used.

<FIG> illustrates an embodiment of a GPTR identifying some of the primary components of the design. In the GPTR <NUM> shown, a gas coolant, that is a coolant that is gaseous under room temperature and pressure, is used to remove heat generated by fission from the nuclear reactor <NUM>. The heat is then converted into mechanical work using a turbine <NUM> and the mechanical work is used to generate electricity.

In the embodiment shown, the nuclear reactor <NUM> includes some number (two are shown) of fuel columns <NUM> in the form of pressurized tubes containing some amount of nuclear fuel. The fuel columns <NUM> are arranged to bring enough fuel into proximity to achieve criticality and generate heat from the fission of the fuel. The generated heat is removed by the flow of gas coolant, which is the working fluid when in normal operation.

The fuel columns <NUM> are submerged in a pool of liquid moderator contained in a vessel called a calandria <NUM>. In some embodiments, the liquid moderator in the calandria <NUM> is water maintained at a relatively low pressure (e.g., less than <NUM> atm) and a relatively low temperature (e.g., less than <NUM>). Other options for moderator fluid include mixtures of ammonia and organic fluids, such as biphenyl and terphenyl mixtures, Monsanto's various Santowax brand products (Santowax-R, Santowax-OM, etc.), and Monsanto's OS-<NUM> (a mixture of terphenyls treated catalytically with hydrogen to produce <NUM> percent saturated hydrocarbons). For example, in an embodiment, the water is maintained at from <NUM>-<NUM> and <NUM>-<NUM> atm. Depending on the embodiment, the water further may be heavy water (i.e., deuterium oxide), normal or "light" water (i.e., protium oxide), or a mix of both depending on the amount of moderation desired. Note that the liquid moderator may also provide some cooling of the fuel column <NUM>. However, the term "liquid moderator" or, simply, "moderator" will generally be used instead of "liquid coolant" to distinguish the moderating calandria liquid from the gas coolant. Likewise, the term "calandria" <NUM> is used herein to distinguish the vessel from a high pressure reactor vessel commonly found in pressurized water reactors in which the water may exceed <NUM> atm in pressure. The calandria <NUM> and the liquid moderator are discussed in greater detail below.

Control rods (not shown) may also be provided as is known in the art for additional moderation and control of the reactivity of the reactor <NUM>. In an embodiment, the reactor may be shut down via movement of the control rods into and out of the calandria <NUM>. For example, in an embodiment during an emergency control rods may be automatically inserted into the calandria <NUM> thus bringing the reactor subcritical.

As mentioned above, a fuel column <NUM> includes an exterior structural tube containing some amount of nuclear fuel in which the structural tube is capable of holding gas pressurized up to the operating gas pressure. In an embodiment, the pressurized tube is provided with an inlet for receiving the gas coolant and an outlet for discharging the gas coolant, thus allowing coolant to be flowed through the interior of the tube and, thus, to remove heat from the nuclear fuel. During normal operation the gas coolant provides most, if not all, heat removal and, thus, temperature control of the fuel.

In an embodiment, the fuel columns <NUM> may be designed to also allow the fuel to be removed from the column <NUM>. Depending on the embodiment, the nuclear fuel in the fuel columns <NUM> may be in any solid form and any geometry. For example, in an embodiment the fuel column <NUM> may be filled with nuclear fuel particulates or pellets. Forms and geometries that provide good thermal contact between the fuel and the gas coolant may be particularly advantageous. In addition, supercritical fluid coolants may also be advantageous because of their improved ability to penetrate porous structures. Several embodiments of nuclear fuel inserts suitable for use in fuel columns <NUM> are discussed in greater detail below.

As illustrated in <FIG>, a heated coolant stream <NUM> exits the fuel columns <NUM> and is passed to a turbine <NUM>. In an embodiment, for example, the coolant is supercritical carbon dioxide maintained at a pressure from <NUM>-<NUM> MPa and at a temperature from as low as <NUM> to as high as <NUM> such as from <NUM>-<NUM> and from <NUM>-<NUM>. The high pressure of the heated coolant stream <NUM> is used to drive the turbine <NUM> converting the potential and/or thermal energy of the coolant into mechanical rotational energy. In the embodiment shown, the turbine <NUM> is connected by a shaft <NUM> to a generator <NUM>, which in turn converts the mechanical energy into electricity.

After driving the turbine <NUM>, the coolant is discharged at a lower temperature and pressure as a depressurized coolant stream <NUM>. In the embodiment shown, this stream <NUM> is passed through heat exchanger <NUM> which cools the coolant. Heat exchanger <NUM> may be considered a recuperator as the coolant stream <NUM> from the turbine <NUM> is transferring heat to the coolant stream <NUM> from the compressor <NUM> prior to its return to the reactor <NUM>. The cooled stream <NUM> discharged by the heat exchanger <NUM> may be further cooled by passing it through a cooler <NUM>. In an embodiment, the cooler <NUM> may simply be a second heat exchanger that cools the coolant using chilled air or water from an external source. Except were explicitly discussed otherwise, heat exchangers will be generally presented in this disclosure in terms of simple, single pass, shell-and-tube heat exchangers having a set of tubes and with tube sheets at either end. However, it will be understood that, in general, any design of heat exchanger may be used, although some designs may be more suitable than others. For example, in addition to shell and tube heat exchangers, plate, plate and shell, printed circuit, and plate fin heat exchangers may be suitable.

The coolant of the cooled output stream <NUM> from cooler <NUM> is then delivered to a compressor <NUM> where it is repressurized to at or near the operating pressure of the fuel columns. In the embodiment shown, the pressurized coolant stream <NUM> discharged by the compressor <NUM> is preheated by the heat exchanger <NUM> before it is returned <NUM> to the reactor <NUM> to flow through the fuel columns <NUM> and reheated to the exit temperature and pressure.

In the GPTR <NUM>, the energy in the form of heat removed from the nuclear reactor <NUM> is converted into mechanical work via a thermodynamic cycle whose working fluid (the coolant) is used directly as the coolant for a nuclear reactor core. In the embodiment of the GPTR <NUM> illustrated, the thermodynamic cycle is a simple recuperated Brayton cycle that involves compressing the working fluid, adding heat to the compressed fluid, expanding the working fluid to generate the mechanical work and cooling the fluid before repeating the cycle. However, the simple recuperated Brayton cycle is but one thermodynamic cycle that may be used to convert heat into mechanical work and any cycle, now known or later developed, may be adapted for use in a GPTR.

For instance, in the embodiment shown the compressor <NUM> is driven by the same shaft <NUM>, thus receiving its mechanical energy directly from the turbine <NUM>. This is but one example of how the turbine and compressor power cycle may be effected. Other embodiments using different power cycles with different turbine and compressor configurations to convert the energy in the form of high pressure and temperature of the coolant into mechanical energy are discussed below and in the attachments. For example, many different variation of the Brayton cycle have been recently developed each with differing performance attributes that, depending on the operating conditions of a GPTR, may be more or less suited for use in a GPTR. These include the pre-compression modified Brayton cycle, the recompression modified Brayton cycle, the split-expansion modified Brayton cycle, and the partial cooling modified Brayton cycle. Other thermodynamic cycles that could be adapted to use with coolants are also feasible.

Many gases may be used in GPTR embodiments as a reactor coolant. Preferably, gases that are well-understood in the art and whose properties and material interactions have been fully characterized may be used advantageously in various embodiments. Examples of such gases may include, but are not limited to, carbon dioxide (CO<NUM>), nitrogen (N<NUM>), helium, enriched nitrogen (that is, nitrogen in which the isotopic balance is shifted by enriching nitrogen gas, which typically comprises almost <NUM>% <NUM>N, with <NUM>N, to reduce generation of <NUM>C within the core), neon, argon, or mixtures of such gases. In some embodiments, it may be preferable to use gases that deviate more from ideal-gas behavior, thereby allowing exploitation of the thermodynamic characteristics of such gases (in particular, by using supercritical gases).

In the embodiment <NUM> shown, supercritical CO<NUM> (sCO<NUM>) is the coolant and the CO<NUM> is maintained in a supercritical state throughout the closed-loop coolant circuit formed by the turbine <NUM>, compressor <NUM>, GPTR <NUM> and heat exchangers <NUM>, <NUM>. In an alternative embodiment, a condensing sCO<NUM> cycle is used in which CO<NUM> is liquefied in a cooler/condenser when its pressure is below the critical point. The properties of sCO<NUM> may provide efficiency and simplified plant design when used in a direct power cycle with a pressure tube core. Some advantages of this approach include a high thermodynamic efficiency attainable with sCO<NUM> as a working fluid in more moderate temperature ranges than other possible choices such as helium and argon. The ability to use sCO<NUM> efficiently in a cycle peak temperature range possibly between <NUM> and <NUM> permits a wide range of materials and fuels to be used in the fuel columns <NUM>, allowing reduced materials costs and enhanced materials durability. More moderate operating temperatures of sCO<NUM> may also greatly reduce plant size, as less infrastructural mass is needed to absorb dangerous reactor heat in the event of primary coolant loss. At the working pressures and temperatures mentioned, sCO<NUM> allows the direct power and cooling cycle to be very compact and with reduced pressure losses as the fluid has a high density. This further improves the economics of the GPTR.

The stability of sCO<NUM> as a working fluid across a relatively wide range of temperatures and pressures also leads to a great increase in efficiency during the compression phase of direct gas-cooled reactor designs. When analyzed in a Brayton cycle setting, ideal gas cycles such as one using helium show a linear relationship between temperature/pressure increase and compressor work needed to achieve those increases. This linear nature of curve reflects the highly linear density change of ideal gases, which must be accomplished by compressor work, during compression. Near its critical point, as is expected at the point of the Brayton cycle where compression may be applied, sCO<NUM> working fluid has very low compressibility and therefore the density changes during compression are small. Compression is correspondingly quite efficient, and the amount of compressor work needed for a desired result is much lower than for an ideal gas. Other advantages of using sCO<NUM> over other available cooling media are that CO<NUM> is readily available, easily stored in a condensed form, and has low toxicity. As a primary nuclear coolant, sCO<NUM> also does not affect neutron passage or energy state and it shows low corrosive potential, all of which add to economic feasibility by simplifying design complexity, planning and building costs and operating overhead.

Embodiments of the GPTR <NUM> may be designed to allow for passive decay heat removal in the event of a failure of the coolant system resulting in a loss of coolant (LOC) event. In an LOC event, the GPTR <NUM> is immediately shutdown by bringing the reactor below critical, such as by use of control rods or adding liquid poison to the moderator (e.g., by adding borated water to the calandria). After the shutdown, heat, referred to as decay heat, is still generated for some period of time from the decay of fission products in the nuclear fuel created while the reactor was in operation. In an embodiment, the GPTR <NUM> may permit removal of decay heat via various passive means (conduction, natural convection, radiation) from the fuel through the pressure boundary (i.e., the pressure tubes), without causing damage to either. This is discussed in greater detail with reference to <FIG>. In an alternative embodiment, an active cooling system, such as the cooler <NUM> may be provided to maintain the moderator within a specified temperature and pressure range during normal operation, an LOC event or both.

The design of the fuel columns and calandria core may be optimized to enhance the passive heat removal performance of the GPTR. For example, to prevent the pressure boundary from becoming too hot during normal operation insulation may be provided between the fuel and the pressure boundary. The insulation may be further designed to prevent the surface of the fuel columns from getting too hot even if the fuel experiences a drastic temperature rise. In an embodiment, the insulation may be designed to have lower thermal resistance at higher fuel temperatures, thus acting like a thermal regulator or a non-linear thermal resistor. For example, this may be achieved by incorporating a gas-filled gap between two concentric tubes in the fuel columns <NUM>. As the temperature increases, one or both of the tubes may expand, thus reducing the gap between the tubes and thereby reducing the insulating effect of the gas-filled gap.

Other geometries may be used, according to various embodiments, to enhance the passive heat removal means provided by the pressure tube and calandria core design. For example, multiple fuel columns may be arranged in a single ring or generally annular arrangement (as opposed to a simple grid of rows and columns), to prevent interior fuel columns from obtaining a higher temperature during an LOC event.

For the purposes of this application nuclear fuel includes any fissionable material. Fissionable material includes any nuclide capable of undergoing fission when exposed to low-energy thermal neutrons or high-energy neutrons. Furthermore, fissionable material includes any fissile material, any fertile material or combination of fissile and fertile materials. A fissionable material may contain a metal and/or metal alloy. Fuels may be in a ceramic or composite fuel form. In an alternative embodiment, the fuel may be a metal fuel. Depending on the application, fuel may include at least one element chosen from U, Th, Am, Np, and Pu. The term "element" as represented by a chemical symbol herein may refer to one that is found in the Periodic Table--this is not to be confused with the "element" of a "fuel element". In one embodiment, the amount of actinides in the fuel may include at least about <NUM> wt % U-e.g., at least <NUM> wt %, <NUM> wt %, <NUM> wt %, <NUM> wt %, <NUM> wt %, <NUM> wt %, or higher of U (wt % here being the wt % of U relative to the weight of the actinides in the fuel, i.e., excluding light elements such as O, C. The fuel may further include a refractory material, which may include at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, and Hf. In one embodiment, the fuel may include additional burnable poisons, such as boron, gadolinium, or indium.

The moderate working temperatures of the GPTR provide a further economic benefit through the ability to use uranium fuels with lower temperature tolerance levels, such as (but not limited to) the uranium dioxide fuel with known stainless steels or ceramic cladding. The uranium dioxide may be unenriched natural unration (<NUM> wt. % <NUM>U) or, alternatively, may be enriched to any level as desired, for example enriched with from <NUM>-<NUM> wt. Because of the passive cooling performance discussed with reference to <FIG>, very-high-temperature-tolerant fuels such as coated particle fuels are not required, but may permit operation at higher power densities. Various embodiments use fuel and structural materials with sufficient temperature tolerance to permit both normal operation in coolant and elevated temperatures during passive decay heat removal. Example fuels include, but are not limited to: stainless steel-clad actinide oxide fuel; other stainless steel-clad high-temperature fuels (for example, using actinide nitrides, actinide carbides, or actinide silicides); fuels coated with ceramic such as silicon carbide composites; or specialized coated particle fuels such as TRISO fuel (while less economical, due to their cost, they are suitable for use in various embodiments).

<FIG> illustrate embodiments of a reactor core designed for use in a GPTR such as that discussed with reference to <FIG>. In <FIG>, a vertical cross-sectional view of a reactor core <NUM> is shown through a vertical row of pressure tubes <NUM>. The reactor core <NUM> includes a calandria <NUM>, which is a vessel that is filled with liquid moderator <NUM>, such as heavy water (as mentioned above, in some embodiments light water, or a mix of light and heavy water, may be alternatively used). The calandria <NUM> is surrounded by side shields <NUM>, a top shield <NUM>, and a bottom shield <NUM>. In the embodiment shown, the side shields <NUM> take the form of tanks filled with a liquid shielding agent such as water <NUM>. The top shield <NUM> and bottom shield <NUM> may be of any shield material <NUM>. For example, the top shield <NUM> and bottom shield <NUM> may also be liquid tanks or may be solid components made of some solid shielding material <NUM> as shown such as lead. Both liquid and solid shielding materials are known in the art and any suitable material, now known or later developed, may be used in any of the shields around the calandria <NUM>. In an alternative embodiment, one or more shields may be located within the calandria <NUM> proper to provide additional neutron shielding for the calandria's walls.

A plurality of horizontal fuel columns (six are illustrated in <FIG>) in the form of pressure tubes <NUM> passes through side shield tanks <NUM> and the side walls of the calandria <NUM>. Each pressure tube <NUM> has a plurality of fuel inserts <NUM> stacked axially within its interior, such that fuel inserts <NUM> are within the calandria <NUM>. In the embodiment shown, the fuel inserts <NUM> do not extend into the side shields <NUM>.

The calandria <NUM> and the shields <NUM>, <NUM>, <NUM> are structurally supported by a reactor vault structure <NUM>, which is built using materials and techniques familiar to one having ordinary skill in the art. In the embodiment shown, the vault structure <NUM> is protected from neutron damage by the shields <NUM>, <NUM>, <NUM> such that no neutrons exiting the calandria <NUM> can pass through the vault structure <NUM> without first passing through at least some shield material <NUM>, <NUM>.

One or both ends of each fuel column <NUM> may be provided with access points to access the fuel inserts <NUM> within the fuel columns <NUM>. For example, in an embodiment the pressure tubes <NUM> may be opened on both ends allowing fresh fuel inserts to be pushed in from one end and spent fuel inserts <NUM> to be removed from the other simultaneously. This is illustrated in <FIG> by fuel access ports <NUM> at each end of each pressure tube <NUM>. One advantage of the arrangement of pressure tubes <NUM> shown in <FIG> is that fuel insert <NUM> may be shuffled into and out of the reactor core during operation, since each set of fuel inserts <NUM> may be forced out by insertion of a new set (or using a special tool), taking advantage of the straight-through design. In an alternative embodiment, only one end is accessible and fuel is inserted and removed from that end, as illustrated in <FIG>.

In some embodiments, coolant flow within pressure tubes <NUM> may alternate directions; that is, coolant in some fuel columns <NUM> flows from left to right, while coolant in other columns flows from right to left. In the embodiment shown in <FIG>, adjacent pressure tubes <NUM> have alternating flow directions. Coolant flow through the pressure tubes <NUM> is illustrated in <FIG> by dashed arrows <NUM>. In an embodiment, flow may alternate between adjacent columns along a row or column of pressure tubes. In this way, thermal load may be balanced (since hot exit gases will alternate with cooler inlet gases in each local region within the calandria).

Coolant flow through the fuel columns <NUM> may be a single pass configuration in that gas flows through the fuel columns <NUM> once and then is passed to the power recovery equipment such as the turbine of <FIG>. This embodiment includes manifolds and coolant lines (not shown) on both ends of the fuel columns <NUM>, possibly complicating the design and access to the fuel. A double pass configuration, as shown in <FIG>, could also be used in which the coolant passes through two columns <NUM> before being passed to the power recovery equipment. The dual pass embodiment may make the coolant piping less complicated and improve the ease of refueling the reactor core <NUM>. In <FIG>, each set of adjacent pressure tubes <NUM> are connected by a coolant transfer connection <NUM> on the left side of the figure. In the embodiment shown, this <NUM> is the only coolant piping on the left side of the reactor core <NUM>, leaving more room for fuel handling transfer equipment (note shown). More passes than two may also be used.

In yet another embodiment, a "reheat" configuration may be used. In this configuration, coolant at a first temperature and pressure passes through a first set of one or more pressure tubes <NUM>. The coolant is then passed out of the reactor core <NUM> to the power recovery equipment which outputs a lower pressure coolant stream after removing some energy from the coolant to generate power. The lower pressure coolant stream is then returned to the reactor core <NUM> where it is flowed through a second set of one or more pressure tubes <NUM> at the lower pressure to be heated up again before passing the coolant stream through the remainder of the power equipment. In some designs, this reheat approach permits higher efficiency to be attained for a given peak temperature. This configuration takes advantage of the use of discrete fuel columns <NUM> in the calandria, a design element that is not available is some other nuclear power reactor designs and which permits different coolant pressures to be present in the reactor core <NUM>.

<FIG> illustrates a different embodiment of the reactor core in which the fuel columns <NUM> are provided in a crisscrossing lattice. In the embodiment shown, horizontal rows of pressure tubes <NUM> are alternatingly aligned along the cross-section shown perpendicular to the cross-section. In an alternative embodiment, the alternating horizontal rows are neither parallel or perpendicular to each other, but some angle in between. Such an alternating lattice increases the available space for fuel transfer equipment and coolant piping outside of the calandria <NUM> for each pressure tube <NUM> relative to the design shown in <FIG>. An alternative lattice design may also improve the heat transfer characteristics between the pressure tubes <NUM> and the moderator <NUM>.

<FIG> illustrates yet another reactor core embodiment having a different crisscrossing lattice of fuel columns. In the embodiment shown, the pressure tubes <NUM> are diagonally oriented so that one end of each pressure tube <NUM> is higher than the other. This may facilitate the transfer of the fuel inserts <NUM> into and out of the pressure tubes <NUM>. Similar to <FIG>, Such an alternating lattice may increase the available space for fuel transfer equipment and coolant piping outside of the calandria <NUM> for each pressure tube <NUM> relative to the design shown in <FIG> and improve the heat transfer characteristics between the pressure tubes <NUM> and the moderator <NUM>.

<FIG> illustrates an alternative embodiment of a reactor core designed for use in a GPTR such as that discussed with reference to <FIG>. <FIG>, as in <FIG>, illustrates a cross-sectional view of a reactor core <NUM> shown with some of the components of the core <NUM>. In <FIG>, a calandria <NUM> is provided that is filled with a moderating liquid <NUM> as in <FIG>. However, in <FIG> the fuel columns <NUM> are vertically oriented within the calandria <NUM>. This allows the calandria <NUM> to be made using a tank with no penetrations in the sides or the bottom of the calandria <NUM>.

In the embodiment shown, the fuel columns <NUM> are arranged, for example, in columns and rows within the calandria <NUM>. One or more fuel inserts <NUM> are located at or near the bottom of the fuel columns <NUM>. Coolant <NUM> is injected at the top of each column <NUM> and flows down into the bottom <NUM> of the column <NUM>. The coolant then rises through the column <NUM> while in thermal contact with the fuel inserts <NUM>. Heated coolant <NUM> is removed from the top of the columns <NUM>. Interior piping <NUM> may be provided to channel heated coolant <NUM> through the center of the fuel column <NUM> and incoming, cold coolant through an annular region between the pressure boundary and the fuel inserts <NUM>. Conduits or spaces <NUM> may be provided in the fuel inserts <NUM> to allow the coolant to pass through as shown. Each column <NUM> may further be provided with an inlet port <NUM> and an inlet valve <NUM> and an outlet port <NUM> with an outlet valve <NUM>.

<FIG> illustrates yet another alternative embodiment of a reactor core designed for use in a GPTR having vertically oriented fuel columns. <FIG> illustrates a cross-sectional view through a row of vertically-oriented fuel columns <NUM> in a calandria <NUM>. As described above, the calandria <NUM> is a vessel containing a moderator <NUM>. The vertically-oriented fuel columns <NUM> pass through the top and the bottom of the calandria <NUM> and the volume, or pool, of moderator <NUM> contained in between. Shields (not shown) may be provided within or outside the calandria <NUM> as described above.

Coolant passes through the fuel columns <NUM> cooling the fuel inserts <NUM> within each column <NUM>. In the embodiment shown, coolant enters the reactor core <NUM> and is distributed to each column <NUM> via an intake manifold <NUM> located above the fuel columns <NUM>. The coolant is delivered to the top end of the fuels columns <NUM> and flows downwardly through the fuel columns <NUM>, thus removing heat from the fuel inserts <NUM>. Heated coolant exits the bottom end of the fuel columns <NUM> and is collected by an outlet manifold <NUM>. The outlet manifold <NUM> routes the coolant out of the reactor core <NUM> to a power recovery system (not shown). Dashed arrows <NUM> are provided to illustrate flows of coolant through the intake manifold <NUM> and outlet manifold <NUM> and at various other locations in the coolant circuit including the fuel columns <NUM>, manifolds <NUM>, <NUM> and various coolant piping within the containment vessel <NUM>.

<FIG> illustrates several additional components of a reactor core <NUM>. In the embodiment shown, the calandria <NUM> is penetrated, both above and below, by the fuel columns <NUM>. To protect against any possible leakage from the calandria <NUM>, the reactor core <NUM> includes a containment vessel <NUM> which completely contains the calandria <NUM>. Nuclear reactor containment vessels are known in the art and the containment vessel may be of any suitable design and material. In the embodiment shown, the containment vessel <NUM> includes a vessel without penetrations in the bottom or the sides of the vessel <NUM> at least up to some height above the calandria <NUM> and intake manifold <NUM>. In an embodiment, a containment vessel head (not shown) may be provided to enclose the top of the containment vessel <NUM> through which some or all of the necessary penetrations of the vessel <NUM> are provided.

In the embodiment shown, the fuel inserts <NUM> are both inserted and removed from the top of the fuel columns <NUM>. A fuel insert access port <NUM> is provided on each fuel column <NUM>. In an alternative embodiment, the fuel inserts <NUM> may be removed from the bottom of the fuel columns <NUM>. In this embodiment the containment vessel <NUM> is sized to provide for the removal of fuel inserts <NUM> below the calandria <NUM>.

In the embodiment shown in <FIG>, various valves are provided to automate, or otherwise allow for the control of, the flow of coolant through the reactor core <NUM>. An intake valve <NUM> is provided between the intake manifold <NUM> and each fuel column <NUM>. The intake valves <NUM> may include one or more of: check valves preventing upward flow (back flow) of coolant out of the fuel column <NUM>; flow control valves controlling the flow rate of coolant into the top of the fuel column <NUM>; and isolation valves <NUM> that prevent flow of coolant into the fuel column <NUM>. In an embodiment, a single valve may be provided that performs all of the functions described above (i.e., back flow prevention, flow control, and isolation).

The operation of any of the intake valves <NUM> may be automated. For example, check valves may be automatic valves that prevent all back flow. In addition, valves may be automatically controlled based on monitored conditions of the reactor core <NUM> or other reactor components. For example, flow control valves may be automated to increase or decrease flow through a particular column <NUM> based on a temperature associated with the column <NUM>, such as the column temperature or the temperature of coolant exiting the column <NUM>.

In the embodiment shown, the outlet manifold <NUM> includes a number of outlet valves <NUM>. An outlet valve <NUM> is provided at the bottom outlet of each fuel column <NUM>. The outlet valves <NUM> may include one or more of: check valves preventing upward flow (back flow) of coolant out of the fuel column <NUM>; flow control valves controlling the flow rate of coolant out of the bottom of the fuel column <NUM>; and isolation valves <NUM> that prevent flow of coolant out of the bottom of the fuel column <NUM>. In an embodiment, a single valve may be provided that perform all of the functions described above (i.e., back flow prevention, flow control, and isolation).

Similar to the intake valves <NUM>, the operation of any of the outlet valves <NUM> may also be automated. For example, check valves may be automatic valves that prevent all back flow. In addition, valves may be automatically controlled based on monitored conditions of the reactor core <NUM> or other reactor components. For example, flow control valves may be automated to increase or decrease flow through a particular column <NUM> based on a temperature associated with the column <NUM>, such as the column temperature or the temperature of coolant exiting the column <NUM>. The outlet valves <NUM> of the manifold <NUM> may be serially oriented as shown or may be in parallel as the intake valves <NUM> are represented in the inlet manifold <NUM>. Likewise, the intake valves <NUM> may be serially oriented or in parallel.

Flow of coolant into and out of the reactor core <NUM> may be further controlled by containment valves <NUM> in the coolant inlet and outlet piping. These valves <NUM> may be located external to the containment vessel or within the containment vessel or vessel head. For example, in an embodiment the containment valves <NUM> are located at the point in the piping where the coolant enters and exits the building containing the reactor core <NUM>. Again, the containment valves <NUM> may include one or more of: check valves preventing upward flow (back flow) of coolant into and out of the reactor core <NUM>; flow control valves controlling the flow rate of coolant into and out of the reactor core <NUM>; and isolation valves <NUM> that prevent flow of coolant into and out of the reactor core <NUM>.

Further safety and control is provided by one or more moderator pressure relief valves <NUM>. A moderator pressure relief valve <NUM> automatically opens when the moderator <NUM> in the calandria <NUM> reaches a selected pressure. The overpressure may be vented into one or more tanks or other vessels. In the embodiment shown, the moderator pressure relief valve <NUM> vents the pressure into a reflood tank <NUM>. In this embodiment, the reflood tank <NUM> is provided to provide additional cooling capacity in the event of an emergency. In an embodiment, the reflood tank <NUM> contains a reflood fluid <NUM>, such as light water, that can add thermal capacity (by replacing moderator lost to boiling) to the moderator <NUM> in the calandria <NUM>. Alternatively, the reflood fluid <NUM> may be the same as the calandria moderator <NUM>.

In the event of an overpressure condition in the calandria <NUM>, the illustrated pressure relief valve <NUM> and reflood tank <NUM> configuration causes the reflood fluid <NUM> to flow into the calandria <NUM> and replace the original moderator <NUM>. Flow of reflood fluid <NUM> from the reflood tank <NUM> into the calandria <NUM> may be further controlled by a reflood outlet valve <NUM>, as shown. In an embodiment, the valve <NUM> may be a check valve to prevent backflow into the reflood tank <NUM>. In an embodiment, the pressure relief valve <NUM> and reflood outlet valve <NUM> may be controlled to maintain the level of moderator in the calandria <NUM> at a desired level. Solid arrows <NUM> are provided to illustrate direction of flow of reflood fluid into and moderator out of the reflood tank <NUM> at selected locations within the fluid/moderator circuit created by the calandria <NUM> and reflood tank.

The reflood tank <NUM> may be sized to contain a volume of fluid <NUM> sufficient to replace all of the moderator <NUM> in the calandria <NUM>. The reflood tank <NUM> may be within the containment vessel <NUM> as shown or may be outside of the containment vessel <NUM>, such as located vertically above the containment vessel.

Alternatively, the reflood tank <NUM> may be sized to contain a volume of reflood fluid <NUM> sufficient to both replace all of the moderator <NUM> in the calandria <NUM> and to fill the containment vessel <NUM>. In this embodiment, the reflood tank <NUM> may be above the containment vessel <NUM> so that gravity will cause the fluid <NUM> to flow into the containment vessel <NUM> upon the opening of a flow control valve <NUM>.

In yet another embodiment, the pressure relief valve <NUM> may vent pressure into an optional second tank <NUM>. This second tank <NUM> and a second pressure relief valve <NUM> are illustrated as optional via dashed lines in <FIG>. The second tank <NUM> may be a second reflood tank or a simple overflow tank as shown.

The reflood tank <NUM> and the second tank <NUM> may be single tanks as shown or may be multiple, different tanks fluidly connected in serial, parallel, or both. The tanks may be pressure vessels or may be open tanks.

In alternative embodiments, any of the valves illustrated and discussed above may be replaced or supplemented with one or more non-moving flow control components. This may include venturi flow limiters or orifice plates, for example. Such non-moving flow control components may be included at the fuel column inlets and/or outlets or at any location along the sCO<NUM> coolant circuit.

<FIG> illustrates a cross-sectional view of an embodiment of a nuclear fuel insert, such as insert <NUM> of <FIG>, suitable for use in a fuel column. In the embodiment shown, <NUM> individual hollow fuel tubes <NUM> are aligned in columns arranged in an annulus about the longitudinal axis <NUM> of the insert <NUM>. The fuel tubes <NUM> are within an annular void space <NUM> through which coolant flows.

In an embodiment, the center void space <NUM> of the hollow fuel tubes <NUM> is filled with helium, which acts as a buffer for thermal expansion and allows for dimensional change in annular fuel column <NUM> as fission product gases build up over core lifetime. The center void space <NUM> also lowers the peak fuel temperature and provides space for fission product gases to collect. The helium may be flowing (requiring a means for circulation and a cladding - not shown) or the fuel tubes <NUM> may be closed-ended, thus trapping the helium within the tube <NUM>.

A central graphite column <NUM> may be provided as shown to act as a secondary heat sink during passive heating (for example, during shutdown or loss of coolant). In an alternative embodiment, the central column <NUM> may be of any other suitable material or combination of materials such as silicon carbide and other ceramics/composites. The central column <NUM> may or may not be provided with an exterior cladding layer. Similarly, an outer graphite annular sleeve <NUM> may be provided to assist in thermal management and in moderation of fast neutrons and to provide structure and ease of fuel handling. Again, in an alternative embodiment, the sleeve <NUM> may be of any other suitable material or combination of materials such as silicon carbide and other ceramics/composites and may or may not be provided with an outer cladding layer.

The fuel insert may also include one or more structural elements (not shown). For example, one or more structural elements may run through the insert axially, either inside the central column, replacing the central column, or as part of the external sleeve. Such a structural element might be a rod, tube, or cable. The structural element may link different pieces of the fuel insert, provide structural support to the other components, bear the weight of the fuel, and aid in fuel handling.

At the ends of the insert <NUM> and/or at various locations through the insert <NUM>, not shown, a framework or other structural elements may be provided to retain the various components in their relative locations. The ends of the inserts <NUM>, however, will at least provide for coolant flow from one insert's coolant flow region <NUM> to an adjacent insert's coolant flow region <NUM>. This allows multiple inserts <NUM> arranged along a common longitudinal axis to form a coolant flow path through the coolant regions <NUM> of the adjacent inserts <NUM> from one end of the insert assembly to the other.

In an embodiment, the ends of the inserts <NUM> may be provided with complimentary connectors allowing two inserts to be connected together to form an insert assembly. The connectors may prevent leakage of the coolant out of the insert assembly. In an alternative embodiment the connectors are not fluid tight and some coolant leakage may be allowed. In yet another embodiment, no connectors are provided that the inserts are simply maintained in an abutting arrangement with the coolant regions <NUM> of adjacent inserts aligned with each other. The design shown in <FIG> provides for good thermal management and allows a high power density to be maintained.

<FIG> illustrates a cross-sectional view of an alternative embodiment of a nuclear fuel insert suitable for use in a fuel column. Similar to that shown in <FIG>, the insert <NUM> includes a number of individual fuel tubes <NUM>. Again, the fuel tubes <NUM> may be filled with a gas such as helium, filled with a liquid at operational temperature such as a salt, or may be solid rods of nuclear material. The fuel tubes <NUM> are arranged more densely than in <FIG> and <FIG> within a cylindrical void <NUM> of the insert <NUM> through which coolant gas flows.

In one aspect of the embodiment <NUM>, coolant region <NUM> is defined on its outer edge by a graphite tube <NUM>. As shown in <FIG>, the graphite tube <NUM> is provided with an optional internal coating layer <NUM>.

<FIG> illustrates a cross-sectional view of yet another embodiment of a nuclear fuel insert suitable for use in a fuel column. In the embodiment shown, the primary component of the insert <NUM> is a solid rod <NUM> of nuclear fuel. The solid rod <NUM> is provided with some number, seven are shown, of flow channels <NUM> that penetrate the rod. The flow channels <NUM> may be simple, straight paths, circular in cross-section (as shown), that run the length of the insert <NUM>. In alternative embodiments, the flow channels <NUM> may be any shape in cross-section and may even vary in size or shape along the length of the insert. In addition, in alternative embodiments the flow paths may be any shape of path, such as spiral or angular, through the rod <NUM>. Such embodiments may be 3D printed or made in conventionally in multiple sections and then combined to form the rod <NUM>.

Exposed surfaces of the nuclear fuel may be provided with a protective layer <NUM> in the flow channels <NUM>, on the outer surface, or both. In an embodiment, zirconium or an alloy of zirconium may be used as the protective layer <NUM>. The protective layer may be a structural element, such as a tube, or may simply be a non-structural coating or cladding applied to or deposited on the surface to be protected.

The inserts <NUM>, <NUM>, <NUM> discussed above may be any desired length. For example, in an embodiment an insert's length matches the full operational length of fuel to be inserted in a fuel column in the calandria. In this configuration one can either have shorter fuel tubes, optionally linked by a connecting structure, or the fuel tubes themselves can also span the entire length of the fuel insert.

In an alternative embodiment, the inserts' length is selected so that an integral number of fuel inserts are required for each fuel column. Depending on the length, however, intermediate structural components (not shown) may be provided on the inserts or in the fuel column, for example to prevent sagging of the nuclear fuel tubes <NUM>, <NUM> when the longitudinal axis of the insert is horizontally aligned, or to prevent thermal bowing, vibration, and/or wear in both vertically and/or horizontally aligned inserts.

Further, in any embodiment the ends of the inserts can include additional features, such as structural features, shock absorbers, flow control devices, instrumentation, pressure boundaries, and neutron shielding.

<FIG> illustrate cross-sectional views of an embodiment of a nuclear fuel insert that incorporate a gas gap for use in a fuel column. <FIG> shows a fuel insert 800A similar to that shown in <FIG> in that the primary component of the insert 800A is a solid rod <NUM> of nuclear fuel. The solid rod <NUM> is provided with some number, seven are shown, of flow channels <NUM> that penetrate the rod <NUM>. The flow channels <NUM> may be simple, straight paths circular in cross-section (as shown) that run the length of the insert 800A. In alternative embodiments, the flow channels <NUM> may be any shape in cross-section and may even vary in size or shape along the length of the insert. Exposed surfaces of the nuclear fuel may be provided with a protective layer <NUM> in the flow channels <NUM>, on the outer surface, or both. In an embodiment, zirconium or an alloy of zirconium may be used as the protective layer <NUM>. The protective layer may be structural element, such as a tube, or may simply be a coating applied to the surface to be protected.

Unlike the insert 700A of <FIG>, the insert 800A of <FIG> incorporates the gas gap <NUM> into the insert 800A so that it can be used in a pressure tube (not shown) that need not have a gas gap. In this embodiment, each insert 800A will have its own, independent gas gap <NUM> with a trapped gas layer.

<FIG> shows a cross-sectional view of an alternative embodiment of a nuclear fuel insert similar to that shown in <FIG> but incorporating a gas gap or other insulating structure. The insert 800B includes a number of individual fuel tubes <NUM>. Again, the fuel tubes <NUM> may be filed with a gas such as helium, filled with a liquid at operational temperature such as a salt, or may be solid rods of nuclear material. In this embodiment, they are illustrated as solid rods of fuel. The fuel tubes <NUM> are arranged within a cylindrical void <NUM> of the insert 800B through which coolant gas flows. In one aspect of the embodiment 800B, coolant region <NUM> is defined on its outer edge by a tube <NUM> of graphite, zirconium alloy, or other material.

In <FIG>, the gas gap <NUM> is between the exterior surface of the rod <NUM> (or the protective layer <NUM> that forms the exterior surface of the rod, if such as layer is used) and an outer tube <NUM>, which is enclosed to prevent flow of gas within the gas gap <NUM>. In <FIG>, the gas gap <NUM> is between the exterior surface of the graphite tube <NUM> and the outer tube <NUM>, which is enclosed at both ends to prevent the gas from escaping the gas gap <NUM>. These embodiments allow the pressure tubes to be a more simple construction and, thus, potentially less expensive, and permits regular replacement of the insulation when the fuel is replaced.

In an alternative configuration, the outer tube <NUM> can be omitted in either embodiment described above, and the pressure tube can serve the same function as the outer tube (i.e. forming the outer boundary of the gas gap/insulating layers).

The gas gap <NUM> may be similar to that described above. In the embodiment shown, the outer tube <NUM> is separated from the other internal components of the inserts 800A, 800B by the gas gap <NUM>. The outer tube <NUM> may be of any suitable material such as, for example, graphite or a zirconium alloy. The gas gap <NUM> is an annular region filled with stagnant gas, such as high pressure CO<NUM>. Other suitable insulating gases include nitrogen (N<NUM>), helium, enriched nitrogen, and argon. The stagnant gas can also be connected to the coolant system and use the same gas. The gas gap <NUM> acts as a thermally insulating region between outer tube <NUM> and the internal components of the fuel insert. The thermal performance of a fuel column can be controlled to meet a desired specification through the selection of the insulating gas and the thickness of the gas gap <NUM>. This allows the fuel columns, as a whole, to be designed to specific LOC events so that sufficient heat transfer is obtained through the pressure tube to allow for passive cooling during the LOC event.

A standoff structure may be provided within the gas gap <NUM> as shown. In an embodiment, the standoff structure may be made from an embossed sheet of thin structural metal material, as described above, to ensure that the width of the gap is maintained throughout the length of the insert. The standoff structure may be of any suitable design including, but not limited to, ribs, fins, or protuberances provided on the exterior of the rod <NUM> or the graphite tube <NUM>. In an alternative embodiment, the standoff structures may be some number of solid, insulating spheres evenly spaced about the exterior of the rod <NUM> or the graphite tube <NUM> and between the rod <NUM> or the graphite tube <NUM> and the interior of the tube <NUM>.

In an embodiment, the standoff structure may be flexible so that the thickness of the gas gap <NUM> is allowed to shrink as the temperature of the internal components increases. This, in turn, reduces the insulating effect of the gas gap <NUM> which increases the thermal flux from the interior of a fuel column to the pressure tube and, thus, to the liquid moderator in the calandria.

<FIG> illustrates a cross-sectional view of an alternative embodiment of a nuclear fuel insert incorporating a standoff structure suitable for use in a fuel column. Similar to that shown in <FIG>, the insert <NUM> includes a number of individual fuel tubes <NUM>. Again, the fuel tubes <NUM> may be filed with a gas such as helium, filled with a liquid at operational temperature such as a salt, or may be solid rods of nuclear material.

In one aspect of the embodiment <NUM>, coolant region <NUM> is defined on its outer edge by a graphite or zirconium alloy tube <NUM>. As shown in <FIG>, the graphite tube or zirconium alloy <NUM> is provided with an optional internal coating layer <NUM>.

The insert <NUM> mainly differs from that of <FIG> by the additional of several standoff elements <NUM> around the circumference of the insert <NUM>. When assembled into a pressure tube that does not have an integrated gas gap, a gas gap will be created by the standoff elements <NUM>. Any type or shape of standoff structure <NUM> may be used. For example, in an embodiment the standoff structure <NUM> is a spiral winding that runs the length of the insert <NUM>. In yet another embodiment, the standoff structure is a series of discontinuous ribs or protuberances spaced about the exterior of the insert <NUM>.

<FIG> illustrates a cross-sectional view of a pressure tube suitable for containing fuel inserts and acting as the exterior pressure-boundary of a fuel column. <FIG> illustrates only the pressure tube portion <NUM> of the fuel column. Specifically, the pressure tube <NUM> illustrated is exaggerated to show the different components and regions within the pressure tube <NUM>. In the embodiment shown, the pressure tube <NUM> includes an outer, structural tube <NUM>, a gas gap <NUM> separating the structural tube <NUM> from a guide sleeve <NUM>. An optional material layer <NUM> between the gas gap <NUM> and guide sleeve <NUM> may also be provided as shown. The material layer <NUM> is illustrated as a single layer, but may generally be any number of layers of various compositions as needed to maintain the gas gap or reduce thermal conductance. For example, a silica fabric may be used as one layer <NUM>.

The outer component <NUM> of the pressure tube <NUM> is the structural pressure boundary and is formed by a structural tube <NUM> of material such as steel, a zirconium or aluminum alloy, a ceramic, or a composite material. In an embodiment, the structural tube <NUM> is made of a material such as HT-<NUM> steel, or a high-temperature ferritic, martensitic, or stainless steel. Relative to the thickness of the other components of the pressure tube <NUM>, the structural tube <NUM> is likely to be thicker than the other layers as it has to withstand the high-pressure differential between the high pressure of the coolant in within the pressure tube <NUM> and the low-pressure liquid moderator outside of the pressure tube.

The structural tube <NUM> may further be provided with a cladding (not shown) to prevent interaction of the structural material with the liquid moderator in the calandria which will be in contact with the exterior surface of the pressure tube <NUM> when in use.

The structural tube <NUM> is separated from the other internal components of the pressure tube <NUM> by a gas gap <NUM>. The gas gap <NUM> is an annular region filled with stagnant gas, such as high pressure CO<NUM>. Other suitable insulating gases include nitrogen, helium, enriched nitrogen, and argon. The gas gap <NUM> acts as a thermally insulating region between structural tube <NUM> and the internal components of the fuel column. The thermal performance of the pressure tube <NUM> can be controlled to meet a desired specification through the selection of the insulating gas and the thickness of the gas gap <NUM>. This allows the fuel columns, as a whole, to be designed to specific LOC events so that sufficient heat transfer is obtained through the pressure tube <NUM> to allow for passive cooling during the LOC event.

A standoff structure (not shown) may be provided within the gas gap <NUM>, such as a tube made from an embossed sheet or sheets of thin structural metal material, or a tube made from a porous ceramic or aerogel material, to ensure that the width of the gap is maintained throughout the length of the pressure tube <NUM>. In an embodiment, the standoff structure may be flexible so that the thickness of the gas gap <NUM> is allowed to shrink as the temperature of the internal components increases. This, in turn, reduces the insulating effect of the gas gap <NUM> which increases the thermal flux from the interior of a fuel column to the structural tube <NUM> and, thus, to the liquid moderator in the calandria.

The guide sleeve <NUM> is situated within the inner diameter of pressure tube <NUM>. It is provided to contact and guide the fuel inserts as they are installed and removed and retain the coolant as it flows through the central region <NUM>. As mentioned above, one or more protective layers <NUM> maybe be provided between the guide sleeve <NUM> and the gas gap <NUM>, in particular between any stand-off structure within the gas gap <NUM> and the guide sleeve <NUM> to prevent damage as the guide sleeve expands and contracts against the stand-off structure.

<FIG> shows a detailed cross-sectional view of a portion of the interface <NUM> between a nuclear fuel insert and a pressure tube. In the embodiment shown, the pressure tube includes an outer structural tube <NUM> attached to the exterior surface which is in contact with the moderator <NUM> (e.g., water) in the calandria. Next to the interior surface of the structural tube <NUM> is a protective layer <NUM>. In an embodiment, the protective layer <NUM> is a thin zirconium alloy wrap. Next is the gas gap <NUM>, this time illustrated with the standoff structure <NUM> in the gas gap <NUM>. As discussed above, the standoff structure <NUM> may be an embossed or corrugated zirconium alloy sheet.

Next, a second thin protective layer <NUM> is provided, for example of silica fabric <NUM>. The second thin protective layer <NUM> prevents contact between the standoff structure <NUM> and a guide tube <NUM>. In an embodiment, the guide tube <NUM> is made of a material such as zirconium alloy or stainless steel. The gas gap <NUM> within which sheet <NUM> is disposed, which region lies between silica fabric <NUM> and zirconium wrap <NUM>, provides a static gas gap for thermal insulation between pressure tube <NUM> and the fuel insert.

Next, a second, thin gas gap <NUM> is shown between the guide tube <NUM> and a graphite sleeve <NUM>, which forms the exterior of the fuel insert. This second, thin gas gap <NUM> represents the clearance fit between a removable fuel insert and the pressure tube. Depending on the amount of clearance between the two and the positioning of the two, the second, thin gas gap <NUM> may vary in thickness and, in some locations, the guide tube <NUM> and the graphite sleeve <NUM> may be in direct contact.

Graphite sleeve <NUM> surrounds a cylindrical void <NUM> within which the fuel tubes <NUM> are arranged. Again, the fuel tubes <NUM> are illustrated as having an inner region <NUM>, which may contain a different material such as a stagnant gas, a liquid, or a solid. In an embodiment, this region may be filled with the same coolant as flowing through the coolant region <NUM> of the insert. In an embodiment, the fuel tubes <NUM> may be porous and penetrate into the inner region <NUM>. In this embodiment, the ends of the fuel tubes may or may not have openings to facilitate a strong flow of coolant through the center <NUM> of the tubes <NUM>, in addition to the flow in the main gas flow region <NUM> outside of the fuel tubes <NUM>.

The exterior surface <NUM> of the fuel tubes <NUM> is exposed to the coolant flow in the fuel insert. Coolant flows through void <NUM> as it cools the fuel tubes <NUM> and thereby gains heat to be used in the direct power cycle.

In the above embodiments of the pressure tubes, the gas gaps or other insulating structures are provided as part of the pressure tubes. In alternative embodiments, the gas gap or standoff structures, which create the gas gap when assembled, could be incorporated into the fuel insert instead.

<FIG> are provided to illustrate the thermal performance of an embodiment of a fuel column in a reactor core during an LOC event. <FIG> illustrates an embodiment of a fuel column <NUM> and <FIG> illustrates the modelled thermal performance of that fuel column <NUM> during an LOC event. The exterior of the fuel column <NUM> is in contact with calandria water <NUM>.

The fuel column <NUM> includes a pressure tube that includes an inner guide tube <NUM>, an embossed zirconium alloy sheet <NUM>, and a structural tube <NUM> (with an insulating static gas gap <NUM> between guide tube <NUM> and the structural tube <NUM> that contains the standoff sheet <NUM>).

Within the pressure tube is a fuel insert that includes a central graphite rod <NUM> surrounded by void <NUM> through which coolant flows. Within the coolant flow region <NUM> is an annulus of fuel tubes <NUM> arranged in a ring. An optional outer graphite annulus <NUM> forms the exterior of the fuel insert. A thin static gas gap <NUM> between the fuel insert and the pressure tube is also provided in this embodiment for modeling purposes.

In such an arrangement, during an LOC event a passive thermal conductance path <NUM> is established between fuel tubes <NUM> (where fission product decay heat and residual heat from operations are present and temperatures are at their highest). The large volume of cool moderator <NUM> allows heat from fuel tubes <NUM> to be removed through cooling <NUM> by moderator <NUM>.

<FIG> illustrates the results of modelling an LOC event. The model was developed as a finite element model of a representative reactor using the software suite ABAQUS™ FEA by Dassault Systems Simulia Corporation. The representative reactor design used has a <NUM>-fuel column single ring bundle. The model assumed the peak power and peak temperature conditions occurred at the same time. The operational initial coolant temperature was <NUM>, the initial moderator temperature was <NUM>, and the fuel temperature at the exterior surface of the pressure tube was <NUM> at the time of the LOC. The average volumetric power density in the fuel was <NUM>×<NUM><NUM> W/m<NUM>. The cross-sectional area of each fuel column was <NUM>×<NUM>-<NUM> m<NUM>. The model was designed to approximate a worst-case scenario and ignored edge effects, modeling only the radial heat dissipation.

The individual fuel columns were modeled as an insert of UO<NUM> fuel, within a pressure tube of stainless steel (having the properties of SS316), a zirconium-alloy pressure tube (with properties of alpha phase zircalloy-<NUM>), a graphite layer, and an insulating layer with its thermal conductance set to permit no more than <NUM>% thermal power loss during operation.

In the model, a loss of coolant after sustained power operations was simulated in which radiative and conductive passive heat transfer to the calandria moderator <NUM> are the only available mechanisms to remove decay heat from fuel. Graph <NUM> illustrates a computed thermal performance of the embodiment upon a total loss of coolant within the fuel column <NUM>. This loss of coolant was modeled by the instantaneous replacement of coolant with a vacuum at time t=<NUM>, interrupting sustained full power operation conditions. The x-axis <NUM> shows time elapsed from loss of coolant. The y-axis <NUM> shows peak fuel temperature in degrees Celsius.

Shortly after loss of coolant (and subsequent reactor shutdown, which ends the addition of heat from fission), fuel temperature increases rapidly <NUM> to a peak <NUM> of about <NUM> (depending on the type of fuel used, this is well below the temperature at which fuel failure will occur). The temperature then drops rapidly <NUM> as residual heat from reactor operations is removed, achieving a local minimum <NUM> about four minutes after loss of coolant; at this point the buildup of decay heat from decay of fission products causes the temperature to gradually increase <NUM> before slowly dropping over many hours as decay heat generation drops off. As can be seen, the modeling indicates that the GPTR embodiment of <FIG> achieves a robust passive heat removal capability that avoids fuel failure after a total loss of coolant, relying on entirely passive heat removal.

An advantage of GPTR designs described herein is the separation of nuclear engineering requirements centered on reactivity control from thermodynamics requirements centered on driving the power cycle and removing heat in loss of coolant situations. Another advantage of the GPTR designs described herein is the separation between the systems used to drive the power cycle (using the high pressure, high temperature coolant gas), and the systems used to safely remove decay heat (using the low pressure, low temperature moderator water). Unlike pressurized water reactors, for example, where the primary coolant is also the chief moderator, in the GPTR embodiments described herein the coolant is essentially nonreactive in a nuclear sense (that is, has very low reactivity worth). This allows the thermal design to be optimized separately from the reactivity management. That is, in the GPTR designs herein, there is a relatively small effect on the reactivity of the GPTR in the event of a loss of the coolant. Moderation is performed primarily by low-temperature, low-pressure calandria fluid (e.g., heavy water). Because there is no significant change in moderator temperature during reactor operations, there is little effect on overall reactivity from the calandria moderator's negative thermal coefficient of reactivity (αT).

<FIG> illustrates a cross-sectional view of an alternative embodiment of a fuel column designed to receive cooled coolant and remove heated coolant from the same end of the fuel column such as may be used with the reactor core illustrated in <FIG>. The fuel column <NUM> includes a pressure tube <NUM>, such as the pressure tube <NUM> shown in <FIG>. Within the pressure tube <NUM>, is a fuel insert <NUM> contained within an interior insert retaining tube <NUM>. The interior insert retaining tube <NUM> is separated from the interior surface of the pressure tube <NUM> by an annular region <NUM> that acts as a coolant flow path. In an embodiment, this region <NUM> receives the incoming, cool coolant and conveys the coolant to the distal end of the fuel column <NUM>.

The fuel insert <NUM> is similar to that shown in <FIG>. The insert <NUM> includes a number of fuel tubes <NUM> arranged within the annular region <NUM> of the insert's tube <NUM>. The fuel insert's tube <NUM> may be graphite, as discussed above or a multilayer construction of different materials.

In an embodiment, the interior insert retaining tube <NUM> may be contiguous and forms a gas barrier between the cool coolant in the exterior flow region <NUM> and the heated coolant flowing through the interior region <NUM> of the insert. In an alternative embodiment, the outer tube <NUM> of the fuel insert <NUM> acts as the gas barrier.

<FIG> illustrates a cross-sectional view of yet another alternative embodiment of a fuel column provided with a moderator around the outside of the fuel column. In this embodiment, the moderator is a sheath of moderating light water that provides an extra layer of moderating material separate and independent from the moderation provided by the calandria's moderator. The fuel column <NUM> includes a pressure tube <NUM>, such as the pressure tube <NUM> shown in <FIG>. Within the pressure tube <NUM>, is a fuel insert <NUM> contained within an interior insert retaining tube <NUM>. The interior insert retaining tube <NUM> is separated from the interior surface of the pressure tube <NUM> by an annular region <NUM> that contains the light water. In an alternative embodiment, any other gaseous, liquid, or solid moderator may be used.

In the embodiment shown, the moderator in region <NUM> may be stagnant and trapped within the fuel column <NUM>. However, in alternative embodiments, moderator in region <NUM> actively or passively circulated within the column <NUM>, or flowing through the region <NUM> and the column <NUM>.

One aspect of designing a power conversion cycle for the GPTR is the mismatch between the power cycle's peak pressure and the pressures of the GPTR. Peak pressures for CO<NUM> power cycles typically range from <NUM>-<NUM> psi (<NUM>-<NUM> MPa). Increasing peak pressure in the power cycle generally improves cycle efficiency and increases compactness in the power generating system. However, increasing pressure in the GPTR fuel columns will increase the stored energy present, which will require larger amounts of structure to safely retain the pressure and also increase the expected rate of corrosion caused by the coolant. While it is not preferable to reduce the power cycle pressure, it is possible to modify the Brayton cycle and where the GPTR is incorporated into the cycle to achieve an efficient power cycle.

<FIG> illustrates a schematic of an embodiment of a split-expansion modified Brayton cycle incorporating a GPTR. In the embodiment shown, the pressure of the sCO<NUM> provided to the GPTR is reduced by employing "split expansion" by placing a pre-expansion turbine T1 <NUM> upstream of the GPTR <NUM> in addition to the turbine T2 <NUM> following the GPTR <NUM> (hence "split-expansion" in that the expansion turbines are separated by the GPTR). This allows the pressure of the coolant in the GPTR <NUM> to be less than the maximum operating pressure of the cycle <NUM>.

In the split-expansion Brayton cycle <NUM> shown, a turbine T1 <NUM> is provided before the GPTR <NUM> and a second turbine T2 <NUM> is provided after the GPTR. The output of the second turbine <NUM> is passed through a high-temperature recuperator <NUM> and a low-temperature recuperator <NUM> after which the coolant stream is split. The split streams are then passed to two different, independent recompression legs. The first leg further cools the stream using a cooler <NUM> and the cooled stream <NUM> is then passed to the first of two compressors <NUM>, a low-temperature compressor designated compressor C4. The compressed output <NUM> is passed to the low-temperature recuperator <NUM>. The second leg is passed directly to the second compressor <NUM>, a high-temperature compressor designated C3. The output of the recompression legs is recombined at the inlet to the high temperature recuperator <NUM> and then fed back into the first turbine T1 <NUM>.

In the embodiment shown, the compressors and turbines are on the same shaft <NUM>. This, however, is optional as illustrated in <FIG>, <FIG>, and <FIG> below. The shaft <NUM> is also shown operating an electrical generator <NUM> although any power recovery system may be used.

The GPTR <NUM> may be of any configuration or embodiment described above. The GPTR <NUM> is shown as including a calandria <NUM> filled with a moderator and having some number, two are shown, of fuel columns <NUM> containing nuclear fuel. The high-pressure turbine T1's outlet coolant stream <NUM> is passed to the GPTR <NUM> where it flows through the fuel columns <NUM>. A heated coolant stream <NUM> exits the GPTR <NUM> and is passed to the inlet of the low-pressure turbine T2, <NUM>.

Note that a split-expansion embodiment of the simple recuperated Brayton cycle shown in <FIG> may also be used. In this embodiment, in addition to the single turbine <NUM> shown, a second turbine (similar to the turbine T2 <NUM> of <FIG>) upstream of the GPTR <NUM> is provided. Again, similar to <FIG> the two turbines may be on the same shaft or, similar to <FIG> and <FIG> below, the different turbines may be on different shafts one driving the compressor <NUM> and one driving the generator <NUM>.

<FIG> illustrates yet another embodiment <NUM> of a modified Brayton cycle, referred to herein as the pre-expansion modified Brayton cycle that incorporates a GPTR. In the embodiment shown, a sole turbine <NUM> is placed ahead of the GPTR <NUM> so that lower pressure sCO<NUM> is delivered to the GPTR <NUM> for use as the primary coolant, but the pressure in the GPTR is low enough that there is no second turbine after coolant exits the GPTR. The single turbine T1 <NUM> drives the electric generator <NUM> via the same shaft <NUM> that drives the two compressors <NUM>, <NUM>.

In the embodiment shown, the heated sCO<NUM> from the GPTR <NUM> is passed through two heat exchangers. The first, a high-temperature recuperator <NUM>, heats the pressurized sCO<NUM> prior to its delivery to the turbine <NUM> and the second, a low-temperature recuperator <NUM>, which heats the output <NUM> of one of the split streams. The sCO<NUM> stream, after passing through the second heat exchanger, is then split into two streams. The two streams are passed to different recompression legs (as described with reference to <FIG>) one with a cooler <NUM> that outputs a low-temperature coolant stream <NUM> and one that is passed directly to the high-temperature compressor <NUM>. The output of the two recompression legs is ultimately recombined and passed to the first heat exchanger <NUM> before going into the turbine <NUM>, again as described with reference to <FIG>.

<FIG> illustrates an alternative schematic of an embodiment of a split-expansion modified Brayton cycle incorporating a GPTR. In this embodiment, the cycle <NUM> is designed such that one turbine (turbine T1 <NUM>) produces enough power to drive the compressors and the other turbine powers the generator on separate shafts. This permits the turbine driving the compressors to be varied in speed, since it does not need to be synchronized to the generator's frequency.

In the split-expansion Brayton cycle <NUM> shown, a turbine T1 <NUM> is provided before the GPTR <NUM> and a second turbine T2 <NUM> is provided after the GPTR. The second turbine T2 <NUM> drives an electrical generator <NUM> via a generator shaft <NUM>. While the shaft <NUM> is shown operating an electrical generator <NUM>, any power recovery system may be used.

The first turbine T1 <NUM> drives the compressors <NUM>, <NUM> by a second compressor shaft <NUM>. The output of the second turbine <NUM> is passed through a high temperature recuperator <NUM> and a low temperature recuperator <NUM> after which the coolant stream is split. The split streams are then passed to two different, independent recompression legs. The first leg further cools the stream using a cooler <NUM> and the cooled stream <NUM> is then passed to the first of two compressors <NUM>, designated compressor C4. The compressed output <NUM> is passed to the low temperature recuperator <NUM>. The second leg is passed directly to the second compressor <NUM>, designated C3. The output of the recompression legs are recombined at the inlet to the high temperature recuperator <NUM> and then fed back into the first turbine T1 <NUM>.

<FIG> illustrates yet another alternative schematic of an embodiment of a split-expansion modified Brayton cycle incorporating a GPTR. In this embodiment, the cycle <NUM> is designed such that two, independent high-pressure turbines 1808A, 1808B are provided on separate shafts, one for each compressor. This separates the two compressors and permits the turbines driving each compressor and the generator to be independently varied in speed, providing additional operational flexibility.

In the split-expansion Brayton cycle <NUM> shown, the output low-temperature coolant from the high-temperature recuperator <NUM> is split and passed to each of a first high-pressure turbine T1A 1808A and a second high-pressure turbine T1B 1808B. The output coolant streams from each of the high-pressure turbines 1808A and 1808B are combined into a GPTR inlet coolant stream <NUM> and then passed to the GPTR <NUM>.

A low-pressure turbine T2 <NUM> is provided after the GPTR and receives the GPTR output heated coolant stream <NUM>. The low-pressure turbine T2 <NUM> drives an electrical generator <NUM> via a generator shaft <NUM>. While the shaft <NUM> is shown operating an electrical generator <NUM>, any power recovery system may be used.

The first high-pressure turbine T1A 1808A drives the high-temperature compressor <NUM> by a second shaft <NUM>. The second high-pressure turbine T1B 1808B drives the low-temperature compressor <NUM> by a third shaft <NUM>.

The output of the low-pressure turbine <NUM> is passed through a high temperature recuperator <NUM> and a low temperature recuperator <NUM> after which the coolant stream is split. The split streams are then passed to two different, independent recompression legs. The first leg further cools the stream using a cooler <NUM> and the cooled stream <NUM> is then passed to the first of low-temperature compressor <NUM>, designated compressor C4. The compressed output <NUM> is passed to the low-temperature recuperator <NUM>. The second recompression leg is passed directly to the high-temperature compressor <NUM>, designated C3. The output of the recompression legs are recombined at the inlet to the high-temperature recuperator <NUM> and then fed back into the first and second high-temperature turbines 1808A, 1808B.

The GPTR <NUM> may be of any configuration or embodiment described above. The GPTR <NUM> is shown as including a calandria <NUM> filled with a moderator and having some number, two are shown, of fuel columns <NUM> containing nuclear fuel. The high-pressure turbines' combined outlet coolant streams <NUM> are passed to the GPTR <NUM> where it flows through the fuel columns <NUM>. A heated coolant stream <NUM> exits the GPTR <NUM> and is passed to the inlet of the low-pressure turbine T2, <NUM>.

In <FIG> and <FIG> the "downstream" turbine drives a generator while the "upstream" turbine(s) drives the compressors. It should be noted that these turbines can be reversed, e.g. the upstream turbine can be the one driving the generator.

In the Brayton cycle embodiments of <FIG>, <FIG>, <FIG> and <FIG> a depressurized sCO<NUM> stream output by a turbine is used by a GPTR as a coolant. The output of the GPTR is a heated sCO<NUM> stream, which is then ultimately repressurized by one or more compressors, which may be driven by the turbine, to create the pressurized sCO<NUM> stream that is the input to the turbine. The heat from the reactor is mostly recovered from the heated coolant during the repressurization resulting in a thermodynamic cycle whose input is the energy generated by the nuclear fission in the reactor and whose output is the mechanical energy from the turbine.

Claim 1:
A nuclear power plant (<NUM>) comprising:
a reactor core (<NUM>) having a plurality of fuel columns (<NUM>) penetrating a calandria, the calandria (<NUM>) containing a volume of secondary coolant, each fuel column including a structural tube that forms the exterior of the fuel column, at least a portion of the exterior of the fuel column being in contact with the secondary coolant, the structural tube containing a nuclear fuel;
a closed-loop carbon dioxide coolant circuit that routes pressurized carbon dioxide (<NUM>) into the fuel columns and removes a heated carbon dioxide (<NUM>) from the fuel columns, thereby removing energy from the nuclear fuel within the fuel columns;
the closed-loop carbon dioxide coolant circuit including:
at least one turbine (<NUM>) configured to generate mechanical energy from the heated carbon dioxide and discharge depressurized carbon dioxide (<NUM>); and
at least one compressor (<NUM>) configured to compress the depressurized carbon dioxide and discharge pressurized carbon dioxide (<NUM>) to the reactor core,
characterised in that
the pressurized carbon dioxide is in a supercritical state throughout the closed-loop carbon dioxide coolant circuit.