Abstract:
A spent nuclear fuel rod canister includes a submersible pressure vessel including a casing that defines an interior cavity, the casing including a corrosion resistant and heat conductive material with a thermal conductivity of above about 7.0 watts per meter per kelvin; and a rack enclosed within the interior cavity and configured to support one or more spent nuclear fuel rods.

Description:
TECHNICAL FIELD 
     This disclosure generally relates to systems and methods for storing and managing nuclear spent fuel. 
     BACKGROUND 
     Spent fuel pools provide long term decay heat removal from fuel that has been recently discharged from a nuclear reactor. A recently discharged nuclear core typically represents the largest source of heat generation in a spent fuel pool. In the event of a complete loss of power to the nuclear power plant, cooling systems for the spent fuel pool may not be available to remove the fuel&#39;s decay heat. For prolonged nuclear plant station blackout conditions with recently discharged fuel, the potential exists to boil off all of the water in the spent fuel pool thereby overheating and subsequently damaging the spent fuel bundles. This may result in a radioactive release to the environment. 
     SUMMARY 
     This disclosure describes technologies related to systems, apparatus, and methods for handling, storing, and otherwise managing spent fuel rods from a nuclear reactor. In one general implementation, a spent nuclear fuel rod canister includes a submersible pressure vessel including a casing that defines an interior cavity, the casing including a corrosion resistant and heat conductive material with a thermal conductivity of above about 7.0 watts per meter per kelvin; and a rack enclosed within the interior cavity and configured to support one or more spent nuclear fuel rods. 
     A first aspect combinable with the general implementation further includes a first hemispherical enclosure coupled to the casing at a top end of the casing. 
     In a second aspect combinable with any of the previous aspects, the first hemispherical enclosure includes a radiussed interior surface that defines a top portion of the interior cavity. 
     A third aspect combinable with any of the previous aspects further includes a second hemispherical enclosure coupled to the casing at a bottom end of the casing. 
     In a fourth aspect combinable with any of the previous aspects, the second hemispherical enclosure includes a radiussed interior surface that defines a bottom portion of the interior cavity. 
     A fifth aspect combinable with any of the previous aspects further includes a riser that defines a fluid pathway through the riser between a top portion of the interior cavity and a bottom portion of the interior cavity. 
     A sixth aspect combinable with any of the previous aspects further includes an annulus defined between the riser and the casing. 
     A seventh aspect combinable with any of the previous aspects further includes a fuel basket positioned in the interior cavity between the riser and the bottom portion of the interior cavity. 
     In an eighth aspect combinable with any of the previous aspects, the fuel basket includes a spent nuclear fuel rod rack. 
     In a ninth aspect combinable with any of the previous aspects, the fuel basket includes a perforated support plate adjacent a bottom surface of the rack, the fluid pathway fluidly coupled to the bottom portion of the interior cavity through the perforated support plate. 
     A tenth aspect combinable with any of the previous aspects further includes a heat exchanger attached to the casing of the pressure vessel. 
     In an eleventh aspect combinable with any of the previous aspects, the heat exchanger includes at least one conduit that is at least partially disposed exterior to the casing and is in fluid communication with the interior cavity. 
     In a twelfth aspect combinable with any of the previous aspects, the corrosion resistant material includes a high radioactivity conduction material. 
     In a thirteenth aspect combinable with any of the previous aspects, the vessel is free of any radiation shielding material. 
     In another general implementation, a spent nuclear fuel rod management system includes a spent fuel pool containing a heat transfer liquid; and a plurality of spent fuel canisters, where each of the canisters includes a submersible pressure vessel including a casing defining an interior cavity at least partially filled with a liquid coolant; a rack enclosed within the interior cavity; and one or more spent nuclear fuel rods supported in the rack. 
     In a first aspect combinable with the general implementation, the liquid coolant includes water. 
     In a second aspect combinable with any of the previous aspects, the heat transfer fluid includes at least one of water or ambient air. 
     In a third aspect combinable with any of the previous aspects, the heat removal rate of each canister is between about 0.3 MW and 0.8 MW. 
     In another general implementation, a method of dissipating decay heat generated by a spent nuclear fuel rod includes loading at least one spent nuclear fuel rod in a spent fuel canister that includes an inner cavity, the interior cavity at least partially filled with a fluid coolant; submerging the spent fuel canister in a heat transfer fluid contained in a spent fuel pool; transferring decay heat from the spent nuclear fuel rod to the fluid coolant; and transferring the decay heat from the fluid coolant to the heat transfer fluid in the spent fuel pool. 
     In a first aspect combinable with the general implementation, a rate at which heat is transferred from the spent fuel rod is at least as great as a rate at which the spent nuclear fuel rod produces decay heat. 
     A second aspect combinable with any of the previous aspects further includes circulating the fluid coolant within the interior cavity of the spent fuel canister via natural circulation. 
     A third aspect combinable with any of the previous aspects further includes exposing an exterior surface of the spent fuel the canister to ambient air. 
     A fourth aspect combinable with any of the previous aspects further includes based on the exposure to ambient air, phase changing a portion of the fluid coolant from a liquid to a gas in the spent fuel canister; and phase changing the gas back to a liquid condensate on an interior surface of the spent fuel canister based at least in part on heat transfer between the gas and the ambient air. 
     A fifth aspect combinable with any of the previous aspects further includes circulating at least a portion of the liquid condensate on the interior surface to a pool of the fluid coolant in a bottom portion of the canister. 
     In another general implementation, a method of managing spent fuel rods includes removing a first batch of spent fuel rods from a nuclear reactor; at a first time, installing the first batch of spent fuel rods in a spent fuel canister, the first batch of spent fuel rods generating decay heat at a first decay heat rate; submerging the spent fuel canister in a heat transfer fluid to remove decay heat from the first batch of spent fuel rods; removing decay heat from the first batch of spent fuel rods using the spent fuel canister for a time period at a rate greater than the first decay heat rate; at a second time subsequent to the first time, installing a second batch of spent fuel rods in the spent fuel canister, the second batch of spent fuel rods generating decay heat at a second decay heat rate greater than the first decay heat rate; and removing decay heat from the first and second batch of spent fuel rods at a rate at least as great as a sum of the first and second decay heat rates. 
     In a first aspect combinable with the general implementation, installing the first batch of spent fuel rods in a spent fuel canister includes installing the first batch of spent fuel rods in a spent fuel canister directly from the nuclear reactor. 
     A second aspect combinable with any of the previous aspects further includes removing at least a portion of the first batch of spent fuel rods; and installing the portion in a dry cask. 
     Various implementations described in this disclosure may include none, one, some, or all of the following features. For example, decay heat removal from spent nuclear fuel may be achieved through a canister into a pool rather than directly to a pool, thereby increasing an ease of handling of spent nuclear fuel and providing an additional safety barrier to fission product release. Further, in the case of loss of pool liquid or loss of recirculation of pool liquid (e.g., water), such as, due to a loss of power incident, decay heat removal from spent nuclear fuel may be achieved through the canister to ambient air. The decay heat removal rate may be substantially similar or identical to that achieved to the pool during normal operating conditions. In some implementations, a desired decay heat removal may be achieved without any operator action or power needed. 
     The details of one or more implementations of the subject matter described in this specification are set forth in the accompanying drawings and the description below. Other features, aspects, and advantages of the subject matter will become apparent from the description, the drawings, and the claims. 
    
    
     
       BRIEF DESCRIPTION OF THE DRAWINGS 
         FIG. 1  is a block diagram illustrating a system of spent fuel management for a nuclear reactor systems. 
         FIGS. 2A-2C  illustrate schematic views of an example implementation of a spent fuel canister operating in normal conditions having one stack or two stacks of spent fuel rods. 
         FIGS. 3A-3B  illustrate schematic views of example racks for holding spent fuel rods. 
         FIG. 4  illustrates a schematic view of an example implementation of a spent fuel canister operating in abnormal conditions. 
         FIGS. 5A-5B  illustrate schematic views of an example implementation of a spent fuel canister that includes an external heat exchanger and is operating in normal conditions. 
         FIG. 5C  illustrates a schematic view of an example implementation of a spent fuel canister that includes an external heat exchanger and is operating in abnormal conditions. 
         FIGS. 6A-6B  illustrate schematic views of another example implementation of a spent fuel canister that includes an external heat exchanger and is operating in normal conditions. 
         FIG. 6C  illustrates a schematic view of another example implementation of a spent fuel canister that includes an external heat exchanger and is operating in abnormal conditions. 
         FIG. 7  is a flow chart illustrating an example method of dissipating decay heat generated by a spent fuel rod. 
         FIG. 8  is a flow chart illustrating an example method of managing spent fuel rods from a nuclear reactor system. 
     
    
    
     DETAILED DESCRIPTION 
       FIG. 1  is a block diagram illustrating a technique of managing spent fuel  104  from one or more nuclear reactors  152  in a nuclear reactor power system  150 . The technique involves removing spent nuclear fuel rods  104  from nuclear reactors  152  and transferring the spent fuel rods  104  to a spent fuel management system  154  that facilitates removal of residual decay heat produced by the spent fuel rods  104 . Spent fuel management system  154  includes multiple spent fuel canisters  100  submerged in a spent fuel pool  156  filled with fluid  158 . Fluid  158  provides a heat sink for receiving and dissipating the decay heat from spent fuel rods  104 . As described in detail below, canisters  100  can be configured to operate passively, e.g., without operator intervention or supervision, under both normal and abnormal emergency conditions. In some examples, canisters  100  provide a long term decay heat removal solution for spent fuel rods  104 . For example, canisters  100  can be capable of achieving a substantially constant heat removal rate (e.g., a heat removal rate of about 0.3 MW, 0.4 MW, or 0.8 MW) in various normal and abnormal operating conditions. The number of nuclear reactors  152  and canisters  100  in  FIG. 1  are not indicative of any particular implementation or implementation, and are depicted for illustrative purposes only. 
     With respect to nuclear reactors  152 , a reactor core  20  is positioned at a bottom portion of a cylinder-shaped or capsule-shaped reactor vessel  70 . Reactor core  20  includes a quantity of nuclear fuel rods (e.g., fissile material that produces a controlled nuclear reaction) and optionally one or more control rods (not shown). In some implementations, nuclear reactors  152  are designed with passive operating systems employing the laws of physics to ensure that safe operation of the nuclear reactor  152  is maintained during normal operation or even in an emergency condition without operator intervention or supervision, at least for some predefined period of time. A cylinder-shaped or capsule-shaped containment vessel  10  surrounds reactor vessel  70  and is partially or completely submerged in a reactor pool, such as below waterline  90 , within reactor bay  5 . The volume between reactor vessel  70  and containment vessel  10  may be partially or completely evacuated to reduce heat transfer from reactor vessel  70  to the reactor pool. However, in other implementations, the volume between reactor vessel  70  and containment vessel  10  may be at least partially filled with a gas and/or a liquid that increases heat transfer between the reactor and containment vessels. 
     In a particular implementation, reactor core  20  is submerged within a liquid, such as water, which may include boron or other additives, which rises into channel  30  after making contact with a surface of the reactor core. The upward motion of heated coolant is represented by arrows  40  within channel  30 . The coolant travels over the top of heat exchangers  50  and  60  and is drawn downward by density difference along the inner walls of reactor vessel  70  thus allowing the coolant to impart heat to heat exchangers  50  and  60 . After reaching a bottom portion of the reactor vessel, contact with reactor core  20  results in heating the coolant, which again rises through channel  30 . 
     Although heat exchangers  50  and  60  are shown as two distinct elements in  FIG. 1 , heat exchangers  50  and  60  may represent any number of helical coils that wrap around at least a portion of channel  30 . 
     Normal operation of the nuclear reactor module proceeds in a manner wherein heated coolant rises through channel  30  and makes contact with heat exchangers  50  and  60 . After contacting heat exchangers  50  and  60 , the coolant sinks towards the bottom of reactor vessel  110  in a manner that induces a thermal siphoning process. In the example of  FIG. 1 , coolant within reactor vessel  70  remains at a pressure above atmospheric pressure, thus allowing the coolant to maintain a high temperature without vaporizing (e.g., boiling). 
     As coolant within heat exchangers  50  and  60  increases in temperature, the coolant may begin to boil. As the coolant within heat exchangers  50  and  60  begins to boil, vaporized coolant, such as steam, may be used to drive one or more turbines that convert the thermal potential energy of steam into electrical energy. After condensing, coolant is returned to locations near the base of heat exchangers  50  and  60 . 
       FIGS. 2A-2C  illustrate schematic views of an example implementation of a spent fuel canister  200  operating in normal conditions having one stack or two stacks of spent fuel rods. Canister  200  includes a submersible vessel  202  that contains spent fuel rods  204  and coolant  206  surrounding the spent fuel rods  204 . As shown schematically in  FIG. 2A , canister  200  (filled to a coolant level  201 ) is supported in a spent fuel pool  256  filled with fluid  258  (e.g., water or some other suitable coolant). In some implementations, the fluid  258  in spent fuel pool  256  (filled to fluid level  203 ) is continuously or intermittently circulated by pumps or other hardware to improve heat transfer between vessel  202  and the fluid  258 . Circulation of the fluid  258 , in some aspects may increase the effectiveness of convective heat transfer between the canister  200  and the fluid  258 . 
     Vessel  202 , in the example implementation, facilitates the dissipation of decay heat from multiple spent fuel rods  204 . In this example, vessel  202  is an elongated capsule-shaped container, having a cylindrical main body with two elliptical or hemispherical heads on either end (e.g., the top head  205  and the bottom head  207 ). The shape of vessel  202 , in this example provides a relatively large amount of available surface area (e.g., relative to the available volume) to facilitate convective heat transfer with both the coolant  206  contained within the vessel  202  and the fluid  258  surrounding the vessel  256  in the spent fuel pool  256 . The shape of the vessel  202  also may facilitate gravity driven natural circulation of the contained coolant  206 . In some examples, vessel  202  defines an outer diameter of between about 7 and 12 ft. and a length of about 72 ft. In some examples, vessel  202  defines a surface area of about 1600 ft. 2  Vessel  202  can be sized to lengths and diameters that can be accommodated in typical commercial nuclear spent fuel pools (e.g., 30 ft. to 50 ft. in length). 
     Vessel  202 , in this example, is hermetically sealed and capable of pressurization to a specified design limit (e.g., 400-500 psia). As discussed below, the design limit pressure of vessel  202  may be particularly significant to vessel heat removal in abnormal operating conditions. The cylindrical shell  208  of vessel  202 , in this example, is a thin-walled construction fashioned from a corrosion resistant and heat conductive material (e.g., steel). In general, cylindrical shell  208  conducts heat and withstands pressure, thermal, radiation, and seismic induced stresses. The cylindrical shell  208  can be fabricated using materials approved for use in nuclear reactor pressure vessels. For example, in some implementations, cylindrical shell  208  includes a steel base material such as SA302 GR B, SA533 GR B, Class 1, SA 508 Class 2, or SA 508 Class 3 that may be clad with TYPE 308L, 309L TYPE 304 austenitic stainless steel. Other base materials can be implemented such as 16MnD5, 20MnMoNi55, 22NiMoCr3 7, 15Kh2MFA(A), 15Kh2NMFA(A) with Sv 07Kh25N13 and/or Sv 08Kh19N10G2B austenitic cladding. In some examples, cylindrical shell  208  does not provide any shielding to block or otherwise inhibit potentially harmful radiation generated by spent fuel rods  204 . However, in some other examples, cylindrical shell  208  is provided with radiation shielding. Cylindrical shell  208  can be fabricated using rolled plate or ring forgings. The wall thickness of cylindrical shell  208  can be between about 1.5 and 4.5 inches. In any event, the material and thickness of cylindrical shell  208  provides sufficient strength to withstand stresses associated with the design limit pressurization. 
     Spent fuel rods  204  are secured in place near the bottom of vessel  202  inside the riser channel  216  and supported by a lower support plate  214  (e.g., as also shown in  FIG. 2B ) and lower support structure  211 . As shown, the lower support plate  214  and riser channel  216  form a “basket” which cradles spent fuel rods  204  and facilitates natural circulation of coolant  206 . In this example, fuel barrel support/shield  210  includes a fuel barrel and radiation shield that supports a plurality of individual racks  212 . It is attached to lower support plate  214  and channel riser  216 . Channel riser  216  is supported by upper support ring  218  and upper support structure  213 . Racks  212  receive respective spent fuel rods  204  and maintain them in a relatively stable, e.g., non-critical, condition. For example, racks  212  can be fashioned from a material that includes a neutron absorber (e.g., boron) to inhibit criticality events.  FIG. 2A  shows a single stack of spent fuel  204  whereas  FIG. 2C  shows a double stack of spent fuel  204 . 
       FIG. 3A  shows a first example fuel barrel support/shield structure  310   a  with a particular number (e.g., 37) of available racks  312   a  to accommodate respective spent fuel rods.  FIG. 3B  shows a second example fuel barrel support/shield structure  310   b  with another number (e.g., 97) of fuel accommodating racks  312   b . Support structure  310   b  is significantly larger than support structure  310   a , and therefore may require a larger vessel. For example, support structure  310   a  can be incorporated in a vessel having a 7 ft. outer diameter, while support structure  310   b  can be incorporated in a vessel having a 12 ft. outer diameter. The racks can be arranged to accommodate a wide variety of fuel types such as those typical of boiling water reactors (e.g., 8×8, 9×9, or 10×10 fuel assemblies) or the larger pressurized water reactor fuel assemblies (e.g., 17×17 fuel bundles). 
     In these illustrations, racks  312   a  and  312   b  are rectilinear in cross-section defining an open area of about 11 and 28 ft 2  respectively. Of course, other suitable shapes (e.g., circular, hexagonal, triangular, etc.) sizes can also be implemented. Further, as shown, racks  312   a  and  312   b  are arranged in a symmetrical, tightly packed honeycomb configuration. In some examples, this geometric configuration is provided for the dual purposes of heat removal and criticality mitigation. However, other suitable configurations can also be effectively implemented. For instance, racks  312   a  and  312   b  can be spaced apart from one another (as opposed to tightly packed), or arranged in some other symmetrical configuration (e.g., a quadrilateral configuration), as opposed to a honeycomb shape. 
     Turning back to  FIG. 2A , upper support ring  218  and lower support plate  214  forms the base of support for the riser channel  216 . In addition, lower support plate  214  may have sufficient strength to bear the weight of spent fuel rods  204 . Lower support plate  214  allows coolant  206  to flow upward past spent fuel rods  204  for convective heat transfer from the spent fuel rods  204  to the coolant. For example, lower support plate  214  can include small perforations or large openings that allow naturally circulating coolant  206  to flow up through the support plate and past spent fuel rods  204 . 
     The illustrated riser  216  extends upward from lower support plate  214  to surround the fuel barrel support/shield  210  and the spent fuel rods  204  supported in racks  212 . As shown, riser  216  extends from a point near the top of the lower support plate  214  to the top of the upper support ring  218 , a point that is approximately halfway to the vessel&#39;s upper head flange  219 . For example, riser  216  can have a height of about 30 ft. In some examples, riser  216  is cylindrical in shape with a rounded shaped exit, so as to reduce form losses in the naturally circulating coolant  206 . 
     The example riser  216  defines a hollow bore  220  that serves to direct coolant  206  upward through the interior of vessel  202 , and a narrow annulus  222  that directs coolant downward along the inner wall of vessel  202 . Upper support ring  218  peels radially inward from the cylindrical shell  208  to the top of riser  216 . Similar to support plate  214 , upper support ring  218  also includes perforations or large openings that allow naturally circulating coolant  206  to pass downward through the upper support ring  218  and through annulus  222 . 
     Vessel  202  may initially be filled with an amount of liquid coolant  206 . In particular, the vessel  202  is filled with at least enough coolant  206  to place the liquid level  201  above the top of the upper support ring  218 . In some examples, vessel  202  is filled with about 35 m 3  of liquid coolant  206 . The coolant can include water and/or some additional type of coolant. For instance, coolant  206  under natural circulation conditions may generate a convective heat transfer coefficient of between about 1000-2500 (W/m 2 K) on the inside surface of cylindrical shell  208 . Coolant  206  can be engineered to undergo a liquid-to-gas phase change under certain conditions (e.g., when convective heat transfer to the ambient fluid  258  in the spent fuel pool  256  has significantly decreased) to maintain the heat removal rate at a substantially constant level in abnormal operating conditions, as explained in detail below. 
     In operating under normal conditions as shown in  FIG. 2A  (e.g., no loss of power or loss of fluid  258 ) vessel  202  is submerged in the spent fuel pool fluid  258 . Natural circulation of the coolant  206  inside of vessel  202  is established by the buoyancy force generated as a result of the density and elevation differences between hot coolant  206  in contact with the spent fuel  204  and cooler coolant  206  in annulus  222 . That is, when coolant  206 , in contact with the spent fuel  204 , is heated by the decay heat emanating from spent fuel rods  204 , the coolant  206  becomes less dense and begins to rise. The rising coolant  206  is directed upward through racks  212  holding spent fuel rods  204 . As the coolant  206  flows up past the spent fuel rods  204 , it receives even more heat, which makes it continue to flow upward. Riser  216  directs the heated coolant  206  upward through bore  220 , away from spent fuel rods  204  and toward the exit of the channel riser  216  near the top of the upper support ring  218 . Coolant  206  emerging from riser  216  is cooled down through convective heat transfer with the inner surface of vessel  202 . The heat is conducted through the wall of vessel  202  then transferred by convection to the spent fuel pool fluid  258 . The cooled coolant  206  becomes denser and is therefore drawn downward by gravity. The sinking coolant  206  is directed trough the perforated upper support ring  218  of support structure  210  and through annulus  222 , through the perforated lower support plate  214  and ultimately returning to the lower head  207  of vessel  202 . 
       FIG. 4  illustrates a schematic view of an example implementation of spent fuel canister  200  operating in abnormal conditions. In some implementations, spent fuel canister  200  is designed to operate in abnormal operating conditions, while maintaining a substantially constant rate of decay heat removal. In some aspects, the abnormal operating condition is an emergency situation where spent fuel pool  256  has been drained or the fluid  258  has evaporated (as shown in  FIG. 4 ). However, other types of abnormal operating conditions may also occur (e.g., loss of fluid circulation in the spent fuel pool  256 ). In such abnormal operating conditions, an amount of convective heat transfer between vessel  202  and the surrounding ambient environment may be significantly reduced. The reduced rate of heat transfer ultimately causes liquid coolant  206  in contact with the spent fuel  204  to undergo a liquid-to-gas phase change. A low density, two-phase coolant mixture  206   c  rises up through the spent fuel  204  and exits the top of the riser channel  216 . At the top of the riser  216 , the gas phase coolant  206   a  and the liquid phase coolant  206   b  separate from the two-phase coolant  206   c  by gravity. The liquid phase coolant  206   b  travels downward through the perforated upper support ring  218  into the annulus  222 . The gas phase coolant  206   a  continues to travel upward in the vessel  202  to the upper head  205 . When the gas phase coolant  206   a  comes in contact with the inside wall of the vessel  202 , it exchanges heat with the wall to produce a condensate  206   d . The condensate  206   d  may be in the form of a liquid film or droplets that travel downward along the inside wall of the vessel  202 . The condensate  206   d  collects in the region above the upper support ring  218  and mixes with the downward flowing liquid coolant  206   b . The condensate  206   d  and the liquid phase coolant  206   b  travel downward through the annulus, through the perforated lower support plate  214  and lower head  207  plenum and back upward through the spent fuel racks  212 . 
     In this example, the canister can transition from liquid cooling (e.g., water) to air cooling in the spent fuel pool  256  without the need for operator actions or external power. As noted above, the heat removal rate of the air cooled canister  200  may be substantially equal to that of the liquid cooled canister  200 . In particular, the liquid-to-gas phase change may cause the inner cavity of vessel  202  to pressurize. Pressurization of vessel  202  increases the saturation temperature within the vessel  202 , and thus raises the temperature of its outer surface. The increased outer surface temperature of vessel  202  increases both the thermal radiation heat transfer rate to the surroundings and the free convection heat transfer rate with the ambient air  260  (as opposed to liquid  258  in the spent fuel pool during normal operating conditions) to a point where the overall heat removal rate of canister  200  is acceptable. For example, the large surface area and high surface temperature of vessel  202  may be sufficient to remove heat from the canister  200  to the ambient air  260  at substantially the same rate as with the fuel pool fluid  258 . 
       FIGS. 5A-5B  illustrate schematic views of an example implementation of a spent fuel canister  400  that includes an external heat exchanger  424  and is operating in normal conditions. As shown, heat exchanger  424  includes a horizontal upper tube header  223   a  and a horizontal lower tube header  223   b  joined together by a series of c-shaped vertical heat exchanger tubes  226 . The heat exchanger tubes can be 2 to 4 inches in diameter and 15-20 feet in length. The upper tube header  223   a , in this example, is connected to cylindrical shell  208  below the coolant level  201  and above the upper support ring  218  by header conduit  225   a . The lower tube header  223   b  is connected to annulus  222  by header conduits  225   b . In some examples, header conduits  225   a  and  225   b  are sloped such that liquid flowing through the conduits is always in the downward direction. The heat exchanger  424  is designed to withstand full pressure and temperatures during normal and abnormal conditions. 
     As shown in  FIG. 5A , during normal conditions, hot liquid coolant  206  rises through the bore  220  to the outlet of the riser  216 . Approximately half of the liquid coolant  206  enters the upper header conduits  225   a  into heat exchanger  424  where it transfers heat to the spent fuel pool fluid  258 . The remaining half of the liquid coolant travels through the perforated upper support ring  218  into the annulus  222  where it transfers heat to the spent fuel pool fluid  258  by convection and conduction heat transfer through the vessel  202  walls. The flow paths for the coolant  206 , in this example, are established by natural circulation created by the buoyancy force established by the density difference of the coolant in the bore  220  and the annulus  222  and the relative elevation of their thermal centers. 
       FIG. 5C  illustrates a schematic view of an example implementation of a spent fuel canister  400  that includes an external heat exchanger  424  and is operating in abnormal conditions. In this example, although similar to that illustrated in  FIG. 4 , the addition of heat exchanger  424  provides additional surface area for natural circulation cooling. Convection heat transfer inside the tubes can increase the heat removal rate capacity of the canister thereby reducing the overall height of the canister. In the present example, a sixty-five tube heat exchanger of 16 ft. tube length can reduce the canister height by at about 30% (e.g., from 72 feet to 50 feet) while rejecting the same amount of heat, 0.35 MW to the ambient air  206 . In some examples, heat exchanger  424  is a sixty-five tube heat exchanger or an approximately 150 tube heat exchanger. The number and lengths of heat exchanger tubes  226  can be selected to provide a wide range of desired heat removal rates. 
       FIGS. 6A-6B  illustrate schematic views of another example implementation of a spent fuel canister  500  that includes an external heat exchanger  525  and is operating in normal conditions. As shown, heat exchanger  524  includes a horizontal upper tube header  223   a , a horizontal lower tube header  223   b  joined together by a series of c-shaped vertical heat exchanger tubes  226 . The heat exchanger tubes can be 2 to 4 inches in diameter and 15-20 feet in length. In the illustrated example, the heat exchanger  525  is connected to cylindrical shell  208  between the level  201  and the upper support ring  218  by header conduit  225   a . The lower tube header  223   b  is connected to annulus  222  by header conduits  225   b . Header conduits  225   a  and  225   b  are sloped such that liquid flowing through the conduits is always in the downward direction. The heat exchanger  524 , in some aspects, is designed to withstand full pressure and temperatures during normal and abnormal conditions. During normal conditions, the heat transfer mechanism may be identical or substantially similar to the same as those described for  FIG. 2A . 
       FIG. 6C  shows canister  500  operating under abnormal conditions, rejecting heat to ambient air  206 . The liquid phase coolant behaves as described previously for  FIG. 4 . However, because heat exchanger  524  is connected to the gas phase region of the canister, (e.g., through riser  216 ) a portion of the gas phase coolant  206   a  is condensed inside the heat exchanger tubes. This creates a low pressure region inside the tubes  526  which draws additional gas phase coolant  206   a  into the tubes. The condensate  206   d  inside the tubes  526  falls by gravity through the tubes  526  into the cylindrical shell. The condensate mixes with the two-phase coolant  206   c  in the region above the upper support ring  218 . The liquid phase coolant  206   b  travels downward by gravity through the perforated upper support ring  218  into the annulus  222 , through the perforated lower support plate  214 , through the plenum formed by the lower head  207 . It flows upward through the spent fuel racks  212  thereby cooling the spent fuel  204 . 
     Another implementation of the present disclosure features various methods of dissipating decay heat generated by a spent fuel rod.  FIG. 7  illustrates an example method  700  for dissipating decay heat. The method includes, at step  702 , submerging a spent fuel canister in a heat transfer fluid contained in a spent fuel pool. As described above, the spent fuel canister can include a cylindrical shell defining an interior cavity which contains the spent fuel rod. At step  704 , decay heat is transferred from the spent fuel rod to liquid coolant contained within the canister. In some implementations, the coolant is circulated within the canister via natural circulation to facilitate heat transfer. At step  706 , the decay heat is transferred from the coolant, through a wall of the canister, to the heat transfer fluid of the spent fuel pool. A rate at which heat is transferred from the spent fuel rod is at least as great as a rate at which the spent fuel rod produces decay heat. 
     Method  700  can also optionally include, at step  708 , exposing the canister to ambient air due to a loss of spent fuel pool fluid. At step  710 , based on the exposure to ambient air, a portion of the coolant inside the canister is phase changed from a liquid to a gas. At step  712 , heat is transferred, through a wall of the canister, from the gas phase coolant to the ambient air. At step  714 , the gas phase coolant is condensed back to a liquid and circulated (e.g., via natural circulation) within the canister. 
     Yet another implementation of the present disclosure features various methods of managing spent fuel rods by cycling them through spent fuel canisters.  FIG. 8  illustrates an example method  800  for managing spent fuel rods. The method includes, at step  802 , removing a first batch of spent fuel rods from a nuclear reactor. At step  804 , the first batch of spent fuel rods is installed in a spent fuel canister (e.g., spent fuel canister  100 ) at a first time (T 1 ). At step  806 , the spent fuel canister is submerged in a heat transfer fluid (such as contained in spent fuel pool  156 ). At step  808 , the canister is used to remove decay heat from the first batch of spent fuel rods for a time period (T). At step  810 , a second batch of spent fuel rods is installed within the spent fuel canister at a second time (T 2 ). The heat removal rate of the spent fuel canister is at least as great as the combined decay heat rate of the first and second batches of spent fuel rods at T 2 . As discussed in context of the first and second examples below, the example method of  FIG. 8  can be used to continuously manage spent fuel from a nuclear reactor. 
     In some aspects, an example spent fuel management system (e.g., spent fuel management system  154 ) that includes a spent fuel pool and multiple spent fuel canisters according to the present disclosure (e.g., spent fuel canister  100 ,  200 ,  400 , and/or  500 ) manages spent fuel from nuclear reactors (e.g., 1-12 nuclear reactors  152 ) each effectively refueled once every twenty-four months, with a spent fuel batch of one-half core, approximately 18 fuel assemblies being removed every two months. Each batch of spent fuel produces approximately 0.2 MW of decay power after twenty days, and 0.1 MW of decay power after six months. Spent fuel that has decayed for six months can be discharged from the spent fuel canisters into, for example, a typical liquid coolant filled, non-pressurized, spent fuel pool. After an additional period of cooling, for example 5-10 years, the spent fuel can be discharged to a dry cask. In this example, there is sufficient liquid coolant  158  in the spent fuel pool  156  to provide 20 days of cooling before transitioning to cooling by ambient air. The system includes two spent fuel canisters, each capable of achieving at least 0.5 MW of decay heat removal when fully immersed in spent fuel pool coolant  158  and 0.35 MW decay heat removal after the 20 day transition cooling period. Table 1 below illustrates an example linear sequence for canister loading and unloading to accommodate spent fuel from the nuclear reactor. In Table 1, “T” is in months and “B#” represents a particular batch of spent fuel. A “+” indicates that the batch is loaded into the canister and a “−” indicates that the batch is removed. 
     
       
         
               
               
               
               
               
               
               
             
           
               
                 TABLE 1 
               
               
                   
               
             
             
               
                 Canister # 
                 T = 0 
                 T = 2 
                 T = 4 
                 T = 6 
                 T = 8 
                 T = 10 
               
               
                   
               
               
                 Canister 1 
                 +B1 
                   
                 +B3 
                   
                 −B1  
                   
               
               
                   
                 0.35 MW 
                   
                 0.5 MW 
                   
                 +B5  
                   
               
               
                   
                   
                   
                   
                   
                 0.5 MW 
                   
               
               
                 Canister 2 
                   
                 +B2 
                   
                 +B4 
                   
                 −B2  
               
               
                   
                   
                 0.35 MW  
                   
                 0.5 MW 
                   
                 +B6 
               
               
                   
                   
                   
                   
                   
                   
                 0.5 MW 
               
               
                   
               
               
                 T = 12 
                 T = 14 
                 T = 16 
                 T = 18 
                 T = 20 
                 T = 22 
                 T =24 
               
               
                   
               
               
                 −B3 
                   
                 −B5 
                   
                 −B7  
                   
                 −B9  
               
               
                 +B7 
                   
                 +B9 
                   
                 +B11 
                   
                 +B13 
               
               
                 0.5 MW 
                   
                 0.5 MW 
                   
                 0.5 MW 
                   
                 0.5 MW 
               
               
                   
                 −B4 
                   
                 −B6  
                   
                 −B8  
                   
               
               
                   
                 +B8 
                   
                 +B10 
                   
                 +B12 
                   
               
               
                   
                 0.5 MW 
                   
                 0.5 MW 
                   
                 0.5 MW 
               
               
                   
               
             
          
         
       
     
     In the example sequence presented in Table 1, all of the spent fuel batches would have decayed for eight months prior to discharge. This approach, in some aspects, eliminates the potential risks associated with having higher power density spent fuel placed directly next to lower power density spent fuel. The higher power density spent fuel presents the greater risk of zirconium cladding ignition in air in the event of a loss of spent fuel pool water  158  which could potentially ignite the lower power density spent fuel. 
     In another example spent fuel management system, the system may manage spent fuel from nuclear reactors (e.g. 1-12 nuclear reactors  152 ) each effectively refueled once every twenty-four months, with a spent fuel batch of one-half core being removed every two months. Each batch of spent fuel provides 0.2 MW of decay power after twenty days, and 0.1 MW of decay power after six months. Spent fuel that has decayed for six months can be discharged from the spent fuel canisters into, for example, a typical liquid coolant filled, non-pressurized, spent fuel pool. After an additional period of cooling, for example 5-10 years, the spent fuel can be discharged to a dry cask. The system includes a single spent fuel canister capable of achieving at least 0.65 MW decay heat removal when fully immersed in spent fuel pool coolant  158  and 0.45 MW decay heat removal after the 20 day transition cooling period. Table 2 below illustrates a linear sequence for canister loading and unloading to accommodate spent fuel from the nuclear reactor using the larger spent fuel canister. 
     
       
         
               
               
               
               
               
               
               
             
           
               
                 TABLE 2 
               
               
                   
               
             
             
               
                 Canister # 
                 T = 0 
                 T = 2 
                 T = 4 
                 T = 6 
                 T = 8 
                 T = 10 
               
               
                   
               
               
                 Canister 
                 +B1 
                 +B2 
                 +B3 
                 −B1 
                 −B2 
                 −B3 
               
               
                 1 
                 0.35  
                 0.5  
                 0.65  
                 +B4 
                 +B5 
                 +B6 
               
               
                   
                 MW 
                 MW 
                 MW 
                 0.65  
                 0.65  
                 0.65  
               
               
                   
                   
                   
                   
                 MW 
                 MW 
                 MW 
               
               
                   
               
               
                 T = 12 
                 T = 14 
                 T = 16 
                 T = 18 
                 T = 20 
                 T = 22 
                 T = 24 
               
               
                   
               
               
                 −B4 
                 −B5 
                 −B6 
                 −B7  
                 −B8  
                 −B9  
                 −B10 
               
               
                 +B7 
                 +B8 
                 +B9 
                 +B10 
                 +B11 
                 +B12 
                 +B13 
               
               
                 0.65  
                 0.65  
                 0.65  
                 0.65  
                 0.65  
                 0.65  
                 0.65  
               
               
                 MW 
                 MW 
                 MW 
                 MW 
                 MW 
                 MW 
                 MW 
               
               
                   
               
             
          
         
       
     
     Note that this larger spent fuel canister, in some aspects, provides sufficient space to accommodate a six month discharge of the spent fuel batches. 
     In another example spent fuel management system, the system may manage spent fuel from a single nuclear reactor effectively refueled once every forty-eight months, with a spent fuel batch of one-full core (e.g. 37 assemblies) being removed and replaced. Each batch of spent fuel produces 0.4 MW of decay power after twenty days and 0.2 MW of decay power after six months. Spent fuel that has decayed for six months can be discharged from the spent fuel canisters into, for example, a typical liquid coolant filled, non-pressurized, spent fuel pool. After an additional period of cooling, for example 5-10 years, the spent fuel can be discharged to a dry cask. The system includes a single spent fuel canister capable of achieving at least 0.85 MW decay heat removal when fully immersed in spent fuel pool coolant  158  and 0.6 MW decay heat removal after the 20 day transition cooling period. Table 3 below illustrates a linear sequence for canister loading and unloading to accommodate spent fuel from the nuclear reactor using the larger spent fuel canister. 
     
       
         
               
               
               
               
               
               
               
               
             
           
               
                 TABLE 3 
               
               
                   
               
               
                 Canister # 
                 T = 0 
                 T = 4 yrs 
                 T = 8 yrs 
                 T = 12 yrs 
                 T = 16 yrs 
                 T = 18 yrs 
                 T = 24 yrs 
               
               
                   
               
             
             
               
                 Canister 
                 +B1 
                 +B2 
                 −B1 
                 −B2 
                 −B3 
                 −B4 
                 −B5 
               
               
                 1 
                 0.7 MW 
                 0.85 MW 
                 +B3 
                 +B4 
                 +B5 
                 +B6 
                 +B7 
               
               
                   
                   
                   
                 0.85 MW 
                 0.85 MW 
                 0.85 MW 
                 0.85 MW 
                 0.85 MW 
               
               
                   
               
             
          
         
       
     
     The use of terminology such as “front,” “back,” “top,” “bottom,” “over,” “above,” and “below” throughout the specification and claims is for describing the relative positions of various components of the system and other elements described herein. Similarly, the use of any horizontal or vertical terms to describe elements is for describing relative orientations of the various components of the system and other elements described herein. Unless otherwise stated explicitly, the use of such terminology does not imply a particular position or orientation of the system or any other components relative to the direction of the Earth gravitational force, or the Earth ground surface, or other particular position or orientation that the system other elements may be placed in during operation, manufacturing, and transportation. 
     A number of implementations have been described. Nevertheless, it will be understood that various modifications may be made. For example, advantageous results may be achieved if the steps of the disclosed techniques were performed in a different sequence, if components in the disclosed systems were combined in a different manner, or if the components were replaced or supplemented by other components. Accordingly, other implementations are within the scope of the following claims.