WO2012075010A1

WO2012075010A1 - Heat transfer systems and methods for a fast reactor

Info

Publication number: WO2012075010A1
Application number: PCT/US2011/062400
Authority: WO
Inventors: Frank Campbell
Original assignee: Fluor Technologies Corporation
Priority date: 2010-11-30
Filing date: 2011-11-29
Publication date: 2012-06-07

Abstract

Heat transfer systems and methods are described in which superheated steam can be generated using heat from a molten sodium primary coolant of a fast reactor. The system can include a primary cooling for a fast reactor core, which includes the molten sodium. The system can also include a superheated steam generator that is thermally coupled to the primary cooling circuit by first and second independent heat transfer circuits, which utilize coolants that are substantially non-reactive with water.

Description

HEAT TRANSFER SYSTEMS AND METHODS

FOR A FAST REACTOR

[0001] This application claims the benefit of priority to U.S. provisional application having serial no. 61/417992 filed on November 30, 2010. This and all other extrinsic materials discussed herein are incorporated by reference in their entirety. Where a definition or use of a term in an incorporated reference is inconsistent or contrary to the definition of that term provided herein, the definition of that term provided herein applies and the definition of that term in the reference does not apply.

Field of the Invention

[0002] The field of the invention is superheated steam generators, and specifically to improved power cycles for nuclear power plants.

Background

[0003] Nuclear power generation can be roughly divided into two types of reactors: thermal neutron or light water reactors and fast neutron or fast reactors.

[0004] The majority of nuclear power plants have light water reactors. A typical light water reactor can produce steam having a temperature of about 550°F (about 287.8 °C) and a pressure between 900 to 1,000 psia (about 62.1 - 69.0 bar). When this steam expands in a steam turbine, a significant fraction of the flow condenses into water droplets, which limits the steam turbine design, efficiency and power production. In addition, light water reactors utilize only 5% of the potentially fissionable atoms in a typical fuel load. It is well known that the degraded and depleted fuel removed from the reactor after about three years still contains about 95% of its total recoverable energy content.

[0005] For example, a 900 MW electric light water reactor will generate more than 100 tons of spent fuel per year. In contrast, a fast reactor having the same electrical capacity typically produces only about one ton of spent fuel per year, and can recover more than 99% of the energy contained in spent light water reactor fuel. Furthermore, after the energy in the spent light water reactor fuel is depleted, fast reactors can burn the depleted uranium to recover more than 99% of the energy contained in the uranium ore. [0006] The range of designs for fast reactors includes the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR), and the Sodium-cooled Fast Reactor (SFR). Examples of various fast reactor configurations are discussed in U.K. Patent no. 1151683 to Georges, et al. U.K. Patent no. 1481544 to Cachera; U.K. Patent no. 1534681 to Taylor; U.S. Patent no. 5289511 to Yamamoto; U.S. Patent no. 6672258 to Hayashida; U.S. Patent Appl. no. 2003/0231731 to Ikegami (publ. Dec. 2003); U.S. Patent Appl. no. 2009/0122944 to Namba, et al. (publ. May 2009); and WIPO Appl. no. 2007/136261 to Van Uitert (publ. Nov. 2007).

[0007] Compared to light water reactors, fast neutron reactor cores generate a large amount of heat in a small space. There are a limited number of suitable coolants for fast reactor cores. Coolants that have been previously used in a fast reactor include mercury, sodium and sodium- potassium alloys, lead, bismuth-lead eutectics, various molten salts and molten salt eutectics. Metals such as indium, tin, and zinc have been found to cause dissolution and severe

embrittlement of steel alloys. The core coolant for a fast reactor must have low neutron absorption and low moderation of neutrons. The core coolant should not cause excessive corrosion of the reactor containment or structural materials, should have good heat transfer characteristics, should have melting and boiling points that are suitable for the reactor's operating temperature, and should not have hazardous neutron activation products or radioactive decay byproducts. Liquid metals typically have the characteristics required for use as a fast reactor core coolant.

[0008] Mercury was used in the very first liquid metal cooled nuclear reactor because of its high neutron cross section, low boiling point and high vapor pressure even at room temperature. However, it is highly toxic, produces noxious fumes when heated, and has a relatively low thermal conductivity.

[0009] Typical LFRs use a molten lead or lead-bismuth eutectic (LBE) as a primary coolant. Lead has low neutron absorption and is an effective radiation shield against gamma rays. The melting point of lead is 327.5°C. The 1749°C boiling point of lead provides safety advantages as it can efficiently cool a reactor core even under abnormal operating conditions. However; because lead has a high melting point with a high vapor pressure, it is difficult to operate and maintain a lead cooled reactor. If the temperature drops too low, the lead will solidify in the reactor or reactor heat transfer systems. Although alloying the lead with bismuth can lower the melting point of lead (e.g., 255 °F (124 °C) for a lead-bismuth eutectic), these alloys become extremely corrosive at high temperatures and neutron activated with long half-life radioactive isotopes.

[0010] MSRs typically use a fluoride salt mixture for the primary coolant. Molten fluoride salts have been extensively investigated as reactor coolants because they have very high thermal and radiation stability, low vapor pressures, good thermal conductivity, and low neutron capture. Molten fluoride salt eutectics have melting points above 400°C (752°F). Low melting point fluoride salts are generally too viscous to be used as a heat transfer coolant. A high melting point fluoride salt will require reactor operating temperatures above 600°C. Fluoride salts tend to be highly corrosive to reactor core components at these high temperatures. Because of this, molten salt reactor designs typically use nuclear fuel dissolved in molten fluoride salt. Fluoride combines ionically with almost any fission or transmutation product, even radioactive noble gases. The fluoride salts of actinides and radioactive fission products are generally not soluble in water. Therefore, radioactive dispersion due to an accident is unlikely. However, fluoride salts naturally produce hydrofluoric acid when in contact with water. Moreover, MSRs require exotic nickel alloys to resist corrosion by the high temperature fluoride salts.

[0011] Typical SFRs use sodium metal as a primary coolant. See, e.g., U.K. pat. no. 1534681 to Taylor; U.S. pat. no. 3054741 to Tatlock et al; U.S. pat. no. 4056439 to Robin; U.S. pat. nos. 4600554 and 4608224 to Brachet et al; U.S. pat. no. 4905757 to Boardman; U.S. pat. appl. no. 2009/0122944 to Namba, et al (publ. May 2009); and EPO pat. no. 316120 to GE. Sodium is widely recognized as a fast reactor coolant. Sodium has a melting point of 98 °C and a boiling point of 883 °C. Sodium does not corrode steel to any significant degree and is compatible with many nuclear fuel assemblies. Sodium also does not substantially slow fast neutrons and conducts heat very well. However, sodium ignites spontaneously on contact with air and reacts violently with water, producing hydrogen gas. Liquid sodium ignites in air above 200°C producing low flame with modest heat evolution. Sodium burning is accompanied by production of dense sodium oxide fumes making fire fighting difficult. Neutron activation of sodium produces sodium-24. This isotope is highly radioactive. However, sodium-24 half-life is only 15 hours, so it is not a long-term hazard. [0012] Typical SFRs range in size from small modular systems that can produce 50 MWe to large reactors producing 1,500 MWe. SFRs are advantageous over LFRs and MFRs in that the sodium coolant has a large thermal inertia, low melting point, high boiling point, low vapor pressure, and low corrosion potential. However, current SFR configurations have many disadvantages. Sodium cooled fast reactors typically have an intermediate sodium coolant loop between the reactor and the steam generator. This intermediate sodium coolant loop acts as a barrier between the radioactive sodium in the primary coolant system and the steam system in the power plant, and ensures that any fire resulting from accidental mixing of sodium metal and water will be limited to the secondary heat exchanger and not directly affect the reactor or release radioactive sodium.

[0013] The primary coolant system in a SFR can be arranged in a pool or in a loop. In a loop, primary coolant pumps are used to transfer heat from the reactor vessel to intermediate loop heat exchangers located outside of the reactor vessel. In a pool configuration, the pumps and heat exchangers are located within the reactor vessel around the core.

[0014] SFRs generally have an intermediate sodium coolant loop between the reactor and the steam generator, which increases the possibility of an exothermic reaction between the sodium and water or steam within a steam generator. High temperatures produced by a sodium and water reaction could propagate leaks within steam generator tubes by material erosion. The generation of high pressure in the sodium side of the steam generator would lead to additional leaks and flow disruption. Hydrogen created by the sodium and water reaction could create an explosion hazard. This possibility, while allowing for superheated steam, has become a serious impediment to commercialization and operation of SFRs. In addition, the use of sodium in piping or pumps outside of the reactor vessel creates similar hazards and attendant maintenance problems.

[0015] U.K. Patent no. 1481544 to Cachera, U.K. Patent no. 1151683 to Georges, et al, U.S. Patent no. 5289511 to Yamamoto, and U.S. Patent no. 6561265 to Ohira, et al. discuss utilizing an intermediary coolant that is non-reactive with water to eliminate this risk. However, such coolant configurations are unable to produce superheated steam because the temperature of the intermediate coolant is insufficient to generate superheated steam. [0016] Current configurations of SFRs are also disadvantageous because the high temperature of sodium coolant entering the reactor can contribute to the formation of sodium vapor voids.

Sodium voids in fuel element passages can result in random reactivity changes and local overheating of fuel elements. If the fuel becomes hot enough, the cladding could rupture or melt, releasing radioactive fission products from the fuel. Cavitation due to the collapse of sodium voids is more severe than water cavitation and can produce severe damage to the fuel elements and other structural parts of the SFR.

[0017] Thus, there is still a need for a SFR that eliminates one or more of the disadvantages of prior art SFRs, while allowing for generation of superheated steam.

Summary of the Invention

[0018] The inventive subject matter provides apparatus, systems and methods for heat transfer in a fast reactor that includes a primary cooling circuit configured to cool a fast reactor core. The primary cooling circuit preferably contains molten sodium as a primary coolant. A superheated steam generator can be coupled to the primary cooling circuit by first and second independent heat transfer circuits. Each of the first and second heat transfer circuits advantageously utilize respective first and second heat transfer media that are each substantially non-reactive with water and compatible with sodium.

[0019] The primary cooling circuit in the sodium-cooled fast reactor can be arranged in a pool or a loop. A pool configuration is preferable so that the primary coolant does not circulate out from the reactor containment vessel. In addition, there is less risk that the primary coolant will solidify during system shutdowns.

[0020] The first and second intermediate heat transfer circuits can advantageously be used to isolate the reactor vessel, reactor core and primary sodium coolant from the steam cycle. The primary and intermediate loop coolants provide a large thermal mass that decouples the reactor from steam cycle perturbations. Not all heat transfer media can work with fast reactor cores and intermediate heat transfer circuits because the chosen media should not cause corrosion of the loop components or piping, should have good heat transfer characteristics, should have melting and boiling points that are suitable for the heat exchange service, and should be compatible with the primary coolant, water, and air. Preferred heat transfer media for the first and second heat transfer circuits are molten salts and liquid metal eutectics, respectively. More preferably, the first heat transfer circuit utilizes a chloride salt as a heat transfer medium, and the second heat transfer circuit utilizes a lead-bismuth eutectic as a heat transfer medium.

[0021] The pressure in both the primary and intermediate heat transfer circuits is preferably low, and the pressure in the reactor vessel is preferably near atmospheric. The pressure in the intermediate loop will likely be somewhat higher than the pressure of the primary loop because of the pump head. This pressure difference can advantageously ensure that any leakage will flow from the intermediate heat transfer circuit to the primary heat transfer circuit, and thereby ensure that radioactivity remains within the reactor vessel.

[0022] Methods of producing superheated steam from molten sodium primary coolant of a fast reactor are also contemplated. In some contemplated embodiments, a first heat transfer circuit can be used to cool molten sodium to a first temperature to thereby convert steam to a superheated steam via heat exchange with the first heat transfer circuit. The cooled molten sodium can be further cooled to a second temperature using a second heat transfer circuit, which can be used to convert water to steam. Each of the first and second heat transfer circuits preferably uses heat transfer media that are substantially non-reactive with water.

[0023] Various objects, features, aspects and advantages of the inventive subject matter will become more apparent from the following detailed description of preferred embodiments, along with the accompanying drawing figures in which like numerals represent like components.

Brief Description of the Drawing

[0024] Fig. 1 is a schematic of one embodiment of a heat transfer system for a fast reactor arranged as a rectangular primary coolant pool.

[0025] Fig. 2 is a flow diagram of one embodiment of a heat transfer system for a fast reactor. [0026] Fig. 3 is a flow diagram of one embodiment of a superheated steam generator. [0027] Fig. 4 is a flow diagram of one embodiment of a steam turbine power generator. [0028] Fig. 5 is a schematic of an embodiment of a fast reactor arranged as a cylindrical primary coolant pool heat transfer system.

[0029] Fig. 6 is a flow diagram of an exemplary embodiment of a steam turbine power generator.

[0030] Figs. 7-8 are flowcharts of various embodiments of methods for producing superheated steam from molten sodium primary coolant of a fast reactor.

Detailed Description

[0031] The following discussion provides many example embodiments of the inventive subject matter. Although each embodiment represents a single combination of inventive elements, the inventive subject matter is considered to include all possible combinations of the disclosed elements. Thus if one embodiment comprises elements A, B, and C, and a second embodiment comprises elements B and D, then the inventive subject matter is also considered to include other remaining combinations of A, B, C, or D, even if not explicitly disclosed.

[0032] In contrast to prior art fast reactors, the inventive subject matter allows for the generation of superheated steam through the selection of heat exchange media, which have properties that allow the media to absorb a sufficient amount of heat from the media of a fast reactor's primary cooling circuit to collectively heat a water stream and produce superheated stream. The inventive subject matter also eliminates many of the operational and safety problems inherent in those prior art fast reactors, while allowing for safe and efficient extraction of thermal energy from a fast reactor through superheated steam generation.

[0033] Contemplated systems allow heat exchange media to be used that mitigate the operational and safety concerns associated with fast reactors in the prior art. Because of the unique reactor vessel configuration, the sodium coolant can follow a defined path where its temperature is uniformly changed, and thus the hot sodium coolant does not heat or intermix with colder sodium coolant. This ensures that only cooled sodium can enter the reactor core, which decreases the risk of cavitation and sodium voids. In addition, the one or more intermediate coolant loops can use coolants that are substantially nonreactive with sodium and water. This advantageously eliminates the possibility of sodium-water reactions, and maintenance problems associated with the use of sodium as an intermediate coolant.

[0034] Unless the context dictates the contrary, all ranges set forth herein should be interpreted as being inclusive of their endpoints and open-ended ranges should be interpreted to include only commercially practical values. Similarly, all lists of values should be considered as inclusive of intermediate values unless the context indicates the contrary.

[0035] The utilization of efficient and sustainable nuclear fuel cycles based on fast reactors taught herein allow vastly more of the energy in the earth's readily available uranium supply to be used to produce electricity. Such cycles can greatly reduce the creation of long-lived reactor waste and support nuclear power generation indefinitely. In this way, nuclear power can offer a reliable electrical supply that does not contribute to greenhouse gas buildup in the atmosphere.

[0036] In Figure 1, one embodiment of a circulating sodium-cooled reactor ("CSCFR") 100 is shown, which includes a primary cooling circuit 102 configured to cool a fast reactor core 104. The primary coolant in the primary cooling circuit 102 is preferably reactor grade sodium, although a LBE, a fluoride salt, or other commercially suitable coolant(s) could be used.

[0037] Liquid sodium is preferred over other coolants because of its short radioactive half-life, and its lower melting point of 98 °C and high boiling point of 883 °C. In addition, liquid sodium does not corrode steel to any significant degree and is compatible with many nuclear fuel assemblies. Liquid sodium also does not substantially slow fast neutrons and conducts heat very well.

[0038] The CSCFR 100 can include first and second independent heat transfer circuits 106 and 108, respectively, which thermally couple the primary cooling circuit 102 to a superheated steam generator (shown in Figure 3). In preferred embodiments, the first and second heat transfer circuits 106 and 108 utilize respective first and second intermediate heat transfer media that are each substantially non-reactive with water.

[0039] Each of the first and second independent heat transfer circuits 106 and 108 is preferably exposed to a different temperature range of the primary sodium coolant. These defined temperature ranges to which the first and second heat transfer media are exposed allow for the intermediate heat transfer media to be specifically selected for that heat exchange service. The first heat transfer medium is preferably a molten salt, and more preferably, a molten salt eutectic, and most preferably, a chloride salt. The second heat transfer medium is preferably a liquid metal eutectic, and more preferably LBE. However, any commercially-suitable medium could be used that has an appropriate boiling point, melting point, and low corrosiveness at the defined temperature range. Sodium is advantageously eliminated as an intermediate heat transfer medium, which thereby eliminates operational, safety and maintenance problems resulting from sodium's extreme reactivity with air and water that previously prevented commercial use of sodium-cooled fast reactors.

[0040] The CSCFR 100 can further include cooling pumps 110 and 112, and reactor cooling pumps 114 and 116, which collectively direct the coolant from the reactor core 104 past the first and second independent heat transfer circuits 106 and 108, respectively, and back to the reactor core 104, as shown by arrows. Although a lesser or greater number of pumps could be used, dual cooling pumps 110 and 112 and dual reactor cooling pumps 114 and 116 are preferably used as a redundancy measure to increase the safety of the CSCFR 100. However, in alternative embodiments, the primary cooling circuit 102 can be configured to allow movement of the sodium coolant by convection only.

[0041] The reactor core 104, first and second heat transfer circuits 106 and 108, and pumps 110, 112, 114, and 116 in the CSCFR 100 can be arranged in a pool configuration in a rectangular reactor vessel 101 with a central baffle 120, as shown in Figure 1. Of course, one of ordinary skill in the art would understand that the CSCFR 100 could have any commercially suitable shape. The pool configuration advantageously creates a coolant flow pattern in which the primary coolant is directed in both vertical and horizontal directions, which produces a high temperature coolant over a wide temperature range. In addition, heat transfer is more efficient because a greater amount of heat can be removed from the sodium by the different heat transfer circuits 106 and 108 before reheating the sodium in the reactor core 104. Furthermore, the defined coolant flow pattern ensures that the temperature of the sodium is uniformly changed, and thus prevents the hot sodium coolant from heating or intermixing with colder sodium coolant. Such pattern thereby ensures that only cooled sodium can enter the reactor core, which decreases the risk of cavitation and sodium voids. [0042] Each of the first and second heat transfer circuits 106 and 108 can include any commercially suitable heat exchanger including, for example, tube-in-shell exchangers and bayonet-type heat exchangers.

[0043] In addition to the benefits of eliminating sodium as a intermediate heat exchange medium, the dual intermediate heat exchange circuits 106 and 108 offer redundancy for reactor core 104 heat transfer. The lower temperature second intermediate heat exchanger circuit 108 is ideal for start-up, shutdown and residual heat removal. Redundant circulation pumps 110, 112, 114, and 116 also provide additional safety. Preferably, these pumps are located in cooler regions of the reactor coolant flow as shown. Circulation pumps 110 and 112 are located at the outlet of the first heat transfer circuit 106. Circulation pumps 114 and 116 are located at the outlet of the second heat transfer circuit 108. These locations reduce the possibility of pump suction induced cavitation, and help reduce the operating temperature of the pump components thereby increasing allowable material stresses. Reactor vessel baffling 120 and gates (not shown) can provide an alternative mechanism to cool the reactor core 104 by convective heat transfer in the event of circulation pump loss. Reactor vessel baffling 120 provides radiation shielding and thermal insulation.

[0044] Furthermore, the reactor core primary cooling circuit 102 and the first and second heat exchange circuits 106 and 108 are preferably operated at near atmospheric pressure, and have no steam, which eliminates potential pressure explosions. In addition, the large masses of sodium, and first and second intermediate heat transfer media, provide thermal storage to decouple the reactor core 104 from plant operational upsets. Even in the unlikely event of an accident resulting in fuel cladding rupture, most radioactive fission products would stay within the molten sodium rather than disperse into the atmosphere. Furthermore, with a readily available supply of salt on hand, it is possible to capture sodium and fission products by reaction with the salt.

[0045] Figure 2 illustrates an embodiment of a heat transfer system 200 for a fast reactor core 204. System 200 includes a primary cooling circuit 202 configured to cool the fast reactor core 204. Preferably, the primary cooling circuit 202 is arranged in a loop configuration and utilizes molten sodium as a primary coolant. [0046] System 200 can include dual cooling pumps 210 and 212, which are preferably arranged in parallel, and dual reactor cooling pumps 214 and 216 that are also arranged in parallel.

Collectively, the pumps 210, 212, 214, and 216 direct the coolant from the reactor core 204 to a first heat exchanger 220 of a first heat transfer circuit 206, to a second heat exchanger 222 of a second heat transfer circuit 208, and back to the reactor core 204. The sets of dual pumps add additional redundancy to the system 200. Each of the pumps 210, 212, 214, and 216 can include a motor 211, 213, 215, and 217, respectively.

[0047] Preferably, the first and second heat exchange circuits 206 and 208 are independent of each other, and include respective first and second heat transfer media that are each substantially non-reactive with water. Each of the first and second heat transfer circuits 206 and 208 can include any commercially suitable heat exchangers including, for example, shell and tube type heat exchangers.

[0048] To further increase the safety of system 200, a safety injection conduit 218 can be fluidly coupled to the fast reactor core 204, such that additional coolant can be directly injected into the core 204, if needed, should the primary cooling circuit 202 fail. Alternatively, safety injection conduit 218 can inject salt directly into the reactor core to react with and bind to any fission products or fuels in the event of a fuel element rupture. In addition, dual heat transfer circuits 206 and 208 advantageously provide redundant heat transfer circuits, such that if circuit 206 or 208 fails, the remaining circuit can continue to cool the reactor core 204.

[0049] In an exemplary embodiment, the temperature of the sodium coolant could be about 545 °C (about 1016 °F) as it exits the reactor core 204. From the reactor core 204, the sodium can flow through the first exchanger 220, where the sodium's temperature may be reduced to about 412 °C through heat exchange contact with a first medium, preferably LiCl-KCl salt eutectic. As the sodium is cooled, the temperature of the first medium may be increased from about 385°C to 520°C.

[0050] The parallel cooling pumps 210 and 212 can be used to pump the cooler sodium coolant through the second heat exchanger 222. At the outlet of the second heat exchanger 222, it is contemplated that the sodium's temperature could drop to about 345 °C as a result of heat exchange with the second medium, preferably LBE. Such heat exchange can in turn heat the second medium from about 318 °C to about 386 °C, for example. Thus, through heat exchange contact with the first and second media, the sodium's temperature can be reduced approximately 200 °C, bringing the sodium's temperature to about 50°C lower than the 395°C reactor core inlet temperature typical of SFRs of the prior art. While specific temperatures are given in the above example, these temperatures are merely exemplary, and could vary depending on the

configuration of system 200.

[0051] In Figure 3, a superheated steam generator 302 is shown that interacts with first and second intermediate heat exchange circuits 306 and 308 of a SFR, such as that shown in Figure 2, for example. Water 370 can interact with the heated second medium of the second

intermediate heat exchange circuit 308 via a third heat exchanger 322 to produce steam 372. The steam 372 can then flow to a fourth heat exchanger 320 of a first intermediate heat exchange circuit 306 where the steam 372 is further heated by a first intermediate heat exchange medium to produce a superheated steam 374.

[0052] The first heat exchange medium can be pumped within the first heat exchange circuit 306 by pump 332. The cooled first heat exchange medium can flow from the fourth heat exchanger 320 to a first heat exchanger 310, where the first heat exchange medium can be heated by a primary sodium coolant of a fast reactor (not shown). The second heat exchange medium can be pumped through the second heat exchange circuit 308 by pump 330. The cooled second heat exchange medium can flow from the third heat exchanger 322 to a second heat exchanger 312, where the second heat exchange medium can be heated by the primary sodium coolant of the fast reactor (not shown).

[0053] From the fourth heat exchanger 320, the superheated steam 374 can flow to a high pressure turbine ("HPT") 340. At least a portion of the superheated steam 376 can be collected after passing through the HPT 340 and be reheated via a fifth heat exchanger 324 through heat exchange with the first heat exchange medium. Alternatively, the fourth and fifth exchangers 320 and 324 could comprise a single exchanger through which the first heat exchange medium can flow. The reheated steam 378 can then flow to an intermediate pressure turbine ("IPT") 360.

[0054] In some contemplated embodiments, a feed water stream can enter the third heat exchanger 322 at a temperature of approximately 289°C (553 °F). Saturated steam can exit the exchanger 322 at a temperature of approximately 357°C (675° F) and enter a fourth heat exchanger 320, where the steam is superheated to a temperature of approximately 493°C (918 °F). While specific temperatures are given, these temperatures are merely exemplary, and could vary depending on the configuration of the steam generator 302, the first and second

intermediate heat exchange circuits 306 and 308, and the SFR.

[0055] It is contemplated that the SFR can produce steam at pressures of up to approximately 2,500 psia (172.4 bar) and temperatures over 900 °F (482.2 °C), which significantly improves power generation efficiency. The arrangement of the intermediate first and second heat exchange circuits 306 and 308 allows the use of a superheater and reheater that operate with a first heat exchange medium having a high temperature. The superheater and reheater increase the temperature of steam substantially above its saturation temperature, and the steam thereby has greater superheat which increases turbine efficiency.

[0056] Figure 4 illustrates an embodiment of a power generator system 400, which shows the use of superheated steam generated by a circulating sodium-cooled reactor ("CSCFR"). Power generator system 400 receives superheated steam 474 from a superheated steam generator (not shown). At least a portion of the superheated steam 474 can flow into HPT 440, which causes HPT 440 to rotate and thereby generate power in generator 401 and causes expansion of the superheated steam 474. At least a portion of the fluid output of the HPT 440 can be fed as first and second streams 476 and 477 into one or more reheaters or feed water heaters ("FWHs"). For example, the second stream 477 can be fed into FWHs 482 to produce heated feed water stream 470 returning to the steam generator, and the first stream 476 can be heated in at least one reheater to produce superheated steam for use in an IPT.

[0057] In preferred embodiments, the first stream 476 is fed into a molten salt reheater, and the second stream 477 is fed to a LBE evaporator 470 as a feed water stream.

[0058] After passing through reheater, the reheated first stream 478 of the fluid output of the HPT 440 can then be fed into an IPT 460, which in turn causes rotation of the IPT 460. Various portions 461-464 of the fluid output of the IPT 460 can be fed into one or more FWHs 480, and an optional deaerator 482. These portions 461-464 can be collected and combined downstream, and fed to the LBE evaporator 470. It is contemplated that a separate portion of the fluid output of the IPT 460 can be fed into a low pressure turbine ("LPT") 450, to thereby cause rotation of the LPT 450. A portion 451 of the fluid output of the LPT 450 can be fed into exchanger 484, and then FWHs 480 where portion 451 can be combined with portions 461-464 and second stream 477 before the combined stream is fed to LBE evaporator 470.

[0059] The second stream 477 preferably can be used as a heat exchange medium to heat at least a portion of the fluid outputs of one or both of the IPT 460 and LPT 450 by heat exchange in one or more of the FWHs 480 or other heat exchangers. In a similar fashion, at least a portion of the fluid output of the IPT 460 can preferably be used as a heat exchange medium to heat at least a portion of the fluid output of the LPT 450, by heat exchange in one or more of the FWHs 480 or other heat exchangers. In this manner, the feed water stream can advantageously be preheated using leftover energy present in fluid outputs of the turbines before being fed into LBE evaporator 470, which increases the overall efficiency of the system 500. Of course, it is contemplated that the number and arrangement of the FWHs can be varied depending upon the specific configuration of system 400.

[0060] In Figure 5, another embodiment of a circulating sodium-cooled reactor ("CSCFR") 500 is shown, which includes a primary cooling circuit 502 configured to cool a fast reactor core 504. The primary coolant in the primary cooling circuit 502 is preferably reactor grade sodium. The reactor core 504, first and second heat transfer circuits 506 and 508, and pumps 514 and 516 in the CSCFR 500 can be arranged in a pool configuration in cylindrical reactor vessel 501 with a central baffle 520, as shown in Figure 5. The cooling pumps 514 and 516, which collectively direct the coolant from the reactor core 504 past the first and second independent heat transfer circuits 506 and 508, respectively, and back to the reactor core 504, as shown by arrows.

Cooling pumps 514 and 516 are driven by cooling pump motors 515 and 517 respectively. The cylindrical pool configuration advantageously creates a coolant flow pattern in which the heated primary coolant is directed vertically, flows radically toward the vessel outer wall, cooling as it flows down through the first and second heat transfer circuits 506 and 508. This cylindrical arrangement allows movement of the sodium coolant by convection only. [0061] Figure 6 illustrates an exemplary embodiment of the power generator system of Figure 5 generated using Steam Pro™ sold by Thermoflow™. With respect to the remaining numerals in Figure 6, the same considerations for like components with like numerals of Figure 4 apply.

[0062] Figure 7 illustrates a method 700 of producing superheated steam from a molten sodium primary coolant of a fast reactor. In step 710, the molten sodium can be cooled to a first temperature using a first heat transfer circuit to thereby convert steam to a superheated steam. The first heat transfer circuit can include a bayonet-type heat exchanger in step 715.

[0063] In step 720, the molten sodium can be further cooled to a second temperature using a second heat transfer circuit to thereby convert water to steam that can be superheated by the first heat transfer circuit. The second heat transfer circuit can include a tube-in-shell heat exchanger in step 725. It is contemplated that the temperature of the primary coolant can be reduced about 200 °C by the first and second heat transfer circuits.

[0064] Preferably, in step 730 the first and second heat transfer circuits use respective first and second heat transfer media that are each substantially non-reactive with water, and more preferably, in step 735, the first and second heat transfer media comprise a molten salt and a liquid metal eutectic, respectively. In step 737, the first heat transfer medium can be a chloride salt. In step 739, the second heat transfer medium can be a lead-bismuth eutectic.

[0065] In step 740, the molten sodium is cooled through convection only.

[0066] In Figure 8, a method 800 of generating superheated steam using heat from a molten sodium primary coolant of a fast reactor is illustrated. In step 810, saturated steam can be provided to a first heat exchanger. In step 815, water can be converted to the saturated steam by heat exchange with a second intermediary heat exchange medium that is substantially non- reactive with water. In step 817, the second intermediary heat exchange medium can be a liquid metal eutectic.

[0067] In step 820, the saturated steam can be converted to the superheated steam by heat exchange with an intermediary heat exchange medium that is substantially non-reactive with water. In step 825, the intermediary heat exchange medium can be a molten salt. [0068] In step 830, the saturated steam can have a temperature and pressure that is sufficient to allow generation of superheated steam from the intermediary heat exchange medium.

[0069] As used herein, and unless the context dictates otherwise, the term "coupled to" is intended to include both direct coupling (in which two elements that are coupled to each other contact each other) and indirect coupling (in which at least one additional element is located between the two elements). Therefore, the terms "coupled to" and "coupled with" are used synonymously.

[0070] It should be apparent to those skilled in the art that many more modifications besides those already described are possible without departing from the inventive concepts herein. The inventive subject matter, therefore, is not to be restricted except in the scope of the appended claims. Moreover, in interpreting both the specification and the claims, all terms should be interpreted in the broadest possible manner consistent with the context. In particular, the terms "comprises" and "comprising" should be interpreted as referring to elements, components, or steps in a non-exclusive manner, indicating that the referenced elements, components, or steps may be present, or utilized, or combined with other elements, components, or steps that are not expressly referenced. Where the specification claims refers to at least one of something selected from the group consisting of A, B, C .... and N, the text should be interpreted as requiring only one element from the group, not A plus N, or B plus N, etc.

Claims

CLAIMS What is claimed is:

1. A heat transfer system, comprising:

a primary cooling circuit configured to cool a fast reactor core, in which molten sodium is a primary coolant;

a superheated steam generator;

wherein the primary cooling circuit and the superheated steam generator are thermally coupled to each other by first and second independent heat transfer circuits; and wherein the first and second heat transfer circuits use respective first and second heat transfer media that are each substantially non-reactive with water.

2. The heat transfer system of claim 1 , wherein a temperature of the primary coolant is reduced by about 200 °C by the first and second independent heat transfer circuits.

3. The heat transfer system of claim 1, wherein the first and second heat transfer media comprise a molten salt and a liquid metal eutectic, respectively.

4. The heat transfer system of claim 3, wherein the first heat transfer medium comprises a chloride salt.

5. The heat transfer system of claim 3, wherein the second heat transfer medium comprises a lead-bismuth eutectic.

6. The heat transfer system of claim 1 , wherein the first heat transfer circuit comprises a bayonet-type heat exchanger.

7. The heat transfer system of claim 1, wherein the second heat transfer circuit comprises a tube-in- shell heat exchanger.

8. The heat transfer system of claim 1, wherein the primary cooling circuit is configured to allow movement of the primary coolant by convection only.

9. A method of producing superheated steam from a molten sodium primary coolant of a fast reactor, comprising: cooling the molten sodium to a first temperature using a first heat transfer circuit to thereby convert a steam to a superheated steam; and

further cooling the cooled molten sodium to a second temperature using a second heat transfer circuit to thereby convert water to the steam;

wherein the first and second heat transfer circuits use respective first and second heat transfer media that are each substantially non-reactive with the water.

10. The method of claim 9, wherein a temperature of the primary coolant is reduced about 200 °C by the first and second heat transfer circuits.

11. The method of claim 9, wherein the first and second heat transfer media comprise a molten salt and a liquid metal eutectic, respectively.

12. The method of claim 1 1, wherein the first heat transfer medium comprises a chloride salt.

13. The method of claim 1 1, wherein the second heat transfer medium comprises a lead-bismuth eutectic.

14. The method of claim 9, wherein the first heat transfer circuit comprises a bayonet-type heat exchanger.

15. The method of claim 9, wherein the second heat transfer circuit comprises a tube-in- shell heat exchanger.

16. The method of claim 9, wherein the step of cooling the molten sodium is through convection only.

17. A method of generating superheated steam using heat from a molten sodium primary coolant of a fast reactor, comprising:

providing saturated steam to a first heat exchanger;

converting the saturated steam to the superheated steam by heat exchange with an

intermediary heat exchange medium that is substantially non-reactive with water; and

wherein the saturated steam has a temperature and pressure that is sufficient to allow generation of superheated steam from the intermediary heat exchange medium.

18. The method of claim 17, wherein the step of providing saturated steam comprises converting water to the saturated steam by heat exchange with a second intermediary heat exchange medium that is substantially non-reactive with water.

19. The method of claim 17, wherein the intermediary heat exchange medium comprises a molten salt.

20. The method of claim 17, wherein the second intermediary heat exchange medium comprises a liquid metal eutectic.