Patent Publication Number: US-11393598-B2

Title: Nuclear reactor vessel support system

Description:
CROSS-REFERENCE TO RELATED PATENT APPLICATIONS 
     This application is a continuation of U.S. patent application Ser. No. 15/413,770 filed Jan. 24, 2017, which is a continuation of U.S. patent application Ser. No. 13/577,163, filed Aug. 3, 2012, which is a national stage entry of International Application No. PCT/US2011/023952, filed Feb. 7, 2011, which claims the benefit of U.S. Provisional Patent Application No. 61/416,954, filed Nov. 24, 2010, U.S. Provisional Patent Application No. 61/333,551, filed May 11, 2010, and U.S. Provisional Patent Application No. 61/302,069, filed Feb. 5, 2010, the entireties of which are herein incorporated by reference. 
    
    
     FIELD OF THE INVENTION 
     The present invention relates generally to nuclear reactor systems, and specifically to nuclear reactor systems that utilize natural circulation of the primary coolant in a single-phase, such as pressurized water reactors (“PWRs”). 
     BACKGROUND OF THE INVENTION 
     Over recent years, a substantial amount of interest has grown in developing commercially viable PWRs that utilize the phenomenon of natural circulation (also known as thermosiphon effect) to circulate the primary coolant to both cool the nuclear reactor and to vaporize a secondary coolant into motive vapor. 
     CAREM (Argentina) is a 100 MW(e) PWR reactor design with an integrated self-pressurized primary system through which the primary coolant circulation is achieved by natural circulation. The CAREM design incorporates several passive safety systems. The entire primary system including the core, steam generators, primary coolant and steam dome are contained inside a single pressure vessel. The strong negative temperature coefficient of reactivity enhances the self-controlling features. The reactor is practically self-controlled and need for control rod movement is minimized. In order to keep a strong negative temperature coefficient of reactivity during the whole operational cycle, it is not necessary to utilize soluble boron for burn-up compensation. Reactivity compensation for burn-up is obtained with burnable poisons, i.e. gadolinium oxide dispersed in the uranium di-oxide fuel. Primary coolant enters the core from the lower plenum. After being heated the primary coolant exits the core and flows up through the riser to the upper dome. In the upper part, the primary coolant leaves the riser through lateral windows to the external region, then flows down through modular steam generators, decreasing its enthalpy by giving up heat to the secondary coolant in the steam generator. Finally, the primary coolant exits the internal steam generators and flows down through the down-comer to the lower plenum, closing the circuit. CAREM uses once-through straight tube steam generators. Twelve steam generators are arranged in an annular array inside the pressure vessel above the core. The primary coolant flows through the inside of the tubes, and the secondary coolant flows across the outside of the tubes. A shell and two tube plates form the barrier between primary and secondary coolant flow circuits. 
     AST-500 (Russia) is a 500 MW(th) reactor design intended to generate low temperature heat for district heating and hot water supply to cities. AST-500 is a pressurized water reactor with integral layout of the primary components and natural circulation of the primary coolant. Features of the AST-500 reactor include natural circulation of the primary coolant under reduced working parameters and specific features of the integral reactor, such as a built-in steam-gas pressurizer, in-reactor heat exchangers for emergency heat removal, and an external guard vessel. 
     V-500 SKDI *(Russia) is a 500 MW(e) light water integral reactor design with natural circulation of the primary coolant in a vessel with a diameter less than 5 m. The reactor core and the steam generators are contained within the steel pressure vessel (i.e., the reactor pressure vessel). The core has 121 shroudless fuel assemblies having 18 control rod clusters. Thirty six fuel assemblies have burnable poison rods. The hot primary coolant moves from the core through the riser and upper shroud windows into the steam generators located in the downcomer. The coolant flows due to the difference in coolant densities in the downcomer and riser. The pressurizer is connected by two pipelines, to the reactor pressure vessel and the water clean up system. 
     The NHR-200 (China) is a design for providing heat for district heating, industrial processes and seawater desalination. The reactor power is 200 MW(th). The reactor core is located at the bottom of the reactor pressure vessel (RPV). The system pressure is maintained by N2 and steam. The reactor vessel is cylindrical. The RPV is 4.8 m in diameter, 14 m in height, and 197 tons in weight. The guard vessel consists of a cylindrical portion with a diameter of 5 m and an upper cone portion with maximum 7 m in diameter. The guard vessel is 15.1 m in height and 233 tons in weight. The core is cooled by natural circulation in the range from full power operation to residual heat removal. There is a long riser on the core outlet to enhance the natural circulation capacity. The height of the riser is about 6 m. Even in case of interruption of natural circulation in the primary circuit due to a LOCA the residual heat of the core can be transmitted by steam condensed at the uncovered tube surface of the primary heat exchanger. 
     While the aforementioned PWRs utilize natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant, all of these natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service but also subjects the equipment to corrosive conditions. Furthermore, locating the heat exchange equipment within the reactor pressure vessel results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. However, prior to the present invention, the location of the heat exchange equipment within the reactor pressure vessel was likely deemed necessary to achieve the natural circulation of the primary coolant in the PWR cycle. 
     A drawback of other PWRs that exist in the art is the fact that the reactor pressure vessels have penetrations at both the top portion of the reactor pressure vessel and at the bottom portion of the reactor pressure vessel. Still another drawback of existing PWRs is the fact that a substantial length of piping and a large number of joints are used carry the primary coolant from the reactor pressure vessel to the heat exchange equipment, thereby increasing the danger of failure due to a pipe break scenario. 
     BRIEF SUMMARY OF THE INVENTION 
     These, and other drawbacks, are remedied by the present invention. A nuclear reactor system is presented herein that, in one embodiment, utilizes natural circulation (i.e., thermosiphon) to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system, wherein the heat exchange sub-system is located outside of the nuclear reactor pressure vessel. In some embodiments, the heat exchange sub-system is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heat exchange sub-system that is used to vaporize a secondary coolant. In another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground. In certain embodiment, the inventive nuclear reactor system is a PWR system. 
     In one embodiment, the invention can be a natural circulation nuclear reactor system comprising: a reactor pressure vessel having an internal cavity; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel; a heat exchange sub-system located outside of the reactor pressure vessel; a closed-loop primary coolant circuit that flows a primary coolant through the reactor pressure vessel to cool the reactor core and through the heat exchange sub-system to transfer heat to a secondary coolant; and wherein operation of the reactor core causes natural circulation of the primary coolant through the closed-loop primary coolant circuit in a single phase. 
     In another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis, a major portion of the axial length of the reactor pressure vessel located below a ground level; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor and below the ground level; the reactor pressure vessel comprising a primary coolant outlet port located above the ground level; the reactor pressure vessel comprising a primary coolant inlet port located above the ground level; a heat exchange sub-system located outside of the reactor pressure vessel and above the ground level, an incoming hot leg of the heat exchange system fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the heat exchange system fluidly coupled to the primary coolant inlet port; and wherein the major portion of the reactor pressure vessel is free of penetrations. 
     In yet another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor; a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant outlet port in fluid communication with a top portion of the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant inlet port in fluid communication with a top portion of the primary downcomer riser passageway; at least one steam generator located outside of the reactor pressure vessel, an incoming hot leg of the steam generator fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the steam generator fluidly coupled to the primary coolant inlet port; and wherein the steam generator does not cause any substantial pressure drop in a flow of the primary coolant through the steam generator resulting from an increase in elevation. 
     Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. 
    
    
     
       BRIEF DESCRIPTION OF THE DRAWINGS 
       The present invention will become more fully understood from the detailed description and the accompanying drawings, wherein: 
         FIG. 1  is a schematic of a natural circulation nuclear reactor system according to one embodiment of the present invention. 
         FIG. 2  is a schematic of an embodiment of a heat exchange sub-system that can be used in the natural circulation reactor system of  FIG. 1 . 
         FIG. 3A  is a schematic top view of a single-pass horizontal steam generator in accordance with an embodiment of the present invention. 
         FIG. 3B  is a schematic side view of the single-pass horizontal steam generator of  FIG. 3A . 
         FIG. 4  is a side view of a portion of the natural circulation nuclear reactor system of  FIG. 1  according to one structural embodiment. 
         FIG. 5  is an elevated isometric view of a portion of the natural circulation nuclear reactor system of  FIG. 1  according to one structural embodiment. 
     
    
    
     DETAILED DESCRIPTION OF THE DRAWINGS 
     The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. 
     Prior to discussing  FIGS. 1-5  in detail, an overview of one specific embodiment of the inventive natural circulation reactor system, and its operation, will be set forth. Those skilled in the art will appreciate that the overview is directed to one very specific embodiment and that the details thereof are not limiting of the present invention in all embodiments. Furthermore, those skilled in the art will appreciate how the overview applies to the subsequent detailed discussion of  FIGS. 1-5 . 
     I. Overview of One Potential Commercial Embodiment 
     The inventive nuclear reactor system, in one potential commercial embodiment, is a 145 MWe nuclear reactor designed to provide an economical and safe source of clean energy from nuclear fission. Strengths of the inventive nuclear reactor system include its inherent safety and simplicity of operation. The operational simplicity of the inventive nuclear reactor system and the modest outlay required to establish and commission it will make it possible to deliver the fruits of pollution-free nuclear energy to the vast mass of humanity around the globe that does not presently have access to a reliable source of power or to a robust electrical energy delivery system. Competitive with large nuclear reactors on a per-megawatt basis, the inventive nuclear reactor system is tailored to add generation capacity to the installed base incrementally with incremental capital outlays. Due to its inherent operational simplicity, the inventive nuclear reactor system requires a minimal cadre of trained personnel to run the plant. Multiple units of the inventive nuclear reactor system can be clustered at one location or geographically dispersed without a significant increase in the per-megawatt construction cost. Geographical dispersal and underground configuration serve as natural antidotes to post-9/11 concerns. The modest power output of the inventive nuclear reactor system makes it a viable candidate source of reliable electrical energy or for providing heating steam to a city or process steam as a cogeneration plant serving an industrial plant. 
     As a passive small modular reactor of the PWR genre with safety, ease of maintenance and superb security, the inventive nuclear reactor system is ideally suited to serve as a reliable power source to strategic national assets of any country. Design features of the inventive nuclear reactor system that speak to its inherent safety and reliability are: 
     1. Reactor Core Deep Underground 
     The reactor core resides deep underground in a thick-walled reactor pressure vessel (RPV) made of an ASME Code material that has decades of proven efficacy in maintaining reactor integrity in large PWR and BWR reactors. All surfaces wetted by the reactor coolant are made of stainless steel or Inconel, which eliminates a major source of crud accumulation in the reactor vessel. 
     2. Natural Circulation of the Reactor Coolant 
     The inventive nuclear reactor system does not rely on any active components, such as a reactor coolant pump, for circulating the primary coolant through the closed-loop primary coolant circuit, which includes flow through the reactor core and the heat exchange sub-system. Instead, the flow of the primary coolant through the reactor pressure vessel, the horizontal steam generators, and other miscellaneous equipment occurs by the pressure head created by density differences in the flowing water in the hot and cold segments of the closed-loop primary coolant circuit. The reliability of gravity as a motive force underpins inherent safety of the inventive nuclear reactor system. The movement of the primary coolant requires no pumps, valves, or moving machinery of any kind, in certain embodiments. 
     3. No Reliance on Off-Site Power 
     Offsite power is not essential for shutting down the inventive nuclear reactor system. The rejection of reactor residual heat during the shutdown also occurs by natural circulation. Thus, the need for an emergency shutdown power supply at the site—a major concern for nuclear plants—is eliminated. 
     4. Assurance of a Large Inventory of Water Around and Over the Reactor Core 
     The reactor pressure vessel of the inventive nuclear reactor system has no penetrations in its below-ground portion, which can be the bottom 100 feet, which means that the reactor core will remain submerged in a large inventory of water. All penetrations in the reactor pressure vessel are located in the above-ground portion, or top portion, of the reactor pressure vessel and are small in size. The absence of large piping in the closed-circuit primary coolant circuit precludes the potential of a “large break” LOCA event. 
     5. All Critical Components Readily Accessible 
     Both the heat exchange sub-system, which includes the steam generators, and the control rod drive system are located outside the reactor pressure vessel at a level that facilitates easy access, making their preventive maintenance and repair a conveniently executed activity. Each of the steam generators is a horizontal pressure vessel with built-in design features to conveniently access and plug tubes. 
     6. Demineralized Water 
     The primary coolant (which can also be referred to as the reactor coolant) is demineralized water, which promotes criticality safety because of its strong negative reactivity gradient with rise in temperature. Elimination of borated water also simplifies the nuclear steam supply system (NSSS) by eliminating the systems and equipment needed to maintain and control boron levels in the primary coolant. Pure water and corrosion resistant primary coolant loop help minimize crud buildup in the reactor pressure vessel. 
     7. Modularity 
     One can build only one of the inventive nuclear reactor systems at a site, or a large number thereof. Clustering a number of inventive nuclear reactor systems at one site will reduce the overall O&amp;M costs. 
     8. Long Operating Cycle 
     The inventive nuclear reactor system will operate for approximately 3.5 years before requiring refueling. 
     9. Short Construction Life Cycle 
     Virtually all components of the inventive nuclear reactor system are shop fabricated. Site work is limited to reinforced concrete construction and a limited amount of welding to assemble the shop-built equipment and parts. As a result, it is possible to complete the construction of one of the inventive nuclear reactor systems in 24 months from the first shovel in the ground. 
     10. Efficient Steam Cycle 
     A pair of two horizontal steam generators are arranged in series and integrally welded to the reactor pressure vessel. The efficiency of the power cycle of the inventive nuclear reactor system, and its compactness, is further enhanced by superheaters that are integrally welded to the horizontal steam generators. The superheaters, one attached to each steam generator, increases cycle efficiency and also protect both the high pressure and low pressure turbines from the deleterious effect of moist steam. 
     11. Integral Pressurizer 
     The design of the reactor pressure vessel incorporates an integral pressurizer that occupies the upper reaches of the reactor pressure vessel. The pressurizer serves to control the pressure in the reactor vessel. 
     12. Suitable for Water-Challenged Sites 
     The inventive nuclear reactor system can be installed at sites with limited water availability, such as creeks and small rivers that are inadequate for large reactors. The inventive nuclear reactor system can be operated equally well in a water-challenged region by using air-cooled condenser technology to reject the plant&#39;s waste heat. Using air in lieu of water, of course, results in a moderate increase in the plant&#39;s cost. 
     12. System Parameters in the Safe and Proven Range 
     The operating pressure and temperature within the reactor pressure vessel is in the proven range for PWRs. Lower core power density than that used in large PWRs for improved thermal-hydraulic control (please see table below) and an improved margin to departure-from-nucleate boiling in the reactor core. 
     
       
         
           
               
               
               
             
               
                   
                   
               
               
                   
                 Exemplary System Parameters 
                 Data 
               
               
                   
                   
               
             
            
               
                   
               
            
           
           
               
               
               
            
               
                   
                 Number of fuel assemblies in the core 
                 32 
               
               
                   
                 Nominal thermal power, MWt 
                 446 
               
               
                   
                 Nominal recirculation rate, MLb per hour 
                 5.46 
               
               
                   
                 Reactor water outlet temperature, deg. F. 
                 580 
               
               
                   
                 Reactor water inlet temperature, deg. F. 
                 333 
               
               
                   
                 Reactor pressure, pounds per sq. inch 
                 2.250 
               
               
                   
                 Water in the RV cavity, gallons 
                 30.00 
               
               
                   
                   
               
            
           
         
       
     
     13. Minimized Piping Runs and Minimum Use of Active Components to Enhance Reliability and Cost Competitiveness 
     The amount of piping in the close-loop primary coolant circuit and the secondary coolant circuit in the inventive nuclear reactor system is the least of any nuclear plant design on the market, as is the number of pumps and valves. 
     14. In-Service Inspection 
     All weld seams in the primary system including those in the reactor pressure vessel wall are available at all times for inspection. In particular, the weld seams in the reactor pressure vessel can be inspected by operating a manipulator equipped in-service inspection device in the reactor well during power generation. Thus, inventive nuclear reactor system exceeds the in-service inspection capability expected of nuclear plants under ASME Code Section XI. 
     15. Earthquake Hardened Design 
     Virtually all major equipment in the inventive nuclear reactor system are either underground or horizontally mounted to withstand strong seismic motions. This includes the reactor pressure vessel, the fuel pool, the reactor water storage tank (all underground) and the horizontal steam generators, the horizontal superheaters, and the horizontal kettle reboiler that are floor mounted. 
     16. Aircraft Impact Proof Containment 
     The containment structure of the inventive nuclear react system is designed to withstand the impact of a crashing fighter plane or a commercial liner without sustaining a thru-wall breach. 
     II. Detail 
     Referring now to  FIG. 1 , a natural circulation nuclear reactor system  1000  (hereinafter the “reactor system  1000 ”) is illustrated according to one embodiment of the present invention. The reactor system  1000  generally comprises a reactor pressure vessel  100  and a heat exchange sub-system  200 . The reactor pressure vessel  100  contains a primary coolant  101  that is used to cool the rector core  102  and to heat a secondary coolant within the heat exchange sub-system  200 . The reactor pressure vessel  100  is fluidly coupled to an incoming hot leg  201  of the heat exchange sub-system  200  via a primary coolant outlet port  103 . Similarly, the reactor pressure vessel  100  is also fluidly coupled to an outgoing cool leg  202  of the heat exchange sub-system  200  via a primary coolant inlet port  104 . As a result, a closed-loop primary coolant circuit  300  is formed through which the primary coolant  101  flows in a single-phase. As discussed in greater detail below, the flow of the primary coolant  101  through the closed-loop primary coolant circuit is a natural circulation flow induced by the heat given off by the normal operation of the reactor core  102 . 
     In certain embodiments, the internal cavity  105  of the reactor pressure vessel  100  is maintained under sufficient pressure to maintain the primary coolant  101  in a liquid-phase despite the high temperature within the rector pressure vessel  100 . In the exemplified embodiment, a pressure control sub-system  50  (commonly referred to in the art as a pressurizer) is located within a top region of the reactor pressure vessel  100  and is configured to control the pressure of the internal cavity  105  of the reactor pressure vessel  100 . The pressure control sub-system  50  is integral with the removable head  106  of the reactor pressure vessel  100  to prevent line break concerns and to provide a more compact reactor system  1000 . Pressurizers are well known in the art and any standard pressurizer could be used as the pressure control sub-system  50 . In one embodiment, the internal cavity  105  of the reactor pressure vessel  100  is maintained at a pressure in a range of 2000 psia to 2500 psia. In one more specific embodiment, the internal cavity  105  of the reactor pressure vessel  100  is maintained at a pressure between 2200 psia to 2300 psia. Of course, the exact pressure maintained in the internal cavity  105  of the reactor pressure vessel  100  is not to be limiting of the invention unless specifically claimed. 
     The reactor pressure vessel  100  is an elongated tubular pressure vessel formed by a thick wall made of an acceptable ASME material, such as stainless steel. The reactor pressure vessel  100  extends from a bottom end  107  to a top end  108  along a substantially vertical axis A-A, thereby defining an axial length of the reactor pressure vessel  100 . In one embodiment, the reactor pressure vessel  100  has an axial length of over 100 feet to facilitate an adequate level of turbulence in the recirculating primary coolant  101  from the natural circulation (also referred to as thermosiphon action in the art). In certain other embodiments, the reactor pressure vessel  100  has an axial length in a range between 100 feet to 150 feet. Of course, the invention is not so limited in certain alternate embodiments. 
     The reactor pressure vessel  100  generally comprises a domed head  106  and a body  109 . The domed head  106  is detachably coupled to a top end of the body  109  so as to be removable therefrom for refueling and maintenance. The domed head  106  can be coupled to the body  109  through the use of any suitable fastener, including bolts, clamps, or the like. In the exemplified embodiment, the body  109  comprises an upper flange  110  and the domed head  106  comprises a lower flange  111  that provided mating structures through which bolts  114  ( FIG. 4 ) extend to couple the domed head  106  to the body  109 . When the domed had  106  is coupled to the body  109 , a hermetic seal is formed therebetween via the use of a gasket or other suitably contoured interface. 
     The body  109  of the reactor pressure vessel  100  comprises an upstanding tubular wall  112  and a domed bottom  113  that hermetically seals the bottom end  107  of the reactor pressure vessel  100 . The tubular wall  112  has a circular transverse cross-sectional profile in the illustrated embodiment but can take on other shapes as desired. In the exemplified embodiment, the domed bottom  113  is integral and unitary with respect to the tubular wall  112 . Of course, in other embodiments, the domed bottom  113  may be a separate structure that is secured to the tubular wall  112  via a welding or other hermetic connection technique, such as the flanged technique described above for the domed head  106  and the body  109 . Integral and unitary construct of the domed bottom  113  and the body  109  is, however, preferable in certain embodiments as it eliminates seams and/or interfaces that could present rupture potential. 
     The reactor pressure vessel  100  forms an internal cavity  105  in which a reactor core  102  is housed. The reactor core  102  comprises nuclear fuel, in the form of fuel assemblies, as is known in the art. The details of the structure of the reactor core  102  are not limiting of the present invention in and the reactor system  1000  can utilize any type of reactor core or nuclear fuel. The reactor core  102  is positioned in a bottom portion  115  of the reactor pressure vessel  100 . In one embodiment, the reactor core  102  has a core thermal power of 400 MWt to 600 MWt during the operation thereof. 
     In one embodiment, the reactor core  102  is comprised of vertically arrayed fuel assemblies. The spacing between the fuel assemblies is governed by the design objective of keeping the reactivity (neutron multiplication factor) at 1.0 at all locations in the reactor pressure vessel  100 . The criticality control in the axial direction is provided by the built-in neutron poison in the fuel rods (called IFBAs by Westinghouse) and possibly by control rods. 
     A partition  120  is provided within the internal cavity  105  of the reactor pressure vessel  100  that divides the internal cavity into a primary coolant riser passageway  105 A and a primary coolant downcomer passageway  105 B. Both the passageways  105 A,  105 B are axially extending vertical passageways that form part of the closed-loop primary coolant circuit  300 . 
     In the exemplified embodiment, the partition  120  comprises an upstanding tubular wall portion  120 A and a transverse wall portion  120 B. The tubular wall portion  120 A is an annular tube that is mounted within the internal cavity  105  of the reactor pressure vessel  100  so as to be concentrically arranged with respect to the upstanding wall  112  of the reactor pressure vessel  100 . As a result, the primary coolant downcomer passageway  105 B is an annular passageway that circumferentially surrounds the primary coolant riser passageway  105 A. The primary coolant downcomer passageway  105 B is formed between an outer surface  121  of the upstanding tubular wall portion  120 A of the partition  120  and the inner surface  116  of the upstanding wall  112  of the reactor pressure vessel  100 . The primary coolant riser passageway  105 B is formed by the inner surface  122  of the upstanding tubular wall portion  120 A of the partition  120 . 
     The transverse wall portion  120 B is an annular ring-like plate that is connected to a top end of the of the upstanding tubular wall portion  120 A of the partition  120  at one end and to the upstanding wall  112  of the reactor pressure vessel  100  on the other end. The transverse wall portion  120 B acts a separator element that prohibits cross-flow of the primary coolant  101  between the primary coolant riser passageway  105 A and the primary coolant downcomer passageway  105 B within the top portion  117  of the reactor pressure vessel  100 . In essence, the transverse wall portion  120 B forms a roof of the primary coolant downcomer passageway  105 B that prevents the heated primary coolant  101  that exits the reactor pressure vessel  100  via the primary coolant outlet port  103  from mixing with the cooled primary coolant  101  that enters the reactor pressure vessel  100  via the primary coolant inlet port  104 , and vice-versa. Cross-flow of the primary coolant  101  between the primary coolant riser passageway  105 A and the primary coolant downcomer passageway  105 B is prohibited by the upstanding tubular wall portion  120 A of the partition  120 . 
     In addition to physically separating the flow of the heated and cooled primary coolant  101  within the primary coolant downcomer and riser passageways  105 A,  105 B as discussed above, the partition  120  also thermally insulates the cooled primary coolant  101  within the primary coolant downcomer passageway  105 B from the heated primary coolant  101  within the primary coolant riser passageway  105 A. Stated simply, one does not want heat to transfer freely through the partition  120 . Thus, it is preferred that the partition  120  be an insulating partition in the sense that its effective coefficient of thermal conductivity (measured radially from the primary coolant riser passageway  105 A to the primary coolant downcomer passageway  105 B) is less than the coefficient of thermal conductivity of the primary coolant  101 . 
     Making the effective coefficient of thermal conductivity of the partition  120  less than the coefficient of thermal conductivity of the primary coolant  101  ensures that the primary coolant  101  in the primary coolant downcomer passageway  105 B remains cooler than the primary coolant  101  in the primary riser passageway  105 A, thereby maximizing the natural circulation rate of the primary coolant  101  through the closed-loop primary coolant circuit  300 . In a very simple construction, this can be achieved by creating the partition  120  out of a single solid material that has a low coefficient of thermal conductivity. However, it must be considered that the material should neither degrade nor deform under the operating temperatures and pressures of the reactor pressure vessel  100 . In such an embodiment, the effective coefficient of thermal conductivity is simply the coefficient of thermal conductivity of the single solid material. 
     In the exemplified embodiment, the low coefficient of thermal conductivity of the partition  120  is achieved by making the partition  120  as a multi-layer construction. As exemplified, the partition  120  comprises an insulating layer  124  that is sandwiched between two outer layers  125 A,  125 B. In one embodiment, the insulating layer  124  is a refractory material while the outer layers  125 A,  125 B are stainless steel or another corrosion resistant material. In certain embodiments, the insulating layer  124  is full encased in the outer layers  125 A,  125 B. 
     The internal cavity  115  of the reactor pressure vessel  100  also comprises a plenum  118  at the bottom portion  115  of the reactor pressure vessel  100  that allows cross-flow of the primary coolant  101  from the primary coolant downcomer passageway  105 B to the primary coolant riser passageway  105 A. In the exemplified embodiment, the plenum  118  is created by the fact that the bottom end  123  of the upstanding tubular wall portion  120 A of the partition  120  is spaced from the inner surface  119  of the domed bottom  113 , thereby creating an open passageway. In alternate embodiments, the partition  120  may extend all the way to the inner surface  119  of the domed bottom  113 . In such embodiments, the plenum  118  will be formed by providing a plurality of apertures/openings in the partition  120  so as to allow the desired cross-flow. 
     The internal cavity  105  further comprises a plenum  126  at the top portion  117  of the reactor pressure vessel  100 . The plenum  126  allows the heated primary coolant  101  that is rising within the primary coolant riser passageway  105 A to gather in the top portion  117  of the reactor pressure vessel  100  and then flow transversely outward from the vertical axis A-A and through the primary coolant outlet port  103 . 
     The reactor core  102  is located within the primary coolant riser passageway  105 A above the bottom plenum  118 . During operation of the reactor core  102 , thermal energy produced by the reactor core  102  is transferred into the primary coolant  101  in the primary coolant riser passageway  105 A adjacent the reactor core  102 , thereby becoming heated. This heated primary reactor coolant  101  rises upward within the primary coolant riser passageway  105 A due to its decreased density. This heated primary coolant  101  gather in the top plenum  126  and exits the reactor pressure vessel  100  via the primary coolant outlet port  103  where it enters the heat exchange sub-system  200  as the incoming hot leg  201 . In one embodiment, the heated primary coolant  101  entering the hot leg  201  of the heat exchanger has a temperature of at least 570° F., and in another embodiment a temperature in a range of 570° F. to 620° F. 
     This heated primary coolant  101  passes through the heat exchange sub-system  200  where its thermal energy is transferred to a secondary coolant (described below in greater detail with respect to  FIG. 2 ), thereby becoming cooled and exiting the heat exchange sub-system  200  via the cold leg  202 . When exiting the cold leg  202  of the heat exchange sub-system, this cooled primary coolant  101  has a temperature in a range of 300° F. to 400° F. in one embodiment. In another embodiment, the heat exchange sub-system  200  is designed so that the temperature differential between the heated primary coolant in the hot leg  201  and the cooled primary coolant in the cold leg is at least 220° F. 
     The cooled primary coolant  101  exiting the cold leg of the heat exchange sub-system  200  then enters the reactor pressure vessel  100  via the primary coolant inlet port  104 , thereby flowing into a top portion  127  of the primary coolant downcomer passageway  105 B. Once inside the primary coolant downcomer passageway  105 B, the cooled primary coolant  101  (which has a greater density than the heated primary coolant  101  in the primary coolant riser passageway  105 A) flows downward through the primary coolant downcomer passageway  105 B into the bottom plenum  118  where it is drawn back up into the primary coolant riser passageway  105 A and heated again by the reactor core  102 , thereby completing a cycle through the closed-loop primary circuit  300 . 
     As discussed above, operation of the reactor core  102  causes natural circulation of the primary coolant  101  through the closed-circuit primary coolant circuit  300  by creating a riser water column within the primary coolant riser passageway  105 A and a downcomer water column within the primary coolant downcomer passageway  105 B. In one embodiment, the riser water column and the downcomer water column have a vertical height in a range of 80 ft. to 150 ft., and more preferably from 80 ft. to 120 ft. The vigorousness of the natural circulation (or thermosiphon flow) is determined by the height of the two water columns (fixed by the reactor design), and the difference between the bulk temperature of the two water columns (in water the SES and the downcomer space). For example, water at 2200 psia and 580° F. has density of 44.6 lb/cubic feet. This density increases to 60.5 lb/cubic feet if the temperature reduces to 250° F. The hot and cold water columns 60 feet high will generate a pressure head of 6.6 psi which is available to drive natural circulation of the primary coolant  101  through the closed-loop primary coolant circuit  300 . A 90 feet high column will generate 50% greater head (i.e., 9.9 psi). 
     As a result of the natural circulation of the primary coolant  101  achieved by the water columns and gravity, the reactor system  1000  is free of active equipment, such as pumps or fans, for forcing circulation of the primary coolant through the closed-loop primary coolant circuit. 
     In the embodiment illustrated in  FIG. 1 , the primary coolant outlet port  103  is at a slightly lower elevation (1-3 ft.) than the primary coolant inlet port  104 . However, in other embodiments, the primary coolant outlet port  103  and the primary coolant inlet port  104  will be at substantially the same elevation (see  FIGS. 4 and 5 ). When the primary coolant outlet port  103  and the primary coolant inlet port  104  are at substantially the same elevation the partition  120  will be appropriately designed. Furthermore, as used herein, the term port includes mere apertures or openings. 
     In one embodiment, the primary coolant  101  is a liquid that has a negative reactivity coefficient. Thus, the chain reaction in the reactor core  102  would stop automatically if the heat rejection path to the heat exchange sub-system  200  is lost in a hypothetical scenario. Thus, the reactor system  1000  is inherently safe. In one specific embodiment, the primary coolant  101  is demineralized water. All systems and controls used to maintain boron concentration in the reactor vessel in a typical PWR are eliminated from the reactor system  1000 . Moreover, the use of demineralized water as the primary coolant  101  and the existence of the corrosion resistant surfaces of the reactor pressure vessel  100  help maintain crud buildup to a minimum. The reactivity control in the reactor core  102  is maintained by a set of control elements (burnable poisons) that are suspended vertically and occupy strategic locations in and around the fuel assemblies to homogenize and control the neutron flux. 
     Referring now to  FIGS. 1, 4 and 5  concurrently, it can be seen that a major portion  130  of the axial length of the reactor pressure vessel  100  located below a ground level  400  while a minor portion  131  of the axial length of the reactor pressure vessel  100  extends above the ground level  400 . As such, the reactor core  102  is located deep below the ground level  400  while the heat exchange sub-system  200  is located above the ground level  400 . In one embodiment, the heat exchange sub-system  200  is at an elevation that is 80 ft. to 150 ft, and preferably 80 ft. to 120 ft., greater than the elevation of the reactor core  102 . 
     The minor portion  131  of the reactor pressure vessel  100  includes a top portion  132  of the body  109  and the domed head  106 . The primary coolant outlet port  103  and the primary coolant inlet port  104  are located on the minor portion  131  of the reactor pressure vessel  100  that is above the ground level  400 . More specifically, the primary coolant outlet port  103  and the primary coolant inlet port  104  are located on the top portion  132  of the body  109  of the reactor pressure vessel  100  that is above the ground level  400 . 
     The major portion  130  includes a majority of the body  109  and the domed bottom  113 . In certain embodiment, the major portion  130  of the reactor pressure vessel  130  is at least 75% of the axial length of the reactor pressure vessel  100 . In other embodiments, the major portion  130  of the reactor pressure vessel  130  is between 60% to 95% of the axial length of the reactor pressure vessel  100 . In another embodiment, the major portion  130  of the reactor pressure vessel  130  is between 75% to 95% of the axial length of the reactor pressure vessel  100 . 
     The reactor pressure vessel  100  comprises a reactor flange assembly  150  comprising a first reactor flange  151  and a second reactor flange  153 . The top portion  132  of the body  109  of the reactor pressure vessel  100  is welded to the reactor flange assembly  150 , which is a massive upper forging. The reactor flange assembly  150  also provides the location for the primary coolant inlet port  104  and the primary coolant outlet port  103  ( FIGS. 4 and 5 ), and the connections to the heat exchange sub-system  200  (and for the engineered safety systems to deal with various postulated accident scenarios). This reactor flange assembly  150  contains vertical welded lugs  152  to support the weight of the reactor pressure vessel  100  in the reactor well  410  in a vertically oriented cantilevered manner ( FIGS. 1 and 4 ). As a result, the reactor pressure vessel  100  is spaced from the wall surfaces  411  and floor surface  412  of the reactor well  410 , thereby allowing the reactor pressure vessel  100  to radially and axially expand as the reactor core  102  heats up during operation and causes thermal expansion of the reactor pressure vessel  100 . 
     Furthermore, the major portion  130  of the reactor pressure vessel  100  is free of penetrations. In other words, the major portion  130  of the reactor pressure vessel  100  comprises no apertures, holes, opening or other penetrations that are either open or to which pipes or other conduits are attached. All penetrations (such as the primary coolant inlet and outlet ports  103 ,  104 ) in the reactor pressure vessel  100  are located in the above-ground minor portion  131 , and more specifically in the top portion  132  of the body  109  of the reactor pressure vessel  100 . In one embodiment, it is further preferred that the major portion  130  be a unitary construct with no connections, joints, or welds. 
     The bottom portion  115  of the reactor pressure vessel  100  is laterally restrained by a lateral seismic restraint system  160  that spans the space between the body  109  of the reactor pressure vessel  100  and the wall surfaces  411  of the reactor well  410  to withstand seismic events. The seismic restraint system  160 , which comprises a plurality of resiliently compressible struts  161 , allows for free axial and diametral thermal expansion of the reactor vessel. The bottom of the reactor well  410  contains engineered features to flood it with water to provide defense-in-depth against a (hypothetical, non-mechanistic) accident that produces a rapid rise in the enthalpy of the reactor&#39;s contents. Because the reactor system  1000  is designed to prevent loss of the primary coolant  101  by leaks or breaks and the reactor well  410  can be flooded at will, burn-through of the reactor pressure vessel  100  by molten fuel (corium) can be ruled out as a credible postulate. This inherently safe aspect simplifies the design and analysis of the reactor system  1000 . 
     Referring now to  FIGS. 2 and 4-5  concurrently, an embodiment of the heat exchange sub-system  200  is illustrated. While a specific embodiment of the heat exchange sub-system  200  will be described herein, it is to be understood that, in alternate embodiments, one or more of components can be omitted as desired. For example, in certain embodiments, one or both of the horizontal superheaters  205 ,  206  may be omitted. In certain other embodiments, one of the horizontal steam generators  203 ,  204  may be omitted and/or combined into the other one of the horizontal steam generators  203 , 204 . Moreover, additional equipment may be incorporated as necessary so long as the natural circulation of the primary coolant  101  through the closed-loop primary coolant circuit  300  is not prohibited through the introduction of substantial head loss. 
     As mentioned above, the heat exchange subsystem  200  comprises an incoming hot leg  201  that introduces heated primary coolant into the portion of the closed-loop primary coolant circuit  300  that passes through the heat exchange sub-system  200  and an outgoing cold leg  202  that removes cooled primary coolant from the portion of the closed-loop primary coolant circuit  300  that passes through the heat exchange sub-system  200 . In order to minimize (and in some embodiments eliminate) pressure loss in the closed-loop primary coolant circuit  300  caused by an increase in the elevation of the primary coolant flow, the steam generators  203 ,  204  and the superheaters  205 ,  206  are all of the horizontal genre (i.e., the tubes which carry the primary coolant extend substantially horizontal through the shell-side fluid) and are in horizontal alignment with each other where possible. 
     Within the heat exchange sub-system  200 , the primary coolant flow of the closed-loop primary coolant circuit  300  is divided into two paths  211 ,  212  at a flow divider  215 . The flow divider  210  can be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. The first path  211 , which carries the majority of the primary coolant flow, travels through the first horizontal steam generator  203  and then through the second horizontal steam generator  204 . Meanwhile, the second path  212 , which carries a minority of the primary coolant flow, travels through the first horizontal superheater  205  and then through the second horizontal superheater  206 . After passing through the first and second horizontal steam generators  203 ,  204  and the first and second horizontal superheaters  205 ,  206 , the first and second paths  211 ,  212  converge in a flow converger  216 , which combines the primary coolant flows of the first and second paths  211 ,  212  and directs the combined flow to the outgoing cold leg  202 . As with the flow divider  215 , the flow converger  216  may be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. 
     In one embodiment, 10% to 15% of the incoming primary coolant flow that enters the heat exchange sub-system  200  via the hot leg  201  is directed into the second path  212  while the remaining 85% to 90% of the incoming primary coolant is directed into the first path  211 . In one specific example, the incoming primary coolant that enters the heat exchange sub-system  200  via the hot leg  201  has a flow rate of 5 to 7 million lbs./hr. In this example, 0.6 to 1 million lbs./hr. of the primary coolant is directed into the second path  212  while the remainder of the primary coolant flow is directed into the first path  211 . 
     The first and second horizontal steam generators  203 ,  204  are operbaly coupled in series to one another along the first path  211  of the closed-loop primary coolant circuit  300 . Both of the horizontal steam generators  203 ,  204  are horizontally disposed shell-and-tube heat exchangers. The first horizontal steam generator  203  is a high pressure steam generator while the second horizontal steam generator  204  is a low pressure steam generator (in comparison to the high pressure steam generator). The high first steam generator  203  is located upstream of the second horizontal steam generator  204  along the closed-loop primary coolant circuit  300 . Similarly, the first and second horizontal superheaters  205 ,  206  are operbaly coupled in series to one another along the second path  212  of the closed-loop primary coolant circuit  300 . The first horizontal superheater  205  is a high pressure superheater while the second horizontal superheater  206  is a low pressure superheater (in comparison to the high pressure superheater). The high first steam superheater  205  is located upstream of the second horizontal superheater  206  along the closed-loop primary coolant circuit  300 . Furthermore, the first and second superheaters  205 ,  206  are located in parallel to the first and second horizontal steam generators  203 ,  204  along the closed-loop primary coolant circuit  300 . 
     Furthermore, the first and second horizontal steam generators  203 ,  204  are interconnected by a return header so that the hot primary coolant entering the first horizontal steam generator  203  heats the secondary coolant to make steam for the high-pressure turbine  220  and then proceeds to the second horizontal steam generator  204  with minimal pressure loss to make steam for the low-pressure turbine  221 . 
     The flow of the primary coolant in the first path  211  is used to convert a secondary coolant flowing through the shell-side of the first and second horizontal steam generators  203 ,  204  from liquid-phase to gas-phase through the transfer of heat form the primary coolant to the secondary coolant within the first and second horizontal steam generators  203 ,  204 . Because the flow of the primary coolant through the first and horizontal second steam generators  203 ,  204  is substantially horizontal in nature, the flow of the primary coolant through the first path  211  does not cause any substantial pressure drop in the closed-loop primary coolant circuit  300  resulting from an increase in elevation. Moreover, because of the horizontal alignment of the first and second horizontal steam generators  203 ,  204  with each other and the primary coolant outlet and inlet ports  103 ,  104  of the reactor pressure vessel  100  ( FIG. 5 ), the primary coolant flow that travels along the first path  211  from the primary coolant outlet port  103  of the reactor pressure vessel  100  to the primary coolant inlet port  104  of the reactor pressure vessel  100  does not cause any substantial pressure drop in the closed-loop primary coolant circuit  300  resulting from an increase in elevation. While the achievement of substantial zero pressure drop in the closed-loop primary coolant circuit  300  resulting from an increase in elevation is exemplified in terms of a horizontal flow, it is possible that such substantial zero pressure drop can be achieved by a decline in elevation as the primary coolant flows downstream in the closed-loop primary coolant circuit  300 . 
     The flow of the primary coolant in the second path  212  is used to superheat the vapor-phase of the secondary coolant exiting the first and second horizontal steam generators  203 ,  204  via the first and second horizontal superheaters  205 ,  206  respectively, thereby further drying the vapor-phase of the secondary coolant. The use if the horizontal superheaters enhance the thermodynamic efficiency of the turbine cycle, carried out on the high pressure turbine  220  and the low pressure turbine  221 . 
     The first and second horizontal superheaters  205 ,  206  are horizontally disposed shell-and-tube heat exchanger positioned directly above (and in series) with the first and second steam generators  203 ,  204  ( FIG. 5 ). However, due to the slight increase in the elevation of the superheaters  205 ,  206  resulting from their location above the first and second horizontal steam generators  203 ,  204 , the flow of the primary coolant in the second path  212  does cause some pressure drop in the closed-loop primary coolant circuit  300  resulting from an increase in elevation. However, because only a small amount (10% to 15%) of the total primary coolant that flows through the heat exchange subsystem  200  is directed into the second path  212  and through the horizontal superheaters  205 ,  206 , the pressure drop does not significantly affect the desired natural circulation. Moreover, the increase in elevation is negligible when compared to the height of the flow driving water columns. In such an embodiment, at least 85% of the flow of the primary coolant through the heat exchange sub-system  200  is still entirely horizontal from the primary coolant outlet  103  to the primary coolant inlet  104  and does not cause any substantial pressure drop in the closed-loop primary coolant circuit  300  due to increase in elevation. Further, in certain alternate embodiments, the horizontal superheaters  205 ,  206  could be eliminated and/or repositioned to be in horizontal alignment with the horizontal steam generators  203 ,  204 . 
     As shown in  FIG. 5 , the first and second horizontal steam generators  203 ,  204  are coupled directly to the each other and to the reactor pressure vessel  100 . More specifically, the inlet of the first horizontal steam generator  203  is coupled directly to the primary coolant outlet port  103  of the reactor pressure vessel  100  while the outlet of the first horizontal steam generator  203  is coupled directly to the inlet of the second horizontal steam generator  204 . The outlet of the second horizontal steam generator  204 , is in turn, coupled directly to the primary coolant inlet port  104  of the reactor pressure vessel  100 . The first and second horizontal steam generators  203 ,  204  are arranged so as to extend substantially parallel to one another, thereby collectively forming a generally U-shaped structure. Thus, the first path  211  also takes on a generally U-shape In certain embodiments, the first and second horizontal steam generators  203 ,  204  are integrally welded to the reactor vessel  100  and to each other. 
     Referring now to  FIGS. 2 and 3A -B, each of the first and second horizontal steam generators  203 ,  204  comprise a preheating zone  208 ,  210  and a boiling zone  207 ,  209 . Both of the first and second horizontal steam generators  203 ,  204  are of the single-pass type in which the primary coolant flow of the first path  211  is the tube-side fluid. Each of the single-pass tubes  330  extend substantially horizontally through the preheating zones  208 ,  210  and the boiling zones  207 ,  209 . The secondary coolant circuit has a main feedwater intake  501  and a return to condenser exit  502  into and out of the heat exchange sub-system  200  respectively. 
     The secondary coolant, which is in the liquid-phase  505 , enters each of the first and second horizontal steam generators  203 ,  204  along line  503 . The incoming liquid phase  505  of the secondary coolant is preheated within the preheater zones  208 ,  210  of the first and second horizontal steam generators  203 ,  204 . The secondary coolant in liquid-phase  505  flows through a tortuous path as shell-side fluid in the preheater zones  208 ,  210  and then enters the boiling zones  207 ,  209 , where it is further heated by the primary coolant flow passing through the tubes  330 . In the boiling zones  207 ,  209 , the liquid-phase secondary coolant  505  vaporizes and exits the first and second horizontal steam generators  203 ,  204  as high pressure and low pressure steam  504  that is respectively supplied to the high and low pressure turbines  220 ,  221 . 
     The shells of the horizontal steam generators  203 ,  204  and the horizontal superheaters  205 ,  206  provide additional barriers against potential large-break LOCAs, as do the turning plenum and the eccentric flanges that join the steam generators  203 ,  204  to the reactor pressure vessel  100 , as shown in  FIGS. 4 and 5 . All systems connected to the reactor vessel  100  use a similar approach to ensure that there is no potential for a large-break LOCA that could rapidly drain the water from the reactor vessel  100  and uncover the reactor core  102 . As long as the reactor core  102  is covered under all potential conditions of operation and hypothetical accident, the release of radioactive material to the public is minimal. 
     As explained in the foregoing, the reactor system  1000  is an intrinsically safe reactor which, in the event of a problem external to the reactor containment building or within containment, is designed to automatically shut down in a safe mode with natural circulation cooling. Nevertheless, to instill maximum confidence, a number of redundant safety systems can be engineered to protect public health and safety under hypothetical accident scenarios that are unknown or unknowable, i.e., cannot be mechanistically postulated. In the case of an abnormal condition when the normal heat transport path through the steam generators are not available, then the pressure in the reactor vessel  100  will begin to increase. In such a case rupture discs will breach allowing the reactor coolant to flow into a kettle reboiler located overhead. The kettle will have a large inventory of water that will serve to extract the heat from the reactor coolant until the system shuts down. Diverse systems perform duplicate or overlapping functions using different physical principles and equipment to ensure that a common-mode failure is impossible. 
     As used throughout, ranges are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, all references cited herein are hereby incorporated by referenced in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. 
     While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims.