Patent Number: 053352521
Section: summary

BACKGROUND OF THE INVENTION The present invention relates to heat exchange apparatus transferring heat from a reactor primary coolant, typically helium or carbon dioxide, to a secondary fluid medium, typically water and steam, and more particularly to a novel superheating arrangement in which a reheater tube bundle located within a nuclear pressure vessel works in conjunction with superheater tube bundles which are located outside of the nuclear pressure vessel. The reheater absorbs sufficient heat from the reactor gas coolant to supply required reheat steam to the reheat turbine, and in addition the reheater absorbs excess heat at higher temperature than required to meet the reheat turbine inlet steam conditions. The excess heat contained in the reheat steam flow is transferred regeneratively to the external superheater tube bundles to raise the superheat temperature of the main steam flow to the temperature required by the main steam turbine. While it is understood that various fluids can be used for the reactor primary coolant and the secondary fluid medium, the descriptions which follow shall employ the terminology reactor gas coolant to describe the reactor primary coolant and water and or steam to describe the secondary fluid medium. It is desirable to remove heat from gas cooled nuclear reactors by circulating superheated steam at maximum temperature to maximize volumetric and thermal efficiency. This is typically done with tubular heat exchangers specifically referred to as steam generators. A steam generator is comprised of a series of high pressure main steam tube bundles which supply steam to the high pressure main steam turbine, and a lower pressure reheat tube bundle which supplies steam to the lower pressure reheat turbine. Within the nuclear pressure vessel the main steam tube bundle is comprised of an economizer/evaporator tube bundle stage in which feedwater is raised in temperature and evaporated to steam, and an initial superheater tube bundle stage in which the main steam flow is superheated to a desired level for exit from the nuclear pressure vessel. The intermediate superheater and the finishing superheater tube bundles are contained in separate pressure vessels located outside of the nuclear pressure vessel. Steam exiting from the initial superheater tube bundle stage is raised in temperature in the intermediate superheater tube bundle until stress limitations on the heat transfer tube material require a higher grade tube material. Accordingly, the finishing superheater tube bundle, pressure vessel and other components are constructed of materials having design stress limits high enough to accomodate the final steam temperature required by the main steam turbine. A bi-metallic weld is provided between the intermediate superheater tube bundle and the finishing superheater tube bundle. Inlet and outlet penetrations in the walls of the various pressure vessels provide for passage of water and steam flow to and from the respective tube bundles. A steam generator can be designed to make steam at subcritical (less than 3206.2 psia) pressure or supercritical (greater than 3206.2 psia) pressure. In a subcritical system water changes to steam with heat addition at constant temperature and with water density exceeding steam density, while in a supercritical system the phase change is temperature dependant, occuring without a change in density. By employing a supercritical main steam system reheat steam pressure can be raised above reactor gas coolant pressure such that radiation bearing reactor gas coolant cannot leak into reheat steam. Because of space limitations in the nuclear pressure vessel a once-through steam generator is preferred over a drum type steam generator in gas cooled reactors. However, once-through steam generators have certain inherent problems when utilized in gas cooled reactors. In prior designs utilizing once-through steam generators in gas cooled reactors parallel tube circuits were continuous from feedwater inlet to finishing superheater outlet so that steam temperature could not be equalized among tube circuits by the use of mixing headers. Also the lack of intermediate mixing headers and confinement in the nuclear pressure vessel precluded the use of water recirculation to provide flow stability (positive upward flow in all tube circuits) during low load and start-up operation of the plant. As a result flow resistance in the form of orifices at tube circuit inlets had to be provided. Orifices imposed a large pressure drop penalty and had a predisposition to foul by build-up of deposits from impurities in the feedwater. Special feedwater demineralizer systems had to be employed to reduce fouling of the otherwise non-maintainable orifices. Another problem with the use of once-through steam generators in gas cooled reactors was protection of the bi-metallic weld which had to be located within the nuclear pressure vessel in the tubing connecting the intermediate and finishing super-heater stages. Because the bi-metallic welds were not maintainable the use of special insulation and temperature sensors was required. It has been accepted practice with once-through steam generators in gas cooled reactors to plan for plugging of failed tubes because access for replacement of these tubes was not available. The potential for tube failure was high due to vibration and wear of tubes, blockage of tubes from orifice fouling, thermal stress at bi-metallic welds, and over heating due to low flow instability, poor gas and or water/steam flow distribution, and gas hot streaks and unmixed tube side flow. The inability to provide recirculation flow during low load and start-up operation also limited main steam outlet pressure such that it was substantially lower than reactor gas coolant pressure. High safety gas cooled reactor designs eliminated reheating from the steam generator system because of potential leakage of radiation bearing reactor gas coolant into reheat steam, leading to further reduction of main steam outlet pressure. As a result the plant was deprived of several economic advantages including thermal and volumetric efficiency and the use of standard turbine equipment. In general the difficulties with once-through steam generators and the lack of reheaters have prevented gas cooled reactors from realizing the very high temperature capability of the graphite core. The advantages of the present invention, namely a steam generator heat removal system having once-through capability at normal load combined with capability for water recirculation at low load and during start-up, working in conjunction with a balanced pressure reheater will avail gas cooled reactors of highest temperature potential. SUMMARY OF THE INVENTION One of the primary objectives of the present invention is to provide a novel steam generator for gas cooled reactors which can maximize thermal and volumetric efficiencies of the plant, is relatively compact, and provides greater ease of fabrication, installation and inspection than heretofore obtainable with known gas cooled reactor steam generator heat removal systems. A more particular object of the present invention is to provide a novel steam generator for transferring heat from a reactor gas coolant to a secondary fluid medium which may be at subcritical (less than 3206.2 psia) pressure or supercritical (greater than 3206.2 psia) pressure. The reheater portion of the steam generator is located inside of the nuclear pressure vessel, and is capable of absorbing sufficient heat to meet the requirements of the reheat system as well as the heat requirements of the finishing superheater and the intermediate superheater tube bundles, which are located outside of the nuclear pressure vessel. The excess heat absorbed by the reheater which is over and above the heat required to produce steam for the reheat turbine is transferred to the finishing superheater and intermediate superheater tube bundles by flowing steam initially at the maximum temperature attained in the reheater tube bundle, first through the shell side of the finishing superheater tube bundle, then through the shell side of the intermediate superheater tube bundle, with superheated steam meeting the requirements of the main steam turbine being produced at the tube side outlet of the finishing superheater tube bundle, before shell side steam flow continues to the reheat turbine. The initial superheater tube bundle stage, located inside of the nuclear pressure vessel, and the intermediate superheater tube bundle located, outside of the nuclear pressure vessel, are sized relative to each other so as to produce desired steam temperature and moisture requirements at the tube side outlet of the initial superheater tube bundle stage. Also, superheated steam can be diverted from the tube side outlet of the initial superheater tube bundle stage to plant feedwater heaters during continuous plant operation as a means to increase the ratio of reheat steam flow to main steam flow through the intermediate superheater and the finishing superheater tube bundles, thereby reducing the maximum steam temperature requirement at the tube side outlet of the reheater tube bundle. Reheat steam pressure is selected to be higher than reactor gas coolant pressure during continuous plant operation to prevent leakage of radioactive material into reheat steam. In summary the steam generator of the present invention includes a reheater tube bundle designed for subcritical pressure just above reactor gas coolant pressure, located inside of the nuclear pressure vessel, and a main steam system designed for subcritical or supercritical pressure comprising an economizer/evaporator tube bundle stage and an initial superheater tube bundle stage located inside of the nuclear pressure vessel, and a finishing superheater and an intermediate superheater tube bundle, contained in separate pressure vessels, located outside of the nuclear pressure vessel. A feature of the steam generator in accordance with the present invention lies in the ability to design the main steam system for subcritical or supercritical pressure operation. Another feature of the steam generator in accordance with the present invention lies in the ability to employ a reheater, thereby absorbing more heat from the reactor gas coolant and operating with higher main steam pressure than with prior steam generator designs. The reheater is designed for very high temperature to regeneratively superheat main steam flow through the intermediate superheater and the finishing superheater tube bundles which are located outside of the nuclear pressure vessel. Because reheat pressure is approximately equal to reactor gas coolant pressure, creep stresses in the reheater heat transfer tubes and other reheater pressure parts are negligible. Also, the very low pressure differential across reheater heat transfer tubes and other reheater pressure parts makes modern high temperature materials, such as graphite, ceramics and special alloy steels, feasible fop fabrication of the reheaters. The inclusion of a reheater in the steam generator system allows for significantly higher main steam system pressure and for the use of standard reheat and main steam turbines, thereby improving overall plant economics. Another feature of the steam generator in accordance with the present invention lies in the provision of water recirculation for tube side flow stability (positive up flow in all parallel main steam tube circuits) during start-up and low load operation, in which water is pumped from the initial superheater tube bundle stage outlet to the economizer/evaporator tube bundle stage inlet. This feature allows for enlargement or total elimination of flow stabilizing orifices as used in prior steam generators, resulting in reduction of steam side pressure loss at full load flow. Another feature of the steam generator in accordance with the present invention lies in the provision of a water cooling heat exchanger at the recirculation pump inlet to produce desired water temperature and to condense excess steam at the recirculation pump inlet. Another feature of the steam generator in accordance with the present invention lies in the ability to divert superheated steam from the outlet of the initial superheater tube bundle stage to plant feedwater heaters to increase the ratio of reheat steam flow to main steam flow through the finishing superheater and intermediate superheater tube bundles, thereby reducing the maximum steam temperature requirement at the reheater tube bundle outlet. Another feature of the steam generator in accordance with the present invention lies in the ability to size the heat transfer tube surface area of the initial superheater tube bundle stage with respect to the heat transfer tube surface area of the intermediate superheater tube bundle to produce liquid flow at the tube side outlet of the initial superheater tube bundle stage during low load and start-up operation of the plant. Another feature of the steam generator in accordance with the present invention lies in the ability to size the heat transfer tube surface area of the initial superheater tube bundle stage with respect to the heat transfer tube surface area of the intermediate superheater and the finishing superheater tube bundles to achieve the desired reheater tube bundle heat duty. Another feature of the steam generator in accordance with the present invention lies in the addition of steam side mixing locations in the piping between the initial superheater tube bundle stage and the intermediate superheater tube bundle, and between the intermediate superheater and the finishing superheater tube bundles, which act to equalize the temperature of steam emerging from the intermediate superheater and from the finishing superheater tube bundles. Equalized steam temperatures at tube bundle outlets promotes steam side flow stability and reduces tube overheating. Another feature of the steam generator in accordance with the present invention lies in the provision of a bypass system utilized during plant start-up, in which water is passed through a pressure reducing valve to a flash tank from which low pressure superheated steam is diverted to the tube side of the reheater tube bundle, and water is diverted to other plant systems. Still another feature of the steam generator in accordance with the present invention lies in locating the finishing superheater and the intermediate superheater tube bundles outside of the nuclear pressure vessel. The bi-metallic weld is then located outside of the nuclear pressure vessel where it is readily monitored and maintained. Tube replacement instead of tube plugging or tube bundle replacement is possible. Furthermore, reliability of steam generator components including the bi-metallic weld is improved. Further objects, advantages and features of the present invention, together with the organization and manner of operation thereof, will become apparent from the foregoing detailed description of the invention when taken in conjunction with the accompanying drawing wherein like reference numerals designate like elements throughout the several views.