Abstract: The invention provides a reactor containment vessel vent system capable of continuously releasing steam generated in a reactor containment vessel to the atmosphere even when a power supply is lost. In the reactor containment vessel vent system (15), the noble gas filter (23) that allows steam to pass through but does not allow radioactive noble gases to pass through among vent gas discharged from the reactor containment vessel (1) is provided at a most downstream portion of the vent line. An immediate upstream portion of the noble gas filter (23) and the reactor containment vessel (1) are connected to each other by the return pipe (24a, 24b) via the intermediate vessel (100). Further, when the radioactive noble gases having pressure equal to or higher than predetermined pressure stays in the immediate upstream portion of the noble gas filter (23), the staying radioactive noble gases flows into the intermediate vessel (100) by the relief valve (25).

Abstract: A nuclear reactor support system that, in one embodiment, includes a reactor vessel, a reactor core disposed within the reactor vessel, an upper portion of the reactor vessel located above a ground plane and a lower portion of the reactor vessel located below the ground plane. The support system further includes a first flange fixedly attached to the upper portion of the reactor vessel and contacting the ground plane, the first flange supporting the reactor vessel, a second flange fixedly attached to the upper portion of the reactor vessel above the ground plane, the second flange spaced vertically apart from the first flange, and a plurality of welded lugs extending vertically between the first and second flanges. The first flange supports the entire weight of the reactor vessel in a cantilevered manner.

Abstract: According to the method, either CO2-free air or pure nitrogen N2 is pumped into the cooling circuit selectively depending on system parameters. To this end, the method ensures that the air injection rate is high enough that, under normal conditions, the hydrogen concentration in the tank and in the riser remains below 2% H2. On air injection, the oxygen O2 (>2 ppm) in the cooling water reacts with the copper in the cooling ducts and a layer of copper oxide forms on the inner walls of said ducts. No reaction is triggered by the injection of nitrogen N2. The CO2 content in the injection air and, at the same time, also the H2 content in the exhaust air are continuously measured and monitored, and an alarm is triggered if adjustable limit values are exceeded. The equipment for performing the method comprises an electronic control unit (65) with an input field and display as a control box, and a pump and a pipe circuit for drawing air in from the riser.

Abstract: Configurations of molten fuel salt reactors are described that include an auxiliary cooling system which shared part of the primary coolant loop but allows for passive cooling of decay heat from the reactor. Furthermore, different pump configurations for circulating molten fuel through the reactor core and one or more in vessel heat exchangers are described.

Abstract: A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant, may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

Abstract: A nuclear steam supply system includes an elongated reactor vessel having an internal cavity with a central axis, a reactor core having nuclear fuel disposed within the internal cavity, and a steam generating vessel having at least one heat exchanger section, the steam generating vessel being fluidicly coupled to the reactor vessel. The reactor vessel includes a shell having an upper flange portion and a head having a head flange portion. The upper flange portion is coupled to the head flange portion, wherein the upper flange portion extends into the internal cavity, and the head flange portion extends outward from the internal cavity. Primary coolant flow between the steam generating vessel and reactor vessel occurs via a fluid coupling comprising direct welding between forged outer nozzles of each vessel and welded inner nozzles between each vessel inside the outer nozzles.

Abstract: A method for storing the energy of a nuclear power plant in which the nuclear core is cooled by gases or liquid heat transfer media. The hot heat transfer liquid is stored directly in storage tanks. When needed, it is used for heating a power plant. The heat of a compressed gas heat transfer medium such as helium is stored by passing the compressed gas through tanks filled with heat-resistant solids and recovered by passing the same type of gas in a second circuit in a reverse direction. Through the hot tanks to the power plant and back. This Abstract is not intended to define the invention disclosed in the specification, nor intended to limit the scope of the invention in any way.

Abstract: A steam power plant is suggested having, parallel to the low-pressure passage (VW1 to VW2), a heat reservoir (A) which is loaded with preheated condensate in weak-load times. This preheated condensate is taken from the heat reservoir (A) for generating peak-load and inserted downstream of the low-pressure preheater passage (VW1 to VW2) into the condensate line that contacts the feed water container (8). Thus it is possible to quickly control the power generation of the power plant in a wide range without significantly having to change the heating output of the boiler of the steam generator (1). A steam power plant equipped according to the invention can thus be operated with bigger load modifications and also provide more control energy.

Abstract: A solar-nuclear hybrid plant includes a nuclear energy loop, a solar energy loop, and an electrical generation loop. The three loops are closed loops. The nuclear energy loop transfers energy to the electrical generation loop through a steam generator. The solar energy loop transfers energy to either the nuclear energy loop and/or the electrical generation loop through one or more heat exchangers in the appropriate loop. This cogeneration plant has higher efficiency compared to a nuclear power plant alone.

Abstract: An embodiment of the present invention takes the form of a system that may reduce the level of flow-induced vibration (FIV) experienced by a jet pump assembly or other similar object within a pressure vessel. Essentially, an embodiment of the present invention may reduce the slip-joint leakage, which may be a cause of the FIVs, by adding a flow-limiting component to an outlet of the slip joint. This component may take the form of a collar, channel, and/or other component that may be connectable to a component of the jet pump assembly. After installation, an embodiment of the present invention may lower the amplitude of, and/or change the frequency of, the FIVs experienced by the jet pump assembly.

Abstract: A nozzle apparatus of a jet pump includes a nozzle base member, and a plurality of nozzles installed to the nozzle base member and forming a plurality of narrowing portions, in which a fluid passage cross-sectional area of a driving fluid passage formed in the nozzle is reduced.

Abstract: A jet pump which can restrain self-excited vibration in a connection portion between an inlet mixer pipe and a diffuser pipe without inhibiting a structural deformation due to thermal expansion and the like includes a slip joint structure connecting the inlet mixer pipe and the diffuser pipe to each other by inserting the inlet mixer pipe into an upper end opening of the diffuser pipe with a clearance left therebetween; and a self vibration damping structure configured such that when the clearance defined by an outer pipe wall of the inlet mixer pipe and an inner pipe wall of the diffuser pipe is widening or narrowing due to vibration of the inlet mixer pipe or the diffuser pipe, a flow path resistance inside a clearance flow path for pumped coolant water defined by the clearance is not smaller than a fluid inertia force all over the clearance flow path.

Abstract: The present invention relates to a nuclear reactor, in particular a pool-type nuclear reactor cooled with liquid metal (for example, a heavy metal such as lead or lead-bismuth eutectic) or with sodium or molten salts, having a core formed by a bundle of fuel elements and immersed in a primary fluid circulating between the core and at least one heat exchanger; the fuel elements extend along respective parallel longitudinal axes and have respective bottom active parts immersed in the primary fluid to constitute the core, and respective service parts that extend at the top from the active parts and emerge from the primary fluid; the fuel elements are mechanically supported via respective top end heads anchored to supporting structures and can be operated via handling machines.

Abstract: Nuclear reactor systems and methods are described having many unique features tailored to address the special conditions and needs of emerging markets. The fast neutron spectrum nuclear reactor system may include a reactor having a reactor tank. A reactor core may be located within the reactor tank. The reactor core may include a fuel column of metal or cermet fuel using liquid sodium as a heat transfer medium. A pump may circulate the liquid sodium through a heat exchanger. The system may include a balance of plant with no nuclear safety function. The reactor may be modular, and may produce approximately 100 MWe.

Abstract: A jet pump has a plurality of nozzles installed to a nozzle base, a throat and a diffuser. A first nozzle straight-tube portion, a first nozzle narrowing portion, a second nozzle straight-tube portion, a second nozzle narrowing portion, and a nozzle lower end portion formed in those nozzles are disposed in this order from the nozzle base to a ejection outlet. A narrowing angle of the second nozzle narrowing portion is larger than of the first nozzle narrowing portion. The jet pump forms, in a lower end portion of the throat, a flow passage narrowing portion having a flow passage cross-sectional area that gradually diminishes. This flow passage narrowing portion is inserted into an upper end portion of the diffuser.

Abstract: Hard stops are useable in an operating nuclear reactor to separate and bias restrainer brackets and inlet mixers. Hard stops include a lip clamp that clamps to a restrainer bracket and a wedge member that biases against the inlet mixer. The wedge member and lip clamp are engaged such that the two components can slide against one another to bias the restrainer bracket and inlet mixer. The lip clamp includes a clamp arm and an engagement member to clamp opposite sides of the restrainer bracket. Ratchet assemblies maintain selective positioning various components of the hard stops. Hard stops may be used in several different numbers, positions, and configurations in repair or modification systems. Hard stops may be installed by determining location on an outside of a restrainer bracket for the hard stop, securing the hard stop at the location, and biasing the hard stop between two components at the location.

Abstract: Example embodiments relate to a method and apparatus for reducing electrostatic deposition of charged particles on wetted surfaces that are exposed, periodically or substantially continuously, to high velocity fluid flow within a coolant flow path in a nuclear reactor. The method may include depositing a first or base dielectric layer and a second or outer dielectric layer on a conductive surface that forms a portion of a high velocity flow path to attain the apparatus. The first dielectric layer material is selected to provide improved adhesion and insulation to the conductive surface and the second dielectric layer material is selected to provide suitable adhesion to the first dielectric layer and improved corrosion and/or mechanical resistance in the anticipated operating environment.

Abstract: A reactor vessel includes a plenum and a reactor core with first and second sets of channels. A blanket salt flows through the first set of channels, and a fuel salt flows through the second set of channels. The plenum receives the blanket salt from the first set of channels. The blanket salt provides a breed-stock for a fission reaction in the fuel salt and transfers heat generated by the fission reaction without mixing with the fuel salt.

Abstract: Apparatus for the production of carbon dioxide from limestone includes a nuclear reactor (10) for generating heat and a rotary kiln (12). The rotary kiln (12) has an inlet (28) for the introduction of limestone and an outlet (30) for the release of carbon dioxide. A heat transfer arrangement is provided for transferring heat from the nuclear reactor (10) to the interior of the rotary kiln (12). The heat transfer arrangement includes feed and return primary conduits (17,18) for passing a heat transfer fluid (14) through the nuclear reactor (10) so that heat may be extracted from the nuclear reactor (10) for transfer to the interior of the rotary kiln (12). Limestone in the rotary kiln (12) is thereby heated to a temperature sufficient for the release of carbon dioxide.

Abstract: A jet pump sensing line support clamp may be used for sensing line repair, replacement, and damage prevention or reduction. The clamp may affix to jet pump sensing line supports and confine the individual jet pump sensing lines. The clamp may provide for further access or securing of the lines in the support through the clamp. Methods of installing the clamp may include attaching and tightening the clamp against the sensing lines.

Abstract: An embodiment of the present invention provides an apparatus and system for repairing and/or maintaining a position of a first component in relation to a second component. The first and second components may be located within a reactor pressure vessel of a nuclear powerplant. The apparatus and system may attach at least one bearing plate to a horizontal surface of the first component. The apparatus and system includes structure to apply a pre-load to the first component. This may assist in maintaining the position of the first component relative to the second component.

Abstract: A nuclear reactor in which a primary coolant is contained, the primary coolant moves upwardly from the core by an operation thereof. An annular steam generator is arranged in an upper side of the core into which the upwardly moving primary coolant flows and transfers heat in the primary coolant into water therein to generate a steam. A passage structure defines a coolant passage for the primary coolant to an outside of the core. The heat-transferred primary coolant in the annular steam generator flows downwardly in the coolant passage so as to flow into the core, thereby moving upwardly. A reactor vessel is arranged to surround the coolant passage so as to contain the core, the annular steam generator and the passage means therein.

Abstract: A fast breeder reactor type nuclear power plant system including a reactor vessel provided with a core and a pipe of primary loop coolant for supplying primary loop coolant to the reactor vessel. One or more bending parts are formed on at least the pipe of primary loop coolant of the pipes, and a part of the bending part on a downstream side is provided with a flow path having a non-circular sectional configuration wherein the negative side of the bending part is formed in either a planar or flat shape.

Abstract: Methods and apparatuses are provided for the removal and transportation of thermal energy from a heat source to a distant complex for use in thermochemical cycles or other processes. In one embodiment, an apparatus includes a hybrid heat pipes/thermosyphon intermediate heat exchanger (HPTIHX) system that is divided into three distinct sections, namely: an evaporation chamber, a condensation chamber, and a working fluid transport section of liquid and vapor counter-current flows.

Abstract: A fast breeder reactor type nuclear power plant system including a reactor vessel provided with a core and a pipe of primary loop coolant for supplying primary loop coolant to the reactor vessel. One or more bending parts are formed on at least the pipe of primary loop coolant of the pipes, and a part of the bending part on a downstream side is provided with a flow path having a non-circular sectional configuration wherein the negative side of the bending part is formed in either a planar or flat shape.

Abstract: A recombination apparatus is provided to an off-gas system of a boiling water nuclear plant. An off-gas system pipe connected to a condenser is connected to the recombination apparatus. A catalyst layer filled with a catalyst for recombining hydrogen and oxygen is disposed in the recombination apparatus. The recombination catalyst has a percentage of the number of Pt particles whose diameters are in a range from more than 1 nm to not more than 3 nm to the numbers of Pt particles whose diameters are in a range from more than 0 nm to not more than 20 nm, falling within a range from 20 to 100%. The condenser discharges gas containing an organosilicon compound (ex. D5), hydrogen, and oxygen, which is introduced to the recombination apparatus. Use of the above recombination catalyst can improve the performance of recombining hydrogen and oxygen more than conventional catalysts and the initial performance of the catalyst can be maintained for a longer period of time.

Abstract: Methods and apparatuses are provided for the removal and transportation of thermal energy from a heat source to a distant complex for use in thermochemical cycles or other processes. In one embodiment, an apparatus includes a hybrid heat pipes/thermosyphon intermediate heat exchanger (HPTIHX) system that is divided into three distinct sections, namely: an evaporation chamber, a condensation chamber, and a working fluid transport section of liquid and vapor counter-current flows.

Abstract: The present invention generally relates to a nuclear reactor, comprising a reactor core, neutron reflectors surrounding the core on all sides, uranium oxide as reactor fuel and a coolant that is liquid at proces conditions and solid at room temperature. The invention is characterized in that the nuclear reactor comprises a thermal isolation which is impermeable for the coolant, which is of a different material than the coolant and which has a melting point higher than the temperature in the reactor, in such a way that the thermal isolation is surrounding the reactor core and the neutron reflectors on all sides in order to provide a leak-tight containment function. Therefore the phenomenon is being used that the temperature on the outside of the installation is always far below the melting point of the coolant, in such a way that the coolant solidifies on the outside of the installation under all possible circumstances.

Abstract: A fast breeder reactor type nuclear power plant system including a reactor vessel provided with a core and a pipe of primary loop coolant for supplying primary loop coolant to the reactor vessel. One or more bending parts are formed on at least the pipe of primary loop coolant of the pipes, and a part of the bending part on a downstream side is provided with a flow path having a non-circular sectional configuration wherein the negative side of the bending part is formed in either a planar or flat shape.

Abstract: The group of inventions relates to nuclear techniques and to power industry. The nuclear power plant contains a nuclear reactor (1) cooled by liquid metal, for example lead. Heated-up liquid metal is cooling in heat exchanger <<liquid metal-gas>> (3). This heat exchanger is embodied as not-mixing heat exchanger. The pump (2) moves cooled liquid metal through the filters into the reactor (1). The compressor (4) moves gas through heat exchanger <<liquid metal-gas>> (3) into gas turbines (8, 9, 10, 22, 23, 24). After gas turbines this gas vaporizes and overheats the water in heat exchangers <<gas-water>> (5, 6, 7) and moves back to compressor (3). Produced steam moves the steam turbines (12, 13), condenses into the water by means of condenser (15) and circulation pump (16) and then is moved by the pumps (17, 20) back to heat exchangers <<gas-water>> (5, 6, 7).

Abstract: A compact pressurised water nuclear reactor comprises a primary circuit fully integrated into the reactor vessel (10). Thus, a single steam generator (12) forms the cover of the vessel (10) and the pressuriser (30) and the primary pumps (28) are housed in the vessel (10). The same is true for the control mechanisms of the control rods (40). Finally, a venturi system (44) is also provided in the vessel (10) to create water circulation if there is a failure of the primary pumps (28).

Abstract: A nuclear reactor in which a secondary or tertiary coolant system of a nuclear steam supply system is simplified, comprising: a reactor vessel (2) which integrates a reactor core (1); a first coolant (3) which is stored in the reactor vessel (2) and heated by the reactor core (1) to convect; a first heat transfer tube (4) which is arranged in the reactor vessel (2) and comes into contact with the first coolant (3); and a second coolant (5) which is supplied from the outside of the reactor vessel (2) to the first heat transfer tube (4), cools the first coolant (3) and led to the outside of the reactor vessel (2).

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving apertures formed therein. Each aperture has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: An improved method of increasing the power output of an existing nuclear power plant includes increasing the thermal power output of the plant's nuclear island and constructing of an auxiliary BOP to handle the increased thermal power. The thermal power of the nuclear island can be increased such as by increasing the thermal power of the plant's reactor, by replacing the plant's steam generator with one that is more efficient, and by increasing the flow rate and/or change in temperature of a coolant in a secondary cooling loop of the plant. The thermal power of the reactor can be increased such as by replacing existing cylindrical fuel rods with fuel rods having a relatively greater surface area to volume ratio and/or by increasing the flow rate and/or the change in temperature of a coolant of a primary cooling loop. The auxiliary BOP can be constructed while the plant is in operation, and can then be connected with the nuclear island during a maintenance operation on the reactor.

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving aperture formed therein. Each apertures has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: Passive emergency cooling in response to a loss of coolant accident (LOCA) in a PWR, having an integral reactor pressure vessel incorporating the steam generators and housed in a small high pressure containment vessel, is provided by circulating cooling water through the steam generators and heat exchangers in an external tank to cool the reactor vessel at a rate sufficient to lower the pressure in the reactor vessel below that in containment to reverse mass flow out of the reactor vessel and keep the reactor core covered without the addition of makeup water. Suppression tanks inside the small high pressure containment structure limit peak blowdown pressure in containment and provide flood-up water and gravity fed makeup water to cool the core. Diverse cooling is provided by natural circulation of air, and if needed, water, over the spherical containment structure.

Abstract: The invention provides a cooling system which includes at least one and preferably a plurality of coolant chambers arranged around a heat source, typically a nuclear reactor. A coolant inlet pipe enters the or each coolant chamber at a high level and extends downwardly through the coolant chamber to a discharge end positioned at a low level within the coolant chamber. At least one anti-siphon bleed opening is provided in that portion of the coolant pipe which is positioned at the highest level within the coolant chamber.

Abstract: The subject of the present invention is to provide a nuclear reactor plant of which is a direct cycle nuclear reactor using a carbon dioxide as a coolant such that a heat evacuation for liquefying coolant is reduced while a compressive work is reduced by using a condensation capability of a carbon dioxide for enhancing a cycle efficiency. The nuclear reactor plant is comprised of a nuclear reactor 1, a turbine 2, and wherein, the coolant of supper critical state is heated by a heat of a nuclear reactor to directly drive a turbine, a gaseous coolant discharged from said turbine is chilled and compressed after said turbine is driven for keeping in a critical state, and then said coolant is circulated again into said nuclear reactor, and wherein, a carbon dioxide is used as said coolant, and a predetermined ratio of gaseous coolant discharged from said turbine is liquefied for being compressed in a liquid state while a rest of gaseous coolant is compressed in a gaseous state.

Abstract: The subject of the present invention is to provide a nuclear reactor plant of which is a direct cycle nuclear reactor using a carbon dioxide as a coolant such that a heat evacuation for liquefying coolant is reduced while a compressive work is reduced by using a condensation capability of a carbon dioxide for enhancing a cycle efficiency.

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving apertures formed therein. Each apertures has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: A turbine generator has a water cooling system for cooling the inside of a stator coil with water, wherein the water cooling system includes a water storage tank for storing water; pipes connecting the water storage tank and the stator coil, the pipes being provided in an inlet side and in an outlet side; a decarbonated air injection unit connected to the inlet side of the pipe by a pipe or a decarbonated air injection unit connected to the water storage tank; a heater for heating the water, the stator coil having an oxide film on a surface in contact with the water, the oxide film being formed before starting operation of the generator by high temperature water injected with the decarbonated air and heated by the heater; and a control unit for effecting control so that the decarbonated air is always and continuously injected into the cooling water during operation of the generator.

Abstract: A feedwater control system and method for a pressurized water reactor steam generating system having first and second output signals. The first output signal is determined by first and second input signals, and, when combined with a third input signal automatically controls at least one feedwater pump and first designated valves which regulate water flow from the one or more feedwater pumps to a steam generator when a steam generator steam load and reactor are operating at a first predetermined power level. The first input signal is determined by a downcomer feedwater flow differential pressure. The second input signal is determined by a reactor power level. The third input signal is determined by a steam generator level. The second output signal, determined by a steam generator water level, automatically controls at least one startup feedwater control valve when the steam generator steam load and the reactor are operating at a second predetermined power level.

Abstract: A power supply system for driving reactor coolant recirculation pumps within an advanced boiling water reactor is configured in such a manner that: at least one normal-operation busbar branches off from either a main on-site power line or from a power line for auxiliary power generation that branches off from the main on-site power line, with a house transformer therebetween; at least one MG set is connected to one of these normal-operation busbars; one static adjustable-frequency power supply device is connected to each MG set; and a plurality of recirculation pumps is connected to each static adjustable-frequency power supply device.

Abstract: The disclosed power plant can attain an extremely high thermal efficiency, as compared with that of the conventional power plant. The power plant comprises a steam system and a mixed medium system. The steam system comprises a heat source (1) for generating steam; a steam turbine (3) driven by the steam generated by the heat source; a steam condenser (81) for forming condensed water by condensing exhaust of the steam turbine; and a condensed water feeding pump (9) for feeding the water condensed by the steam condenser to the heat source.

Abstract: In a nuclear reactor water cleanup pump, a purge fluid is continuously delivered during at least the operation of the pump into the cavities containing the bearings and the electrical motor at a desired flow rate. The desired flow rate being based on the heat loss from said pump and a differential temperature defined as being the difference between the predetermined temperature of the purge fluid delivered into the pump and the desired temperature for the purge fluid exiting the pump.

Abstract: The surface of the component (1, 2) to be protected is scanned, a jet of semitransferred arc plasma (17) into which a metal powder is introduced. The process can be applied to the production of a layer of coating (18) on the outer surface of a region of welding connection (3) between a nozzle (1) and a primary pipe (2) of a pressurized water nuclear reactor or to the coating of the inner surface or the connecting surface of an adapter passing through the reactor vessel head.

Abstract: A method and apparatus for improving the efficiency and performance of a nuclear electrical generation system that comprises the addition of steam handling equipment to an existing plant that results in a surprising increase in plant performance. More particularly, a gas turbine electrical generation system with heat recovery boiler is installed along with a micro-jet high pressure and a low pressure mixer superheater. Depending upon plant characteristics, the existing moisture separator reheater (MSR) can be either augmented or done away with. The instant invention enables a reduction in T.sub.hot without a derating of the reactor unit, and improves efficiency of the plant's electrical conversion cycle. Coupled with this advantage is a possible extension of the plant's fuel cycle length due to an increased electrical conversion efficiency. The reduction in T.sub.hot further allows for a surprising extension of steam generator life.

Abstract: A steam generator for a nuclear power station in which the flow of secondary fluid is deflected and distributed by a block articulated on a fixed horizontal pivoting spindle, such as to retain maintenance access to the tube place by tooling introduced through orifices in the outer casing, and to eliminate dead zones difficult to access for cleaning and servicing the tubes.

Abstract: A nuclear internal pump is disclosed in which a flywheel and a thrust disc acting as an auxiliary impeller are fixed on a motor shaft. Fixed orifices X and Y are formed respectively between outer circumference section at larger and smaller diameter sides of the thrust disc and larger and smaller diameter rings disposed to face the outer circumference section respectively, thereby to form a pressurized chamber between the two fixed orifices X and Y. The thrust disc is equipped with an axial thrust reducing device which has two sets of channels for circulating motor cooling water and for boosting the pressure in the pressurized chamber.

Abstract: A method and apparatus for improving the efficiency and performance a of nuclear electrical generation system that comprises the addition of steam handling equipment to an existing plant that results in a surprising increase in plant performance. More particularly, a gas turbine electrical generation system with heat recovery boiler is installed along with a high pressure and a low pressure mixer superheater. Depending upon plant characteristics, the existing moisture separator reheater (MSR) can be either augmented or done away with. The instant invention enables a reduction in T.sub.hot without a derating of the reactor unit, and improves efficiency of the plant's electrical conversion cycle. Coupled with this advantage is a possible extension of the plant's fuel cycle length due to an increased electrical conversion efficiency. The reduction in T.sub.hot further allows for a surprising extension of steam generator life.