Abstract: A system and a method of detecting a condensate line blockage in an air conditioning unit includes receiving a call for cooling indication from a thermostat associated with the air conditioning unit and detecting a pattern of a plurality of loss of power (LOP) to the air conditioning unit events, the LOP events caused by a trip of a condensate drain line switch. The method also includes analyzing the detected pattern for at least one of a frequency of LOP events in a predetermined time period and an elapsed time between successive LOP events, and initiating a condensate line blockage alert when the analyzed pattern meets predetermined condensate line blockage criteria.

Abstract: Disclosed is a method for managing stoppage of a pressurized-water nuclear reactor integrated into a submerged module for producing electrical power, in case of detection of a primary/secondary leak in a steam generator equipped with a safety valve, which generator is connected to the reactor and associated with a standby cooling unit. The method includes: detecting a primary/secondary leak in the steam generator; automatically stopping the reactor and isolating the broken steam generator; bringing the corresponding standby cooling unit online; monitoring the primary pressure and, once the primary pressure has passed below the set pressure of the safety valves of the steam generators, isolating the standby cooling unit of the broken steam generator; and continuing to passively cool the reactor with the remaining steam generators and cooling unit.

Abstract: A device and a method determines the concentration of a vapor in a volume, in particular for determining or controlling the mass flow of the vapor being conveyed through the volume by a carrier gas. The device comprises a sensor, which supplies a sensor signal that is dependent on the concentration or partial pressure of the vapor. The sensor has an oscillatory body that can be brought to oscillation, the oscillation frequency of which is influenced by a mass accumulation formed on a surface of the oscillating body by the condensed vapor. The oscillating body has a temperature control unit, by means of which the oscillating body can be brought to a temperature below the condensation temperature of the vapor. An evaluation unit determines the concentration or the partial pressure of the vapor from the temporal change of the oscillator frequency.

Abstract: A method for controlling energy storage device includes determining whether the operating frequency of the system falls within a dead band range corresponding to a preset frequency range, determining whether an SOC level indicating the charging amount of the energy storage device falls within a preset maintaining range, in a case where the operating frequency falls within the dead band range, and adjusting the SOC level so that the SOC level falls within the maintaining range in a case where the operating frequency of the system falls within the dead band range and the SOC level is outside the preset maintaining range, wherein the maintaining range represents a range between a preset lower limit and a preset upper limit.

Abstract: A steam power plant with a low-pressure turbine is suggested with a second low-pressure turbine on a separated shaft line including a separate generator. The second low-pressure turbine is connected to an additional condensing system without cooling water consumption, thus allowing to maintain the power output at a high level, even if the main condensing system has a reduced capacity due to cooling water restrictions.

Abstract: [Problem] To build a secure and high-efficiency power generation system while achieving a further reduction in the size of the entire system. [Solution] A compact nuclear power generation system is provided with a reactor (3) comprising a core (2) which uses metal fuel containing either or both of uranium-235/238 and plutonium-239, a reactor vessel (1) which houses the core (2), metal sodium which is filled into the reactor vessel (1) and heated by the core (2), and a neutron reflector (9) which maintains the effective multiplication factor of neutrons emitted from the core (2) at approximately one or more to bring the core into a critical state. A main heat exchanger (15) is installed outside the reactor (3). The metal sodium heated by the reactor (3) is supplied to the main heat exchanger (15), and supercritical carbon dioxide that is subjected to heat exchange with the heated metal sodium circulates therethrough.

Abstract: According to an embodiment, a pressurized water reactor plant has a primary system which includes: a reactor vessel for housing a reactor core which is cooled by a primary coolant, a single steam generator, a hot leg pipe for connecting the reactor vessel and the steam generator, cold leg pipes, at least two primary coolant pumps, and a pressurizer for pressurizing the primary coolant pressure boundary in which the primary coolant flows. The plant also has: a passive cooling and depressurization system which is a primary depressurization means for equalizing the primary system pressure to the secondary system pressure at the time of a tube rupture accident of the steam generator, and a reactor containment vessel containing the primary system and cooling the primary system by air cooling. Thus, a compact pressurized water rector with high economic efficiency, safety, and reliability can be provided.

Abstract: A radiation exposure reduction method includes the steps of: injecting iron into a coolant which flows from the reactor cooling system to the reactor core in a nuclear power plant; and ferritizing and fixing radionuclides or parent nuclides thereof, which are contained in the coolant, on the surface of a reactor core structure, wherein an iron citrate which is soluble organic iron, or iron oxalate or iron fumarate which has a particle diameter of 3 ?m or less, is used as the iron to be injected into the coolant.

Abstract: The invention relates to a nuclear reactor primary circuit comprising a primary pipeline (30), which defines an internal volume (32) and in which a primary nuclear reactor coolant downwardly runs, an additional pipeline (26) which is branched to the primary pipeline (30) and defines an internal volume communicating with the internal volume (32) of the primary pipeline (30) and a cuff (36) whose first end (50) is connected to the additional pipeline (26) and the second free end (52) is positioned in the internal volume (32) of the primary pipeline (30). According to said invention, the second end (52) is delimited by a free peripherial edge (53) comprising at least one upstream and downstream sections (56, 58) which are oriented towards the upstream, wherein the upstream section (56) penetrates into the internal volume (32) deeper from the primary pipeline (30) than the downstream section (58).

Abstract: A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

Abstract: An improvement to an SFR reactor of the integrated type wherein each of the outlet windows of the intermediate exchangers is surrounded in an enclosure in fluid communication with a pipe shaped into a toroid; and each of the inlets of the pump group which pumps the sodium from the hot area to the cold area through the intermediate exchangers is also in fluid communication with the toroid, such that the primary sodium originating from the hot area and exiting from the intermediate exchangers flows through the toroid and is then directed to the cold area by the said pump group.

Abstract: In implementing an induction heating stress improvement (IHSI) method in a nuclear plant, cooling characteristic with respect to the inner surface of piping is improved by appropriate structure/layout of nozzles, and further, the cooling effect when applying IHSI to a real machine is verified by installing a thermometer, as well as air in the portion subjected to IHSI is removed by heating the piping prior to the execution of IHSI.

Abstract: An auxiliary feedwater valve control apparatus of a steam generator that operates an auxiliary feedwater valve in an auxiliary feed water system provided as a protective system of a main feed water system that feeds secondary cooling water to a steam generator, includes a water-level detection means that detects a water level of secondary cooling water in the steam generator, a water-level-deviation calculation means that calculates a deviation between a preset target water level and a water level detected by the water-level detection means, a valve-operation setting means that sets an aperture of the auxiliary feedwater valve corresponding to a deviation of the water level, and a valve drive means that outputs a signal for driving the auxiliary feedwater valve corresponding to setting by the valve-operation setting means.

Abstract: A nuclear power plant and method of operation for augmenting a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle. The plant is equipped, for example, with a reactor; a steam loop comprising high and low pressure turbines; a condenser for condensing steam discharged therefrom the low pressure turbine; a feedwater heater for heating feedwater supplied from the condenser; and a feedwater loop for leading feedwater discharged from the feedwater heater to the reactor. The operation method includes decreasing a ratio of extraction steam which is led to the feedwater heater from a steam loop in the second operation cycle.

Abstract: A U-tube steam generator having a dual system for collecting loose parts and sludge. A loose parts collector having a water overflow edge is disposed between a feedwater inlet and a tube bundle of the steam generator. A sludge collector having a water outlet is disposed downstream of the overflow edge of the loose parts collector and maintains a pressure differential between a water inlet of the sludge collector and the water outlet.

Abstract: A nuclear power plant and method of operation for augmenting a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle. The plant is equipped, for example, with a reactor; a steam loop comprising high and low pressure turbines; a condenser for condensing steam discharged therefrom the low pressure turbine; a feedwater heater for heating feedwater supplied from the condenser; and a feedwater loop for leading feedwater discharged from the feedwater heater to the reactor. The operation method includes decreasing a ratio of extraction steam which is led to the feedwater heater from a steam loop in the second operation cycle.

Abstract: A method of modeling a heat exchanger is disclosed and comprises assigning input temperatures, assumed output temperatures, and a set of flow rates, inputting the parameters into a set of equations arranged to calculate a heat transfer coefficient, inputting parameters into a second set of equations arranged to calculate output temperatures, substituting actual output temperatures for the assumed output temperatures, and again calculating the heat transfer cooefficient. The new heat transfer coefficient is then used to obtain revised actual output temperatures, and the initial actual output temperatures and the revised actual output temperatures are compared to determine whether they differ by less than a desired variance. If not, a new iteration is performed until the output temperatures converge.

Abstract: A method is provided for determining and controlling feed water level in a steam generator of a nuclear power plant, whereby the water level sensor is calibrated in terms of the pressure drop across structural components in the feed water path. The water level differential pressure sensor is calibrated so that maximum water level is indicated as the level corresponding to the upper tap level plus the pressure drop across the foregoing structural components at maximum power plant power.

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: 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: 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: A steam turbine control device. A main steam system runs between a nuclear reactor and a steam turbine and includes a main steam isolation valve (MSIV), a main steam control valve, and a turbine by-pass system with a turbine by-pass valve A calculating means generates a main steam pressure control signal and a reactor dome pressure control signals. A pressure control signal changeover means changes a pressure control signal from the reactor dome pressure signal to the main steam pressure signal when the MSIV closes to restrain the abrupt decrease of steam in the main steam system effectively.

Abstract: There is provided a nuclear power plant having a steam turbine controller. The nuclear power plant includes a main steam supply system and a turbine by-pass system. The main steam supply system guides steam from heat generated by the nuclear reactor to a steam turbine. The turbine by-pass system is branched from the main steam supply system at a main steam header. A main steam control valve is equipped with the main steam supply system and adjusts steam pressure in the main steam supply system supplied to the steam turbine. A turbine by-pass valve is used to by-pass steam to the turbine by-pass system. A regulating controller generates first and second opening/closing signals for the main steam control valve and the turbine by-pass valve, and an ON-OFF controller generates a third opening/closing signal for the turbine by-pass valve. The third opening/closing signal has priority over the second opening/closing signal.

Abstract: A steam turbine power plant comprises a high-pressure turbine to be driven by steam generated by a steam generator, a first reheater for separating moisture from steam discharged from the high-pressure turbine and heating the steam, an intermediate-pressure turbine to be driven by the steam heated by the first reheater, a second reheater for heating exhaust steam discharged from the intermediate-pressure turbine, and a low-pressure turbine to be driven by the steam heated by the second reheater. The intermediate-pressure turbine discharges dry steam.

Abstract: A valve control apparatus for generating an anticipatory output response for controlling a fluid flow in a fluid flow line includes a valve provided inside the fluid flow line, at least one sensor for providing at least one input signal, said at least one input signal responsive to properties of said fluid flow in advance of said valve, and a letdown back-pressure controller receiving said at least one input signal and providing a tunable output signal for selectively opening and closing said valve to vary flow in the flow line responsive to said at least one input signal.

Abstract: A method of controlling the system pressure in a power generating system, having a turbine-generator and a BWR, that modulates the core thermal power of the reactor while maintaining the main turbine control valves in a constant steady position is described. The core thermal power may be adjusted by adjusting the control rod density within the reactor core or by adjusting the flow rate through the reactor which may be accomplished by modulating the speed of variable frequency recirculation pumps or by modulating recirculation flow control valves. The method includes transferring the power generation system from normal turbine control valve modulation pressure control to core thermal power modulation pressure control. Additionally the method includes modifying the bypass valve closure bias and the power control bias to accommodate the variances from core power modulation pressure control over normal pressure control.

Abstract: A feedwater control system for a pressurized water steam generating system, and a method for using such, includes a steam generator, at least one feedwater pump which receives an output signal from the feedwater control system, an economizer feedwater control valve which receives the output signal and automatically controls flow of feedwater to the steam generator when a power level of the feedwater control system is at a first predetermined range, preferably between about 10 and 100%, and a downcomer feedwater control valve which receives the output signal and automatically controls the flow of feedwater to the steam generator when the power level is at a second predetermined range, preferably between 0 and about 10%. A first input signal is provided, determined by the feedwater flow of the steam generator, along with a second input signal, determined by the steam flow of the steam generator, and a third input signal, determined by the level of the steam 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 fluid tight mechanical clamp for creating a compressed graphite seal between a resistance temperature detector nozzle and a NSSS piping hot or cold leg pipe is provided which includes a split graphite seal ring, a peripheral seal back-up ring, a split load sleeve to axially compress and radially expand the split seal ring, a reactor plate and load ring with bolts to drive the load sleeve and compress the seal, and a blowout retainer to prevent the nozzle from blowing out of its seat in the pipe wall in the event of a weld failure. Extensive welding procedures to repair nozzle leaks are thus eliminated.

Abstract: Apparatus and method for removing a wall portion from a wall of a tubular member. The wall portion has a distal end portion and a proximal end portion. The wall portion is connected to the wall of the tubular member by a pair of ligatures, which remain after use of a suitable cutting tool. The ligatures are aligned and disposed on either side of the wall portion to define a pivot axis. The apparatus has a probe body insertable into the tubular member, which probe body includes an expandable bladder for pushing the distal end portion of the wall portion in order to pivot the wall portion about the ligatures (i.e., about the pivot axis) as the bladder expands. As the wall portion pivots, the distal end portion of the wall portion rotates outwardly from the wall of the tubular member as the proximal end portion of the wall portion rotates inwardly into the tubular member. A movable elongate mandrel engages the proximal end portion of the wall portion.

Abstract: In a steam generator or boiler of the type having a pressure vessel having a zone in which heated water and steam can be separated, an outlet for the flow of pressurized steam and an outlet for the flow of liquid, a riser section in which fluid passes for heating therein and flow into the vessel zone, and a downcomer to receive the recirculated liquid from the vessel zone and feedwater for flow to the inlet of the riser section, the system includes feedwater control apparatus for sensing the mass flow of liquid in the downcomer, determining the liquid mass in the vessel zone, downcomer and riser section and controlling the feedwater rate in relation to such mass and the respective power conditions of the system, thereby providing better stability in the system operation. Trips and other problems caused by shrink and swell are thereby avoided and other benefits achieved.

Abstract: A method for producing electrical power from steam generated by a nuclear reactor comprising the steps of: providing a nuclear reactor engaged to a steam generator for generating steam when heated aqueous product is passed therethrough; and passing heated aqueous product through the steam generator to produce steam. The method additionally comprises passing the produced steam through a fossil fired or steam to steam superheater to superheat the produced steam; and passing the superheated produced steam through a first turbine to expand the superheated produced steam and produce steam. The produced steam from the first turbine is subsequently reheated to obtain a reheated steam. The obtained reheated steam is then passed through a second turbine coupled to a generator in order to expand the obtained reheated steam and generate electrical power with the generator. An apparatus to accomplish the method for producing electrical power from steam generated by a nuclear reactor.

Abstract: A main steam isolation valve of a reactor power plant comprises a valve body provided with inlet and outlet portions through which a steam flows, a cylindrical valve disk accommodated in the valve body to be reciprocatingly movable therein along an inner peripheral surface of the valve body for opening and closing the steam flow inlet portion and a driving mechanism secured to the valve body and operatively connected to the valve disk for reciprocatingly moving the valve disk in the valve body. A coupling member is applied to an end opening of the valve body for holding the valve disk when the valve disk is shifted to a position fully opening the inlet portion. A tubular wall member is integrally formed with the valve body so as to surround the valve disk with a gap therebetween when the valve disk is fully opened. A guide rib is further disposed to the valve body for causing asymmetric steam flow in the inlet portion of the steam flow of the valve body.

Abstract: A nuclear reactor is joined to a steam turbine by a main steamline for discharging steam thereto. A plurality of flow control valves regulate flow to the turbine, and a bypass valve selectively bypasses a portion of the steam around the turbine to its condenser. A pressure regulator and turbine controller are operatively joined to the control valves and the bypass valve for controlling steamflow to the turbine. An apparatus for detecting failure of one of the control valves is operatively connected to the bypass valve, and upon failure of one of the control valves to channel sufficient flowrate, the bypass valve is automatically opened to reduce reactor pressure rise. The failure detecting apparatus also provides a reduction demand signal for reducing reactor power for allowing the bypass valve to close.

Abstract: A system for improving the performance of nuclear power plant feedwater control systems and simplifying steam generator low water level reactor protection logic includes a plurality of water level channels for measuring steam generator water level and generating a plurality of signals representative thereof. The median steam generator water level signal is selected from among the plurality of steam generator water level signals. The median steam generator water level signal is then communicated to the feedwater control system through an output interface.

Abstract: An electrical generating system in which exhaust steam from the turbine is recycled through the turbine to reduce feedwater requirements and condenser complexity. The turbine is an axial flow dual shaft compressor/turbine unit in which a coaxial shaft couples the turbine section to the compressor section. A portion of the energy extracted from the steam is used to power the generator and the remainer is used to power the compressor. The turbine exhaust steam is directed through compressor section and the compressed steam is then mixed with turbine inlet steam by a steam jet ejector. The compressor may be powered from either one or both shafts from the turbine.

Abstract: A steam generating system including a heat energy source, a feedwater heating section including at least one heat exchange unit (2) having a feedwater inlet, a device (10) for supplying feedwater to the feedwater inlet of the heat exchange unit (2), and a steam generating section composed of at least two steam generating heat exchange units (4) each having a feedwater inlet connected for receiving heated feedwater from the feedwater heating section and each connected to the heat energy source for transferring heat from the energy source to the feedwater.

Abstract: The steam generator water level, the feed water temperature, and the reactor power are continuously measured and a level signal (76), a temperature signal (78), and a power signal (80) commensurate with these measurements are respectively generated. From the power signal and the temperature signal, a target flow signal (84) is generated corresponding to the feedwater flow rate that would produce a stable steam generator water level at the steady state condition of the power signal and the temperature signal. The level signal is compared to a level signal setpoint (54) and a resulting level error signal (86) is generated. The level error signal is adjusted by a gain factor (88) that is dependent on the temperature signal to produce an adjusted level error signal (90). The feedwater flow through the bypass valve is then controlled in response to the target flow signal (84') and the adjusted level error signal.

Abstract: A device for measuring cycles of random reciprocal motion of an object, such as a control valve which moves in discrete movements of varying length between an opened and closed position, includes a cylindrical travel translator member biased within a housing by a coil spring to maintain tension on a cable connecting the travel translator member to the object, so that the travel translator member moves in a direction corresponding to and a distance proportional to that of the object. An elastomeric wheel frictionally engages the travel translator member to rotate a shaft which operates a mechanical counter, and through a cam, a limit switch which generates pulses representative of the random cycles of the object.

Abstract: A system for variably delaying reactor trip and auxiliary feedwater system actuation resulting from a low water condition in a steam generator includes a plurality of sensors for determining both whether a low water condition exists in each steam generator and the power output level of the reactor. The length of the variable time delay is determined based upon the severity of the steam generator low water level condition and the reactor power output level. The reactor is then tripped and the auxiliary feedwater system actuated at the expiration of the time delay period if the steam generator water level is not restored.

Abstract: A water level control for a steam generator of a pressurized water type of nuclear steam supply system varies the water level demand as a function of power. The control also varies the water feed rate after a reactor trip initially as a function of reactor coolant average temperature.

Abstract: This invention relates to a control system for heat supply as hot water from nuclear power plants equipped with condensation units. This leads to the increased efficiency of the nuclear power plant on the basis of heat extraction from the turbine and its utilization for district heating with hot water as agent. The system requires a turbine nonadjustable bleeding through which a steam flow is extracted. A flow nozzle for steam flow measurement is provided on the steam transport line, a pressure transducer which sends a signal to a comparing and control unit which receives also a signal from another pressure transducer located on the turbine first stage. The comparing and control unit actuates the motorized steam and condensate valves, performing the turbine protection in case that the steam flow extracted at the nonadjustable bleeding exceeds the allowed limit.

Abstract: Excessive primary cooldown and start up of the auxiliary feedwater system in a pressurized water reactor (PWR) following a reactor trip are avoided by switching to primary coolant T.sub.avg programmed, main feedwater flow control in place of normal level programmed control. The Low T.sub.avg set point at which the main feedwater isolation valves are closed is lowered below the no load T.sub.avg temperature set point and the flow rate is decreased gradually to zero at a T.sub.avg between these two set points. In addition, the low-low level set point at which the auxiliary feedwater system is started is lowered if possible, and/or the level signal is filtered, or the start signal delayed, to increase the margin to the low-low set point.

Abstract: A full range steam generator level control system and method is provided by a weighted linear combination of a pair of inputs from a low power and a high power controller to actuate feedwater regulation valves.

Abstract: A drain recovery system for the condensate feedwater system of a nuclear power plant having condensate pumps for boosting the condensate from a condenser, and feedwater heaters for heating the condensate from the condensate pumps.

Abstract: A low capacity nuclear reactor with spherical fuel elements in a subterranean configuration. The reactor is characterized by a compact construction and extensive elimination of active operating devices, such as charging installation, gas purification devices, regulating systems and safety systems. It is particularly suitable for the generation of thermal energy for heating purposes. A core vessel contains a pile of fuel elements also contains a portion of the graphite bottom and side reflector. Channels are provided in the part of the side reflector located inside the core vessel. Trim and shutdown rods are contained in the channels. The roof reflector rests on the stationary pile of fuel elements. The entire core vessel together with the components contained therein is removed following burnoff of the fuel elements. The heat generated in the core is transferred to a cooling system mounted on the inside of a pressure vessel enclosing the nuclear reactor by a cooling gas which flows from top to bottom.

Abstract: The overflow in the secondary of a steam generator of a nuclear-reactor plant, which occurs when one or more primary tubes conducting the coolant are ruptured, is controlled. The secondary of the steam generator, which contains the water that is converted to steam, is connected through valves to the pressurizer relief tank. The level of the liquid in the shell is monitored. An alarm is sounded when the level of the liquid in the generator reaches a predetermined height alerting the operator. When the level reaches a greater height, the valves are opened and the liquid in the shell is discharged into the pressurizer relief tank. The liquid in the tank is discharged into the containment sump when a disc is ruptured. The time taken after the rupture is 30 minutes as mandated by safety regulations. If, after the liquid in the shell reaches the level at which the valves are opened, the level drops to a height at which steam may be injected into the pressurizer relief tank, the valves are automatically closed.

Abstract: A steam generator recirculating system for use with the secondary coolant of the steam generators of a pressurized water nuclear reactor is comprised of two separate recirculation loops. Each loop contains a common discharge line into which a secondary coolant discharge line and blowdown discharge line from at least one steam generator feeds. The common discharge line contains a recirculating pump and a heat exchanger, with bypasses provided about each. The common discharge lines discharge into branch feedlines back to the steam generators. The system may also be used as a blowdown system leading to a steam generator blowdown processing system and may include a water purification means and used to purify the secondary coolant water. Cross-connecting lines are provided to permit optional use of a particular loop with various steam generators.

Abstract: A method for regulating the power supplied to a steam turbine by a pressurized water nuclear reactor driving an electric current generator supplying a network by utilizing the movement of control rods in the reactor and the by-passing of a fraction of the steam produced by the reactor out from the turbine circuit. In the case of electrical uncoupling between the generator and the network or load rejection, the parameter determining the position of the control rods is selected as the greater of two values of a reference temperature. The by-passing of the steam is kept in operation and the operation of the reactor remains automatic. The opening signal for the by-pass valves is modified by a gain signal determined from a signal representing the power level. The invention is particularly applicable to the regulation of the power of pressurized water nuclear power stations.

Abstract: A system for controlling the feedwater system in a nuclear power plant is disclosed. The system utilizes a logic calculations subsystem (50) which selects the operational mode of the system to be monitored from a plurality of operational modes. A disturbance estimate calculations subsystem (52) contains a dynamic mathematic model of the feedwater system and compares measured system variables with corresponding dynamic model variables contained therein to produce signals representative of deviations therebetween. A modified feedforward calculations subsystem (54) utilizes these signals and other system parameters as inputs to a steady-state mathematical model to produce control signals to determine valve positions and pump speeds so that the control valve flows, minimum control valve pressure drop, and feedwater pump flows are at their respective demanded values.