Abstract: A decay heat removal system for a liquid metal reactor, in which a decay heat exchanger (DHX) is installed concentrically with an intermediate heat-exchanger (IHX) in the same cylinder which separates the DHX and IHX from the reactor pool fluid, and serves to remove the reactor core decay heat. The cylinder surrounds the IHX and the DHX, and has an opened top portion protruded out of the level of the fluid in a hot pool, a bottom portion connected to a cold pool and a guide pipe for allowing the passage of the fluid from the hot pool into the IHX. The decay heat removal system can remove decay heat immediately after occurrence of an accident, thereby improves the safety of a nuclear plant.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a reactor pressure vessel which circulates cooling water using the density difference of the coolant inside, a feed water pipe which is connected to the reactor pressure vessel and supplies cooling water, a power control section which controls the reactor power using a control rod. The feed water pipe has an ultrasonic thermometer sensor. Driving of the control rod is controlled by the power control section based on the feed water temperature detected by the feed water thermometer. The reactor power control apparatus can detect the temperature of the feed water and perform drive control of the control rod preferentially, and obtain stable reactor power.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a power control apparatus for controlling reactor power and a pressure control apparatus for controlling reactor pressure. A power adjustment demand signal S4 is input from the power control apparatus into the pressure control apparatus. The pressure control apparatus controls degree of opening of a inlet port steam control valve provided in a moisture separation heater that is connected to a low-pressure turbine, based on the power adjustment demand signal S4. A reactor power control apparatus of the natural circulation reactor can suppress reactor power variation and obtain stable power.

Abstract: This invention relates to a nuclear reactor plant which includes a nuclear heat source (12), and a cooling system (10). The cooling system (10) includes at least two cooling circuits (26), each of which includes a plurality of coolant chambers (18) each having an inlet (40) and an outlet (42), the coolant chambers (18) being arranged around the nuclear heat source (12). The cooling system (10) further includes pump means (52) for pumping coolant to and from the coolant chambers (18), the volumetric capacity of the coolant chambers (18) being sufficiently large such that, when in a passive mode, the temperature of a water coolant in the coolant chambers (18) will remain below boiling point for at least eight hours.

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 head assembly for a reactor pressure vessel includes a removable closure head, an array of control rod drive mechanisms, a seismic support platform, a missile shield assembly and a cooling system for drawing atmospheric air across electromagnetic coil stack assemblies. The cooling system includes: a lower shroud surrounding the coil stack assemblies with an end open to the atmosphere around the CRDMs; a plurality of extending internal ducts disposed within the array of CRDMs upwardly to an upper plenum disposed above the seismic support platform and a plurality of fan assemblies disposed on the upper plenum. The missile shield is disposed within the upper plenum. The cooling system effectively cools the coil stack assemblies during fuel cycles and does not hinder access to the CRDMs during an outage. The head assembly has lift legs for transporting either the entire head assembly as an integral unit or the structure above the seismic support platform a subassembly.

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: A plurality of input signals are pre-distorted for combination into a first combined signal having a constant envelop. The input signals are combined into a second combined signal, and a similarity between each of the input signals and the second combined signal is measured. Ones of the input signals are selected and attenuated based on the similarity measurement. The attenuated input signals and non-attenuated signals are output for combination to form the first combined signal.

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: At least one injector condenser (20) is used to supply secondary feedwater to the steam generator (1) during a phase in the course of which the temperature and the pressure of the coolant of the reactor coolant system (2) of the nuclear reactor are varied between hot shutdown conditions of the nuclear reactor and conditions making it possible to bring into service the residual heat cooling system (RRA). The injector condenser (20) is fed with steam withdrawn from an upper part of the steam generator (1), at a first inlet, and with feedwater from a storage tank (10), at a second inlet. High-temperature pressurized feedwater is supplied to the steam generator (1) by means of one injector condenser (20) outlet. The steam generator (1) is fed without using an additional pump for withdrawing feedwater from the storage tank (10) and for injecting feedwater into the secondary part (3) of the steam generator (1).

Abstract: A compact metal containment vessel, for a boiling water nuclear reactor, includes in one exemplary embodiment, a bottom head, a removable top head, and a substantially cylindrical sidewall extending from the bottom head to the top head. The bottom head, top head and cylindrical sidewall define a containment cavity sized to receive and enclose a reactor pressure vessel. The containment vessel has a pressure rating of at least about 50 atmospheres atm.

Abstract: The present invention provides a boiling water reactor nuclear power plant in which a reactor core support plate, upper grid plate, and a reactor core consisting of fuel assemblies supported by these plates are provided in the inner base portion of a nuclear reactor pressure vessel. Control rod guide tubes and a reactor core shroud are positioned over the upper grid plate, and a control rod drive mechanism is provided further above same, whereby the control rods can be inserted from above the reactor core, and natural circulation of cooling water inside the reactor can be achieved by means of a chimney effect of the control rod guide tubes. According to the above structure, there can be provided a compact and economical nuclear power plant.

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: In an indirect cycle nuclear reactor, a size of the reactor containment vessel is decreased by removing decay heat inside the reactor pressure vessel without using any active component to improve the economic feasibility. A main steam pipe communicating with a heat exchanger of the indirect cycle nuclear reactor is branched in a position upstream of a main steam isolation valve to connect the branched pipe to a heat exchanger in a pressure suppression pool through an isolation valve. A feed water pipe is also branched in a position upstream of an isolation valve to connect the branched pipe to the heat exchanger through the isolation valve. Decay heat is dissipated from the heat exchanger into the pressure suppression pool, and condensed water condensed by heat dissipation is returned to the heat exchanger to cool the inside of the pressure vessel.

Abstract: A head assembly of a reactor pressure vessel in an ice condenser plant is retrofitted to permit the assembly to be removed during a later refueling operation in reduced time and with reduced exposure to radiation by operating personnel. The portions of the ductwork ventilation system that originally provided cooling air to CRDMs are removed below the seismic support platform. The original CRDM cooling shroud that surrounds the lower portion of the CRDM assemblies is extended to the seismic support platform. A plenum is mounted on the seismic support platform and in air flow communication with interior portion of the extended CRDM cooling shroud. The plenum fits under the missile shield. Spool pieces are connected between the plenum and the portion of the ductwork adjacent to the seismic support platform.

Abstract: The present invention is used to reduce thermal load itself, being the cause to generate stress, which develops near liquid surface in a nuclear reactor wall and to contribute to further improvement of safety. A partition member (5) is arranged above a coolant liquid surface (9) in an annulus space (3) between a reactor vessel (1) and a guard vessel (2), a low-temperature gas is circulated through the annulus space above the partition member to cool down, the gas is circulated through the annulus space from under the coolant liquid surface to the partition member, and the high-temperature gas heated under the coolant liquid surface is used to raise the temperature above the coolant liquid surface.

Abstract: A compact metal containment vessel, for a boiling water nuclear reactor, includes in one exemplary embodiment, a bottom head, a removable top head, and a substantially cylindrical sidewall extending from the bottom head to the top head. The bottom head, top head and cylindrical sidewall define a containment cavity sized to receive and enclose a reactor pressure vessel. The containment vessel has a pressure rating of at least about 50 atmospheres atm.

Abstract: A liquid metal nuclear reactor is described. The reactor includes a concrete reactor silo, and at least one primary vessel located in the reactor silo and coupled to a reactor shield deck. Each primary vessel is substantially surrounded by a containment vessel in a spaced apart relationship. The reactor also includes a heat removal system which includes a guard vessel substantially surrounding each containment vessel in a spaced apart relationship, at least one inlet conduit in fluid communication with the ambient atmosphere outside the nuclear reactor, and at least one outlet conduit in fluid communication with the ambient atmosphere outside the nuclear reactor. A fluid flow heat transferring flowpath is formed by the inlet conduits, the space intermediate the guard vessel of each primary vessel and the containment vessel of each primary vessel and the outlet conduits.

Abstract: A blow-off device for live steam includes a filter for radioactive substances in order to increase operational safety in a nuclear power plant. The filter is preferably integrated in a housing of a muffler or sound absorber and is disposed on a slide which is movable along an axis under the effect of gravitational force and in particular also serves as a closure for an outlet opening of the muffler or sound absorber. A method of blowing off live steam is also provided.

Abstract: A method for controlling heat exchange in a nuclear reactor. The reactor contains at least one thermal valve, at least one heat exchanger having a coolant flowing therein, with the heat exchanger being immersed in a pool containing a fluid. The heat exchanger is confined by a container having an upper part with an opening therein and a lower part having means for introducing the fluid through such lower part as well as means for partially or totally opening or closing said opening in the upper part and means for partially or totally opening or closing opening in the lower part. The method comprises the steps of closing the opening in the upper part of the container to thereby vaporize said fluid, in order to cause a cessation of heat exchange between the coolant and the fluid; and opening the opening in the upper part of the container to thereby cause the fluid to be heated and to rise by convection, thereby permitting heat exchange to occur between the coolant and the fluid.

Abstract: Modified passive containment cooling systems for cooling a reactor core of a boiling water nuclear reactor are described. The reactor core is positioned in a reactor pressure vessel which is located in a drywell of the nuclear reactor. The passive containment cooling system (PCCS), in one form, includes an IC/PCC pool, a GDCS pool, a suppression pool, and a condensate drain tank. The IC/PCC and the GDCS pool each are substantially isolated from the drywell, and the suppression pool is separated from the drywell by a wall having a spill-over hole therein. An equalizing line extends between the suppression pool and the RPV and is configured to transport water from the suppression pool to the RPV. The condensate drain tank is positioned in the drywell and includes a base wall having a sidewall extending therefrom to define a fluid retaining cavity. A steam inlet line extends from within the drywall to a set of passive containment cooling condensers, which condense steam generated within the drywall to water.

Abstract: An augmented cooling system for a CEDM in a nuclear reactor includes self actuated louvers or flap valve louvers which allow hot air from the head lift rig to vent and cooler ambient air to enter the rig in the event that pressure is lost for the forced cooled air which normally cools the CEDM.

Abstract: A nuclear reactor coolant system includes a primary coolant circuit connected to a secondary coolant circuit. A desired amount of coolant is bled from the primary circuit to the secondary circuit for purification and controlling chemical composition of the coolant. Variable speed charging pumps are provided in the secondary circuit to pump coolant back into the primary circuit at variable pressures and flow rates. A pressurizer level control system is also disclosed for controlling pump speeds.

Abstract: Modified passive containment cooling systems for cooling a reactor core of a boiling water nuclear reactor are described. The passive containment cooling system (PCCS), in one form, includes a vent line coupled to the vacuum breaker. The vent line includes a first end, a second end, and a passage extending between the first and second ends for transporting noncondensibles between the first and second ends. The first end is coupled to PCCS condenser, and the second end is submerged in a suppression pool. A branch extends from an intermediate portion of the vent line and is coupled to the vacuum breaker. The branch includes a first end, a second end, and a passage extending between the first and second ends. The first end of the branch is coupled to the intermediate portion of the vent line so that the branch passage is in communication with the vent line passage.

Abstract: An integrated head assembly for a nuclear reactor pressure vessel includes a closure head with CRDM assemblies extending upwardly from the head. A CRDM seismic support disposed above the closure head provides lateral support for the CRDM assemblies. The seismic support also has air flow holes which communicate with the interior of a shroud enclosing the CRDMs and extending upwardly from the closure head to the seismic support. The shroud has an air port in direct air flow communication with the surrounding atmosphere so that the closure head, shroud and seismic support define an air flow passageway directly communicating with the surrounding atmosphere via the air port and without the extensive ductwork and, therefore, with substantially less power. A missile shield disposed above and in air flow communication with the seismic support provides a lightweight, compact physical barrier for removing energy from dislocated objects such as CRDM drive rods.

Abstract: A method for safeguarding discharge of residual heat from a nuclear power station reactor upon a lowered filling level in a primary circuit of a reactor cooling system, includes initially shutting down the reactor and running through an initial cooling and pressure reduction phase in the primary circuit. Then an aftercooling system is cut in for taking over heat discharge from the primary circuit, when heat discharge is no longer guaranteed by a steam generator plant. Then complete pressure relief and a lowering of the filling level in the cooling system to a mid-loop level of a main coolant conduit take place. A coolant reservoir is present for a required refilling of the primary circuit and aftercooling system.

Abstract: An integrated head assembly for a nuclear reactor pressure vessel includes a closure head with CRDM assemblies extending upwardly from the head. A CRDM seismic support disposed above the closure head provides lateral support for the CRDM assemblies. The seismic support also has air flow holes which communicate with the interior of a shroud enclosing the CRDMs and extending upwardly from the closure head to the seismic support. The shroud has an air port in direct air flow communication with the surrounding atmosphere so that the closure head, shroud and seismic support define an air flow passageway directly communicating with the surrounding atmosphere via the air port and without the extensive ductwork and, therefore, with substantially less power. A missile shield disposed above and in air flow communication with the seismic support provides a lightweight, compact physical barrier for removing energy from dislocated objects such as CRDM drive rods.

Abstract: Systems, methods, and apparatus for preventing steam leakage between a drywell and a wetwell in a nuclear reactor are described. In accordance with one embodiment of the present invention, the vacuum breaker of the nuclear reactor is coupled to a vacuum breaker condensing system which includes a condenser and a steam inlet pipe. The steam inlet pipe is substantially hollow and includes a first end, a second end, and a loop seal between the first and second ends. The first end of the pipe is positioned adjacent the drywell and the second end of the pipe is coupled to the vacuum breaker. The condenser is positioned proximate the steam inlet pipe and includes an inlet, an outlet, and a plurality of condenser tubes. The condenser inlet and condenser outlet are each coupled to a pool of water, e.g., the Gravity Driven Cooling System pool, and configured to draw water from the pool of water and through the condenser tubes to substantially condense steam flowing through the steam inlet pipe.

Abstract: A passive emergency water system for cooling the atmosphere inside containment in a water-cooled nuclear reactor comprises an elevated in-containment water reservoir connected to an in-containment heat exchanger. The heat exchanger promotes natural convection of containment atmosphere and heat transfer. Heat from containment atmosphere is transferred to the reservoir by a convective return flowpath. The heat exchanger is preferably an elevated tube bank Baffle walls can be used inside containment to promote circulation of containment atmosphere. In one embodiment, the reservoir tank is closed with respect to containment atmosphere and is vented through the containment wall to the external atmosphere. In an alternative embodiment, the reservoir tank can be of a sufficient volume to absorb heat from containment as sensible heat without boiling.

Abstract: A passive containment cooling system for a nuclear reactor and a method of cooling a pressurized water reactor thereby has an isolation condensing means for cooling and condensing steam and accumulating and cooling noncondensable gas released from within a containment of a nuclear reactor. A steam-driven air ejector which is driven by steam of high temperature and pressure generated from a steam generator is used to exhaust the noncondensable gas from the isolation condensing means, to flow the noncondensable gas into the containment. A steam generator makeup water storage tank makes up the water inventory of the steam generator and provides security for steam generator water inventory lost by operation of the steam-driven air ejector. This leads to the removal of early decay heat by dumping steam from the steam generator into the tank and injection of makeup water into the steam generator during the early stage of an accident.

Abstract: In the event of an incident in a pressurized water reactor with a safety feed and borating system, it is necessary that sufficient primary cooling water be available in the reactor cooling system or be resupplied, and it must also be possible to render the reactor subcritical with added boron and to maintain it in that state. The invention is a safety feed and borating system for a pressurized water reactor with series-connected safety feed and additional borating pumps. The additional borating pump is connected downstream of the safety feed pump and it has a parallel-connected bypass line with a check valve. Both pumps have a common minimum quantity line with different and reversible throttle restrictions. Thus the feed level and throughput of the entire system can be adapted optimally to existing demands. It is no longer necessary to provide a separate high-pressure borating system.

Abstract: An integrated head assembly for a nuclear reactor pressure vessel includes a closure head with CRDM assemblies extending upwardly from the head. A CRDM seismic support disposed above the closure head provides lateral support for the CRDM assemblies. The seismic support also has air flow holes which communicate with the interior of a shroud enclosing the CRDMs and extending upwardly from the closure head to the seismic support. The shroud has an air port in direct air flow communication with the surrounding atmosphere so that the closure head, shroud and seismic support define an air flow passageway directly communicating with the surrounding atmosphere via the air port and without the extensive ductwork and, therefore, with substantially less power.

Abstract: A transport/storage container holding spent radioactive fuel elements is cooled by withdrawing gases from inside the container to create a subatmospheric pressure therein, feeding coolant water from a reservoir into the container, withdrawing steam from the container, and condensing the steam outside the container into condensate water. A common pump is used both to withdraw gases and steam from the container. The steam is condensed by contacting it with water drawn from the reservoir. In addition the water drawn from the reservoir is passed countercurrent to the steam in a condenser and the condensate water is fed back to the reservoir.

Abstract: An improved system for passively removing hydrogen inside containment in a nuclear reactor in the event of a loss of coolant accident by means of catalytic hydrogen recombination. A baffle wall located inside containment is used to establish within containment an air upflow path and an air downflow path in convective exchange. The air upflow is past the area of the coolant lines and is effective to entrain hydrogen from the break and the air downflow path is in the area adjacent the containment wall. The air upflow path downstream of the coolant lines is ducted so as to confine the entrained hydrogen to the ducted air upflow path and catalytic hydrogen recombiners are located in the ducted air upflow path for recombining said entrained hydrogen with oxygen in said ducted air upflow.

Abstract: A system (10) for dissipating heat from the interior space (50) of a containment structure (3, 6) for a nuclear reactor (2), for an indefinite time period, comprises a first heat exchanger (15, 15') outside said containment structure (3, 6) which is submerged vertically in a pool (11) associated with the exterior of the top wall (4, 7) of the structure (3, 6), and a second heat exchanger (25) placed in the interior space (50), said first and second heat exchangers (15, 25) being fluid connected to each other in a closed loop circuit (28) with piping (23, 24) containing a thermal carrier fluid. Said pool (11 is provided with a covering (90) defining a first duct (12) in communication with an outside air intake, and a second duct (13) communication with a chimney (14), the interconnection of said ducts (12, 13) being inhibited by the water present in the pool (11) when filled to a predetermined level (9).

Abstract: The device includes a corium collector arranged below the vessel of the reactor, with circulation channels for cooling the collector and the corium. These channels are connected to a water feed and cooling-fluid removal circuit including a water storage tank, a tank for collecting fluid removed from the channels and a steam ejector which receives steam from the cooling fluid originating from the collection tank and sucks water from the storage tank and injects the water into the channels.

Abstract: A nuclear reactor plant has a nuclear reactor vessel in a containment vessel. Air ducts, sump discharge lines, service water lines and other non-process non-critical lines extending through the wall of the containment vessel can be isolated passively (i.e., without instrumentation and control systems or power) should a postulated event which raises the atmospheric temperature within the containment vessel occur. The passive containment isolation system includes an isolation valve disposed in the non-process line and an actuator responsive to the atmospheric temperature within the containment vessel. The actuator is operatively connected with the isolation valve for closing the non-process line in response to the atmospheric temperature. The actuator is preferably disposed in or adjacent a containment sump at the bottom of the reactor vessel or adjacent the reactor vessel cavity where the temperature of the local atmosphere will most rapidly rise.

Abstract: A passive emergency water system for cooling the atmosphere inside containment in a water-cooled nuclear reactor comprises an elevated water reservoir connected to an in-containment heat exchanger. The heat exchanger promotes natural convection of containment atmosphere and heat transfer. Heat from containment atmosphere is transferred to the reservoir by the in-containment heat exchanger via a convective return flowpath. The heat exchanger is preferably an elevated tube bank and the reservoir is preferably external to containment. Baffle walls can be used inside containment to promote circulation of containment atmosphere.

Abstract: A containment spray system for a light-water reactor includes a water trough being disposed in a safety tank. An immersion pump disposed in the vicinity of the bottom of the water trough, a spray branch and an outlet-side spray nozzle array, are connected to the water trough for injecting water into the containment in finely dispersed form in the event of an operational incident.

Abstract: A reactor container is composed of a dry well and a wet well and, in the reactor container, there is provided with a dry well cooling system, the dry well cooling system including, in one aspect, an in-dry-well heat exchanger disposed in the dry well of the reactor container, an in-dry-well blower connected to a primary side of the in-dry-well heat exchanger, a circulation pipe connected to a secondary side of the in-dry-well heat exchanger, a normal cooling system connected to the secondary side of the in-dry-well heat exchanger through the circulation pipe and including an equipment cooling pump, an equipment cooling heat exchanger and a seawater pump, and a standby cooling system connected to and branched from the circulation pipe.

Abstract: The performance of a compact heat exchanger in which a flow of air is employed to cool a flow of water can be enhanced by spraying water as a fine mist into the stream of coolant air. The water droplets, preferably less than 100 microns in diameter, coat the heat exchanger surface on the air side of the heat exchanger and provide evaporative cooling. The preferred form of heat exchange surface has strip fins.

Abstract: A passive nuclear power plant includes a reactor vessel enclosed by a containment shell. An in-containment cooling system piped with an out-of-containment heat sink, includes an in-containment heat exchanger vertically extending adjacent the sidewall of the containment shell for inducing natural circulation of the air in the containment shell. The heat exchanger has substantially parallel water-conducting pipes with cooling fins vertically extending therefrom for transferring heat from the naturally circulating air and for condensing steam from the atmosphere.

Abstract: A nuclear reactor, in particular a pressurized water reactor, has a containment, a containment shell surrounding the containment and a concrete construction of a reactor building surrounding the containment shell. A heat dissipation system for the nuclear reactor includes a sump volume disposed in a lower region of the containment shell for receiving coolant. A sump cooler is disposed inside the sump volume, has cooling tubes with a primary side and a secondary side and has feed and return lines. The primary side of the cooling tubes is covered at least when the sump volume is largely filled with coolant. An intermediate cooler has a tertiary side and is connected through the feed and return lines of the sump cooler to the secondary side of the cooling tubes. A heat sink is disposed outside the reactor building and is connected to the intermediate cooler on the tertiary side.

Abstract: A system for cooling water from a spent fuel pool of a nuclear power generating plant. The system includes a heat exchanger in which a flow of air is employed to cool a flow of pool water. The performance of the heat exchanger is enhanced by spraying water as a fine mist into the flow of coolant air. The water droplets, preferably less than 100 microns in diameter, coat the heat exchange surface on the air side of the heat exchanger and provide evaporative cooling. The preferred form of heat exchange surface has strip fins.

Abstract: The currently used atomic reactors has to be refueled periodically. During this refueling, the checking and repair of the major sections of the atomic reactor are carried out, and therefore, a low water level operation, i.e., mid-loop operation, is carried out for removing the residual heat. According to the present invention, a round-about pipe conduit is additionally installed between a suction pipe conduit and a discharge pipe conduit of the residual heating removing pump, and a flow rate adjusting valve is installed on the round-about pipe conduit. Thus the flow passing through the pump is maintained at the normal operation level during the mid-loop operation, while, as the residual heat is decreased, the round-about flow rate is gradually increased until the suction rate from the hot leg can be maintained at a proper level, thereby preventing the introduction of air into the residual heat removing pump.

Abstract: A jet condenser injects steam from a steam generator into a mixing tube, where it combines with coolant water from a heat exchanger. The steam completely condenses Upon mixing with the water because the ratio of the mass flow rate of water to that of steam is relatively high at any given combination of temperatures and pressures. The mixture of coolant water and condensate then enters a divergent tube. This action enhances dynamic natural convection in the condenser loop, which provides the forces that move the water through the loop. A small portion of the water is returned to the steam generator; the remainder is diverted to the heat exchanger. The mass flow of the steam in the jet condenser is equal to that of the condensate returned to the steam generator, thereby maintaining a constant coolant water inventory. During an accident, radioactive fission products would be retained within the closed jet condenser loop.

Abstract: A pressured-water nuclear reactor comprising an apparatus for directly removing the residual power from the core of the reactor is provided. Advantageously, upon shut-down of the nuclear reactor of the present invention, residual power of the reactor may be safely removed therefrom.

Abstract: A nuclear reactor installation includes a reactor pressure vessel and a reactor core in the reactor pressure vessel. A supporting and protective structure supporting the reactor pressure vessel and surrounding the reactor pressure vessel on the bottom and laterally, has a bottom region and a circumferential wall. A core catcher device for the reactor core has a collecting basin for a core melt being installed below the reactor pressure vessel. The collecting basin has a bottom wall and a jacket wall being respectively separated from the bottom region and the circumferential wall of the supporting and protective structure by a spacing gap. Cooling channels are disposed in the spacing gap at the bottom wall and the jacket wall for exterior cooling of the collecting basin with a cooling liquid. Turbulence bodies are disposed in a surface region of the bottom wall for generating a turbulent flow of the cooling liquid flowing from the inside to the outside over the bottom wall toward the jacket wall.

Abstract: A shutdown cooling system for a nuclear reactor operates during lapse of normal power and emergency power and has an independent power source fo removing residual heat while cooling the seals of the main reactor coolant pump. A high pressure pump delivers cooling water to the reactor cooling pump seals, a low pressure pump circulates core coolant, and a cooling mechanism discharges the decay heat. Electrical power for the pumps and associated valves and controls is provided by a dedicated power source apart from the regular residual heat removal apparatus of the reactor, and apart from the emergency generators provided for regular power failure backup.

Abstract: A reactor core with each fuel rod surrounded by an individual cylindrical channel is disclosed. This individual channel on each fuel rod provides thermal hydraulic and heat transfer advantages to enable all fuel rods within the fuel bundle to uniformly approach their own thermal limits. The preferred fuel rod pitch is a triangular pitch between the individual fuel rods as they are discretely surrounded by their own channel. The new triangular geometry provides for more uniform (flat) power distributions within all fuel rods--and hence all groups of fuel rods. Bypass flow is introduced uniformly between the fuel rod channels, rather than heterogeneously in the channel gap and water rods as in present BWR fuel designs. Individual fuel rod channels can be orificed differently, as required, to match inlet flow to fuel rod power output to maintain uniformity between all fuel rods as they approach their respective thermal limits.