Abstract: A method, system, and apparatus for the selective transfer of thermoelectrically generated electric power to operation systems of a nuclear reactor system including thermoelectrically converting nuclear reactor generated heat to electrical energy and selectively transferring the electrical energy to at least one operation system of the nuclear reactor system.

Abstract: A method and system for the thermoelectric conversion of nuclear reactor generated heat including upon a nuclear reactor system shutdown event, thermoelectrically converting nuclear reactor generated heat to electrical energy and supplying the electrical energy to a mechanical pump of the nuclear reactor system.

Abstract: A method, system, and apparatus for the thermoelectric conversion of nuclear reactor generated heat including thermoelectrically converting gas cooled nuclear reactor generated heat to electrical energy and supplying the electrical energy to an operation system of the nuclear reactor system.

Abstract: A method and system for the thermoelectric conversion of nuclear reactor generated heat including upon a nuclear reactor system shutdown event, thermoelectrically converting nuclear reactor generated heat to electrical energy and supplying the electrical energy to a mechanical pump of the nuclear reactor system.

Abstract: A method, system, and apparatus for the thermoelectric conversion of nuclear reactor generated heat including thermoelectrically converting gas cooled nuclear reactor generated heat to electrical energy and supplying the electrical energy to an operation system of the nuclear reactor system.

Abstract: A system for evacuating the residual heat from a nuclear reactor cooled with liquid metal or molten salts has two types of heat exchangers immersed in the primary fluid of the reactor: heat exchangers with higher power density, which use boiling water as secondary cooling fluid and are particularly suitable for evacuating the residual heat in the first days after turning-off of the reactor; and heat exchangers operating with atmospheric air or with water and suitable for evacuating the residual heat for indefinite periods of time.

Abstract: Disclosed herein is a fully passive decay heat removal system utilizing a partially immersed heat exchanger, the system comprising: a hot pool; an intermediate heat exchanger which heat-exchanges with the sodium of the hot pool; a cold pool; a support barrel extending vertically through the boundary between the hot pool and the cold pool; a sodium-sodium decay heat exchanger received in the support barrel; a sodium-air heat exchanger provided at a position higher than the sodium-sodium decay heat exchanger; an intermediate sodium loop connecting the sodium-sodium decay heat exchanger with the sodium-air heat exchanger; and a primary pump, wherein a portion of the effective heat transfer tube of the sodium-sodium decay heat exchanger is immersed in the cold pool, particularly in a normal operating state, and the surface of the lower end of a shroud for the sodium-sodium decay heat exchanger, the lower end being immersed in the sodium of the cold pool, has perforated holes.

Abstract: A containment vessel of a boiling-water nuclear power plant and a method of operating a condenser in a nuclear power plant, include a drain pipe which connects a top region of the containment vessel to a condensing chamber disposed in the containment vessel. The drain pipe draws off noncondensible gases from the surroundings of a building condenser in the containment vessel and thus maintains reliability of performance of the building condenser. The noncondensible gases flow automatically into the condensing chamber through the drain pipe. As a result, the building condenser may have a simple and cost-effective structure.

Abstract: A containment of a nuclear power plant has a pressure chamber, a condensation chamber, and a substantially vertically running condensation tube. The upper end of the tube is connected to the pressure chamber and the lower end of the tube is immersed in a cooling liquid in the condensation chamber. The lower end of the condensation tube has an elbow and an outlet nozzle. The elbow has an elbow angle which is such that the lower end of the elbow is immersed obliquely in the cooling liquid in the condensation chamber, and the outlet nozzle has an outlet opening which is substantially shielded with respect to the base of the condensation chamber. This renders it possible to significantly reduce the pressure loads on the base and the walls of the condensation chamber in the event of an emergency.

Abstract: Disclosed herein is a decay heat removal system, including: a decay heat exchanger that absorbs decay heat generated by a nuclear reactor; a heat pipe heat exchanger that receives the decay heat from the decay heat exchanger through a sodium loop for heat removal and then discharges the decay heat to the outside; and a sodium-air heat exchanger that is connected to the heat pipe heat exchanger through the sodium loop and discharges the decay heat transferred thereto through the sodium loop to the outside. According to the decay heat removal system, a heat removal capability can be realized by the heat pipe heat exchanger at such a high temperature at which the safety of a nuclear reactor is under threat, and a cooling effect can be obtained through the sodium-air heat exchanger at a temperature lower than that temperature.

Abstract: A power module assembly includes a reactor core immersed in a coolant and a reactor vessel housing the coolant and the reactor core. An internal dry containment vessel submerged in liquid substantially surrounds the reactor vessel in a gaseous environment. During an over-pressurization event the reactor vessel is configured to release the coolant into the containment vessel and remove a decay heat of the reactor core through condensation of the coolant on an inner surface of the containment vessel.

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 DHX, and has an open top portion protruding 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 improving the safety of a nuclear plant.

Abstract: A direct pool cooling type passive safety grade decay heat removal method and system for removing core decay heat in a pool type liquid metal reactor when a normal heat removal system breaks down. In the liquid metal reactor comprising a reactor vessel, the interior of which is partitioned into a hot pool above a core and a cold pool around the core so that liquid level difference between the hot pool and the cold pool is maintained by a primary pumping head under normal steady-state conditions, is disposed at least one circular vertical tube in such a manner that the sodium in the circular vertical tube is maintained with the same liquid level as the liquid level of the sodium in the cold pool.

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 DHX, and has an open top portion protruding 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 improving the safety of a nuclear plant.

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: 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: 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: 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 nuclear reactor containment cooling system includes a containment vessel having a drywell and a wetwell, a cooling condenser submerged in a cooling pool of water located outside the containment vessel, a vent line extending from the condenser to a suppression pool disposed in the wetwell, and at least one drain line extending from the condenser to a condensate drain tank located in the drywell. An end of the drain line is vertically submerged below the surface of a pool of water in the drain tank. To enhance flow, a blower can be located in the drain line. The containment cooling system can include a drywell gas recirculation subsystem coupled to the vent line, and including a suction pipe coupled to the vent line, at least one valve located in the suction pipe, at least one blower coupled to the suction line, and a discharge pipe in flow communication with the drywell.

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: 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 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 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: Combination primary containment cooling system and residual heat removal steam condensers (PCCS-CND) operable in a containment cooling mode and in a reactor vessel cooling mode are described. In the containment cooling mode, the PCCS-CND interfaces with the primary containment vessel (PCV) through an isolation valve which can be normally closed or open. In the normally closed valve position, and upon receipt of a high drywell pressure signal, the valve opens allowing the steam in the PCV to flow to the PCCS-CND where it condenses, and the decay heat is transferred to the condenser pool. In the normally open valve position, the PCCS-CND is in the standby containment cooling mode of operation. In the reactor pressure vessel (RPV) cooling mode, the PCCS-CND interfaces with the RPV through isolation valves.

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: 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: 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: A depressurization system (1) for pressurized steam operated plant employing injection of cold water contained in a superelevated tank (5), said injection being triggered rapidly and in a manner independent of the magnitude of the damage and of the presence of incondensible gases, by means of a natural circulation of steam in a closed loop of pipes (19, 20, 21, 22) around said plant (2, 3, 4), caused by the condensation of steam in a condenser (10), placed on said pipes, during the occurence of an accident, comprises a first siphon-type hydraulic shutoff (32), on said pipe loop (19, 20, 21, 22) and arranged close to the inlet collector (14) of said condenser (10), a second siphon-type hydraulic shutoff (29), on the cold water delivery pipe (26) and placed close to the pressurized steam operated plant (2, 3, 4), said second siphon-type hydraulic shutoff (29) being below said first siphon-type hydraulic shutoff (32).

Abstract: Breaker valve assemblies for a simplified boiling water nuclear reactor are described. The breaker valve assembly, in one form, includes a valve body and a breaker valve. The valve body includes an interior chamber, and an inlet passage extends from the chamber and through an inlet opening to facilitate transporting particles from outside of the valve body to the interior chamber. The breaker valve is positioned in the chamber and is configured to substantially seal the inlet opening. Particularly, the breaker valve includes a disk which is sized to cover the inlet opening. The disk is movably coupled to the valve body and is configured to move substantially concentrically with respect to the valve opening between a first position, where the disk completely covers the inlet opening, and a second position, where the disk does not completely cover the inlet opening.

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 safety system for a nuclear reactor coolant system including a water storage tank, a heat exchanger positioned within the water storage tank and connected to the hot leg, and a core make-up tank having a tank inlet connected to the heat exchanger and a tank outlet connected to a selected location of the nuclear reactor coolant system. Upon the occurrence of an abnormal condition, water flows by natural circulation from the hot leg through the heat exchanger so that heat may be transferred to the water in the water storage tank, and thereafter flows through the core make-up tank and into a selected location of the nuclear reactor coolant system to provide cooling for the reactor core.

Abstract: A reactor cavity flooding system, which is used to immerse the hemispherical lower head of a nuclear reactor vessel by flooding the reactor cavity, is connected to both coolant injection nozzles located at the annulus gap between the lower head and the thermal insulator of a reactor and the discharge loops which are used to drain the hot water of the annulus gap into either the cavity floor or a liquid eductor. The subcooled water at a fire protection system can be directly injected into the annulus gap through twenty-five (25) nozzles at the lowest, middle, and top injection headers by a pump. The hot water heated at the lower head will be drained into either the cavity floor and/or the liquid eductor via two discharge loops that consist of both a suction header in the annulus gap at the equator level of the lower head and four (4) leakage collectors at the outside of four (4) shear keys of a reactor vessel. Drainage and recirculation of the hot water can be achieved in two ways.

Abstract: A drain system for a shutdown cooling system of a pressurized water reactor type of a nuclear power plant (10) has an increased flow rate in the drain pipe (28) and drain pump (33). The hot leg (18) connected drain pipe has a vaned vortex breaker (30) mounted therein between its connection with the hot leg (18) lower region and the drain pump (33). This minimizes vortex-created cavitation in the drain pump thereby permitting the increased flow rate.

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: The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

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: 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 light water reactor, in particular a boiling water reactor, includes a reactor pressure vessel having an interior, a core disposed in a lower half of the pressure vessel, fuel assemblies disposed in the core, and a column of water covering the core and acting as a coolant and a moderator. The column has an initial level range during normal operation. A passively operating safety device is provided. Fluid lines are connected between the safety device and the interior of the pressure vessel. The fluid lines automatically transmit an actuation criterion to the safety device, with at least a drop of a level in the pressure vessel to a value below the initial level range serving as the actuation criterion.

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: In a method of decontaminating a primary loop of a pressurized water reactor, including a steam generator, reactor coolant pump and hot and cold legs in the loop, the primary loop is isolated from a nuclear reactor vessel by closing hot leg and cold leg loop stop valves. Decontamination process water is circulated from one side of the steam generator channel head to the other side of the channel head via a bypass pipe extending between the cold leg and the hot leg without bypassing the water through steam generator tubes extending between the sides of the channel head. The level of the decontamination water in the steam generator is maintained at two to three feet in the tubes.

Abstract: An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Abstract: An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise.