Abstract: Alternative designs for a modular test reactor are presented. In one aspect, a molten fuel salt nuclear reactor includes a vessel defining a reactor volume, the vessel being open-topped and otherwise having no penetrations. A neutron reflector is provided within the vessel and displacing at least some of the reactor volume, the neutron reflector defining a reactor core volume. A plurality of heat exchangers are contained within the vessel above the neutron reflector. A flow guide assembly is provided within the neutron reflector that includes a draft tube draft tube separating a central portion of the reactor core volume from an annular downcomer duct. Fuel salt circulates from the reactor core volume, through the heat exchangers, into the downcomer duct and then back into the reactor core volume.

Abstract: Damper systems selectively reduce coolant fluid flow in nuclear reactor passive cooling systems, including related RVACS. Systems include a damper that blocks the flow in a coolant conduit and is moveable to open, closed, and intermediate positions. The damper blocks the coolant flow when closed to prevent heat loss, vibration, and development of large temperature gradients, and the damper passively opens, to allow full coolant flow, at failure and in transient scenarios. The damper may be moveable by an attachment extending into the coolant channel that holds the damper in a closed position. When a transient occurs, the resulting loss of power and/or overheat causes the attachment to stop holding the damper, which may be driven by gravity, pressure, a spring, or other passive structure into the open position for full coolant flow. A power source and temperature-dependent switch may detect and stop holding the damper closed in such scenarios.

Abstract: A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Abstract: A passive containment cooling system (PCCS) condenser, for reducing some non-condensable gases in the PCCS, includes a first and a second stage condenser that each include channels in fluid communication between an inlet and an outlet header. The inlet header of the first stage condenser is configured to receive a fluid mixture through a first inlet opening. The channels are configured to condense water from the fluid mixture flowing through the channels from the inlet header to the outlet header, respectively, of the first and second stage condenser. The PCCS condenser includes a catalyst in at least one of the outlet header of the first stage condenser or the inlet header of the second stage condenser. The catalyst catalyzes a reaction for forming water from hydrogen and oxygen in the fluid mixture. The outlet header of the second stage condenser is in fluid communication with a combined vent-and-drain line.

Abstract: A passive reactor cavity cooling system according to the present invention includes: a reactor cavity formed between a reactor vessel and a containment structure enclosing the reactor vessel; a first cooling system to control external air to sequentially pass through an air falling pipe and an air rising pipe provided in the reactor cavity, so that residual heat of a core transferred to the reactor cavity is discharged to the atmosphere; a second cooling system having a water cooling pipe disposed in an inner space of the containment structure or in a wall of the containment structure to discharge the residual heat of the core transferred to the reactor cavity to outside; and a functional conductor having an insulating property in a normal operation temperature range of the reactor and a heat transfer property in an accident occurrence temperature range of the reactor which is a higher temperature environment than the normal operation temperature range, wherein the air falling pipe and the water cooling pipe

Abstract: A reservoir tank for use in a nuclear reactor building includes a reservoir tank in which cooling water is stored. A partition part is disposed in the reservoir tank and partitions the inside of the reservoir tank into a first storage tank and a second storage tank so as to separate the cooling water in the reservoir tank. An inflow part is disposed in the partition part and allows the cooling water of the second storage tank to naturally flow into the first storage tank by the difference between the water levels of the first and second storage tanks when the water level of the first storage tank is lowered. This configuration is advantageous in that a nuclear reactor does not need to be pressurized and reheated when the nuclear reactor building is cooled due to a severe accident.

Abstract: In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer.

Abstract: A system for evacuating the residual heat from a nuclear reactor comprises: a first heat exchanger, which transfers heat from a primary fluid of the reactor to a secondary fluid; a second heat exchanger, where the secondary fluid is cooled by an auxiliary fluid which crosses a cooling duct; and a control portion, subject to thermal expansion by effect of the heating, induced by an increase in the temperature of the primary fluid beyond a preset threshold, of the secondary fluid in the control portion; the control portion being connected to a mechanical actuator device moved by the thermal expansion of the control portion to open the cooling duct and allow the passage of auxiliary fluid into the cooling duct and through the second heat exchanger.

Abstract: A cooling system for a reactor module includes a reactor pressure vessel that houses primary coolant and a steam generator that lowers a temperature of the reactor pressure vessel by transferring heat from the primary coolant to a secondary coolant. The steam generator releases at least a portion of the secondary coolant as steam. Additionally, the cooling system includes a containment vessel that at least partially surrounds the reactor vessel in a containment region. The containment region is dry during normal operation of the reactor module. A controller introduces a source of water into the containment region in response to a non-emergency shut down of the reactor module. The source of water is located external to the containment vessel, and the water is introduced into the containment region after the steam generator has initially lowered the temperature of the reactor pressure vessel in response to releasing the steam.

Abstract: A residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant includes a natural ventilation apparatus and a forced ventilation apparatus, comprising a cold air intake chamber, a hot air removal chamber, a pipeline, a ventilation heat shield cylinder, a heat removal fan, and an air cooling equipment having certain connecting relationships and being correspondingly arranged in a storeroom, an operating room and a ventilation equipment room. The system doesn't require storing spent fuel in a pool storage manner. The safety of the spent fuel doesn't rely on power equipment, thus not only reducing routine maintenance, saving energy, but also has inherent safety. Furthermore, the system can be used to cool spent fuel storage canisters within spent fuel storage facility of pebble bed high temperature gas-cooled reactor nuclear power plant, and discharge residual heat of spent fuel storage canisters to the external environment.

Abstract: A power module assembly may include a reactor vessel containing a primary coolant and one or more inlets configured to draw a secondary coolant from the containment cooling pool in response to a loss of power and/or a loss of coolant. One or more outlets may be submerged in the containment cooling pool and may be configured to vent the secondary coolant into the containment cooling pool. A heat exchanger may be configured to remove heat from the primary coolant, wherein the heat may be removed by circulating the secondary coolant from the containment cooling pool through the heat exchanger via natural circulation.

Abstract: In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer.

Abstract: A pressurized water nuclear reactor (PWR) includes a pressure vessel having a lower portion containing a nuclear reactor core comprising a fissile material and an upper portion defining an internal pressurizer volume. A condenser is secured to, and optionally supported by, the upper portion of the pressure vessel. A condenser inlet is in fluid communication with the internal pressurizer volume. A heat sink is in fluid communication with the condenser such that the condenser operates as a passive heat exchanger to condense steam from the internal pressurizer volume into condensate while rejecting heat to the heat sink. A condenser outlet connects with the pressure vessel to return condensate to the pressure vessel. A single metal forging having a first end welded to the pressure vessel and a second end welded to the condenser inlet may provide the fluid communication between the condenser inlet and the internal pressurizer volume.

Abstract: The underwater electricity production module according to the invention includes means in the form of an elongated cylindrical box (12) in which means are integrated forming an electricity production unit including means forming a nuclear boiler (30), associated with electricity production means (37) connected to an external electricity distribution station (7) by electrical cables (6), is characterized in that the nuclear boiler-forming means (30) include a secondary circuit (36) associated with the electricity production means (37) and a secondary backup circuit (60) in parallel on that secondary circuit and including at least one secondary passive heat exchanger (61) placed outside the underwater module (12) in the marine environment.

Abstract: A steam generator system for a pressurized water reactor which employs an external to containment steam drum and recirculation loop piping. The steam generator system changes the arrangement of a typical pressurized water reactor recirculation steam generator by relocating the functions of steam separation and feedwater preheating outside of the reactor coolant system. The steam generator system and thermal hydraulic conditions are selected in order to minimize the size of the steam generator heat exchanger component volume inside of the containment. The external steam drum component can be isolated in accident conditions when desired and is used as a source of secondary fluid inventory for improved decay heat removal capability and tolerance for loss of feedwater events. Thus, the steam generator component volume inside of the containment is reduced and the amount of maintenance required for the reactor coolant system components are similarly reduced.

Abstract: In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer.

Abstract: Disclosed is a decay heat removal system for cooling the decay heat of a reactor core and the spent fuel. The decay heat removal system including: a first heat pipe which is placed in an upper plenum of the reactor vessel and arranged in upward and downward directions corresponding to a position of an insertion hole formed on a top of the nuclear fuel assemblies; a control rod drive mechanism which is connected to an upper portion of the first heat pipe and drives the first heat pipe to move up and down so that the first heat pipe can be selectively inserted in a control rod insertion hole of the reactor core arranged in the nuclear reactor vessel; and a second heat pipe which is coupled to and in close contact with a bottom surface of the reactor vessel and removes the decay heat generated in the reactor core.

Abstract: A system for removing the residual power of a pressurised water nuclear reactor, includes a reserve of water, a steam generator, wherein the primary water heated by the core either circulates in a forced manner during power operation, or circulates naturally when the primary pump is stopped, and a condenser housed in the containment vessel. The condenser includes a recovery unit for recovering the condensed water and a condenser link to ensure the circulation of water in a closed circuit between the reserve and the condenser. The system further includes a device for circulating the secondary water between the steam generator and the condenser, the device being activated without an external supply of electrical energy, when an operating parameter characteristic of excessive heating of the primary water reaches a certain threshold, such that the primary water heated by the core and circulating in the steam generator vaporises the secondary water.

Abstract: A liquid metal cooled nuclear reactor includes a reactor vessel, a containment, an air flow path, and an injection unit. The vessel has a reactor core and a coolant for the reactor core. The containment surrounds an outside of the vessel. The air flow path removes heat by flowing air around the containment. The injection unit injects filler in a gap between the vessel and the containment.

Abstract: In a nuclear reactor core, each of a plurality of pressure tubes contains fuel elements spaced apart to permit coolant to flow through spaces between adjacent fuel elements. Each fuel element comprises fuel pellets in cladding, e.g., sapphire, having a melting temperature of at least 1900° C. and does not form significant hydrogen if exposed to high temperature steam. Each pressure tube has an internal insulator sleeve, e.g., fused silica, that has relatively low thermal conductivity over a range of normal operating temperatures and relatively high thermal radiation transmission at temperatures higher than said normal operating temperature range. When coolant is absent from said spaces, the insulator sleeve transmits to the pressure tube at least about 10%, but preferably more than about 40% of thermal radiation from the fuel for conduction through the pressure tube to the moderator and fuel temperature remains within safe limits after the reactor is shut down.

Abstract: A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger.

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: The present invention relates to passive cooling systems and methods for cooling a spent fuel pool in a nuclear power plant in the absence of onsite and offsite power, e.g., in a station blackout event. The systems include a gap formed along the periphery of the spent fuel pool, a heat sink, one or more thermal conductive members, a water supply system for delivering water to at least partially fill the gap and conduct heat generated from the spent fuel pool through the gap to at least one thermal conductive member for transporting heat to the heat sink, and a thermal switch mechanism for activating and deactivating the water supply system. In particular, the passive spent fuel pool cooling systems and methods of the invention are useful when the active spent fuel pool cooling system is unavailable or inoperable.

Abstract: A boiling water reactor has a reactor pressure vessel and a through piping. The reactor pressure vessel includes a main body trunk and an openable upper lid covering an upper open end of the main body trunk from above. The through piping penetrates lateral side of the main body trunk and has an opening section at a same level with or higher than the upper open end of the main body trunk in the reactor pressure vessel. The through piping may be connected to the sump arranged outside the reactor pressure vessel in the dry well. The through piping may be further connected to the suppression pool in the wet well and/or to the water level gauge in the dry well.

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 spring mounting device in a cladding tube for nuclear fuel, comprising a spring distributor, a spring loader in the cladding tube, the distributor supplying the springs to the loader, said loader comprising a longitudinal slide to receive the spring, a pusher in order to set the spring into place in the cladding tube and is able to move in the slide, means of displacement of said pusher, said device comprising mechanical means in order to associate the actuation of the means of displacement with that of the spring distributor.

Abstract: Apparatus for passively generating electric power during a nuclear power station blackout by utilizing the temperature difference between the hot inlet of a residual heat removal circuit and the surrounding containment environment. A heat engine, such as a thermoelectric generator, a Sterling Cycle Engine or Rankine Cycle Engine, is coupled in heat exchange relationship with an uninsulated portion of the inlet to a passive residual heat removal heat exchanger and/or passive residual heat removal heat exchanger channel head to generate the power required to operate essential equipment needed to maintain the nuclear power station in a safe condition during a loss of normal onsite and offsite power.

Abstract: A pressurized water nuclear reactor (PWR) has an internal pressurizer volume containing a steam bubble and is surrounded by a containment structure. A condenser is disposed inside the containment structure and is operatively connected with an external heat sink disposed outside of the containment structure. A valve assembly operatively connects the PWR with the condenser responsive to an abnormal operation signal such that the condenser condenses steam from the steam bubble while rejecting heat to the external heat sink and returns the condensed water to the PWR. A quench tank contains water with dissolved neutron poison. A valved tank pressurizing path selectively connects the steam bubble to the quench tank to pressurize the quench tank, and a valved soluble poison delivery path selectively connects the quench tank to the PWR such that the quench tank under pressure from the steam bubble discharges water with dissolved neutron poison into the PWR.

Abstract: Disclosed is a heat exchanger for a passive residual heat removal system, which improves heat transfer efficiency by expanding a heat transfer area. A heat exchange tube includes a first member connected to a steam pipe through which steam generated from a steam generator of a nuclear reactor circulates, and a second member connected to both of the first member and a feed water pipe used to supply water to the steam generator provided in the nuclear reactor, and the first member has the shape different from that of the second member, thereby expanding the heat transfer area so that the heat transfer efficiency is improved.

Abstract: A heat transfer system for a nuclear plant may include a piping system that includes first and second connectors, heat exchanger, pump, and power source. The heat transfer system may not be connected to the plant during normal power operations. The power source may be independent of a normal electrical power distribution system for the plant and may be configured to power the pump. The piping system may be configured to connect the heat exchanger and pump. The connectors may be configured to connect the heat transfer system to a fluid system of the plant. When the connectors connect the heat transfer system to the fluid system, the heat transfer system may be configured to receive fluid from the fluid system of the plant via the first connector, to pump the fluid through the heat exchanger, and to return the fluid to the fluid system via the second connector.

Abstract: A nuclear reactor includes a nuclear reactor core disposed in a pressure vessel and immersed in primary coolant water at an operating pressure higher than atmospheric pressure. A containment structure contains the nuclear reactor. A reactor coolant inventory and purification system (RCI) is connected with the pressure vessel by make-up and letdown lines. The RCI includes a high pressure heat exchanger configured to operate responsive to a safety event at the operating pressure to remove heat from the primary coolant water in the pressure vessel. An auxiliary condenser located outside containment also removes heat. The RCI also includes a pump configured to inject make up water into the pressure vessel via the make-up line against the operating pressure. An emergency core cooling system (ECC) operates to depressurize the nuclear reactor only if the RCI and auxiliary condenser are unable to manage the safety event.

Abstract: A passive containment air cooling system for a nuclear power plant that enhances air flow over a metal containment that houses the reactor system to improve heat transfer out of the containment. The heat transfer is improved by employing swirl vanes to mix the air as it rises over the walls of the containment due to natural circulation and a vortex engine proximate an exit along the cooling air path to increase the quantity of air drawn along the containment.

Abstract: An enhanced passive containment air cooling system for a nuclear power plant that increases the heat transfer surface on the exterior of the nuclear plant's containment vessel. The increased surface area is created by forming a tortuous path in or on at least a substantial part of the exterior surface of the containment vessel over which a cooling fluid can flow and follow the tortuous path. The tortuous path is formed from a series of indentations and protrusions in or on the exterior surface that form a circuitous path for the cooling fluid.

Abstract: Disclosed is an advanced process that relates to the enhanced production of energy using the integration of multiple thermal cycles (Brayton and Rankine) that employ multiple fuels, multiple working fluids, turbines and equipment. The method includes providing a nuclear reactor, reactor working fluid, heat exchangers, compressors, and multiple turbines to drive compressors that pressurize a humidified working fluid that is combusted with fuel fired in at least one gas turbine. The turbine(s) provide for electrical energy, processes or other mechanical loads.

Abstract: Most modern integrated circuit transceivers, especially wireless LAN, utilize a direct conversion radio architecture, which is highly advantageous from the perspectives of cost and flexibility, there exist several performance impairments, including gain and phase imbalances between the in-phase (I) and quadrature (Q) of a transmitter or receiver. Disclosed herein is a signal processing methodology and system for compensation of I/Q imbalance for a direct conversion packet-switched OFDM communications system. The imbalance, which accounts for transmitter I/Q imbalance, RX I/Q imbalance, phase/frequency error, and dispersive multipath fading. Both frequency dependent I/Q imbalance and frequency independent cases are considered, covering both wideband and narrowband modulation. The proposed estimation algorithms operate within the fully compliant framework of existing multi-user OFDM radio standards (WLAN, LTE, WimAX).

Abstract: A method and apparatus for providing an alternative cooling system for the suppression pool of a Boiling Water Reactor (BWR) nuclear reactor. The cooling system is operated to cool the suppression pool in the event of a plant accident when normal plant electricity is not available for the conventional residual heat removal system and pumps. The cooling system may also be used to supplement the cooling of the suppression pool via the residual heat removal system. The cooling system is operated and controlled from a remote location, which is ideal during a plant emergency.

Abstract: An emergency temporary spent fuel pool cooling system for a nuclear power generating facility that has a permanently installed primary loop within the nuclear containment and a mobile temporary secondary loop. The secondary loop is housed in transport vehicles that can be stored off site and is connectable in heat exchange relationship with the primary loop through quick disconnect couplings that are accessible on the outside of the reactor containment. The transport vehicles also include self-contained power and compressed air sources for powering and controlling the entire emergency cooling system. The system also has a make-up water injection capability for refueling the spent fuel pool and secondary loop.

Abstract: A liquid metal cooled nuclear reactor includes a reactor vessel, a containment, an air flow path, and an injection unit. The vessel has a reactor core and a coolant for the reactor core. The containment surrounds an outside of the vessel. The air flow path removes heat by flowing air around the containment. The injection unit injects filler in a gap between the vessel and the containment.

Abstract: A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger.

Abstract: An apparatus for cooling a spent fuel pool having a heat exchanger includes a cooling water pool positioned above the spent fuel pool; a floating device configured to be elevated according to a water level of a cooling water in the spent fuel pool; and an emergency cooling water supply pipe configured to form a path through which the cooling water of the cooling water pool is moved to the spent fuel pool and configured to include a floating valve that opens or closes a flow passage of the cooling water in connection with the elevation of the floating device.

Abstract: A nuclear plant auxiliary backup power system that uses decay heat following a plant shutdown to produce electrical power through a dedicated steam turbine/generator set. The decay heat produces a hot operating gaseous fluid which is used as a backup to run an appropriately sized turbine that powers an electrical generator. The turbine is configured to utilize a portion of the existing nuclear plant secondary system and exhausts the turbine exhaust to the ambient atmosphere. The system functions to both remove reactor decay heat and provide electrical power for plant systems to enable an orderly shutdown in the event traditional sources of electric power are unavailable.

Abstract: Various embodiments of a decay heat conversion to electricity system and related methods are disclosed. According to one exemplary embodiment, a decay heat conversion to electricity system may include a spent fuel rack configured to pressurize spent fuel bundles to obtain superheated vapor to drive a turbine-driven pump and fast alternator all submerged with the spent fuel rack and positioned at the bottom of the spent fuel pool for conversion of electricity distributed outside of the spent fuel pool via cables without impairing spent fuel pool operations.

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 nuclear reactor includes a reactor vessel, a containment vessel that surrounds the reactor vessel, and a condenser that receives coolant from within the reactor vessel. The containment vessel and the condenser are at least partially submerged within a common reactor pool of liquid.

Abstract: A power module assembly may include a reactor vessel containing a primary coolant and one or more inlets configured to draw a secondary coolant from the containment cooling pool in response to a loss of power and/or a loss of coolant. One or more outlets may be submerged in the containment cooling pool and may be configured to vent the secondary coolant into the containment cooling pool. A heat exchanger may be configured to remove heat from the primary coolant, wherein the heat may be removed by circulating the secondary coolant from the containment cooling pool through the heat exchanger via natural circulation.

Abstract: A boiling water reactor has a reactor pressure vessel and a through piping. The reactor pressure vessel includes a main body trunk and an openable upper lid covering an upper open end of the main body trunk from above. The through piping penetrates lateral side of the main body trunk and has an opening section at a same level with or higher than the upper open end of the main body trunk in the reactor pressure vessel. The through piping may be connected to the sump arranged outside the reactor pressure vessel in the dry well. The through piping may be further connected to the suppression pool in the wet well and/or to the water level gauge in the dry well.

Abstract: An emergency core cooling system comprises first and second safety divisions for an active emergency core cooling system. Each of the first and second safety divisions is provided with a high-pressure core cooling system and a low-pressure core cooling system, which also acts as a residual heat removal system.

Abstract: A protective screen for the screening off of a suction space and a suction duct connected to it in a cooling system, include at least one screen wall element which has a suction side and an outflow side. The screen wall element is built up from a plurality of modular rectangular cassette units, which respectively contain a plurality of suction pockets open towards the suction side, with the screen pockets being surrounded by outflow gaps which are open towards the outflow side.

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 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.