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

Abstract: A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an internal pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the internal pressurizer. A central riser may be disposed concentrically inside the pressure vessel, and the RCP impels primary coolant downward into a downcomer annulus between the central ser and the pressure vessel. A steam generator may be disposed in the downcomer annulus and spaced apart from with the impeller by an outlet plenum, A manway may access the outlet plenum so tube plugging can be performed on the steam generator via access through the manway without removing the RCP.

Abstract: Example embodiment clamps may be used to clamp two components together, with a degree of freedom of motion between the clamped components, including both rotational and translational motion. Example embodiment clamps may clamp a BWR jet pump sensing line to a diffuser as a repair or installation of a sensing line support. Example clamps may include a ball subassembly that holds a component and allows rotation and twisting of the component within the clamp. Example clamps may further include jaws holding the ball subassembly and a biasing element that permits tightening of the clamp.

Abstract: A method, system, and apparatus are disclosed for a complex shape structure for liquid lithium first walls of fusion power reactor environments. In particular, the method involves installing at least one tile on the surface area of the internal walls of the reactor chamber. The tile(s) is manufactured from a high-temperature resistant, porous open-cell material. The method further involves flowing liquid lithium into the tile(s). Further, the method involves circulating the liquid lithium throughout the interior network of the tile(s) to allow for the liquid lithium to reach the external surface of the tile(s) that faces the interior of the reactor chamber. In addition, the method involves outputting the circulated liquid lithium from the tile(s). In one or more embodiments, the reactor chamber is employed in a fusion reactor. In some embodiments, the tile is manufactured from a ceramic material or a metallic foam.

Abstract: A dynamic port that extends from the bottom wall of an oil reservoir that surrounds the lower guide bearing of a reactor coolant pump and is in fluid communication within an oil level gauge. The dynamic port is rotatable into and out of the oil flow path to adjust the dynamic oil level shown by the oil level gauge when the pump is at operating speed to be substantially equal to the static oil level when the motor is at rest.

Abstract: Disclosed herein is a high Cr Ferritic/Martensitic steel comprising 0.04 to 0.13% by weight of carbon, 0.03 to 0.07% by weight of silicon, 0.40 to 0.50% by weight of manganese, 0.40 to 0.50% by weight of nickel, 8.5 to 9.5% by weight of chromium, 0.45 to 0.55% by weight of molybdenum, 0.10 to 0.25% by weight of vanadium, 0.02 to 0.10% by weight of tantalum, 0.21 to 0.25% by weight of niobium, 1.5 to 3.0% by weight of tungsten, 0.015 to 0.025% by weight of nitrogen, 0.01 to 0.02% by weight of boron and iron balance. By regulating the contents of alloying elements such as nitrogen, born, the high Cr Ferritic/Martensitic steel with to superior tensile strength and creep resistance is provided, and can be effectively used as an in-core component material for sodium-cooled fast reactor (SFR).

Abstract: An auxiliary wedge positioning apparatus/assembly 100 for use in a nuclear reactor pressure vessel 12 having riser piping 44 and 144 and a jet pump assembly 46 and 120, the apparatus 100 having a combination slide wedge 105 and spring 103 mounted on a restraint bracket body 102 having a transverse rail 110 with end gull-wing hooked protrusions 111 with sections 150, 152, 156 and wing stability attachment 155, the slide wedge used for placement between the riser piping 44 and 144 and jet pump assembly 46 and 120 to control vibrations during operation of the reactor vessel 12.

Abstract: A fuel assembly is charged in a reactor core of a nuclear reactor using a liquid metal as a coolant, and includes a wrapper tube storing a plurality of fuel pins and including an entrance nozzle for introducing the coolant and an operation handling head, grids disposed in the wrapper tube to support the fuel pins in the radial direction of the wrapper tube, liner tubes inserted in the wrapper tube to fixedly hold the respective grids in the axial direction of the wrapper tube, and a fixing device for fixing the grids and the liner tubes in the radial direction of the wrapper tube.

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

Abstract: A boiling water reactor includes a reactor pressure vessel; a steam pipe for transporting steam generated in the reactor pressure vessel out from a steam dome positioned at an upper part of the reactor pressure vessel; a high pressure turbine connected to the steam pipe and driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam flowing from the high pressure turbine to the feedwater heater; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The boiling water reactor further includes a monitor and control system which controls an opening degree of the bleeding valve based on a magnitude of fluctuating pressure in the main steam line that is detected by the pressure sensor.

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

Abstract: Illustrative embodiments provide a nuclear fission reactor, a vented nuclear fission fuel module, methods therefor and a vented nuclear fission fuel module system.

Abstract: A pressure control unit and method are provided for facilitating single-phase heat transfer within a liquid-based cooling system. The pressure control unit includes a pressure vessel containing system coolant, and a pressurizing mechanism associated with the pressure vessel. A coolant line couples system coolant in the pressure vessel in fluid communication with the coolant loop of the cooling system, and a regulator mechanism couples to the pressurizing mechanism to maintain pressure within the pressure vessel at or above a defined pressure threshold, thus maintaining pressure within the coolant loop above the pressure threshold. The defined pressure threshold is set to facilitate system coolant within the coolant loop remaining single-phase throughout an operational temperature range of the system coolant within the coolant loop.

Abstract: Illustrative embodiments provide a nuclear fission reactor, a vented nuclear fission fuel module, methods therefor and a vented nuclear fission fuel module system.

Abstract: The invention relates to a method for the water filling and air blow-off of a main primary circuit (1) of a nuclear unit after stopping the same, said main primary circuit (1) including a tank (2) located in a pool of a reactor building (4), primary pumps (5), at least one steam generator (6) including a plurality of steam generating tubes (7) located above the level of the tank (2), and a pressuriser (8).

Abstract: A nuclear fission reactor device including a core having an array of fissile material and which is capable of being transported to and from the place of operation using conventional transportation vehicles. In a first embodiment, the fissile material is a uranium hydride enriched 15%-to-20% with U-235. In a second embodiment, the fissile material is a uranium oxide enriched to 18% to 20% with U-235.

Abstract: A steam separator comprises an outer main swirler and an inner auxiliary swirler which is smaller than the main swirler. The swirlers are provided so as to be concentric on the inner wall at the lower side of the first stage inner cylinder. In the steam separator, when the gas-liquid two-phase flow which flows in the vicinity of the axial center of the first stage inner cylinder passes the auxiliary swirler, it is separated into steam and water by the centrifugal force. The separated water (droplets) is introduced into the main swirler. When the separated water (droplets) passes the main swirler, it is separated at the inner wall side of the first stage inner cylinder by the centrifugal force. Pressure loss in a steam separator is reduced and steam separation capability is increased without increasing the moisture from the steam separator.

Abstract: A particle beam target for producing radionuclides includes a target body, a target cavity, parallel grooves, peripheral bores, and radial outflow bores. The parallel grooves are formed in a back side of the target body and include respective first and second groove ends. The peripheral bores extend through the target body from the plurality of grooves generally toward the front side that receives a particle beam. Each groove communicates with a peripheral bore at the first groove end and another peripheral bore at the second groove end. The radial outflow bores extend radially from the plurality of peripheral bores. The target body defines a plurality of liquid coolant flow paths. Each liquid coolant flow path runs from a respective groove to at least one of the first groove end and the second groove end of the respective groove, through at least one peripheral bore, and through at least one radial outflow bore.

Abstract: A direct vessel injection (DVI) nozzle for minimum emergency core coolant (ECC) bypass is disclosed. The DVI nozzle is used in a pressurized light water reactor (PLWR) having a reactor vessel with a reactor coolant system in which a coolant flows into the reactor vessel through a cold leg and passes through a reactor core prior to being discharged to the outside of the reactor vessel through a hot leg. The DVI nozzle, provided to directly inject ECC into the reactor vessel to cool the reactor core during a break in the reactor coolant system, such as a cold leg break (CLB) that may occur in the PLWR, is placed on the reactor vessel at a position horizontally offset from the central axis of the hot leg at an angle of 10° to 30° and is involved within a region defined above the central axis of the hot leg by a distance of 1.5 times the sum of diameters of the hot leg and the DVI nozzle.

Abstract: A technology for accurately measuring with the use of an ultrasonic flowmeter the flow rate through a nuclear reactor water supply piping interposed between a nuclear reactor and a nuclear reactor water supply pump. In a reactor coolant cleaning system (30), piping to be joined with a reactor water supply system (10) is branched, thereby forming an auxiliary path (A) to be similarly joined with the reactor water supply system (10). The auxiliary path (A) is provided with a restriction part (C) having an inside diameter narrower than that of other parts of the auxiliary path (A) to complete a nuclear reactor water supply piping structure. Through this structure, cavitation bubbles are generated from the restriction part (C). An ultrasonic transducer (40) is disposed between the nuclear reactor (1) and the auxiliary path (A) in the piping for composing the reactor water supply system (10), thereby conducting measurement of the flow rate of fluid within the piping.

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

Abstract: A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

Abstract: A jet pump comprising: a nozzle apparatus having a header portion including, inside, a first pipe member forming a suction fluid passage for introducing suction fluid and the header portion surrounding the first pipe member, for introducing driving fluid, and a nozzle portion connected to the header portion, surrounding the first pipe member and forming an annular ejection outlet for ejecting the driving fluid; a jet pump body for mixing the driving fluid and the suction fluid sucked by the ejection of the driving fluid, and discharging the mixed fluid; and a second pipe member having one end connected to the nozzle apparatus, for introducing the driving fluid to the header portion, wherein the first pipe member is disposed through the one end inside a driving fluid passage formed in the second pipe member, and forms an opening portion of the suction fluid passage opened to the outside of the second pipe member; and the driving fluid passage is formed so that the driving fluid flowing toward the one end hi

Abstract: A combinatorial heterogeneous-homogeneous reactor configuration in which an array or groups of homogeneous fuel assemblies are interlinked together in a heterogeneous lattice. The present invention removes the limitation of a homogeneous reactor by providing a reactor concept that utilizes the inherent advantages of homogeneous fuel elements but in a heterogeneous fuel lattice arrangement that limits the power density of any one homogeneous fuel element and yet forms a reactor arrangement that is capable of producing any product demand of interest. The present invention provides a method for producing medical isotopes by the use of a modular reactor core comprised of homogeneous fuel assemblies arranged in a regular rectangular or triangular pitch lattice.

Abstract: An injection system designed to deliver a chemical solution into a reactor through feedwater system taps during normal operating condition of a power reactor is disclosed. The process of delivery is via positive displacement pumps. Injection of chemical is in a concentrated solution form, which is internally diluted by the system prior to discharging from the skid. The injection system minimizes chemical loss due to deposition on the transit line, enables a higher concentrated solution to be used as the injectant, eliminates the time consuming laborious process of chemical dilution, raises chemical solution to the pressure required for injection, prevents solid precipitations out of solution at the injection pump head through the use of a flush solution, and deposits fresh chemical on new crack surfaces that develop during a power reactor start-up, shutdown and operation.

Abstract: A method treats a flow gas that is guided via a catalytic adsorber module to oxidize contaminants carried in the flow gas. The method reliably purifies the flow gas using equipment that is held to a comparatively low level of complexity. To this end, the flow gas is guided in a first purification step via a first catalytic adsorber module to oxidize contaminants carried along therewith, during which molecular or atomic oxygen is added to the flow gas, and the flow gas mixed with the added oxygen is guided in a second purification step via an oxidation catalyst. The flow gas flowing away from the oxidation catalyst is guided in a third purification step via a second catalytic adsorber module to reduce excessive oxygen.

Abstract: A handling method for a natural circulation boiling water reactor having a chimney having cylindrical chimney shell disposed above a core in a reactor pressure vessel, a plurality of square tubes disposed in the chimney shell, and a grid support plate with grid holes supporting and in communication with a lower end portion of the square tubes so that adjacent square tubes are disposed at an interval which exceeds the width of one square tube between them. The method includes steps shifting predetermined ones of the square tubes to the interval between the square tubes, and performing maintenance and inspection of members around a core via the grid holes which appear at positions from which the predetermined square tubes are taken out due to the shift of the predetermined square tubes.

Abstract: Marking of the cuts which have to be made, cutting of the section at two ends, removal of the section which has to be replaced, bevelling of the joint ends of the parts remaining after the section has been cut out from the pipe, adjustment of a new or replacement section for length and bevelling of its joining ends and positioning and narrow bevel welding of the ends joining the replacement section to the ends of the remaining parts of the pipe are performed outside the pipe. Within the pipe operations of machining and inspecting an internal part of the joining ends welded together are performed by remote control in a programmed way by introducing and positioning means for working within the pipe from a component of the primary circuit.

Abstract: Illustrative embodiments provide modular nuclear fission deflagration wave reactors and methods for their operation. Illustrative embodiments and aspects include, without limitation, modular nuclear fission deflagration wave reactors, modular nuclear fission deflagration wave reactor modules, methods of operating a modular nuclear fission deflagration wave reactor, and the like.

Abstract: An energy reactor containment system comprises an outer containment chamber and an inner containment chamber supported within the outer containment chamber to define a space between respective walls of the inner and outer containment chambers. Water is contained within the space for generating steam which feeds turbine generators. Fuel for an energy reaction is suspended within the centre of the inner containment chamber by a suitable mechanism. The structure permits the water or another incompressible fluid to be used both for producing usable steam and for absorbing blast impact due.

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

Abstract: Marking of the cuts which have to be made, cutting of the section at two ends, removal of the section which has to be replaced, bevelling of the joint ends of the parts remaining after the section has been cut out from the pipe, adjustment of a new or replacement section for length and bevelling of its joining ends and positioning and narrow bevel welding of the ends joining the replacement section to the ends of the remaining parts of the pipe are performed outside the pipe. Within the pipe operations of machining and inspecting an internal part of the joining ends welded together are performed by remote control in a programmed way by introducing and positioning means for working within the pipe from a component of the primary circuit.

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 main steam system around a nuclear reactor which comprises two main steam nozzles in a reactor pressure vessel, each of the main nozzles being disposed at a symmetrical position with respect to a plane parallel to steam outlet faces of steam dryers provided inside the reactor pressure vessel and passing through a center of the reactor pressure vessel; and main steam pipes each connected to the main steam nozzles.

Abstract: Method and apparatuses are provided for removing thermal energy from a nuclear reactor, which are fault tolerant. The apparatus includes at least one heat pipe configured to absorb thermal energy produced by the nuclear reactor. In addition, the apparatus includes a first compartment thermally coupled to the at least one heat pipe. The first compartment is configured to contain a first gas. Furthermore, the apparatus includes a second compartment thermally coupled to the at least one heat pipe. The second compartment is configured to contain a second gas and configured to isolate the second gas from the first gas.

Abstract: A fluid pump utilizing a canned rotor and canned stator is provided. The fluid pump has increased insulative properties over past “spool-type” pumps and has an increased ability to cool the stator, making it suitable for high temperature applications. A nuclear reactor is also provided. The reactor comprises a reactor vessel, that contains a nuclear fuel, control rods, reactor coolant and a reactor coolant pump for providing the reactor coolant to a steam generator. In a preferred embodiment, a steam generator is also provided inside the reactor vessel.

Abstract: A method and system of regulating an output voltage of a boiling water reactor nuclear reactor plant recirculation system motor generator are provided. The method includes sensing an alternator output voltage and transmitting an alternator output voltage signal to a voltage regulator circuit, sensing an alternator speed and transmitting an alternator speed signal to the voltage regulator circuit, comparing the alternator output voltage signal to the alternator speed signal with a volts per hertz divider network electrically coupled to the alternator output voltage sensing circuit and the alternator speed sensing device, adjusting a capacitive reactance of the voltage regulator with a lead compensation circuit electrically coupled in series with the volts per hertz divider network, and adjusting a current in a control winding of a saturable reactor.

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

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

Abstract: The 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: Method of controlling the temperatures of an exothermic process carried out in a suspension of solids in a reactor system formed by a wind box (2), a vertical riser (5), which is essentially not cooled, a particle separator (6), at least one set of recycling channels (9), which are not cooled, and at least one cooled set of recycling channels (12). According to the invention, the flow of solids travelling through the recycling channel (9), which is no cooled, as adjusted based on the temperature difference (T2−T1) between the upper and lower parts of a riser (5), which is not cooled, and the flow of solids travelling through the heat exchanger (12) is adjusted based on the temperature (T1) of the lower part or the temperature (T2) of the upper part of the riser tube. The control of the riser temperature is thus exclusively based on the regenerative heat transfer of the solids returned from the heat exchangers.

Abstract: Within an upper plenum of a nuclear reactor, a portion of a heated coolant flows radially outward from a central portion of a core barrel (30) towards outlet nozzles (12) in a region of an upper core plate (21) extending outside of an outer periphery of the core along an inner wall of a core barrel (30). Portions of the coolant flows beneath the outlet nozzles (12). Thus, streams of heated coolant flowing in opposite directions may collide with each other. After collision, the flow directions of the heated coolant are changed to flow upward. Due to the collision, the coolant flow behavior becomes complicated and unstable, making it difficult to measure the temperature of the heated coolant with an outlet pipe (42) connected to the outlet nozzle (12).

Abstract: A clamp assembly prevents separation of a thermal sleeve penetrating a nuclear reactor pressure vessel wall and a jet pump riser elbow welded one to the other and lying within an annular space between the pressure vessel wall and core shroud. The clamp assembly includes a pair of heads having semi-cylindrical recesses. Pins project inwardly from each of the heads in the semi-cylindrical recesses and have axes which lie parallel to one another. By using actuators mounting multiple electrodes, holes may be formed in situ in the adjoining ends of the thermal sleeve and elbow. Once the holes are formed, the heads can be applied in situ with the pins being received through the holes and the heads clamped to one another. Thug, the pins in each of the sleeve and elbow prevent separating movement of the sleeve and elbow in the event of failure of the welded joint therebetween.

Abstract: A method and an apparatus for separating a neutron absorber from a coolant of a nuclear facility are provided. The coolant is evaporated by heating. Discharged coolant vapor is compressed in a compressor while the temperature increases and is used for the evaporation of further coolant. A fraction of the compressed coolant vapor is preferably fed to a condenser. A purging-gas configuration and a sealing-fluid configuration are preferably provided.

Abstract: Downcommer coupling apparatus and methods for replacing a core spray line downcommer pipe coupled to a shroud T-box are described. In one embodiment, the coupling apparatus includes a wedge flange, a wedge, a wedge housing, a pipe seal, a cylindrical pipe, an elbow, a lower flange, and a shroud seal. The downcommer pipe is connected to the coupling apparatus by extending the pipe through the wedge flange, the wedge, and the wedge housing. The wedge has a plurality of flexible thinned segments that extend into the wedge housing to secure the downcommer pipe to the wedge housing. Wedge flange bolts extend through the wedge flange, the wedge, and the wedge housing to rigidly secure the downcommer pipe to the coupling apparatus. Dowel bolts extend through the wedge housing and the downcommer pipe to provide vertical and torsional load transfer from the downcommer pipe to the coupling apparatus. The lower flange is configured to receive the shroud T-box.

Abstract: A nuclear fuel core in a pressurized-water reactor including fuel assemblies, each of which comprise a top tie plate; a bottom tie plate, the top tie plate and the bottom tie plate having through-openings for passing coolant flowing from beneath upwards through each of the fuel assemblies; elongated elements, arranged between the top tie plate and the bottom tie plate; a mixing cross section having a mixing center, the mixing cross section extending through four orthogonally arranged fuel assemblies and having a size corresponding to at least that of two fuel assemblies; and a plurality of spacers for retaining and mutually fixing the elongated elements, the plurality of spacers having flow control members for controlling coolant flow around the mixing center. A nuclear fuel core in a boiling water reactor is also disclosed.

Abstract: A fuel assembly is provided with a coolant ascending path for making coolant rise and a water rod having a coolant descending path for conducting the coolant.A ratio of a flow area in a coolant inlet port of the smallest in coolant ascending path 13 on the downstream side than large diameter tube portion 3E to a flow area of the largest in the axial direction of coolant ascending path 13 in large diameter tube portion 3E is set to be 0.2-20%.In the normal operation, the declination degree from the liquid level in the coolant ascending path, corresponding to the coolant flow rate of the liquid level formed in the coolant ascending path can be controlled. Further, at the time of the excess the change speed of the liquid level can also be controlled.

Abstract: A stator core for an electromagnetic pump includes a plurality of circumferentially abutting tapered laminations extending radially outwardly from a centerline axis to collectively define a radially inner bore and a radially outer circumference. Each of the laminations includes radially inner and outer edges and has a thickness increasing from the inner edge toward the outer edge to provide a substantially continuous path adjacent the circumference.

Abstract: A self-aligning seal system for use in carrying out maintenance services in a nuclear fission reactor plant is disclosed. The self-aligning seal system enables an operator to block off openings in the lower portion of a water filled reactor pressure vessel for maintenance service in an underlying dry field without draining the coolant water from the reactor pressure vessel.