Abstract: A nuclear reactor cooling system with passive cooling capabilities operable during a loss-of-coolant accident (LOCA) without available electric power. The system includes a reactor vessel with nuclear fuel core located in a reactor well. An in-containment water storage tank is fluidly coupled to the reactor well and holds an inventory of cooling water. During a LOCA event, the tank floods the reactor well with water. Eventually, the water heated by decay heat from the reactor vaporizes producing steam. The steam flows to an in-containment heat exchanger and condenses. The condensate is returned to the reactor well in a closed flow loop system in which flow may circulate solely via gravity from changes in phase and density of the water. In one embodiment, the heat exchanger may be an array of heat dissipater ducts mounted on the wall of the inner containment vessel surrounded by a heat sink.

Abstract: A nuclear reactor containment system with passive cooling capabilities. In one embodiment, the system includes an inner containment vessel for housing a nuclear steam supply system and an outer containment enclosure structure. An annular water-filled reservoir may be provided between the containment vessel and containment enclosure structure which provides a heat sink for dissipating thermal energy, in the event of a thermal energy release incident inside the containment vessel, the reactor containment system provides passive water and air cooling systems operable to regulate the heat of the containment vessel and the equipment inside. In one embodiment, cooling water makeup to the system is not required to maintain containment vessel and reactor temperatures within acceptable margins.

Abstract: A pressure vessel has a restraint structure for accommodating thermal cycling thereof, including: a body with a cylindrical section; a skirt connected to the body and having a foot; and a plurality of blocks disposed around the skirt adjacent to the foot. Each block is anchored to a support frame or foundation, and has a base and a flange. Each flange overlaps the foot, thereby vertically linking the blocks and the skirt. A radial clearance is formed between an outer surface of the foot and an inner surface of each base. A vertical clearance is formed between an upper surface of the foot and a lower surface of each flange.

Abstract: In the foundation structure of a reactor containment vessel, a lower foundation plate provided on a ground, an upper foundation plate provided over the lower foundation plate, a base isolation device inserted between the upper foundation plate and the lower foundation plate, and a reactor containment vessel provided on the upper foundation plate are provided. The ends of reverse U-shaped tendons serving as the lower structures of the reactor containment vessel are fixed to the lower surface of the upper foundation plate through the upper foundation plate, which makes it possible to simplify a structure and reduce a construction cost.

Abstract: An apparatus for reinforcing a jet pump riser includes: an elbow upper clamp for covering a riser elbow coupled to a thermal sleeve from an upper side thereof; an elbow lower clamping member for clamping the riser elbow from a lower side thereof; an elbow vertical portion clamping member for covering a vertical portion of the riser elbow; and an elbow horizontal portion clamping member for covering a horizontal portion of the riser elbow. These members are disposed in different orientations with respect to the elbow upper clamp so as to fix the thermal sleeve, the riser elbow and the riser pipe.

Abstract: An assembly for securing a riser brace to a riser pipe includes brackets configured to engage a yoke of the riser brace and clamp bands that are configured to extend around the riser pipe and connect to the brackets. The assembly also includes a connection that connects the assembly as a unit and is configured to adjust the tightness of the assembly around the riser pipe and riser brace.

Abstract: Method of repairing nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: removing a damaged section of one of the one or more submerged lines; and replacing the damaged section of the one of the one or more submerged lines without welding. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: shutting down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and starting up the nuclear reactor. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: cooling down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and heating up the nuclear reactor.

Abstract: A method and apparatus in a steam generator that employs tube support plates within a shroud that is in turn disposed within a shell. The tube support plates are made of a material having a coefficient of thermal expansion lower than that of the shroud. The tube support plates are aligned during fabrication, with minimal clearances between components. Using a tube support displacement system, a controlled misalignment is then imposed on one or more tube support plates, as the steam generator heats up. Displacement to produce misalignment is produced only when the steam generator is heated. The tube support plate displacement system has only two parts, a spring bar and a push rod, that are internal to the steam generator shell and threadably engaged, thereby minimizing the potential of loose parts.

Abstract: A hold-down spring unit for a top nozzle of a nuclear fuel assembly. The hold-down spring unit is coupled to the upper end of the top nozzle of the nuclear fuel assembly. The hold-down spring unit includes a first spring which provides a hold-down force upon the nuclear fuel assembly under start-up conditions or hot full power conditions of a nuclear reactor, and a second spring which provides an additional hold-down force upon the nuclear fuel assembly under start-up conditions of the nuclear reactor. The hold-down margin under start-up conditions or hot full power conditions is reduced, thus enhancing the mechanical and structural stability of the nuclear fuel assembly.

Abstract: A hold-down spring unit for a top nozzle of a nuclear fuel assembly. The hold-down spring unit is coupled to the upper end of the top nozzle of the nuclear fuel assembly. The hold-down spring unit includes a first spring which provides a hold-down force upon the nuclear fuel assembly under start-up conditions and hot full power conditions of a nuclear reactor, and a second spring which provides an additional hold-down force upon the nuclear fuel assembly under start-up conditions of the nuclear reactor. The hold-down margin under start-up conditions and hot full power conditions is reduced, thus enhancing the mechanical and structural stability of the nuclear fuel assembly.

Abstract: An embodiment of the present invention may reduce the level of vibration experienced by a line, such as, but not limiting of, a pipe, a cable, tubing, or the like, that is connected to at least one separate structure. For example, but not limiting of, the structure includes: a reactor pressure vessel, a sparger pipe, steam generator, a pipe, a pressure vessel, a heat exchanger, a pump, a condenser, a tank, or the like. An embodiment of the present invention may provide support and a preload to the line at a new location or may replace an existing support, such as, but not limiting of, a weld; which may alter the natural frequencies to avoid resonance from occurring when the structure(s) is excited.

Abstract: The nuclear reactor comprises a lower stopper sealingly fixed to a lower end of a tubular casing of the pencil that comprises a part which is internal to the casing successively comprising a first cylindrical section having a diameter which is substantially equal to the inner diameter of the tubular casing; a second cylindrical section whose diameter is smaller than the inner diameter of the tubular casing and a third cylindrical section whose inner diameter is smaller than the inner diameter of the tubular casing and greater than the diameter of the second cylindrical section such that 1-2 tenths of a millimetre radial play is created in the gas passable between the outer surface of the third section and the inner surface of the casing.

Abstract: A quick disconnect for a control rod drive mechanism seismic support tie rod system that is remotely operable from a nuclear power plant's operating deck. A wall mounted anchor in the reactor cavity contains one half of a disconnect coupling that interfaces with the other half of the disconnect coupling on the ends of the tie rods employing a remote winching system that is actuated from the top of the reactor head assembly. A latching mechanism is then actuated from the refueling cavity operating deck to lock the tie rod in place and prevent displacement during a seismic or pipe break event. The tie rod may similarly be unlocked from the wall anchor and raised above the reactor head assembly as part of a reactor head disassembly operation to gain access to the core of the reactor vessel for refueling.

Abstract: A fuel assembly 20 is constituted by combining a plurality of fuel rods 21, holding the fuel rods 21 by a plurality of support grids 22, and arranging a lower nozzle 24 and an upper nozzle 23 at opposite ends of the fuel rods 21. A shock-absorbing device 10 for a fuel assembly is fitted to the lower nozzle 24 and the upper nozzle 23. The shock-absorbing device 10 for a fuel assembly is constituted a nozzle support 12 fitted to a depression 24U of the lower nozzle 24 and a depression 23U of the upper nozzle 23, and a buffer 11 combined with the nozzle support 12, with stiffness thereof in a longitudinal direction of the fuel rods 21 being equal to or lower than that of the nozzle support 12.

Abstract: An embodiment of the present invention takes the form of an apparatus or system that may reduce the level of vibration experienced by an inlet riser or other similar object within a reactor pressure vessel. An embodiment of the present invention may eliminate the need for welding the riser brace to the inlet riser. An embodiment of the present invention provides at least one riser brace clamp that generally clamps the riser brace to the inlet riser. After installation, the riser brace clamp may lower the amplitude of, and/or change the frequency of, the vibration experienced by the inlet riser.

Abstract: The tie rod repair apparatus includes a tie rod secured at its upper end to the shroud flange at the top of the shroud. The lower end of the tie rod passes through an opening in the shroud support plate without imposing a load on the plate. The lower end of the tie rod is anchored to the lower end of the core shroud support cylinder such that the compressive load path exerted by the tie rod to restrain the cracked shroud passes directly through the shroud support cylinder and the assembly of shroud cylinders bypassing the shroud support plate.

Abstract: An apparatus and method are provided which are designed to structurally replace welds that attach a riser brace assembly to a jet pump riser pipe. The riser brace assembly may include an upper clamp assembly, a lower clamp assembly, and a plurality of mechanical fasteners to provide clamping forces to the upper clamp assembly and the lower clamp assembly. The upper clamp assembly and the lower clamp assembly sandwich a riser brace support, wherein the riser brace support is provided with at least one through-hole to accommodate the plurality of mechanical fasteners.

Abstract: A BWR jet pump assembly has a wedge slidably mounted on a jet pump that is movable under the force of gravity to seat between a bracket and the jet pump in order to horizontally support the jet pump against vibrations. The wedge or the bracket is repaired in situ by: providing a spacer layer between mating surfaces of the wedge and the bracket; and then lowering the wedge until the spacer layer is between and in contact with the wedge and the bracket.

Abstract: Control device for a cooling circuit which enables a pressure tank of a reactor to be cooled even when a serious failure occurs, without having to resort to the use of external energy. One or more rods is disposed below a bottom cap of the reactor pressure tank. Each rod is connected to an emergency cooling system by a transmission system. The end of each rod opposite the transmission system is arranged at a distance of approximately 3 cm from the bottom cap. A hydraulic system is used advantageously as a transmission system. Each rod is placed vertically on a hydraulic cylinder arranged in a parallel position with respect to the axis of the reactor pressure tank, and each cylinder is connected to the emergency cooling system. When several hydraulic cylinders are used, they are connected to each other by hydraulic lines.

Abstract: A piping support wedge apparatus for a jet pump in a nuclear reactor is provided. In an exemplary embodiment, the wedge apparatus includes a first tapered wedge segment having a first end portion and a second end portion and a second tapered wedge segment having a first end portion and a second end portion. The first and second wedge segments are joined at the first ends portions to form a substantially U-shaped body. The wedge apparatus also includes a slot defined by an area between the first and second wedge segments and extending from the first end portions to the joined second end portions of the wedge segments.

Abstract: A fuel assembly for a pressurized-water reactor includes fuel rods held in cells of spacers, control rod guide tubes, and a headpiece and a foot piece, by which it is fixed to upper and lower core grids, respectively. A spacer having a first part, which lies on a radially outer side with respect to a longitudinal center axis of the fuel assembly, and a second part, which lies on a radially inner side and is completely surrounded by the first part are included to reduce the size of gaps between fuel assemblies. The second part is formed of Zircaloy. The first part is made of a metallic material, and when compared to the Zircaloy of the second part, has a lower growth in the radial direction, caused by neutron radiation and a higher coefficient of thermal expansion.

Abstract: A high energy absorption top nozzle for a nuclear fuel assembly that employs an elongated upper tubular housing and an elongated lower tubular housing slidable within the upper tubular housing. The upper and lower housings are biased away from each other by a plurality of longitudinally extending springs that are restrained by a longitudinally moveable piston whose upward travel is limited within the upper housing. The energy imparted to the nozzle by a control rod scram is mostly absorbed by the springs and the hydraulic affect of the piston within the nozzle.

Abstract: A vibration mitigation clamp apparatus is provided, which is designed to stiffen a riser brace assembly in a nuclear reactor so as to increase the natural vibration frequency of the riser brace assembly. In an embodiment, the vibration mitigation clamp apparatus may include a first plate, a second plate and a wedge assembly. The vibration mitigation clamp apparatus is attached to upper and lower riser brace leaves of the riser brace assembly, at a location near a weld that attaches the leaves to a riser brace block of the riser brace assembly that is affixed to an RPV sidewall. The wedge assembly is expandable to apply forces on inside surfaces of the riser brace leaves, countering clamping forces applied to the first and second plate to fixedly secure the vibration mitigation clamp apparatus on the riser brace assembly.

Abstract: A nuclear steam supply containment system employing prefabricated component supports that are surrounded by steel concrete forms and anchored to the floor liner of a nuclear containment. The prefabricated component supports extend through the concrete forms and a reinforce concrete slab cap to a support elevation for the component. A second inner steel concrete form can be employed to form an annular concrete channel through which the supports extend. The center of the inner form can be filled with a sand-like substance, radiation absorber material or concrete grout.

Abstract: A core spray sparger assembly for supplying coolant to a nuclear reactor. The nuclear reactor includes fuel assemblies, a top guide, coolant supply pipes and a shroud head. The core spray sparger assembly includes at least one coolant manifold, at least one coolant coupling in fluid communication with the coolant manifold, and at least one mounting device configured to couple the coolant manifold to the nuclear reactor. The core spray sparger assembly further includes a plurality of fluid conductors in a parallel array positioned above the fuel assemblies, the fluid conductors in fluid communication with the coolant manifold, and a plurality of nozzles in fluid communication with the fluid conductors.

Abstract: A jacking device includes first, second, and third members (51, 52, 53), and a jacking screw (55). The first and second members (51, 52) have sloping side portions (57). The third member (53) has a plurality of sloped side portions (57) and a threaded opening (67) extending from its top portion (69) through to its bottom portion (59). The third member (53) is movably supported between the first and second members (51, 52). The jacking screw (55) engages the threaded opening (67) in the third member (53).

Abstract: Apparatus for the determination of radiation induced growth due to burnup of a nuclear reactor fuel assembly in a reactor core.

Abstract: An apparatus for reinforcing weakened portions of the top guide assembly in a boiling water reactor. The repair apparatus includes a cruciform lattice segment which reinforces the damaged or weakened region of the top guide beam lattice. This cruciform lattice segment is held in place atop the existing top guide with specially designed straps. The cruciform lattice segment and associated straps are arranged so that a beam segment of the cruciform lattice segment bridges the weakened region in the cracked top guide. Thus, the bridging beam segment transmits loads across the weakened region of the top guide. The straps are designed to avoid interference with removal and installation of the fuel assemblies and the control rod blade which is inserted between the fuel assemblies.

Abstract: A top guide to shroud head interface is described. In one embodiment, the interface includes a top guide that includes a flange that is configured to engage a corresponding flange of the shroud head. The top guide flange includes a plurality of frusto-conical shaped guide pins extending from the top surface of the flange. The shroud head flange includes a plurality of guide pin openings configured to align with the guide pins located on the top guide flange. Each guide pin opening includes a frusto-conical portion that extends through the flange from the bottom surface of the shroud head flange and has a slope equal to the slope of the frusto-conical pins. Each guide pin opening also includes a cylindrical portion that extends from the small base of the frusto-conical portion of the guide pin opening to the top surface of the shroud head flange.

Abstract: A seismically isolated liquid metal reactor power block is described. In one embodiment, the liquid metal reactor power block is a substantially rectangular nuclear plant with two nuclear islands and a common liquid metal service facility commonly supported on a seismic isolation platform near grade level. The seismic isolation platform supports the nuclear islands in a mirror-symmetric arrangement on opposite sides of a shared liquid metal service facility. A containment pressure suppression system connects the primary pressure relief volumes of the containment structures of each nuclear island to that of the other and also to a common pressure relief volume in the liquid metal service facility, thereby reducing the pressure load on each individual containment structure.

Abstract: A seismic load suppressing system for a nuclear reactor that reduces the seismic input to the reactor and the internal components of the reactor including the fuel is described. The seismic load suppressing system includes, in one embodiment, a plurality of seismic isolators positioned circumferentially between the reactor pedestal and a lower pedestal and a plurality of displacement limiters coupled to a lower pedestal skirt and positioned adjacent the reactor pedestal. The seismic isolators include a plurality of alternating layers of a resilient material and steel plates bonded together. Each displacement limiter includes a cantilever beam. The beam includes a beam core and two steel plates laminated on opposing outer surfaces of the beam core. The steel plates are fabricated from carbon steel having a lower yield strength than the beam core.

Abstract: A shroud attachment assembly which provides both structural support for the shroud and a seal between the shroud and shroud support is described. In one specific embodiment, the shroud attachment assembly includes a shroud flange formed integral with the shroud, and a pump deck flange formed integral with the pump deck. The shroud flange extends over the pump deck flange, and a plurality of studs extend through aligned openings in the flanges. In addition, a plurality of wedges are secured to the shroud flange by studs, and the wedges extend between the shroud flange and the pump deck flange. Generally, the studs which extend through aligned openings in the shroud and pump deck flanges transfer vertical loads from the shroud to the pump deck, and the wedges transfer horizontal loads from the shroud to the pump deck.

Abstract: An auxiliary wedge apparatus configured to couple to the restrainer bracket of a jet pump assembly is described. In one embodiment, the auxiliary wedge apparatus includes a support block configured to couple to the restrainer bracket, and a wedge configured to slidingly couple to a wedge channel in the support block, and to engage the inlet mixer to restore a tight rigid fit-up of the jet pump components. The support block includes a wedge channel having tongues depending from the parallel sides of the channel and a hook shaped portion configured to receive the restrainer bracket. The support block also includes a lock screw to couple the support block to the restrainer bracket. The wedge includes grooves in the parallel sides of the wedge configured to slidingly engage the tongues extending in the wedge channel.

Abstract: A neutron absorbing pin including at least, a neutron absorber, a thin-wall pipe surrounding the neutron absorber, and a cladding disposed at a distance from the thin-wall pipe. In the neutron absorbing pin, the difference between the coefficient of thermal expansion (.alpha.1) of the neutron absorber and the coefficient of thermal expansion (.alpha.2) of the thin-wall pipe has an absolute value of .vertline..alpha.2-.alpha.1.vertline..ltoreq.10.times.10.sup.-6 /K.

Abstract: Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

Abstract: A nuclear power plant has a reactor core and a containment chamber for receiving core melt of the reactor core. A cooling system for cooling the containment chamber includes a flooding container to be filled with coolant fluid. A cooling pipe leads from the flooding container to the containment chamber. A passively opening closure element closes the cooling pipe in the flooding container and opens as a function of a level of the coolant fluid.

Abstract: A control rod includes four panels in cruciform section each containing a plurality of absorber tubes. The absorber tubes are mounted between upper and lower mounting structures having generally T-shaped slots opening toward one another. The absorber tubes have generally T-shaped end plugs for reception in the slots. The slots and end plugs are configured to permit limited axial movement of each absorber tube relative to the control rod and to adjacent tubes. The absorber tubes are maintained in tension during all nuclear operations including during a scram.

Abstract: A method and an apparatus for repairing a core plate having a cracked support structure. The method entails the step of providing a vertical compression load path between the core plate and one or more control rod guide tubes. This is accomplished by installing a plurality of load-bearing wedges between the core plate and the step of each guide tube. Each load-bearing wedge is placed between the core plate and the guide tube machined step so that core plate pressure loads are reacted by the weight of the fuel rather than the core plate support structure. The load-bearing wedges can be installed without removing the guide tube or the fuel support casting.

Abstract: In a nuclear reactor including a reactor vessel and a core shroud disposed within the reactor vessel in laterally spaced relation, a seismic support includes a spacer extending between laterally opposed surfaces of the reactor vessel and the core shroud to limit lateral movement of the core shroud relative to the reactor vessel and a deformable member carried by the spacer and having a configuration to deform substantially elastically in response to impacts caused by an operating basis earthquake and substantially plastically at relatively constant load in response to impacts caused by a design basis earthquake. The deformable member can be carried at an end of the spacer adjacent the reactor vessel or between the spacer and a bearing plate, and can be formed of a material having a yield strength less than that of the spacer and/or be of reduced cross-sectional area as compared with the spacer to yield under a desired seismic load.

Abstract: In order to support a shroud against bending, shear, stress and torsion, straps are placed in strategic positions with respect to cracks or like types of weaknesses which have been detected, and fastened in place using a suitable fastening technique. Holes are formed using an EDM technique in cooperation with supports for an EDM placed on the straps, and bolt units which have an expanding portion are inserted into the holes, torqued and expanded in a manner which fastens the strap to the shroud. A temperature responsive bimetallic strain system which produces a clamping force, is created by forming the straps of a material which has a lower expansion coefficient than the shroud material. As the reactor heats, the expansion differential tensions the straps and produces a clamping force.

Abstract: A joint for interfacing a steel head closure and a steel-lined prestressed concrete reactor vessel (PCRV). The steel head closure is bolted to a thick steel annular plate and a thermal barrier is arranged in a volume bounded by the head ring, the liner and the concrete of the PCRV wall. The thermal barrier is made of zirconia sand or other material having a low thermal conductivity, a high bulk modulus equal to that of concrete, and a low shear modulus. Differential thermal expansion of the vessel head closure and PCRV is accommodated by a flexible connection of the PCRV liner to the inner periphery of the head ring. In the alternative, the steel head closure is interfaced to the top of the PCRV via a sliding joint or a thermal stress relief ring. In accordance with a further alternative, thermal stresses in the steel vessel head closure are mitigated by reducing the temperature of the head closure by installing thermal insulation on all internal surfaces thereof.

Abstract: An apparatus for stabilizing a shroud against vertical and lateral deflection. The apparatus forms a vertical load path by which the shroud flange can be anchored to the shroud support plate. In the tensioned state, an upper support assembly exerts a restraining force on the top surface of the shroud flange which opposes separation of the shroud at any assumed failed circumferential weld location. The apparatus further incorporates upper and lower restraint springs which oppose lateral deflection at the elevation of the top guide support ring and at the elevation of the core plate support ring respectively. The vertical restraint assembly is made of material having a coefficient of thermal expansion which is less than a coefficient of thermal expansion of the shroud material. A desired differential thermal expansion of the vertical restraint assembly and the shroud can be attained by the selection of member lengths and materials having appropriate coefficients of thermal expansion.

Abstract: A vertical seismic isolation structure is capable of achieving vertical seismic isolation of nuclear reactor components connected to each other by primary pipings, while suppressing the relative displacement with respect to the primary pipings. A nuclear reactor vessel, circulating pumps and heat exchangers, all of which are connected to each other by primary pipings in which a coolant is circulated, are mounted on a common deck. Vertical seismic isolators each of which has laminated large coned disc springs and is capable of expanding and contracting in only the vertical direction are installed on the top of a concrete wall surrounding the respective nuclear reactor components. The common deck is placed on all the vertical seismic isolators so as to support the entire seismic isolation structure (the common deck, the nuclear reactor components, etc.).

Abstract: A pressure suppressing and aerosol scrubbing system configured to be located between low pressure upper containments associated with modular reactor vessels is described. In one embodiment, the pressure suppression and scrubbing system includes a water tank having connected, partially filled water chambers and a vertical baffle having horizontal vent holes positioned near the bottom of the tank. Each water chamber is connected by a vent/relief line to one of the respective rectangular upper containments. The partially filled water tank assures that the respective containments are isolated from each other under normal operating conditions due to the water trap inherent in the system. In the event that an accident occurs in one of the reactors, the affected containment is heated by the sodium spray and/or pool fire and such heat forces its way through the pressure suppressing and scrubbing system to the unaffected reactor containment.

Abstract: A steam generator originally having hydraulic snubbers damping lateral thermal movement of the steam generator is backfit to laterally support the steam generator without hydraulic damping. Hydraulic fluid is emptied from piston cylinders of all snubbers extending from a steam generator; and then stops are placed around piston rods extending into the piston cylinders of empty snubbers extending from the back side of the steam generator to stop the thermal growth of the steam generator in the operation of the nuclear reactor vessel to generate power. Thus, the costs associated with maintaining, inspecting and testing hydraulic snubbers are entirely eliminated. Also, the backside empty snubbers are used as bumpers and do not incur disposal costs.

Abstract: A strongback for lowering a tie rod into the downcomer annulus of a boiling water reactor during a shroud repair operation. The tie rod strongback is suspended from a cable via a cable adaptor at its upper end. The lower end of the strongback is coupled to a tie rod adaptor, which in turn couples to the top of the tie rod. The strongback is a welded assembly of square tubes, channels for reinforcing the joints of the welded tubes, and upper and lower couplings. In particular, the strongback has mutually parallel first and second rigid linear members which are disposed vertically when the strongback is suspended from a plumb cable. The second rigid linear member is connected to the first rigid linear member by a relatively obliquely disposed third rigid linear member.

Abstract: The present invention relates to a miniaturized nuclear utilizing improved pressure tube structural members. More particularly, the present invention relates to a new miniaturized nuclear reactor utilizing novel structural members that are used to support the loads and stresses of multiple nuclear reactor fuel channel pressure tubes and calandria tubes in a moderator.

Abstract: A method and an apparatus for stabilizing a shroud against lateral deflection. The shroud restraint stabilizer has springs which exert radially inwardly directed forces that resist lateral deflection of the shroud relative to the reactor pressure vessel. Each stabilizer includes a stabilizer preload spring and a seismic restraint spring. Each spring has resilient arms integrally connected to opposing ends of a common main support span. The resilient arms of the stabilizer preload spring form a first leaf spring of relatively lower stiffness and the resilient arms of the seismic restraint spring form a second leaf spring of relatively higher stiffness. The different stiffnesses are achieved by providing the first leaf spring with a longer base than that of the second leaf spring. The pads on the seismic restraint spring are in contact with the reactor pressure vessel, but are not preloaded under operating conditions. The seismic restraint spring is loaded in response to lateral deflection of the shroud, e.g.

Abstract: A pressure vessel of a pressurized-water reactor has an upper part in which coolant is collected in a plenum. The coolant enters the plenum through a grid plate on which tops of the fuel assemblies are supported and the coolant is removed from the plenum laterally. Support columns and control-rod guideways form obstacles to the flow of the coolant which can lead to damming-up below the grid plate with horizontal pressure differences and corresponding horizontal flows. Throttle plates inserted in the top of the fuel assemblies or throttle inserts in apertures in the grid plate produce a uniform pressure in the coolant below the grid plate so that it flows uniformly and in a vertical direction through the individual fuel assemblies.

Abstract: A nuclear power generation complex constructed on a stratum of rock overlain by soil deposits. All major structures in the power generation complex (i.e., reactor, turbine, radwaste and control buildings) are located on a common mat foundation that houses elastomeric bearings or seismic isolators. The foundation includes a concrete slab supported by a plurality of pedestals. Each of the pedestals is embedded in the rock. Preferably, each pedestal is a circular metal shell filled with underwater-type concrete, and the concrete slab is reinforced with steel bars. For example, a reactor building is supported by one set of seismic isolators mounted on the concrete slab, and a turbine building is supported by another set of seismic isolators mounted on the concrete slab. The isolators filter out a great deal of the seismic vibratory inputs. The common mat foundation eliminates differential movements between the buildings.