Abstract: A method and apparatus for repairing nuclear reactor core shrouds having horizontal cracks in heat affected zones of welds joining a horizontal plate to upper and lower cylindrical sections of the shroud includes securing a plurality of braces between an upper shroud flange and a top surface of the horizontal plate near an inner edge of the plate. The braces are installed at a respective plurality of angularly spaced locations within the upper cylindrical section of the shroud and each have an upper notched end for bearing against a bottom edge of the flange and a lower notched end for bearing against the top surface of the horizontal plate and/or an outer peripheral edge of an upper core grid structure mounted on the top surface of the plate.

Abstract: A method and apparatus for repairing nuclear reactor core shrouds includes applying vertical compressive forces to the shroud in situ to urge opposed ends of the shroud toward one another using a plurality of tie-rods, each including a plurality of coaxial longitudinal members separated by a radial gap filled with reactor coolant or some other damping fluid. One of the longitudinal members is secured to upper and lower portions of the shroud for maintaining a sufficient preload to hold the shroud together and is coupled with one or more other longitudinal members in a manner to transmit axial loads thereto upon extending a predetermined axial distance. A plurality of radial spacers can also be positioned at longitudinally spaced locations along the length of the tie-rod to assure control rod insertion by limiting lateral movement of the shroud and can reduce the preload required for preventing the formation of new cracks and stabilizing existing cracks.

Abstract: In a pressure vessel of a nuclear reactor containing a core assembly enclosed within a core shroud, the shroud spaced radially inwardly of a side wall of the pressure vessel, and an annular pump deck welded in a place in an annular radial space between the core shroud and the side wall of the pressure vessel. The shroud provided with a radially outwardly extending flange seated on and removably secured to the pump deck. Wedge elements are interposed between flange portions and the side wall.

Abstract: A boiling water reactor control rod drive support system having radiation shields to protect persons working under the reactor vessel from radiation exposure and to limit the travel of a control rod in the event that a control rod housing is ruptured. The radiation shields include radiation shield rings for surrounding the control rod drive housing and radiation shield cups for surrounding the bottom portion of the control rod drive. The radiation shield rings can be used with support systems having radiation shield cups placed on rows of support bars or with existing support systems.

Abstract: A pressure vessel apparatus for containing fluid under high pressure and temperature including pressure surges caused by in-vessel explosions includes a vessel main body, a vessel top body and a vessel head. Elongated tendons interconnect the vessel main body to the vessel top body and elongated tendons with removable anchors also connect the vessel head to the vessel top body. The tendons with insulated protecting sleeves connecting the vessel main body to the vessel top body are chosen to allow the vessel top body to temporarily separate from the vessel main body before the vessel head separates from the vessel top body. These tendons may be mounted directly, on springs, or on dampers in order to provide optimum (minimum) response to the postulated load cases.

Abstract: An apparatus for stabilizing a core plate against lateral deflection. A plurality of core plate wedge assemblies (e.g., four) are installed in the circumferential gap between the circular outer peripheral edge of the core plate and the circular cylindrical wall of the middle shell of the shroud at respective azimuthal positions. These core plate wedge assemblies are wedged into place with preload to maintain the spacing between the core plate and the shroud, thereby maintaining the alignment of the fuel assemblies. The core plate wedge assemblies further serve to transmit loads from the nuclear fuel core to the core shroud at azimuthal locations where lateral shroud restraints outside the shroud contact the reactor pressure vessel.

Abstract: A device for radially holding the envelope of a tube bundle and spacer plates of a steam generator, constituted by abutments traversing the wall of the bundle envelope and fixed to the wall level with the spacer plates. The said abutments being distributed on the periphery of each spacer plate in order to radially separate the bundle envelope from the pressure envelope of the steam generator. Each abutment comprises a first part (37) fixed to the bundle envelope (6) and contacted with the inner face of the pressure envelope (1), as well as a second part (32) exerting rigid holding stress on the spacer plate.

Abstract: The invention relates to a device for the radial holding or supporting of the envelope of a bundle of tubes and spacer plates of a steam generator or boiler, constituted by stops or abutments traversing the wall of the bundle envelope and fixed to the wall level with the spacer plates. The abutments are distributed over the periphery of each spacer plate in order to radially separate the bundle envelope from the pressure envelope of the steam generator and to exert rigid supporting stresses between the spacer plates and the bundle envelope. The abutments are adapted for sliding on the inner face of the pressure envelope. Each abutment is constituted by a first part which comes into contact with the inner face of the pressure envelope, a second part and a third part coming into contact with the spacer plate. The second part serving as the intermediary between the first and third parts. The first part has a first thread permitting its fixing by screwing to the bundle envelope and a second thread.

Abstract: Pressure vessel apparatus includes upper and lower pressure vessel housings maintained together under pressure by tendons. Bellows are connected to the pressure vessel housings in order to provide leaktight barriers, and to permit motion of the pressure vessel housings. The tendons are divided into two sets, with the tendons of each set of tendons being stressed differently than the tendons of the other set of tendons. The apparatus when utilized as a nuclear reactor pressure vessel includes a core catcher with flotation pool connected to and disposed below the pressure vessel housings and defining a core catcher interior in communication with the pressure vessel interior for receiving core material from the pressure vessel interior resulting from reactor melt-down.

Abstract: The equipment comprises a bottom plate formed with outlet openings for the coolant, a top plate, spacer rods interconnecting the plates, and cluster guides for guiding control clusters. At least some of the guides have respective top lengths that project above the top plate and bottom lengths that extend between the plates. Each bottom length includes a base bearing against a frustoconical seat secured to the bottom plate, and it is subjected to the force of resilient means. Together, the lengths of each guide constitute a one-piece cartridge suitable for insertion in a corresponding column. The guide also includes a cap for receiving the top length, which cap is fixed to the top plate against which the resilient means bear.

Abstract: An adjusting and positioning assembly for a bearing element (55), constituted by a nut (52) carried by a base plate, (40) fixed on a wall of a concrete structure in which the component is arranged, and a screw-jack (54) engaged with and blockable with respect to the nut (52). The transverse holding of the lower part of a steam generator pressurized-water nuclear reactor is provided by an assembly of supports each constituted by an adjustable transverse holding device (24) according to the invention. The position of the bearing element is adjusted in such a manner as to maintain play allowing movement of the steam generator as a result of thermal effects or in the event of an earthquake.

Abstract: A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity.

Abstract: In a method for fabricating an end fitting (12,200) for a nuclear fuel assembly (10), wherein the end fitting includes at least one spring member (24,202) having an active external surface (32,206) for contacting rigid core support structure (100) in the nuclear reactor, the improvement comprising applying a coating to the active surface of the spring member. The coating is selected from the group consisting of nitrides, Cr, TiC, CrC, ZrC and NiTaB. The spring member is preferably Inconel and the coating is one of ZrN or TiN. Each spring member can be formed by nesting a plurality of spring elements (28,34,36 or 208,210) so that the spring elements are in contact with each other and the method includes applying said coating to each of the spring elements at least where they contact each other.

Abstract: The internal structure comprises a bed (5) supporting the core (3) of the reactor and supplying liquid metal to the core, plating (6) resting on the bottom of the main vessel (1) of the reactor and an internal vessel (8) separating the hot collector (12) from the cold collector (13) of the liquid metal. The internal vessel (8) is directly fixed to the upper part of the bed (5) which rests on the plating (6) through the medium of slidable bearing means (21), (22). The bed (5) is connected to means (18) for supplying cooling liquid metal in a zone located at its periphery and outside the internal vessel (8) and the plating (6). Assemblies (31) providing a lateral neutronic protection are fixed to the bed (5) at the periphery of the fuel core (3).

Abstract: A device for restricting the motion of a thimble tube in a nuclear reactor during normal reactor operating conditions. A cylindrical sleeve open at each end has upper, middle, and lower sections. The upper section has an over-expanded area adjacent the top end to aid in proper positioning in the lower core support plate. The middle section is formed from inner and outer sections attached at their ends and formed from material having different coefficients of thermal expansion. The inner portion of the middle section moves from a first normal relaxed state during cold conditions in the reactor to a second flexed state during hot conditions in the reactor to provide a clamping action on the thimble tube received by the sleeve. A baffle in the lower section reduces coolant flow velocity and pressure through the sleeve.

Abstract: A mounting device for a heat exchanger or the like elongated member subject to longitudinal expansion and contraction, that freely permits longitudinal movement sufficient to accomodate anticipated expansion and contraction of the member, but that provides positive restriction on movement in every other direction such as is likely to occur during an earthquake. The device moves on rolling, line-contact bearing surfaces, substantially eliminating the danger of binding or freezing of the device.

Abstract: Vessel structural support system for laterally and vertically supporting a vessel, such as a nuclear steam generator having an exterior bottom surface and a side surface thereon. The system includes a bracket connected to the bottom surface. A support column is pivotally connected to the bracket for vertically supporting the steam generator. The system also includes a base pad assembly connected pivotally to the support column for supporting the support column and the steam generator. The base pad assembly, which is capable of being brought to a level position by turning leveling nuts, is anchored to a floor. The system further includes a male key member attached to the side surface of the steam generator and a female stop member attached to an adjacent wall. The male key member and the female stop member coact to laterally support the steam generator.

Abstract: A device for wedging and positioning the upper support plate of the guides of control clusters in relation to a bearing surface linked to the vessel of a pressurized-water nuclear reactor, comprising an upper ring (22) and a lower ring (23) laid onto the cluster guide support plate (8) and the bearing surface (9), respectively, and between which is seated a stack of preloaded elastic washers (24). The device comprises a hollow connecting shank (26) with an axial channel (44) and passing through the two rings, to engage in a bore (20, 21) formed in the support plate or the bearing surface, the lower ring (23) being connectable to the shank, while the upper ring is mounted slidably relative to the lower ring (23) under the effect of the load exerted on the elastic washers.

Abstract: A device for the centering and fixation of a cluster guide flange and of a core plate of a nuclear reactor, extending parallel to one another in horizontal planes with a predetermined spacing. The device includes at least two axial guiding spindles of flange with respect to the core plate, diametrically opposite and rigidly connected to the plate or to the flange in order to engage into a housing in register formed in the flange or in the plate, respectively, and an assembly of self-locking shoes, adapted for sliding with a clearance within bores formed in the flange in order to be pressed on the plate surface, these shoes being each associated with a position control mechanism, carried by the flange and exerting on the shoes a force having a transverse component so as to cause, by reaction, immobilization of the flange with respect to the plate.

Abstract: A method of securing a cylindrical centering pin for a nuclear fuel assembly in a bore formed in a plate includes introducing an end of the centering pin into the bore until at least part of a coaxial, radially expandable wall portion of the end to be secured protrudes beyond the plate; and generating pressure in a closed hollow space surrounded by the wall portion with a pressure fluid until the wall portion radially expands and anchors the centering pin. According to another method, the end is introduced into the bore until at least part of the end to be secured, which has a periphery and an interior with radial slits formed therein, protrudes beyond the plate; and a mandrel-like tool is partially brought into the slits in the interior of the end of the centering pin and radially expands the end of the centering pin with the tool for anchoring the centering pin.

Abstract: A fuel assembly has a top nozzle which includes a lower adapter plate and a plurality of guide structures thereon. The lower plate has a periphery bounding an interior thereof mounted by guide thimbles. The guide structures are attached to and extend along the periphery of lower plate and upwardly therefrom. The top nozzle also includes upper hold-down plate and a plurality of leaf spring assemblies. The upper plate is mounted to the guide structures for slidable movement relative thereto such that the upper plate can move toward and away from the interior of the lower plate within the space bounded by the guide structures as the upper plate slidably moves along the guide structures. The leaf spring assemblies are interposed between and engaged with the lower and upper plates so as to yieldably support the upper plate in spaced relation above the lower plate and bias the upper plate for movement away from the lower plate.

Abstract: A support system is provided for a pressurizer relief and safety valve system in a nuclear power plant. The pressurizer has a generally cylindrical vessel wall topped with a dome-like member. The valve system includes a manifold generally located above and extending about the periphery of the vessel wall of the pressurizer.The valve support system comprises a plurality of paired lug members that extend outwardly from and are secured to the vessel wall at points spaced around the wall periphery. V-shaped columnar supports extend vertically upward from each of the lug pairs to the valve system manifold.A structural pin extends in the horizontal direction to support a base portion of each of the columnar supports relative to the associated lug pair. Disengageable locking clamps secure the pin to the lugs. A pair of diagonal columns extend upwardly in a V-shape from the support base portion to the manifold. A crossbar connects the diagonal columns at the top of each columnar support.

Abstract: The sleeve (20) for locking the guide tube has an outer surface comprising successivley a cylindrical and a frustoconical part in the axial direction, without any radially protruding part. The upper cylindrical part (25) of the sleeve constitutes a fastening shell of the sleeve (20) in the end fitting of the assembly and the lower frustoconical part (24) a ring for expanding the guide tube. The fastening shell (25) the thickness of which is substantially smaller than the thickness of the ring (24) can be distorted into cavities machined in the end fitting of the assembly. An annular interlocking groove (20) is machined in the upper part of the ring (24).

Abstract: Nuclear devices are detonated periodically inside an underground cavity in which a working fluid absorbs most of the thermal energy released by the device detonation. The kinetic energy contained in the shock wave created within the working fluid is absorbed by shock absorbers located outside the inner wall of the cavity. This wall consists of adjacent plate segments that separate and recoil outwardly as do pieces of shells of fragmentation grenades. The shock wave kinetic energy is transferred to the segments that are constrained in their outwardly-directed flight by shock absorbers. The shock-absorbing system is solidly attached to the rock structure surrounding the cavity. The loading transmitted to the rock is of a much lower intensity than that which the plate segments sense, though of a much longer duration. The momentum trapped in the shock wave is thus absorbed by the rocks but in a manner such that the rock structure is not affected.

Abstract: A snubber monitoring system is provided for a nuclear power plant wherein the snubbers are disposed at various locations within a plant containment structure to support piping and equipment rigidly against seismic and vibratory loading while permitting thermal expansion and contraction thereof. The system includes displacement sensing means and temperature sensing means coupled to each of the snubbers to generate continuous snubber displacement and pipe and equipment temperature signals.A sensor data highway system has cable connected data collectors disposed within the plant containment to collect the snubber sensor signals. A receiver and a programmed computer system are located outside the containment to process the snubber displacement and pipe and equipment temperature signals to provide substantially continuous snubber monitoring. A data highway cable extends through the containment to couple the signal collectors to the receiver and computer system.

Abstract: The wedging members (15) are introduced between the components (4) at a temperature t0 substantially different from the working temperature ts of the stream generator. The wedging member (15) comprise at least one portion (22) of shape memory alloy having a first configuration at the temperature t0 and a second configuration at the temperature ts, such that the wedging parts (15) have an initial transverse dimension less than the width of the separation spaces between the components (4) at the temperature t0, and a transverse dimension substantially greater than the initial dimension at the temperature ts. The components are wedged automatically at the moment when the installation is brought to the working temperature. The invention is applicable in particular to the anti-vibratory bars disposed in the top part or chignon of a steam generator of a pressurized water nuclear reactor.

Abstract: A device and method for augmenting support and suppressing vibration for U-bend tubes and tube spacer supports in a vapor generator vessel. The device is installed in the vapor generator vessel from outside the vessel through dual oppositely oriented openings cut through the wall and inner shroud of the generator vessel. The device extension rods are inserted through one of the openings, and the vibration suppressor blades are each inserted sequentially through the other opening and installed on a support rod and then rotated upwardly into place between the parallel tube batwing spacer supports by using a special push/twist tool, after which the support rod is indexed forward to provide positive engagement between the rod and blade, such as by a rod key and blade slot means. The suppressor blades are retained in place on the step block and rod by a clamp block attached to the step block, and the vessel dual openings are then covered by pressure-tight flange plates.

Abstract: A liquid metal nuclear reactor has a sodium tank, having a closed arcuate bottom and upstanding cylindrical wall for containing a reactor core in a pool of liquid metal coolant, which is disposed in a guard tank having a bottom and upwardly extending cylindrical shell, and a support inside the guard tank adjacent the closed bottom of the sodium tank to support the two tanks in spaced relationship to each other. The sodium tank has an open top spaced from an overlying closure deck, with the cylindrical shell of the guard tank secured to the closure deck. The support is preferably an inner ledge. Support legs are also provided about the bottom periphery of the guard tank adapted to rest on a base mat of a reactor cavity.

Abstract: Concrete support ledges (67) and columns (94) for the upper internals (60) and lower internals (63) structures are poured into the floor (41) and walls (36) of the refueling cavity (33) at the time of initial plant construction. Such concrete supports (67, 94) provide greater stability and do not require any external bracing. A pedestal assembly (100) is associated with each of the concrete supports (67, 94) to provide an adjustable and secure support for the internals structures (60, 63).

Abstract: A building, especially a nuclear plant, includes concrete walls enclosing components such as plant components as a protection against external action. The concrete walls have exposed locations, and the concrete walls have double-layered regions at the exposed locations with double layers defining hollow spaces therebetween. A damping material may fill the hollow space.

Abstract: A method for limiting movement of a thermal shield of a nuclear reactor, and a thermal shield displacement limiter therefor, is provided where at least four and preferably six or more key-way inserts are positioned in pockets machined in the upper portion of a thermal shield, and bolted and doweled to the thermal shield. Each key-way insert is engaged by a limiter key that is bolted and doweled to the core barrel to tangentially restrain movement of the thermal shield relative to the core barrel, while allowing radial and axial movement to the thermal shield relative to the core barrel. Preferably, lock pins extend diametrically through the bolts and dial pins and engage the key-way insert and limiter key, respectively, to lock the bolts and dowels in position, and the engaging surfaces between the key-way inserts and the limiter keys are wear hardened.

Abstract: Core barrel for a high-temperature nuclear reactor, including internal parts formed of an inner layer of graphite blocks stacked on each other and an outer layer of carbon blocks stacked on each other together forming columns, the blocks of the respective layers being staggered in height, offset relative to each other in azimuth and having indentations alternately formed therein causing each two layers of graphite blocks to be supported on a protruding layer of carbon blocks and each two layers of carbon blocks to be supported on a protruding layer of graphite blocks.

Abstract: A method of providing a nuclear steam supply system that is constructed to include a barge or flotation base as an integrated portion thereof. The system is tested for operability and safety at the factory and then towed along navigable coastal or inland waterways, or overseas to foreign locations, to a prepared foundation site at its point of use. The overall unit is so constructed and arranged as to permit the removal of top and side portions to permit passage under low bridges and through narrow locks without requiring disassembly of safety class piping and wiring. The method further provides for a complete nuclear power plant constructed on multiple barges at a separate factory site in parallel with plant site preparation, i.e., in estuaries or caves.

Abstract: The adapter extension of subject invention provides for the coaxial extension of a plunger or rod whose effective length has become shortened by weakness or failure of a back-up spring which causes the plunger or rod to lose its effectiveness as a strut, for example, a situation that occurs primarily in nuclear reactors where the plunger engages a grid, whereas the grid provides lateral support to the upper end of the dry tube. The adapter is thimble-like and fits tightly over the end of the plunger that engages the grid. The adapter contains a spring which has the effect of substituting for or supplementing the original spring behind or below the plunger. The body of the adapter includes a retainer for holding the interior spring captive during assembly and operation of the adapter.

Abstract: A gas-cooled high temperature reactor having a core filled with spherical fuel elements is provided comprising a graphite side reflector including at least one nose-like projection comprised of a plurality of graphite nose stones stacked one upon the other, said nose stones each including at least one vertically disposed continuous cavity aligned with at least one vertically disposed continuous cavity in adjacent nose stones, said cavity adapted to receive discrete absorber material elements introduced into the reactor core, said nose stones further including at least one vertically aligned continuous gap which extends into said nose stones from a front portion thereof and is aligned with a corresponding continuous gap in adjacent nose stones, communication between said continuous gap and said cavity being prevented, said nose stones also comprising in top and bottom surfaces thereof sealing means which enable adjacent nose stones to be stacked in a manner sufficient to seal said at least one cavity in a gas-

Abstract: The pump (8) comprises a vertical body (8a) and a discharge sphere (10) in its lower part. Discharge pipes (14) are connected to the sphere (10). The sphere (10) is suspended from the plating (4) of the reactor through a flexible metal strip (20). The sphere (10) is connected to the strip (20) in a region located below the region of connection between the strip (20) and the plating (4). The invention is in particular applicable to the primary pumps of fast neutron nuclear reactors of the integrated type.

Abstract: A radiation shield (10) is comprised of three, equiangularly extending, arcuate sections (14) which are secured to the nuclear reactor pressure vessel closure head (12) so as to annularly surround the nuclear reactor control rod drive mechanisms (26) and thereby prevent radiation emanating from the mechanisms (26) from reaching maintenance personnel who may be operatively working within the peripheral section (20) of the closure head. The shield sections (14) are bolted together by fasterners (18) and are secured to closure head lift lugs (30) by means of mounting brackets (38) and transversely extending pins (36). Clevis assemblies (34) are also pinned to the mounting brackets (38) and lift lugs (30), and lift rods (32) are threadedly engaged with the clevis assemblies (34). In this manner, when the closure head (12) is lifted from its associated pressure vessel, the radiation shield (10) remains with the closure head (12).

Abstract: A sleeve-like sealing unit, particularly a thimble guide extender is adapted to be installed on and coaxially with a thimble guide mounted on a lower core plate of a nuclear reaction vessel and projects toward and terminates at a distance from an underface of a bottom nozzle forming part of a fuel assembly and supported on a top face of the lower core plate. The extender has a lower fitting adapted to surround the thimble guide at a bottom length portion thereof adjacent the top face of the lower core plate; and an upper fitting being axially displaceable relative to the lower fitting and being adapted to surround the thimble guide at a top length portion thereof. The upper fitting has an axial passage and a terminal opening defined by an outer annular radial face adapted to be in a face-to-face engagement with the underside of the bottom nozzle.

Abstract: A nuclear power plant includes a concrete biological shield, a metallic reactor pressure vessel surrounded by the concrete biological shield, coolant connector pieces connected to the reactor pressure vessel, coolant lines connected to the coolant connector piece, and a support engaging the coolant connector pieces and holding the reactor pressure vessel. The support includes at least one supporting tube surrounding one of the coolant connector pieces, the supporting tube having an inner surface anchored on the one coolant connector piece and an outer surface anchored on the biological shield.

Abstract: Disclosed is a reactor internals hold down spring and upper head region cooling passage. The hold down spring utilizes a plurality of stacks of Belleville spring washers to provide spring load and deflection capability to hold down the reactor vessel internals. Each spring assembly includes a generally leak tight means for passing an adjustable coolant flow to the upper reactor vessel head region in order to maintain that region at inlet coolant temperature. The spring assemblies are angularly spaced about a core barrel support flange to cooperate with coolant flow passage formed therein.

Abstract: A novel concrete core support structure for nuclear reactors is described.

Abstract: The cover plug (4) comprises a cylindrical barrel (8), a support plate (10), a group of vertical tubes (12) for the control rods and for the instrumentation of the core, an apertured transverse plate (28) fixed to the lower end of the barrel (8) and a transverse deflection plate (32) located above the apertured plate (28). The transverse deflection plate (32) is suspended from the barrel (8) by substantially vertical elastically yieldable strips (34) evenly spaced apart on the periphery of the barrel (8). The strips (34) are welded to the barrel (8) by their upper end and to the edge of the deflection plate (32) by their lower end. They are located in the region of openings (35) in the barrel (8) which permit radial movement of the deflection plate. The deflection plate (32) is also integral with tubes (14b) receiving the control rods disposed in a single row coaxial with the barrel (8).

Abstract: Slots are provided in the wrapper of a nuclear steam generator to increase the local flexibility of the wrapper to allow for the difference in thermal expansion between the carbon steel wrapper and the high corrosion resistant alloy tube support members. The radial structural load paths between the tube support members, the wrapper, and the outer shell provided by the radial support means, about which the expansion slots are provided, are not affected by the addition of the expansion slots. Since the slots are preferably provided by means of a high speed industrial laser, they are very thin and do not permit substantial commingling of cooler downcomer water and the boiling mixture in the tube bundle region. The expansion slots may be vertically or horizontally oriented and may be cut into the wrapper on more than one side of the radial support means. The slots may end in small curved portions to decrease stress concentrations and increase fatigue life.

Abstract: Apparatus such as a nuclear fuel sub-assembly for use in a liquid alkali metal environment has bearing surfaces (4) coated with alumina or aluminide particles to reduce wear, fretting and friction. The particles are deposited by plasma or detonation gun using a nickel chromium alloy as a metallic bond.

Abstract: A multiple shell pressure vessel utilizes a number of spaced apart, concentrically disposed pressure vessel shells within and spaced apart from an outer pressure vessel shell, the spaces between each shell being filled with a low melting point, high boiling point material selected from the group, lead, tin, antimony, bismuth, or sodium and potassium, and mixtures thereof and subjected to a pressure whereby the innermost pressure vessel shell wall is maintained in compression while the main vessel is pressurized. Chemical compositions or compounds which contain boron or cadmium may also be added to the filler material. The vessel may include devices for keeping the pressure in the space between the innermost pressure vessel shell and the next innermost shell at a fixed predetermined multiple of the pressure within the innermost pressure vessel shell.

Abstract: A device for relieving thermally induced stresses developed in a nuclear fuel assembly during reactor heatup is disclosed herein. The device generally comprises a stress relieving fastener capable of connecting a fuel assembly top nozzle, which may be stainless steel, to a threaded stud, which may be Zircaloy. The stud is attached to a fuel assembly channel. The fastener includes a threaded nut having a deformable portion for relieving thermally induced stresses developed in the stud by the differential thermal expansion of the top nozzle and stud. In a first embodiment of the nut, the deformable portion comprises a circumferential, deformable ridge which is substantially recessed from the marginal edge of the nut and which is disposed on the bottom most surface of the nut. The ridge contacts the top nozzle when the nut threadedly engages the stud. A second embodiment of the nut is similar to the first embodiment except that the deformable ridge is disposed flush with the marginal edge of the nut.

Abstract: A nuclear fuel assembly for a spectral shift reactor has an upper end piece and a lower end piece secured to a structural element belonging to the upper end piece by guide tubes arranged for receiving a cluster of control elements. The upper end piece further comprises a support member arranged for receiving said cluster at the end of the fall thereof, means for guiding said support member. The support member is guided for movement parallel to the axis of the guide tubes. Springs are contained within the end piece and disposed between the structural member and the plate for braking the cluster at the end of the fall thereof.

Abstract: A novel liquid nuclear reactor is described which comprises a reactor vessel that is connected through upper and lower liquid conduit means to one or more satellite tanks that contain a heat exchanger means and pump means.

Abstract: A mechanism for anti-seismic connection between the pump base and the dome of a nuclear power station is characterized by the fact that the pump base (1) is equipped externally with a flange (3) seated with a slight axial play and significant radial play within a socket (4) which is a fixed cover closed by and able to slide axially with a slight play within a cylindrical reinforcement (7a) of the dome (7). A radial damping mechanism (8) is installed between the flange (3) and the socket (4) to permit slow displacements between these two entities, but to block abrupt displacements. The lower end of the socket (4) comprises a collar (15) equipped with a metallic journal (16) which can slide axially within a bearing (17) firm with the internal wall of the dome (7).

Abstract: An upper tie plate assembly for a nuclear reactor includes the usual tie plate having holes to receive the tie rod of the fuel assembly. A lower side member is welded to each side of the tie plate and there is a spring seat at each corner, connected to the lower side members. An upper side member parallels each lower side member at a higher elevation. At an intermediate elevation, there are leaf springs paralleling the side members, with their ends received within spring seats provided at the corners of the plate. Sliders are mounted on the upper side members and engage the midpoints of the upper surfaces of the springs. As the fuel assembly expands, the sliders, which engage an upper core plate of the reactor, deflect the springs from their normal, upperwardly bowed position, through a flattened position, to a downwardly bowed position. The ends of the springs are mounted within the spring seats in such a manner that they are free to accommodate extension of the springs.