Abstract: A method comprises inserting the fuel rod (4) through the spacer grids (14) aligned along an assembling axis (A) with passing the fuel rod (4) through a lubrication chamber (30) aligned with the spacer grids (14) such that the lubrication chamber (30) is passed through by the fuel rod (4) before the insertion of the fuel rod (4) through one of the spacer grids (14), and circulating a lubrication fluid containing a gas and a lubricant in gaseous phase and/or mist form in the lubrication chamber (30). The lubrication fluid is injected in the lubrication chamber (30) at a temperature strictly higher than ambient temperature, such that lubricant deposits or condensates in liquid phase with forming a lubricant film on an external surface of the fuel rod (4) that is being inserted through said one of the spacer grids (14).

Abstract: An apparatus including a lower end fitting having a top planar surface, a bottom planar surface, a counterbore defined therebetween, an opening extending from the counterbore to the top planar surface, and a lock recess that extends both radially outwardly from the counterbore and extends inwardly into the lower end fitting from the bottom planar surface. The apparatus includes a guide tube having a lower end and an end plug configured to connect with the lower end of the guide tube. The end plug has a threaded shaft sized to pass through the opening of the lower end fitting, and a female lock nut has a deformable side wall configured to be swaged into the recess of the lower end fitting. The lock recess remains visible as viewed from the bottom planar surface of the lower end fitting when the female lock nut is fully disposed in the counterbore.

Abstract: A base station central unit receives, from a base station distributed unit, configuration parameters of at least one periodic radio resource for packet transmissions of a wireless device in a radio resource control (RRC) inactive state or RRC idle state. The base station central unit sends, to the base station distributed unit, a wireless device context release message comprising an RRC release message and an indication that the base station distributed unit maintains the at least one periodic radio resource after the wireless device transitions to the RRC idle state or the RRC inactive state. The base station central unit receives, via the base station distributed unit, data from the wireless device while the wireless device is in the RRC inactive state or the RRC idle state.

Abstract: An article fixing device includes an adjustment mechanism, a protrusion, and a fastening bolt. The adjustment mechanism is joined to a side of a lower surface of the roof panel. The protrusion is provided integrally with the adjustment mechanism and fixes a roof rail. The fastening bolt fastens a support member to the adjustment mechanism. Further, the adjustment mechanism includes a fitting convex member provided integrally with the protrusion, a fitting concave member that that is able to be screwed into the fitting convex member, and a resin ring that is provided in the fitting concave member and that cooperates with rotation of the fastening bolt.

Abstract: Method for long-term dry storage of spent nuclear fuel includes injecting the inert gas into a cavity formed in a cask between inner and outer lids under pressure greater than the pressure of the inert gas in the inner cavity of the cask in which the spent nuclear fuel is located. The cask contains a body with a spent nuclear fuel container placed in it, tight inner and outer lids forming a cavity into which the inert gas is pumped under pressure greater than the pressure in the inner cavity of the cask. The outer lid comprises a labyrinth hole closed with a flange and sealed with a gasket. The flange comprises an angle valve, to which at least two pressure sensors are connected. The angle valve and sensors are covered with a protective cover during long-term storage. The gasket is installed in an annular groove of the lower flange.

Abstract: A block of dynamic memory in a DRAM device is organized to share a common set of bitlines may be erased/destroyed/randomized by concurrently activating multiple (or all) of the wordlines of the block. The data held in the sense amplifiers and cells of an active wordline may be erased by precharging the sense amplifiers and then writing precharge voltages into the cells of the open row. Rows are selectively configured to either be refreshed or not refreshed. The rows that are not refreshed will, after a time, lose their contents thereby reducing the time interval for attack. An external signal can cause the isolation of a memory device or module and initiation of automatic erasure of the memory contents of the device or module using one of the methods disclosed herein. The trigger for the external signal may be one or more of temperature changes/conditions, loss of power, and/or external commands from a controller.

Abstract: A base station distributed unit sends, to a base station central unit, configuration parameters of at least one periodic radio resource for packet transmissions of a wireless device in a radio resource control, RRC, inactive or RRC idle state. The base station distributed unit maintains, in the RRC inactive or RRC idle state of the wireless device, the configuration parameters of the at least one periodic radio resource. The base station distributed unit receives, from the wireless device in the RRC inactive or RRC idle state, one or more transport blocks via the at least one periodic radio resource.

Abstract: A system for monitoring a reactor module housed in a reactor bay may include a mounting structure and one or more extendable attachment mechanisms connected to the mounting structure. Additionally, one or more monitoring devices may be operably coupled to the one or more extendable attachment mechanism, and the one or more extendable attachment mechanisms may be configured to selectively position the one or more monitoring devices at varying distances from a wall of the reactor bay to place the one or monitoring devices in proximity to the reactor module.

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 integral 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 integral pressurizer. The drive shaft passes through a vessel penetration of the pressure vessel that is at least large enough for the impeller to pass through.

Abstract: In a computer-implemented method of designing a nuclear reactor of a given reactor plant, an initial, test reactor core design is generated for the given plant based on a plurality of limits input by a user. The limits include a plurality of transient licensing constraints to be satisfied for operating the given plant. The method includes selecting, from a set of automated tools, one or more automated tools to evaluate the test core design, and operating one of more of selected automated tools to output data for display to the user. The displayed data is related to a core design that satisfies the limits inclusive of the transient licensing constraints.

Abstract: In one embodiment, the in-bundle force measurement device includes an elongated main body, and a sensing structure disposed at a distal end of the main body. The sensing structure is configured to output a measurement indicative of force applied to the sensing structure.

Abstract: The geometric dimensions and shape of a device for removing solid particles from the cooling medium that is circulated in the primary circuit of a nuclear reactor, in particular a boiling water nuclear reactor, are such that the device can be inserted in lieu of a fuel element or fuel assembly into an empty fuel element or assembly position of the reactor core of the nuclear reactor.

Abstract: The present invention provides a method of manufacturing a core shroud for a nuclear plant and a nuclear power plant structure in which a groove portion is easily assembled in the case of manufacturing the core shroud having a weld structure of a nuclear power plant by a laser welding, and it is possible to obtain a weld joint portion in which a plastic distortion region and a residual stress are as small as possible, going with a solidification shrinkage of the weld portion. At a time of welding butted portions of a plurality of members constructing a core shroud, a root face is provided in the butted portion, a length of the root face is set to 25% to 95% of a thickness of the thinner one of the butted portions of a plurality of members, a narrow groove is provided in the other than the root face, and the butted portions are welded by a laser welding using a weld wire.

Abstract: The invention relates to an inexpensively-/easily-decommissionable nuclear power plant, where a nuclear isle of one or more nuclear power-stations is installed in caverns, and further, side by side with them, a center for characterizing, treating and conditioning radioactive wastes and two repositories are installed in suitable caverns, with a final repository being adapted to store low-intermediate level nuclear wastes and a temporary repository being adapted to store spent fuel, high-level long-life radioactive materials and, in case, spare nuclear rods for reactor refueling.

Abstract: The present invention relates generally to nuclear reactors, and more particularly, to nuclear reactors having fuel assemblies that employ support grids. A method of reducing friction and physical contact between a fuel rod and support grid in a nuclear fuel assembly is provided. The method includes applying a lubricant composition to the outer surface of the fuel rod during fuel assembly fabrication and removing the lubricant composition afterward.

Abstract: In a clamping jig and system for inspecting spacer grids for a nuclear fuel assembly, the jig firmly and precisely clamps a variety of grids on a measurement stage moving at high speed so that, among constituent parts of the nuclear fuel assembly, the spacer grids can be inspected using a non-contact measurer. The clamping jig includes a frame having a plurality of holding recesses into which the corresponding spacer grids are inserted, supporting means mounted in each holding recess of the frame to allow each spacer grid to be adjusted from below so it can be maintained in a horizontal state, clamping means installed around each holding recess to clamp two adjacent sides of each spacer grid to a corner of the respective holding recess, and resilient means installed on the sides opposite the clamping means and resiliently supporting the other two sides of each spacer grid.

Abstract: A jet pump sensing line support clamp may be used for sensing line repair, replacement, and damage prevention or reduction. The clamp may affix to jet pump sensing line supports and confine the individual jet pump sensing lines. The clamp may provide for further access or securing of the lines in the support through the clamp. Methods of installing the clamp may include attaching and tightening the clamp against the sensing lines.

Abstract: A reusable up-ending device for loading or unloading nuclear fuel assembly shipping packages. The device has a translating base mounted to a fixed beam structure and two stabilizing arms affixed to a mounting fixture on the base that is secured to the operating floor. The stabilizing arms include a spring dampener and have a simple elbow joint that straightens after the package is up righted. Translation is achieved by means of a hand adjustable lead screw and enables the base to be most advantageously positioned. The shipping package is inserted into pivot mount blocks located at the front of the base and the package is secured at the pivot mount blocks by means of a trunnion and at the support arms by means of a locking pin. Once the package is up-righted the elbows on the support arms are locked in place.

Abstract: Reactor component adapted to be used in fission reactors comprising a core (2) consisting of a first material and a layer (3) consisting of a second material. The layer (3) encloses at least partly the core (2). The reactor component is characterized in that the component (1) comprises and intermediate layer (4) between the core (2) and the layer (3). The intermediate layer (4) has a material gradient that comprises a decrease of the concentration of the first material from the core (2) to the layer (3) and an increase of the concentration of the second material from the core (2) to the layer (3).

Abstract: Provided is a sliding support and a system using the sliding support unit. The sliding support unit may include a fulcrum capture configured to attach to a support flange, a fulcrum support configured to attach to the fulcrum capture, and a baseplate block configured to support the fulcrum support. The system using the sliding support unit may include a pressure vessel, a pedestal bracket, and a plurality of sliding support units.

Abstract: A nut assembly for connecting adjoining poles in a nuclear reactor may include an upper nut and a lower nut. The upper nut may include first threads on an inner surface of the upper nut. The lower nut may include second threads on an inner surface of the lower nut. The lower nut may include third threads on an outer surface of the lower nut. The upper nut may rotate over the lower nut in order to lock the lower nut.

Abstract: A method is provided for removing radioactive internals structural members in the core of a reactor pressure vessel in a containment vessel. The method includes placing a first cask in a first internals assembly, detaching radioactive first internals structural members from second internals structural members in the first internals assembly, placing the detached first internals structural members in the first cask, placing the first internals assembly in a second cask, and removing the second cask containing the first internals assembly and containing the casked detached radioactive first internals members from the containment vessel. The first internal members may be radioactive baffle plates, and the second internals members may be former plates bolted to the radioactive baffle plates. Novel tooling, framework and fixtures facilitate disassembling, moving and storing the radioactive baffle plates.

Abstract: Exemplary embodiments provide automated nuclear fission reactors and methods for their operation. Exemplary embodiments and aspects include, without limitation, re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven surface area/volume ratio neutron absorption, low coolant temperature cores, refueling, and the like.

Abstract: The apparatus for positioning a fuel assembly (10) comprises, on a respective side of the core support plate (19) and the upper core plate (9) directed towards an endpiece (15, 14) of the fuel assembly (10) at least two projecting positioning pins (16, 22) adapted to engage in openings in the lower endpiece (15) or upper endpiece (14). Each of the positioning pins (16, 22) is fixed in a blind opening (24) comprising a threaded bore and a support surface perpendicular to the axis of the blind opening (24) and comprises a flange adapted to come into contact with the support surface when the positioning pin (16, 22) is screwed into the threaded bore. A gudgeon enables the positioning pin to be locked to secure it against unscrewing. The invention also relates to a process and apparatus for replacing damaged positioning pins on the core support plate (19) or on the upper core plate (9).

Abstract: A tie rod having a cruciform cross section is provided with steps for fixing sheaths at tips of cruciform arms of the tie rod; the tips of each of sheaths are fitted onto the steps of the tie rod, each of the sheaths having a U-shaped cross section; and each of the sheaths is fixed to the tie rod by performing a laser welding using a YAG laser beam or a CO2 laser beam with the sheath being fitted onto the tie rod to achieve a continuous weld of at least a part of the tie rod in a longitudinal direction thereof. An axial center position of the beam is shifted from an end face position of the step of the tie rod at least toward an axis center of the tie rod.

Abstract: Methods 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: The maintenance device comprises a robot (2) for carrying out maintenance work in a remote-controlled or programmed way in the maintenance region. The robot (2) comprises a supporting part (2a) connected to a clamping support (3) secured to a means (4) of transporting the maintenance device (1) between an introductory region and the maintenance region. The clamping support (3) comprises clamping shoes (11, 13a, 13b) which can be moved in roughly parallel and opposite directions so as to clamp the maintenance device (1) in the maintenance region. The transport means (4) may be constructed in the form of a carriage comprising two motorized wheelsets. The device and the method of the invention make it possible in particular to carry out machining or inspection work inside primary pipes of a pressurized water nuclear reactor.

Abstract: In order to carry out a radioactivated reactor pressure vessel from a nuclear reactor building of a nuclear power plant, a first opening portion for carrying out the reactor pressure vessel is provided in a roof of the nuclear reactor building. A radiation shield for covering the reactor pressure vessel and shielding radiations is carried into the reactor building through the first opening portion, and installed on a reactor shield wall. A hanger is lowered through the first opening portion and a slit provided on an upper lid of the radiation shield. A portion of a lid of the reactor pressure vessel and a portion of the upper lid of the radiation shield are abutted by hanging up the reactor pressure vessel by the hanger.

Abstract: When a radioactivated reactor pressure vessel is carried out from a nuclear reactor building of a nuclear power plant, a first opening portion for carrying out the reactor pressure vessel is provided in a roof of the nuclear reactor building, a radiation shield for covering the reactor pressure vessel and shielding radiations is carried into the reactor building through the first opening portion, and installed on a reactor shield wall, lowering a hanger through the first opening portion and a slit provided on an upper lid of the radiation shield, a part of a lid of the reactor pressure vessel and a part of the upper lid of the radiation shield are abutted by hanging up the reactor pressure vessel by the hanger, and the reactor pressure vessel and the radiation shield integrated therewith are raised and carried out of the reactor building, whereby the shield can be easily mounted on the large-sized apparatus in a short time and a dose of radiations exposed to a worker when the shield is mounted can be reduced.

Abstract: A new approach to scabbling of surfaces of structural materials is disclosed. A layer of mildly energetic explosive composition is applied to the surface to be scabbled. The explosive composition is then detonated, rubbleizing the surface. Explosive compositions used must sustain a detonation front along the surface to which it is applied and conform closely to the surface being scabbled. Suitable explosive compositions exist which are stable under handling, easy to apply, easy to transport, have limited toxicity, and can be reliably detonated using conventional techniques.

Abstract: A collar clamp assembly is disclosed for repairing and providing a temporary seal around leaking pipe coupling areas of an internal core spray system of a nuclear reactor. The clamp assembly includes first and second housings having respective first and second ends and a pair of longitudinal edges between the ends. The housings have a tongue and groove mating configuration along the longitudinal edges to seal and accommodate variations in the outside diameter of the core spray piping. A plurality of threaded fasteners are used to secure the first and second housings together to form a closed conduit shaped to receive a length of the core spray piping, including the leaking pipe coupling. First and second seals are located at first and second longitudinally spaced positions, respectively, on each of the first and second housings and extend around the closed conduit.

Abstract: A reactor stud cleaning apparatus is presented for cleaning studs, nuts, and washers used to secure reactor heads on nuclear reactors. The cleaning apparatus is a self-contained unit having a housing that includes two independent sealable compartments allowing cleaning of two studs simultaneously. Each sealable compartment has a topside covering with a port that allows reactor studs to be lowered into the apparatus keeping the longitudinal axis of the reactor stud maintained in a substantially vertical position. Inside the sealable compartments is a turn table for vertically mounting a reactor stud and for rotating the stud about its longitudinal axis. Brushes, rotatably mounted inside the compartment, are used to clean the reactor stud. The apparatus also has a cleaning fluid circulation system for circulating solvents or other cleaning agents or fluids over the reactor stud during cleaning.

Abstract: In order to provide an austenitic single crystal stainless steel having preferable stress corrosion cracking resistance, strength, and irradiation induced embrittlement resistance so as to extend the life of a nuclear reactor core structure, which is used under a high radiation dose environment, a method is employed, which comprises the steps of homogeneously dispersing carbides into a parent phase of the austenitic single crystal stainless steel by a two step solution heat treatment, and subsequently effecting an ageing heat treatment after rapid cooling for precipitating fine carbides. Austenitic single crystal stainless steel having preferable stress corrosion cracking resistance, strength, and irradiation induced embrittlement resistance can be provided, and the life of nuclear reactor core structure, which is used under a high radiation dose environment, can be extended.

Abstract: Core spray line riser repair apparatus and methods for axially restraining a reactor core spray line riser to a T-box are described. In one embodiment, the repair apparatus includes a sleeve, a threaded draw bolt extending through the axial bore, and a threaded block. The sleeve includes a substantially cylindrical main body having an axial bore extending therethrough and a flange at one end of the main body. The sleeve is configured to be inserted into the T-box until the flange abuts against the T-box. The draw bolt is configured to extend through the sleeve flange, the sleeve bore, and an opening in the core spray line riser. The draw bolt engages the block to draw the block in tight engagement with the core spray line riser.

Abstract: A method and system for providing nuclear fuel rods with a configuration of isotopic gas tags. The method includes selecting a true location of a first gas tag node, selecting initial locations for the remaining n-1 nodes using target gas tag compositions, generating a set of random gene pools with L nodes, applying a Hopfield network for computing on energy, or cost, for each of the L gene pools and using selected constraints to establish minimum energy states to identify optimal gas tag nodes with each energy compared to a convergence threshold and then upon identifying the gas tag node continuing this procedure until establishing the next gas tag node until all remaining n nodes have been established.

Abstract: A tool for vertically supporting a shroud repair tie rod assembly having a clevis hook by bracing the clevis hook against a clevis pin installed in a gusset plate. This allows the vertical support at the upper end of the tie rod assembly to be removed, so that the upper support structure can be coupled to the tie rod. The lower end of the tie rod assembly is supported on both sides of the gusset plate. The tool for vertically supporting the lower end of the tie rod assembly is designed to fit in the narrow space bounded by a jet pump diffuser, the gusset plate and the reactor pressure vessel. The vertical support tool includes a pole adaptor for coupling to the end of a service pole, a pole adaptor extension having one end connected to the pole adaptor, a mounting block connected to the other end of the pole adaptor extension, and a hydraulic spreader mounted on the mounting block. The hydraulic spreader opens in response to actuation of a hydraulic cylinder.

Abstract: In order to plug a pressure vessel penetration through which a heater was disposed, a stepped diameter hollow cylinder is provided with grafoil seals which are set on an annular shoulder defined by the stepped diameters of the cylinder and compressed against a lower annular edge of a sleeve which is partially machined away. The hollow cylinder is locked in position in the sleeve by a plurality of barb-shaped flanges which are formed at the upper ends of a plurality of flexible finger-like segments defined in the upper end of the hollow cylinder, and which engage the upper end of the sleeve. A rod is disposed in the hollow cylinder and welded in positon to complete the plug.

Abstract: For use in the manufacture of gas tags employed in a gas tagging failure detection system for a nuclear reactor, a plurality of commercial feed gases each having a respective noble gas isotopic composition are blended under computer control to provide various tag gas mixtures having selected isotopic ratios which are optimized for specified defined conditions such as cost. Using a new approach employing a discrete variable structure rather than the known continuous-variable optimization problem, the computer controlled gas tag manufacturing process employs an analytical formalism from condensed matter physics known as stochastic relaxation, which is a special case of simulated annealing, for input feed gas selection.

Abstract: A nuclear fuel channel has four corners and four elongated sidewalls disposed between the corners so as to provide a polygonal transverse cross-section. Each of the channel sidewalls has a longitudinal centerline that coincides with a line of maximum stress that is imposed on the sidewall during in-reactor service. The sidewalls have one or more longitudinal off-center lines that are located between the longitudinal centerline and one of the corners abutting the elongated sidewall. The longitudinal off-center line coincides with a line of minimum stress imposed on an elongated sidewall during in-reactor service. A weld is formed along the length of the off-center line on at least two of the sidewalls. At least one of the off-center line welds is also a seam weld. The disclosed arrangement provides improved resistance to bulging of the channel sidewalls during in-reactor service.

Abstract: Apparatus is provided for simulating a nuclear fuel rod bundle transient condition and includes a vessel for containing a coolant, a pair of heating elements disposed in the vessel and a power supply for supplying power over time to each of the heating elements. The power supply is independently controlled whereby a variable supply of power over time to each heating element is provided, thus simulating the variation in power output and axial flux shape in a nuclear fuel bundle as a function of time.

Abstract: An arrangement of reactor coolant pump seals arranged in a generally annular stack with a plurality of stacking spacers in a test housing for emplacement in a chemical decontamination test loop enables a variety of flow characteristics to be simulated by varying the detailed arrangement of the RCP seals and supporting structure in the test housing. By adjusting parameters such as fluid chemistry, gross flow, pressure, and temperature in a controlled manner with systems incorporated into the test loop, additional information can be obtained regarding the feasibility of any specific decontamination process. The magnitude of process fluid flow across particular surfaces of one or more RCP seal components can be predetermined by selecting the placement, number and dimensions of penetrations in a flow distribution tube through which the process fluid enters the housing.

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: For use in the manufacture of gas tags such as employed in a nuclear reactor gas tagging failure detection system, a method for designing gas tagging compositions utilizes an analytical approach wherein the final composition of a first canister of tag gas as measured by a mass spectrometer is designated as node #1. Lattice locations of tag nodes in multi-dimensional space are then used in calculating the compositions of a node #2 and each subsequent node so as to maximize the distance of each node from any combination of tag components which might be indistinguishable from another tag composition in a reactor fuel assembly. Alternatively, the measured compositions of tag gas numbers 1 and 2 may be used to fix the locations of nodes 1 and 2, with the locations of nodes 3-N then calculated for optimum tag gas composition.

Abstract: A tool for removal of the internal bushings from a piston coupling of a CRD. The tool incorporates a spring-loaded and hardened collet supported on one end of a ram. The collet has a pair of arms with shoulders or projections which latch inside the internal bushing ring to be removed. The tool is inserted into the piston coupling from either end, depending on which internal bushing is to be removed. The tool is inserted until the latching shoulders or projections snap behind the radially inwardly projecting internal bushing, with the contact surfaces of the shoulders or projections in contact with a radial end surface of the bushing. The bushing can then be dislodged, from the inside out, by impacting the other end of the ram with a hammer.

Abstract: Disclosed is a means for increasing coolant flow through the areas adjacent to where the removable top head closure is bolted to the body of pressurized water nuclear reactor. A circumferential plenum for coolant is provided between the upper flange of the core barrel, the inner wall of the pressure vessel body and the upper circular support plate and providing a plurality of orifices for coolant in the upper flange of the core barrel and the upper circular support plate whereby a coolant may be circulated through the core barrel flange, the circumferential plenum and the upper support plate and into the space enclosed by the removable top closure head.

Abstract: A boiling water reactor control rod drive housing support bars having wing assemblies to limit the travel of a control rod in the event that a control rod housing is ruptured. The wing assemblies are placed on support bars located in rows on opposing sides of the control rod drives. Each wing on the wing assemblies supports a quadrant of a control rod drive. The wing assemblies may be rotated from an operation position to a service position to allow replacement of the control rod drives.

Abstract: A boiling water reactor control rod drive housing support bars having radiation shields to limit the travel of a control rod in the event that a control rod housing is ruptured and protect persons working under the reactor vessel from radiation exposure. The radiation shield includes radiation shield cups for surrounding the bottom portion of a control rod drive. The radiation shield cups are placed on support bars located in rows on opposing sides of the control rod drives. The radiation shield may be easily raised and lowered automatically by the control rod drive installation/removal machine.

Abstract: A tool for use in installing O-ring seals in the removable head of a nuclear reactor pressure vessel. Disclosed are two embodiments of the tool. One embodiment is adapted to fit within the stud bolt holes of the flange of the removable head of the pressure vessel. The other embodiment is adapted to be magnetically attached to the underside of the flange of the removable head of the pressure vessel. The tool uses a movable support arm to properly position and support the O-rings in their respective grooves machined in the underside of the flange of the removable head.

Abstract: A device for sealing a penetration in a pressurizer of a nuclear steam supply system includes a hollow cylinder adapted to be inserted into the heater sleeve, in place of the heater. The cylinder includes a nose portion formed with a number of axially extending slots defining therein a number of radially displaceable segments, each of which has an outwardly extending flange, proximate its distal end, which is engageable with the heater sleeve to resist blowout. A rod, inserted into the cylinder, seals the passage therethrough while locking the displaceable nose portion segments in their sleeve-engaging position. An annular soft nickel band on the outside of the cylinder engages the inner surface of the heater sleeve to provide an effective seal therebetween.

Abstract: A latch assembly removably joins a steam separator assembly to a support flange disposed at a top end of a tubular shroud in a nuclear reactor pressure vessel. The assembly includes an annular head having a central portion for supporting the steam separator assembly thereon, and an annular head flange extending around a perimeter thereof for supporting the head to the support flange. A plurality of latches are circumferentially spaced apart around the head flange with each latch having a top end, a latch hook at a bottom end thereof, and a pivot support disposed at an intermediate portion therebetween and pivotally joined to the head flange. The latches are pivoted about the pivot supports for selectively engaging and disengaging the latch hooks with the support flange for fixedly joining the head to the shroud or for allowing removal thereof.