Abstract: A modular submersible repairing system includes a working unit having a tool module capable of repairing structures in a reactor. A scanning/pitching module is capable of being selectively connected to or disconnected from the tool module, and is provided with a scanning/pitching shaft for scanning or pitching the tool module. A submersible fan module is capable of being selectively connected to or disconnected from the scanning/pitching module. A first buoyant module keeps an orientation of the tool module. A base unit includes a manipulator module internally provided with an actuator driving mechanism. An adsorbing module is capable of being detachably mounted on the manipulator module and a wall. A second buoyant module keeps the orientation of the manipulator module. The scanning/pitching module and the manipulator module are provided with a submersible connecting device capable of being operated in water for engagement and disengagement.

Abstract: The invention relates to a guide tube especially for an instrumentation lance extending into a reactor pressure vessel. The guide tube includes a lower tube section and an upper tube section that extends into the interior of the reactor pressure vessel. In order to prevent radioactively contaminated particles from reaching the lower tube section disposed outside the reactor pressure vessel from forming a source of radiation, a separator is arranged in the upper tube section. The separator has a separation chamber in which the particles are deposited and are removed from the water.

Abstract: An improved method of increasing the power output of an existing nuclear power plant includes increasing the thermal power output of the plant's nuclear island and constructing of an auxiliary BOP to handle the increased thermal power. The thermal power of the nuclear island can be increased such as by increasing the thermal power of the plant's reactor, by replacing the plant's steam generator with one that is more efficient, and by increasing the flow rate and/or change in temperature of a coolant in a secondary cooling loop of the plant. The thermal power of the reactor can be increased such as by replacing existing cylindrical fuel rods with fuel rods having a relatively greater surface area to volume ratio and/or by increasing the flow rate and/or the change in temperature of a coolant of a primary cooling loop. The auxiliary BOP can be constructed while the plant is in operation, and can then be connected with the nuclear island during a maintenance operation on the reactor.

Abstract: A method and apparatus is disclosed for smoothly inserting a fuel rod loader through the skeleton of a fuel cell and then for pulling an attached fuel rod into the fuel rod skeleton while keeping the fuel rod under constant tension and preventing the bowing of the fuel rod. The pulling mechanism comprises a smooth collet with a smooth bullet end cap for guiding the puller mechanism into and through the skeleton where the bullet end cap is removed and the fuel rod is attached with a unique ball and slide mechanism actuated by a long puller rod that grabs the end of the fuel rod by a forward motion of the puller rod and is easily removed from the fuel rod by reversing this motion once the fuel rod is in place.

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: Object of this invention is to provide an axial torsional vibration measurement device of the reactor internal pump capable of measuring the axial torsional vibration easily. The vibration detector 2 that detects vibration in the radial direction, and the sound detector 3 that detects vibration in the circumferential direction are fitted on cylindrical motor casing 6, which contains the motor of a reactor internal pump. The dimension of both signals of a vibration waveform signal detected by the vibration detector 2, and a sound waveform signal detected by the sound detector 3 are matched in vibration calculating device 15, to thereby obtain an axial torsion vibration of a motor from the difference of both signals.

Abstract: After a nuclear reactor has been stopped and allowed to cool down following a period of operation, action is taken in a pool of the nuclear reactor on at least one guide tube of the activated upper internal equipment of the nuclear reactor to enlarge (21b, 21c) guide channels (16b, 16c) of the guide tube (8) in at least one continuous guidance zone. The enlargement is preferably performed by machining discontinuous cavities along the length of the guide channel (16b, 16c). Enlargement is preferably performed on the guide tube (8) in place within the upper internal equipment that has been placed on a storage stand in the reactor pool.

Abstract: According to the method, a reactor core is shut down during the operation cycle. One or more fuel bundles of the reactor core are then moved to new positions within the reactor core to increase a total energy, output of the reactor core as compared to continuing operation of the reactor core without the shutting down and moving steps.

Abstract: A modular reactor servicing platform for a nuclear reactor is provided. The modular servicing platform includes a plurality of modular sections coupled together, a support structure coupled to the coupled together modular sections, a center access opening, and a refuel access channel extending from the center access opening to an outer edge of the platform forming a substantially C-shaped servicing platform. Each modular section includes a plurality of interconnecting beams, and at least one floor panel attached to and covering the beams.

Abstract: A reactor servicing platform for a nuclear reactor is provided. The nuclear reactor includes a reactor pressure vessel positioned in a primary containment and at least one refuel bridge. The primary containment includes a refueling floor. The servicing platform, in an exemplary embodiment, includes a frame, a support structure, and a floor covering the frame. The floor includes a reactor access opening sized to permit access to the reactor pressure vessel.

Abstract: A structure in a reactor building to be transported is made small when the structure is carried out from the reactor building and/or when the structure is carried into the reactor building. A beam is used outside a building inside of which a reactor pressure vessel is installed. The beam passes above the reactor pressure vessel and crosses over the building. A lifting machine has a lifting device and is moved on the beam. A shield is provided for shielding radioactive rays radiated from the reactor pressure vessel into the building through an opening portion provided in a roof of the building. The reactor pressure vessel, is moved together with the shield upward through the opening portion using the lifting device.

Abstract: The method facilitates sealing a control rod drive (CRD) housing to a stub tube in a reactor pressure vessel of a nuclear reactor, if any cracks develop proximate a heat affected zone of the stub tube at a CRD housing re-attachment weld. The method includes cutting the CRD housing to separate an upper portion of the CRD housing from a lower portion of the CRD housing. The lower portion remains in place and in substantial alignment with the upper portion of the CRD housing. A lower portion member, which may be the lower portion that was cut, is re-attached to the reactor vessel by a re-attachment weld at a location below potential differential leakage paths along the CRD housing. The heat affected zone (e.g., at the re-attachment weld) is covered with a corrosion resistant material to prevent future leaking, and to protect the weld from inter-granular stress corrosion cracking (IGSCC).

Abstract: The present invention relates to a zirconium alloy having excellent corrosion resistance and mechanical properties and a method for preparing a nuclear fuel cladding tube by zirconium alloy. More particulary, the present invention is directed to a zirconium alloy comprising Zr-aNb-bSn-cFe-dCr-eCu (a=0.05-0.4 wt %, b=0.3-0.7 wt %, c=0.1-0.4 wt %, d=0-0.2 wt % and e=0.01-0.2 wt %, provided that Nb+Sn=0.35-1.0 wt %), and to a method for preparing a zirconium alloy nuclear fuel cladding tube, comprising melting a metal mixture comprising of the zirconium and alloying elements to obtain ingot, forging the ingot at &bgr; phase range, &bgr;-quenching the forged ingot at 1015-1075° C., hot-working the quenched ingot at 600-650° C., cold-working the hot-worked ingot in three to five passes, with intermediate vacuum annealing and final vacuum annealing the worked ingot at 460-540° C.

Abstract: An electric discharge machining sampling apparatus for obtaining samples from a surface of metal components provided. In an exemplary embodiment, the sampling apparatus is operable underwater and includes a base plate and an electrode assembly movably coupled to the base plate. The electrode assembly includes an electric discharge electrode and an electrode holder with the electrode including at least one bore extending therethrough. The sampling apparatus also includes a particle collection assembly operatively coupled to the electrode. Each electrode bore is in flow communication with the particle collection assembly.

Abstract: A method for grouting a decommissioned nuclear reactor containing residual water. A first, water-absorbent grout is used to grout the bottom portion of the reactor vessel. The water-absorbent grout is formed of cement and a water-absorbent particulate material, such as expanded perlite, exfoliated vermiculite or granulated bentonite. Excess water is absorbed by the particulate material, so that the residual water is incorporated in the fill material when it sets. A set retarder may be included to extend the time for uptake of residual water by the particulate material. Foam may also be included to reduce density and increase pumpability. The remaining, upper portion of the reactor vessel is grouted using lightweight cellular cement grout, lacking the water-absorbent particulate material, so that total weight of the fill remains low enough that the grouted vessel can be removed using the facility's installed crane system.

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: A system and method for moving a canister of spent nuclear fuel from a first location, which may be within a nuclear power generation facility to a second location such as a storage cask that is used for short-term or long-term storage of spent nuclear fuel includes a first lifting mechanism for engaging a transfer cask and a second lifting mechanism for engaging and lifting the canister of spent nuclear fuel. Preferably, the method is practiced by first positioning a canister of spent nuclear fuel within the transfer cask and then engaging the transfer cask with the first lifting mechanism. The canister is engaged with the second lifting mechanism. The transfer cask having the canister positioned within is moved to the vicinity of a storage cask, and the canister is lowered from the transfer cask into the storage cask by the second lifting mechanism without disengagement of the first lifting mechanism from the transfer cask.

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: When carrying the large structure out through an aperture portion provided at a roof of a reactor building, a chamber for covering the aperture is provided at the roof so that work of providing the chamber for covering the aperture portion can be performed independent of other work performed in the reactor building. At least a part of the work of using a roof crane, and at least a part of the work of handling fuel assemblies in the reactor building can be performed in parallel with forming the roof aperture.

Abstract: A positioning apparatus for a CRD handling assembly for a nuclear reactor. The positioning apparatus includes at least one linear slide rail, a drive screw coupled to the at least one slide rail, an elevator movably coupled to the drive screw, and at least one linear bearing attached to the elevator. The linear bearing is slidably coupled to the at least one slide rail.

Abstract: Plant parameters and performance of an industrial plant such as a boiling water nuclear reactor plant can be assessed over a global network. Plant parameter and performance data are compiled from a plurality of operating plants, and this data is stored in a categorized historical database. Users are provided access to the historical database over the global network to perform self-assessments and generate comparison graphics showing percentile ranking, trending, etc. Data generated at a particular plant can be downloaded to a server site for more detailed engineering analyses using conventional methods. Reports and commentary on the analyses can be returned to the user via the global network.

Abstract: Object of this invention is to provide an axial torsional vibration measurement device of the reactor internal pump capable of measuring the axial torsional vibration easily. The vibration detector 2 that detects vibration in the radial direction, and the sound detector 3 that detects vibration in the circumferential direction are fitted on cylindrical motor casing 6, which contains the motor of a reactor internal pump. The dimension of both signals of a vibration waveform signal detected by the vibration detector 2, and a sound waveform signal detected by the sound detector 3 are matched in vibration calculating device 15, to thereby obtain an axial torsion vibration of a motor from the difference of both signals.

Abstract: A method for manufacturing a control rod of a nuclear reactor, the control rod comprising a blade including neutron absorbers and a sheath, a tie rod for fixing the blade, a handle and a lower blade fixed to the tie rod and the blade. The method comprises a first step for cutting weep holes for water to cool the neutron absorbers in the sheath and cutting a periphery of the sheath, a second step for bending the sheath cut in the first step to a C-shape, a third step for inserting the neutron absorbers in a bent portion of the sheath formed by bending in the second step, a fourth step for successively welding the blade to the tie rod, the blade to the handle and the blade to the lower blade, and a fifth step for finishing portions welded in the fourth step.

Abstract: A method of carrying equipment out of a nuclear reactor building that, in carrying equipment rendered radioactive out of a nuclear reactor building, makes it possible to shorten the operating time and reduce the number of casks to be used, wherein the structures of a nuclear power plant that are outside the equipment installed in the nuclear power plant are removed and these removed structures are then stored in the space in the equipment and carried out of the building.

Abstract: A method and system utilizing polarized ultrasonic shear waves and other guided waves to detect and characterize flaws oriented parallel to the wave propagation direction and perpendicular to the wave particle motion. As the wave passes, the waves are dampened by these flaws causing a reduction in the received signal amplitude as well as other changes in the signal. This inspection can be applied to nuclear reactor vessel control rod drive mechanisms (CRDMs) and other tubular and plate products. In addition, the same waves are utilized to detect and characterize other types of flaws in CRDMs.

Abstract: One of methods for carrying out a reactor vessel according to the present invention comprises the steps of removing an overhead traveling crane in a reactor containment vessel of a pressurized water reactor, or the steps of, in an area where an overhead traveling crane is installed, operating the overhead traveling crane to move aside for creating a space, through which the reactor vessel is able to pass, and then carrying out the reactor vessel through an opening provided in a top portion of the reactor containment vessel. With the present method, the reactor vessel of the pressurized water reactor can be carried out in a short time with high efficiency.

Abstract: A device and method for removing radioactive crud material deposited on surfaces of a nuclear fuel assembly in a nuclear plant is provided. The device comprises a container arranged to accommodate the nuclear fuel assembly, a first pumping means to pump a fluid, composed of water and ice, through the container in order to release the deposit from the nuclear fuel assembly by abrasion and to transport the released radioactive material out from the container. The first pump means also enables the feeding of a gas or steam or both to the fluid that is fed into and through the container in order to create a turbulence in the fluid. The device further comprises means, arranged in a &ggr;-radiation-dampening medium, disposed to receive the radioactive deposit transported out of the container by the fluid.

Abstract: A modular submersible repairing system includes a working unit having a tool module capable of repairing structures in a reactor. A scanning/pitching module is capable of being selectively connected to or disconnected from the tool module, and is provided with a scanning/pitching shaft for scanning or pitching the tool module. A submersible fan module is capable of being selectively connected to or disconnected from the scanning/pitching module. A first buoyant module keeps an orientation of the tool module. A base unit includes a manipulator module internally provided with an actuator driving mechanism. An adsorbing module is capable of being detachably mounted on the manipulator module and a wall. A second buoyant module keeps the orientation of the manipulator module. The scanning/pitching module and the manipulator module are provided with a submersible connecting device capable of being operated in water for engagement and disengagement.

Abstract: A fuel channel box manufacturing method processes a fuel channel box of a zirconium-base alloy by a beta-quench treatment that heats the fuel channel box by a heating coil. The distance between the heating coil and the opposite walls of the fuel channel box is controlled so that temperature differences between the opposite walls of the fuel channel box are reduced. The fuel channel box of a zirconium-base alloy is heated at a temperature in a temperature range including &bgr;-phase temperatures so that temperature difference between the opposite walls is 50° C. or below. The fuel channel box manufacturing method is capable of manufacturing a fuel channel box that is not subject to significant irradiated bow even if a deflection is produced therein originally when the same is manufactured and of manufacturing the fuel channel box at a high production efficiency.

Abstract: An electric discharge machining sampling apparatus for obtaining samples from a surface of metal components provided. In an exemplary embodiment, the sampling apparatus is operable underwater and includes a base plate and an electrode assembly movably coupled to the base plate. The electrode assembly includes an electric discharge electrode and an electrode holder with the electrode including at least one bore extending therethrough. The sampling apparatus also includes a particle collection assembly operatively coupled to the electrode. Each electrode bore is in flow communication with the particle collection assembly.

Abstract: The invention relates to nuclear engineering and more particularly to a method and apparatus for inspecting and examining a surface inside experimental channels of nuclear and thermonuclear plants, providing examination of the surface both in the course of and after irradiation, the surface being examined in different operation modes of a probe scanning microscope by selecting a probe and a respective mode; the obtained images of different characteristics of the surface are analyzed to judge the degree and dynamics of irradiation effect on the examined surface from the analysis results, wherein the probe scanning microscope comprises a computer with a control console, and a measuring head of the probe scanning microscope, the measuring head including a cylinder housing (1) with dogs (4), a scanning unit, an electronics unit (3), a data path (2), a video camera (5), lights (6) and a retaining mechanism.

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

Abstract: A clamp for a feedwater sparger end bracket assembly in a boiling water reactor vessel comprises first and second clamp members disposed in opposition over a feedwater sparger end bracket assembly and a connector connecting the first and second clamp members to one another over the feedwater sparger end bracket assembly. Each clamp member has a compartment for receiving a portion of the feedwater sparger end bracket assembly, and the clamp constrains the feedwater sparger end bracket assembly against separation in first, second and third directions. Each clamp member includes shoulders and tabs which place the clamp and the feedwater sparger in equilibrium.

Abstract: The core of a pressurized water reactor is assembled in such a way that if the fuel elements on the edge of the core are bent, these fuel elements are oriented such that the bending of the fuel elements points outwards in a convex manner. When the reactor is in operation, forces arise which increase the size of small gaps between the fuel elements at the expense of greater gaps and counteract the bending effect of the fuel elements.

Abstract: The method facilitates sealing a control rod drive (CRD) housing to a stub tube in a reactor pressure vessel of a nuclear reactor, if any cracks develop proximate a heat affected zone of the stub tube at a CRD housing re-attachment weld. The method includes cutting the CRD housing to separate an upper portion of the CRD housing from a lower portion of the CRD housing. The lower portion remains in place and in substantial alignment with the upper portion of the CRD housing. A lower portion member, which may be the lower portion that was cut, is re-attached to the reactor vessel by a re-attachment weld at a location below potential differential leakage paths along the CRD housing. The heat affected zone (e.g., at the re-attachment weld) is covered with a corrosion resistant material to prevent future leaking, and to protect the weld from inter-granular stress corrosion cracking (IGSCC).

Abstract: The basket comprises a bundle of tubes (10) connected to each other by an assembly structure (14) so as to form a regular network of compartments (16). Cross pieces (12) are placed between the tubes (10) so as to define a second wall approximately continuous around each compartment (16) surrounding the first wall materialized by tubes (10) and at least partially in contact with this wall. Thus, baskets with compartments with arbitrarily shaped cross sections, and particularly square, rectangular, hexagonal or circular cross sections, can be made for a limited cost.

Abstract: Nuclear fuels are taken from a nuclear reactor pressure vessel, a rail for running is mounted under the reactor pressure vessel drain piping in parallel with a horizontal portion of the reactor pressure vessel drain piping inside a pedestal positioned on the lower side of the reactor pressure vessel, an inspection apparatus is mounted on the-rail, and the inspection apparatus is moved along the rail to inspect the reactor pressure vessel drain piping, whereby the reactor pressure vessel drain piping can be surely inspected while reducing radiation exposure to an inspector.

Abstract: A modular reactor servicing platform for a nuclear reactor is provided. The modular servicing platform includes a plurality of modular sections coupled together, a support structure coupled to the coupled together modular sections, a center access opening, and a refuel access channel extending from the center access opening to an outer edge of the platform forming a substantially C-shaped servicing platform. Each modular section includes a plurality of interconnecting beams, and at least one floor panel attached to and covering the beams.

Abstract: A stud enclosure for protecting a stud extending upwardly from a nuclear reactor pressure vessel (RPV) flange has a cylindrical can with a capped end and an open end. The capped end has an axially extending hole with a screw extending therein for fastening the stud enclosure to the stud. A seal ring is disposed adjacent the open end of the cylindrical can for sealing the ring on the RPV flange. A gas valve is disposed in the capped end of the cylindrical can for pressurizing the interior portion of the can with air. The stud enclosure is used to protect the RPV studs after the RPV head has been removed in order to permit access to the interior portions of the RPV.

Abstract: The invention relates to a method of repairing the labyrinth seal of a diffuser in a primary pump, the diffuser being of the type comprising a peripheral skirt and a circular end wall having an orifice provided with the labyrinth seal. The method consists in making a bore in the circular end wall having a diameter greater than that of the orifice, in reducing the thickness of the remaining portion of the end wall, in forming a chamfer on the edge of the bore, in positioning a replacement part in said bore, the replacement part including a central orifice and a chamfer on its own peripheral edge, in making a bead of welding between the chamfers, and in making a new labyrinth seal in the central orifice of the replacement part.

Abstract: A method for manufacturing a control rod of a nuclear reactor, the control rod having a blade including neutron absorbers and a sheath, a tie rod for fixing the blade, a handle and a lower-blade fixed to the tie rod and the blade. The method includes a first step for cutting weep holes for water to cool the neutron absorbers in the sheath and cutting a periphery of the sheath, a second step for bending the sheath cut in the first step to a C-shape, a third step for inserting the neutron absorbers in a bent portion of the sheath formed by bending in the second step, a fourth step for successively welding the blade to the tie rod, the blade to the handle and the blade to the lower-blade, and a fifth step for finishing portions welded in the fourth step.

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: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: An automated machine for loading nuclear reactor fuel, BP, or APSR pellets from small trays having rows of pellets into reactor rods of long length wherein the pellets are individually counted and loaded into rows of a loading tray according to the desired number of pellets per row and total number of pellets per tray with the loading tray controls being moved to a que tray where they may be weighted and there from to an unloading tray for sequential row loading into the reactor rod.

Abstract: A head assembly of a reactor pressure vessel in an ice condenser plant is retrofitted to permit the assembly to be removed during a later refueling operation in reduced time and with reduced exposure to radiation by operating personnel. The portions of the ductwork ventilation system that originally provided cooling air to CRDMs are removed below the seismic support platform. The original CRDM cooling shroud that surrounds the lower portion of the CRDM assemblies is extended to the seismic support platform. A plenum is mounted on the seismic support platform and in air flow communication with interior portion of the extended CRDM cooling shroud. The plenum fits under the missile shield. Spool pieces are connected between the plenum and the portion of the ductwork adjacent to the seismic support platform.

Abstract: An opening portion 3 is produced by cutting a recirculating piping 72 connected to a recirculating nozzle 2 of a reactor pressure vessel 1. A repairing device is brought into an annulus 10 of the reactor pressure vessel 1 from the opening 3, thereby repairing the annulus 10.

Abstract: In an exemplary embodiment, an electric discharge machining sample removal apparatus for removing a material sample from a structural component in a boiling water nuclear reactor includes a base plate, a motor mounted on the base plate, and an electrode assembly rotatably coupled to the base plate and operatively coupled to the motor. The electrode assembly includes an outer wall defining a substantially semi-cylindrical hollow cavity. The outer wall includes a conductive first arcuate portion a conductive second arcuate portion and a non-conductive third arcuate portion positioned between and coupled to the first and second arcuate portions.

Abstract: A head assembly of a reactor pressure vessel in an ice condenser plant is retrofitted to permit the assembly to be removed during a later refueling operation in reduced time and with reduced exposure to radiation by operating personnel. The portions of the ductwork ventilation system that originally provided cooling air to CRDMs are removed below the seismic support platform. The original CRDM cooling shroud that surrounds the lower portion of the CRDM assemblies is extended to the seismic support platform. A plenum is mounted on the seismic support platform and in air flow communication with interior portion of the extended CRDM cooling shroud. The plenum fits under the missile shield. Spool pieces are connected between the plenum and the portion of the ductwork adjacent to the seismic support platform.

Abstract: A method of carrying out a reactor vessel according to the present invention includes removing an overhead traveling crane in a reactor containment vessel of a pressurized water reactor. Alternatively, in an area where an overhead traveling crane is installed, operating the overhead traveling crane to move aside for creating a space, through which the reactor vessel is able to pass. Then the reactor vessel is carried out through an opening provided in a top portion of the reactor containment vessel. With the present method, the reactor vessel of the pressurized water reactor can be carried out in a short period of time with high efficiency.