Abstract: A method of jacketing a uranium slug to an aluminum container comprising applying a coating to the exterior of the container, the coating consisting of colloidal graphite in water, permitting the coating to dry, applying an alloy of aluminum and silicon to the interior surface of the container at a temperature between 588° C. and 594° C., inserting the slug into the container in complete contact with the alloy, and quenching the assembly.

Abstract: A method for providing a leak-tight metal enclosure to a fuel matrix penetrated by coolant channels, wherein the mutually contacting surfaces of said metal enclosure and said fuel matrix are metallurgically bonded, comprising placing a metal cladding about the lateral surface of said fuel matrix; disposing metal coolant tubes within said coolant channels; placing a perforated header plate having tubular extensions at each end of the fuel matrix from which the coolant tube ends protrude, said coolant tubes passing through said perforated header plate and said tubular extensions and terminating even with the ends of said extensions; welding, under vacuum, said cladding to said header plates, and the ends of said coolant tubes to the ends of said tubular extensions; exposing the assembly comprising the fuel matrix and enclosure to a gas at high temperature and pressure; and machining said header plates to provide a finished fuel element.

Abstract: A method of jacketing a body of fissionable material within a nonfissionable jacket having a cup-shaped body open at one end and a cap effecting closure at said end comprising the steps of inserting the cup-shaped body into a tightly fitting cup-shaped sleeve; completely submerging said assembly in a bonding bath of a molten metallic bonding material, allowing the inner body to fill with molten bonding material; dipping the fissionable body into the open end of said assembly while under the surface of the bonding bath; closing the open end of the assembly with the cap while under the surface of the bonding bath; removing the complete assembly from the bonding bath; quenching in cold water; and removing the sleeve.

Abstract: A holder for fuel elements in a reactor, in particular in a boiling water reactor, is provided, as well as a method for repairing such a holder. A core grid is demounted from a core shroud and replaced by a new forged core grid for the purpose of repairing the holder. An adapter ring is to be inserted between the new core grid and the core shroud. There is provision for connecting the adapter ring to the core grid without a welding operation, for example by shrink fitting.

Abstract: In a fuel transport container, basket has a basket hole including at least two adjacent inner side walls to be fit to a fuel assembly for housing the fuel assembly element in the basket hole. The fuel assembly element is directly housed in the basket hole of the basket. The fuel assembly housed in the basket hole is slid by bias means toward the two adjacent inner side walls thereof without pushing the fuel assembly thereby biasing the housed fuel assembly so that the fuel assembly element is in contact with the two adjacent inner side walls of the basket hole.

Abstract: In a transport container, at least one fuel assembly containing element including at least one fuel assembly is housed in a container having an inner surface portion to be fit to the at least one fuel assembly containing element. The inner surface portion has a predetermined shape substantially corresponding to a fit portion of the at least one fuel assembly containing element. The at least one fuel assembly containing element is pushed by a support against the inner surface portion of the container along a fixed support direction. Therefore, the fit portion of the at least one fuel assembly containing element is fit to the inner surface portion of the container so that the at least one fuel assembly containing element is fixedly supported to the container.

Abstract: In a fuel protective apparatus having a fuel assembly, the fuel protective apparatus is hoisted up by using a hoisting accessory. The fuel protective apparatus includes a fuel protective container for housing the fuel assembly. The fuel protective container has an opening portion to which the hoisting accessory is accessed. An attachment jig member is fixedly secured to the fuel protective container so as to be engaged to the hoisting accessory accessed to the opening portion.

Abstract: The spacer envelope measurement system includes a generally U-shaped fixture having support plates extensible into positions overlying the spacer for supporting the fixture from the spacer. Extensible heads are carried by the fixture for engaging the spacer, followed by the heads of measurement gauges whereby, upon extending the heads into clamping engagement with the spacer along opposite sides thereof, the measurement gauges accurately determine the lateral dimensions of the irradiated spacer. A spacer bar clamp member is pivotal between positions opening the fixture opposite the base leg of the U-shaped fixture, enabling the bundle to reside within the U-shaped opening and a second position closing the opening for aligning the fixture and the bundle. The measurements are taken under water in a nuclear fuel service pool.

Abstract: The invention provides a method for making Zr alloy nuclear reactor fuel cladding having excellent corrosion resistance and creep properties. The method includes performing hot forging, solution heat treatment, hot extruding, and repeated cycles of annealing and cold rolling of a Zr alloy including, by weight, 0.2 to 1.7% Sn, 0.18 to 0.6% Fe, 0.07 to 0.4% Cr and 0.05 to 1.0% Nb, with the remainder being Zr and incidental impurities, and the incidental nitrogen impurity content being 60 ppm or less, and then performing final stress relief annealing thereon. The annealing is performed at a temperature of 550.degree. C. to 850.degree. C. for 1 to 4 hours such that the accumulated annealing parameter .SIGMA.Ai=.SIGMA.ti.multidot.exp(-40,000/Ti) satisfies relationships -20.ltoreq.log.SIGMA.Ai.ltoreq.-15, and -18-10.multidot.X.sub.Nb .ltoreq.log.SIGMA.Ai.ltoreq.-15-3.75.multidot.(X.sub.Nb -0.

Abstract: In a transport container, at least one fuel assembly is housed in a container having an inner surface portion to be fit to the at least one fuel assembly. The inner surface portion has a predetermined shape substantially corresponding to a fit portion of the at least one fuel assembly. The at least one fuel assembly is pushed against the inner surface portion of the container along a fixedly support direction. Therefore, the fit portion of the at least one fuel assembly is fit to the inner surface portion of the container so that the at least one fuel assembly is fixedly supported to the container.

Abstract: A pyrolytic carbon coated nuclear fuel particle and method of making it. The fuel particle has a core composed of a refractory compound of an actinide metal. The pyrolytic carbon coating surrounds the core so as to provide a void volume therebetween. The coating has an initial density of no greater than 1.45 grams/cm.sup.3 and an anisotropy factor than 3.0 and a final density upon heat treatment above about 2000.degree. C. of greater than 1.7 grams/cm.sup.3 and an anisotropy factor greater than 5.

Abstract: A storage rack for nuclear fuel assemblies and its manufacturing process.A storage rack for nuclear fuel assemblies is formed by assembling main plates (12), spacer plates (14), two end plates (16) and a base plate (18), preferably by welding. Once a first main plate (12(1)), a first end plate (16(1)) and the base plate (18) have been mounted on a tooling, a first series of housings (10) is formed by securing a first row of spacer plates (14(1)) to the first main plate (12(1)), and then mounting a second main plate (12(2)). The other rows of housings (10) are obtained in the same manner. The rack is completed by mounting the second end plate.

Abstract: A nuclear reactor plant includes a reactor pressure vessel having a core barrel supported on a cylindrical holding structure fixed in the reactor pressure vessel. The core barrel has a lower part with at least one elastically resilient segment. A projection disposed on the segment engages under the holding structure in such a way that a wedge-like gap is formed between a lower surface of the holding structure and the projection. A wedge is braced in the gap. A method is also provided for mounting the core barrel.

Abstract: A fuel rod for a light water reactor comprises a cladding tube which comprises a zirconium alloy having a composition including 0.6 to 2.0% by weight of Nb, 0.5 to 1.5% by weight of Sn, 0.05 to 0.3% by weight of Fe, and the balance being Zr and incidental impurities; uranium oxide fuel pellets packed in the cladding tube; and end plugs comprising a zirconium alloy and closing both ends of the cladding tube. The cladding tube is sealed by TIG welding with the end plugs. Grain boundaries in each heat affected zone of the cladding tube, which are adjacent to a bead formed by TIG welding, have structural compositions including 4 to 30% by weight of Nb, and 0.9 to 20% by weight of Fe.

Abstract: The invention is directed to a high-density dispersion nuclear fuel having spherical particles of an uranium alloy dispersed in a nonfissionable matrix. The alloy includes uranium and 4-9 wt % Q, wherein Q is selected from the group consisting of Mo, Nb, and Zr. A process of manufacturing the spherical particles is also disclosed.

Abstract: A method of producing mixed oxide fuel pellets. The method uses oxides of at least two fissile elements and includes a mixing, milling, spheroidising and sintering step to produce a fuel pellet incorporating a neutron poison.

Abstract: A process for fabricating weapons-grade plutonium into mixed-oxide fuel for use in a nuclear reactor. The plutonium is converted into plutonia powder at a site remote from the fuel fabrication facility and then the plutonia powder is pressed and fired into high-density plutonia pellets for transport to the fuel fabrication facility. The high-density plutonia pellets serve as a diversion-resistant and environmentally sound method of shipping the plutonia for use as a feed material. Comminution methods are employed to reduce the plutonia pellets to a fine powder for blending with urania feed to produce mixed oxide acceptable for reactor operation.

Abstract: Apparatus for rectifying the tendency of a fuel assembly (1) to incline from the perpendicular has a top nozzle support (19) and a base assembly (18) for supporting, respectively, a top nozzle (7) and a bottom nozzle (2) of the fuel assembly. Incorporated in the base assembly is an elevating device (78, 79) which is operable to raise the fuel assembly. When raised, a grinding assembly (62) in the base assembly is able to remove material from the lower surfaces of the bottom nozzle so that they lie in a horizontal plane. The elevating device comprises a gimbal mounting arrangement which includes a gimbal (94) movably supported on a gimbal mount (88) by means of cooperating male and female spherically-shaped bearing surfaces. The grinding assembly includes a grinding wheel (66) which is mounted on a rotor (43) arranged for rotation about an axis along which the elevating device raises the fuel assembly so that material can be removed from each of the lower surfaces during one revolution of the rotor.

Abstract: A fuel rod for a light water reactor comprises a cladding tube which comprises a zirconium alloy having a composition including 0.6 to 2.0% by weight of Nb, 0.5 to 1.5% by weight of Sn, 0.05 to 0.3% by weight of Fe, and the balance being Zr and incidental impurities; uranium oxide fuel pellets packed in the cladding tube; and end plugs closing both ends of the cladding tube. The cladding tube is sealed by TIG welding with the end plugs. Precipitates having grain diameters of 0.01 to 0.5 .mu.m and comprise intermetallic compounds containing Zr, Nb and Fe are present at grain boundaries in the structure of heat affected zones of the cladding tube, the heat affected zone being adjacent to a bead formed by TIG welding.

Abstract: A fuel assembly locator for guiding and inserting nuclear fuel assemblies into a grid plate of a reactor vessel during reload. In the preferred embodiment, the fuel assembly locator includes a plurality of generally trapezoidal guides each having a multiplicity of serially increasing tiered slope regions for gradually directing the fuel assembly precisely into a grid plate opening. The increasing slope regions allow greater positional tolerance between the exterior of the fuel assembly and the interior opening of the grid plate opening initially, thus requiring less time and precision to align the fuel assembly with respect to the grid plate opening. The tolerance then gradually decreases since the slope of each successive tier is greater than the previous tier, thus precisely directing the fuel assembly into alignment with the grid plate opening for insertion therein.

Abstract: A capsule containing a tag gas in a zircaloy body portion having a hemispical top curved toward the bottom of the body portion. The hemispherical top has a rupturable portion upon exposure to elevated gas pressure and the capsule is positioned within a fuel element in a nuclear reactor.

Abstract: An apparatus for handling nuclear fuel rods, which include a mixed oxide fuel material, and for constructing fuel assemblies from the fuel rods, includes a store (9) constructed from radiation shielding material. The store includes a number of channels (10), each of which is adapted to receive a magazine (2) containing fuel rods. Removal and insertion of magazines from and into the channels is effected by a handling machine (29) which is remotely and automatically operated. Each of the channels is closed by a plug (11) which can be removed and inserted by the handling machine. The handling machine is able to remove a magazine from a selected channel and to deliver the magazine to either of two fuel assembly lines (16, 21). Each assembly line constructs a different type of fuel assembly from that of the other assembly line, and each assembly line is arranged in a separate shielded enclosure (15, 24).

Abstract: An extraction tool for removing a partial length fuel rod from a fuel bundle assembly, wherein the partial length fuel rod has an upper end plug, the tool comprising an extension rod having a pair of gripper jaws mounted at one end thereof, the gripper jaws shaped and sized to engage the upper end plug; and a locking tube slidably received over the extension rod and engageable with the gripper jaws to lock the gripper jaws to the upper end plug.

Abstract: End plugs for nuclear fuel rods are provided with a spot weld along the interior end face of the end plugs to seal any streamer openings or passages along the axis of the end plugs, thereby maintaining the fuel rods sealed at its opposite ends. The spot welds are applied by an automated laser welding system in which pallets of end plugs are supported on X-Y positioning tables below a laser movable in a vertical direction. By successively locating the end plugs in the pallet below the laser, the spot welds are formed along the axis of the end plugs and along the interior faces thereof sealing the end plugs.

Abstract: The installation includes at least one loading line comprising a framework support, a magazine for storing fuel rods and a traction bench in an arrangement aligned along the longitudinal direction of the housings of the framework of the assembly. The traction bench includes a support having a fixed part and a movable part to be positioned in the loading line in longitudinal alignment with the framework support and the storage magazine, a carriage mounted for movement on the support in the longitudinal direction, a plurality of rod grippers, linkage between the grippers and the carriage, an arrangement selectively controlling displacement or retention of the grippers, a selection unit for selective activation of a subset of grippers and controls for gripping the grippers.

Abstract: A grid keying fixture for installing grid keys in a nuclear fuel assembly spacer grid having a matrix of cells, at least some of the cells including a soft stop in a first axis and a soft stop in a second axis. The grid keying fixture includes a frame assembly having a plurality of fixture pegs affixed to the frame assembly and each arranged and adapted to enter a respective cell of the grid. Each of the fixture pegs is operable to simultaneously deflect the first axis soft stop and the second axis soft stop of the respective cell without bearing on a hard stop in the respective cell opposite one of the first axis and second axis soft stops. In the preferred embodiment, alignment pegs are affixed to the frame assembly for aligning the fixture pegs with the respective cells.

Abstract: A dry transfer system for spent nuclear fuel (SNF) assemblies includes a transfer container with a sliding sleeve for vertical translation therein. The sliding sleeve includes a number of compartments for receiving a corresponding number of fuel assemblies. The container includes an integral and remotely controllable hoist with a number of individually actuated grapples for latching onto a corresponding number of fuel assemblies. The system further includes a loading stand with an elevator for raising and lowering a fuel basket that also includes a number of compartments for moving fuel assemblies up to and in alignment with the sleeve of the transfer container that is landed on the loading stand.

Abstract: A component maintenance system having a network linked to a data acquisition system, a data analysis system, a historical file of component design and inspection information, and a diagnostic system having the capability to perform structural analysis of discontinuities identified by the data acquisition system. The network nodes may be located apart from each other; for example in the maintenance of a nuclear power plant steam generator, the data acquisition system may be located in a high radiation area while the data analysis system is safely located away from the radiation areas. Furthermore, the diagnostic system and the file of component design and inspection history information may be centrally located and shared to support component evaluations being conducted in numerous geographic locations. Network communication facilitates the rapid evaluation of discontinuities, thereby permitting the inspection plan to be modified as the inspection results are obtained.

Abstract: The method serves to manufacture tubes for constituting sheaths for nuclear fuel rods. A bar is made out of a zirconium-based alloy containing 50 ppm to 250 ppm iron, 0.8% to 1.3% by weight niobium, less than 1600 ppm oxygen, less than 200 ppm carbon, and less than 120 ppm silicon. The bar is heated to a temperature in the range 1000.degree. C. to 1200.degree. C. and is quenched in water. A blank is extruded after heating to a temperature in the range 600.degree. C. to 800.degree. C. and cold-rolled in at least four passes in order to obtain a tube, with intermediate heat treatment being performed between passes at temperatures in the range 560.degree. C. to 620.degree. C. A final heat treatment is performed at a temperature in the range 560.degree. C. to 620.degree. C., all of the heat treatments being performed under an inert atmosphere or a vacuum.

Abstract: A nuclear power plant fuel transfer tube quick operating remote closure device (10) comprised of a reducing ring (20) and a cover assembly (60). The reducing ring (20) is bolted to the transfer tube housing flange (12). Rotation of a chuck key advances jaws on the cover assembly (60) which feed closure latches (78) into a groove (40) on the reducing ring (20). The upper surface (80) of each latch (78) has a bevelled surface (82) corresponding to a bevelled surface (42) of the groove (40). As the latches (78) engage the groove bevelled surface (42), the entire cover assembly (60) is uniformly drawn toward the reducing ring (20). Machined seal faces (13, 30, 44, 68) and O rings (36, 74) provide a fluid and air tight seal between the transfer tube housing flange (12) and the sealing ring (20) and between the cover assembly (60) and the sealing ring (20).

Abstract: A method of repairing a nuclear fuel rod assembly having a damaged fuel rod and/or a damaged spacer cell.

Abstract: The guiding device comprises a body, the size of which corresponds to that of a fuel assembly, intended to rest on the lower core plate (20), equipped with holding means intended to be grasped by an assembly-handling machine, extended downwards by means (18) for positioning it on the lower core plate at an assembly location adjacent to the one intended to receive the assembly to be guided, and two wings (30) carried by the bottom part of the body, forming an angle of 90.degree. in a horizontal plane, each one having an inclined guide face (32) ending at the bottom above the location intended to receive the assembly when the device is in place. Each wing (30) can be retracted, starting from a deployed state, towards a position in which it is withdrawn within the horizontal size of the body.

Abstract: The girth welds joining the end plugs and the hollow tubes of nuclear fuel rods are inspected automatically using a technique that averages reflectance values, compares the reflectance values to standards defined as proportions of the average, and counts adjacent pixels outside the standards to analyze for defects exceeding a minimum defect size. The minimum defect size is checked by counting the adjacent pixels in mutually perpendicular directions, such as rows and columns in the collected matrix of pixel data. The maximum count in the two directions can be different, for profiling the maximum acceptable defect as to the direction of its extension relative to the tube and/or weld. Tubes are inspected prior to surface treatments that may conceal defects.

Abstract: The lower part (31) of the pole (10) carries an expansible bush (40) engaged and fixed on the outer surface of the part (31) by one of the ends (41) of the bush and comprising longitudinally extending slots defining resiliently bendable strips. A group of balls (39) mounted to be radially movable in openings (38) in the lower part (31) of the pole (10) and bearing against the bendable strips of the expansion bush are movable between a retracted position and a position for expanding the bush (40), by a cylindrical cam (36) connected to an actuating rod (20). The device permits inserting the locking sleeve (8) in the guide tube (13) without risk of deformation of the end strips of the guide tube (13).

Abstract: A method for retrofitting an irradiated nuclear fuel assembly including removing a portion of the inlet nozzle and installing a debris filter.

Abstract: A fuel rod for a nuclear reactor includes a metal cladding tube filled with nuclear fuel and having ends, an outer surface, and a longitudinal axis. A metal seal plug is welded to one of the ends of the cladding tube at a transition point, defining an annular bead on the outer surface of the cladding tube at the transition point. The annular bead has a cylindrical outer jacket surface with jacket lines being substantially parallel to the longitudinal axis of the cladding tube. The annular bead has material being formed of the metal of the cladding tube and the metal of the seal plug. A welding apparatus for producing a fuel rod includes an electrode having a through bore formed therein for receiving one end of a cladding tube. A counter electrode is displaceable relative to the electrode for holding a seal plug to be welded the one end of the cladding tube. The electrode has a cylindrical step formed therein at an end of the through bore facing toward the counter electrode.

Abstract: A skid for transporting a nuclear fuel transportation cask. The skid comprises a supporting member having parallel spaced-apart plates. The plates are aligned perpendicular to a longitudinal axis of the cask, include a semi-circular trough for mating with the cask, and are connected by longitudinal fins parallel to the longitudinal axis of the cask. The skid also comprises a retaining member including parallel spaced-apart plates. The plates of the retaining member are aligned perpendicular to a longitudinal axis of the cask, include a trough for mating with the cask, and are connected by longitudinal fins parallel to the longitudinal axis of the cask. The fins of the supporting member are spaced apart such that when the cask rests in the trough of the supporting member, the fins of the supporting member are aligned with elongate members of the cask neutron radiation shielding material to transfer the load between the cask and the skid.

Abstract: A transport, storage, monitoring, and retrieval system for casks including intermediate storage of spent fuel rods from nuclear power plants. The system consists of a powered rail transport car which carries a cask standing on its end. A cross-ways traveling powered rail "under car" unit carries the rail transport car and loaded cask transversely on rails underneath to align with a bay rail which the powered rail car traverses to the cask storage location. Bridge cranes off-load the casks and with monitoring equipment perform remote cask inspections. Bay rails and a pair of cross-ways traveling "under car" units serve for movement of both the transport car and the bridge units. Casks are located automatically at specific addresses for inventory control and data logging of radioactive emissions. The system may operate remotely and cask conditions are monitored from automated retrieved collected at an off-site data base.

Abstract: An end plug is inserted into the end of a nuclear fuel pin cladding tube after loading mixed oxide fuel pellets into the tube from a containment area. During the pellet loading operation, a disposable sleeve and an end plug carrier protect the external surfaces of the cladding tube and the end plug from contamination. The sleeve, mounted on the end of the cladding tube, is inserted into a seal arranged in a wall of the containment area. The end plug, located within a recess in the plug carrier, is then inserted into the end of the cladding tube by sliding the carrier through the sleeve. Upon removal of the cladding tube, the sleeve and the carrier are retained in the seal for ejection into the containment area during the next pellet loading operation.

Abstract: A method of estimating the crystal grain size of a nuclear fuel pellet. A plurality types of UO.sub.2 powders are heated at a predetermined temperature raising speed in dry air so as to measure weight change ratios occurring due to the oxidation of each of the UO.sub.2 powders. From a value at which the weight change ratio changes, a temperature at which the composition of the powder from the UO.sub.2+x phase to the U.sub.3 O.sub.7 phase is determined for each type of the powders. Sintered pellets are produced from the plurality types of UO.sub.2 powders in which the arrival temperatures are known. Crystal grain sizes of the plurality types of the sintered pellets produced are measured. According to the U.sub.3 O.sub.7 phase arrival temperature determined and the crystal grain size of the sintered pellet measured, a correlation between the both is recognized. A U.sub.3 O.sub.7 phase arrival temperature of a UO.sub.2 powder of a test sample is determined under the same conditions as described above.

Abstract: A grapple is provided for use with a fuel bundle channel of a nuclear reactor wherein the channel is provided with a pair of gussets in opposite corners of an upper end of the channel, each gusset having a hole therein. The grapple includes a pair of lever arms, each provided with a lifting foot at a lower end of the respective lever arm. The lever arms are movable in opposite directions to a channel lifting position wherein the channel gripper of each lever arm is located under a respective one of the pair of channel gussets. The channel lifting foot of each lever arm includes a pair of laterally spaced lifting elements such that the lifting feet may engage the underside of the gusset on either side of the gusset hole. In addition, each channel lifting foot includes a secondary foot or slider captured between the pair of lifting elements.

Abstract: An apparatus is presented for removing the keys smoothly and reliably from support grids of a fuel assembly. The apparatus performs the task of removing the keys, which had been inserted into the grid through the opening sections of the grid, and enables mechanization and automation of the key removal operation. The apparatus has a robotic arm including a pair of parallel grid support members freely movably disposed at the end of the robotic arm for clamping onto the side edges of the grid. The apparatus further includes a key rotation device disposed on the support arm for engaging with and rotating the keys; a key moving device disposed on the grid support members so as to hold the keys and move the keys in the direction of the key axis.

Abstract: An apparatus is presented for removing the inner and outer keys from a support grid of a fuel assembly automatically or by mechanical methods. The apparatus removes a line of keys from a loaded assembly inserted into a grid through an opening section of the grid to deflect the springs. Both ends of the keys are engaged with the engaging members, and one set of keys are rotated in one direction while the other set of keys are rotated in the opposite direction. The springs are then released to press on the fuel rods firmly in place between the springs and the dimples in the grid cell. The keys are removed from the grid by clamping or holding the keys and moving the keys together with the clamping/holding device in the key axis direction by operating the driving device until the keys are removed completely from the grid.

Abstract: A process of preparing a fuel pellet for a nuclear reactor, comprising washing the gel particle using an organic solvent miscible with water to substitute the organic solvent for the water, removing the organic solvent, moistening again the dry gel particle, followed by press molding and sintering.

Abstract: An apparatus for loading nuclear fuel rods through spacer grid cells of a fuel bundle includes a handle having at least one elongate clip joined thereto, and a flexible tubular sleeve joined to the clip. The clip rigidly supports the sleeve along its longitudinal axis while allowing the sleeve to unfurl perpendicularly thereto for disassembly from a fuel rod. In several embodiments of the invention, the clip may include a slot for bridging adjacent grid cells during assembly, or the clip may be removably fixedly joined to the handle for allowing easy replacement of the sleeves.

Abstract: A method is provided for retrofitting nuclear fuel assemblies in bottom entry fuel assembly type nuclear reactors having a transition piece to guide liquid coolant into a channel entry area. The transition piece is disposed above an entry orifice in a fuel assembly support plate. The transition piece has side walls and a multi-pronged lead-in finger. The method comprises the steps of removing the lead-in finger from the transition piece, machining internal surfaces of the side walls of the transition piece and installing an insert to be disposed within and supported by the machined transition piece.

Abstract: A key member, a method for insertion and/or removal of the fuel rods in a nuclear fuel assembly using the key member and a method of disassembling the nuclear fuel assembly using the same key member are disclosed. The key member has first projections and second projections formed on the opposite faces of an elongated key body. In the insertion or removal of the fuel rods or in the disassembling the nuclear fuel assembly, the key member is inserted into the grid and rotated to bring the first projections and the second projections into engagement with the straps of the grids and the springs on the straps to thereby the springs are deflected in a direction away from the dimples opposing to the springs. In this situation, the insertion and removal of the fuel rods can be carried out.

Abstract: The present invention provides improved nuclear fuel pellets having high thermal conductivity for use in an LWR. This can be achieved by creating a continuous deposition phase of high-thermal conductivity substances in the grain boundaries in the pellets. As a result, the temperature in the center of the fuel rod can be significantly reduced, and the discharge amount of gases generated on the nuclear fission can be efficiently reduced.The present invention also provides a method of manufacturing the above-described nuclear fuel pellets.

Abstract: A storage rack for storing spent nuclear fuel assemblies using a minimum space of a storage pool. The storage rack includes a plurality of tubular cells with a plurality of moderating members of neutron-absorbing material coupled thereto, and a base member coupled to the tubular cells for supporting the tubular cells substantially perpendicular thereto. The tubular cells are constructed of a plurality of prebent panels forming side portions of the cells. At least one of the prebent panels form at least two of the side portions of at least two of the tubular cells which are diagonally arranged relative to each other.

Abstract: In a method of assembling a nuclear fuel assembly, a deflecting jig is inserted into grid cells in each of a plurality of grids. The diameter of the deflecting jig is enlarged to urge a spring of at least one pair of dimples and spring associated with the grid cell to deflect the springs away from the dimple. A plurality of elongated key members are inserted into the grid cells through a plurality of openings defined at intersections between the straps forming walls of the grid cells. Each key member is rotated about its axis to cause hooks of the key member to project from a wall surface of the strap in a direction opposite to the projecting direction of the springs. The key member is then moved in a direction to engage the hooks with the wall surface of the strap. Urging of the spring by the deflecting jig is released to allow the same to be withdrawn from the grid cells and, subsequently, the fuel rods are inserted into the respective grid cells.