Abstract: A method for replacing a damaged sleeve lining a tube passing through a nuclear reactor pressure vessel. The damaged sleeve has an end including a radially enlarged end portion for resting on a support section of the tube for retaining the damaged sleeve in the tube. The method includes removing the damaged sleeve from a tube; providing a sleeve assembly including a first sleeve with a radially variable end and a retainer; installing the sleeve assembly in the tube so the radially variable end of the first sleeve is received by the support section, the radially variable end being in a radially contracted configuration during installation and being in a radially expanded configuration after the sleeve assembly is installed in the tube; and deforming the retainer from an installation configuration to a retention configuration to retain the radially variable end of the first sleeve in the radially expanded configuration.

Abstract: A method of determination of a nuclear core loading pattern defining the disposition of fuel assemblies. The method includes defining at least one potential core loading pattern and calculating predictive bowing of the fuel assemblies at the end of the operation cycle for each potential core loading pattern. The calculation is carried out by an automatic learning algorithm trained on a training data set that includes a plurality of other core loading patterns. The set also includes, for each of the other core loading patterns, measurements of bowing of fuel assemblies at the end of operation cycle. The method also includes evaluating the at least one potential core loading pattern based on the predictive bowing calculations and at least one predetermined criteria. The method further includes selecting one of the potential core loading patterns based at least in part on the evaluating.

Abstract: A method for providing a neutron radiation shield includes a step (a) of providing a wall having an opening; and a step (b) of at least partially filling a space in the opening with a neutron moderator. Step (b) includes blast-injecting plastic granules into the opening with a blower device to form the neutron moderator.

Abstract: A pressurized water reactor includes a primary reactor coolant circuit flown through by a primary reactor coolant during operation, and a chemical and volume control system for the primary reactor coolant. The chemical and volume control system includes, along the direction of flow of the primary reactor coolant, a letdown line, a high-pressure charging pump with a given discharge pressure, and a charging line leading to the primary reactor coolant circuit. The chemical and volume control system further includes a hydrogenation system with a hydrogen supply and a hydrogen feeding line. In order to achieve efficient and fast hydrogen injection into the primary reactor coolant, a high-pressure feeding pump is arranged in the feeding line to provide a gas pressure higher than the discharge pressure of the charging pump. The feeding line discharges into the charging line.

Abstract: A welding torch (18) for a welding system (10) includes a substantially spherical main body (20) that has an integrated coolant chamber (60) which is in fluid communication with at least one coolant port (62). The substantially spherical main body (20) is a joint head (22) of a ball joint (24) of the welding system (10). A welding system (10) includes a torch bracket (16) and the welding torch (18). The torch bracket (16) has a socket (26, 28) of a ball joint (24), and the socket (26, 28) receives the substantially spherical main body (20) of the welding torch (18) in a spherically orientable manner thereby forming the ball joint (24).

Abstract: A method is for controlling a nuclear power plant comprising a pressurized water nuclear reactor. The method includes determining that an obtained waiting period and/or a remaining waiting period is greater than a first predetermined time allowing raising of a Xenon concentration to maximal value. The method further includes, responsive to the determination, moving one or more control rods out of the reactor core for compensating the reactivity loss due to an increase of the Xenon concentration, and moving the one or more control rods into the reactor core to a control rod setpoint for the start of power ramp up before the end of the obtained waiting period and/or remaining waiting period.

Abstract: A method for optimizing an aviation radiation dose comprises a) gathering flight-relevant data including at least one flight schedule; b) gathering radiation data including at least a current radiation field assigned to the at least one flight schedule, historical radiation data assigned to the flight-relevant data and a radiation dose threshold; c) calculating an expected radiation dose based on the flight-relevant data and the radiation data; and d) modifying the flight-relevant data and repeating steps a) to c) at least once in order to obtain optimized flight-relevant data with regard to the expected radiation dose. With each iteration the flight-relevant data is modified.

Abstract: A BWR fuel assembly is elongated along a fuel assembly axis and comprises a lower tie plate, an upper tie plate axially spaced from the lower tie plate, a bundle of fuel rods extending axially between the lower tie plate and the upper tie plate, and a tubular fuel channel extending from the lower tie plate to the upper tie plate with encasing the fuel rods. The fuel assembly comprises an interaction device mounted on the lower tie plate and configured to interact with the fuel channel. The interaction device has an inactive configuration and an active configuration.

Abstract: A system for sterilizing sterilization units by radiation exposure comprises a conveyance system for transporting sterilization units through a sterilizing environment along a conveying path. The conveyance system comprises at least one lifting beam conveyor with at least one stationary supporting beam and at least one movable lifting beam, which is movable with respect to the at least one stationary supporting beam in a longitudinal and a vertical direction. The at least one stationary supporting beam has a central region between two fixed bearings, which is supported via at least one tensile-loaded tensile element, which is fastened to the central region and to at least one vertical strut, which is arranged in the region of at least one of the fixed bearings, in such a manner that the tensile element extends in a direction diagonal to the longitudinal and the vertical direction.

Abstract: A nuclear power plant includes a nuclear reactor and a reactor coolant circuit containing a reactor coolant. The nuclear power plant further includes a degasification system (2) for the reactor coolant. The degasification system (2) is an ultrasonic degasification system comprising a sonotrode cluster (11) with at least one sonotrode (10) arranged in a line of the reactor coolant circuit or in a line which is fluidically connected to the reactor coolant circuit.

Abstract: A CNC machining device (1) comprises a control console (3) and a CNC machine (5). A training assembly (33) for training operation of the CNC machining device includes a training control console (35) substantially identical to the control console (3) of the CNC machining device (1); a digital twin (36) of the CNC machine (5), comprising a simulator (37) configured to simulate the effect of commands from the training control console (35) on the CNC machine (5); and a display device (39) configured for a trainee to view the current state of the simulator (37).

Abstract: A method for converting uranium hexafluoride to uranium dioxide includes steps of hydrolysis of UF6 to uranium oxyfluoride (UO2F2) in a hydrolysis reactor (4) by reaction between gaseous UF6 and dry water vapour injected into the reactor (4), and pyrohydrolysis of UO2F2 to UO2 in a pyrohydrolysis furnace (6) by reaction of UO2F2 with dry water vapour and hydrogen gas (H2) injected into the furnace (6). The hourly mass flowrate of gaseous UF6 supplied to the reactor (4) is between 75 and 130 kg/h, the hourly mass flowrate of dry water vapour supplied to the reactor (4) for hydrolysis is between 15 and 30 kg/h, and the temperature inside the reactor (4) is between 150 and 250° C.

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: A repairing system (2) is used for applying an adhesive to a location to be repaired underwater in a container. The repairing system (2) comprises a movable and remotely controlled repairing robot (10) that is configured to be driven underwater. The repairing robot (10) comprises a repairing module (12) that is configured to be coupled with an adhesive applying module (34). The adhesive applying module (34) is configured to dispense an adhesive on the location to be repaired. Further, a repairing method is described.

Abstract: A method for protecting a nuclear reactor includes reconstructing a maximum linear power density released among the fuel rods of the nuclear fuel assemblies of the core; calculating the thermomechanical state and the burnup fraction of the rods; calculating a mechanical stress or deformation energy density in the cladding of one of the rods by using the said reconstructed maximum linear power density, the calculated thermomechanical states and the calculated burnup fractions, by means of a meta-model of a thermomechanical code; comparing the calculated mechanical stress or the calculated deformation energy density with a respective threshold; and stopping the nuclear reactor if the calculated mechanical stress or the calculated deformation energy density exceeds the respective threshold.

Abstract: An operating facility (1) for operating a mine by in-situ leaching comprises a supply manifold (9) for supplying leach liquor to the injector well heads (3), a production manifold (15) configured to discharge leachate to a leachate treatment plant (17), a treatment manifold (19), a facility (20) for providing a treatment solution configured to inject the treatment solution into the treatment manifold (19) and/or into the supply manifold (9); and for each producing well, a production pipe (21) equipped with a production cut-off device (25) and a treatment pipe (23) equipped with a treatment cut-off device (27), fluidically connecting the corresponding producing well head (5) in parallel, respectively, to the production manifold (15) and to the treatment manifold (19).

Abstract: A system (1) includes a module (30) for calculating a design of experiment comprising a plurality of software tasks to be performed in order to solve a predetermined physical problem. The software tasks of the design of experiment have a first priority level. The system also includes a module (4) for scheduling the execution of software tasks by the computing system, configured to check for the presence of at least one software task with a second priority level higher than the first priority level waiting to be executed; in case of the presence of at least one such software task, obtaining freed-up computational resources for executing said at least one software task of a second priority level; in the absence of at least one such software task, allocate at least some of the software tasks of the first priority level to the available computational resources.

Abstract: A tube inspection unit (10) includes an eddy current probe (14). The probe (14) has a plurality of inductors and a plurality of receivers. The receivers are magnetoresistances having a substantially linear functional zone. The inductors of the probe (14) are linked to a controller (26). The controller (26) is programmed to inject, into the inductors, a voltage with a sinusoidal component and a nonzero direct-current component, such that the receivers have a polarization center situated inside the substantially linear functional zone.

Abstract: A method for conditioning of spent ion exchange resins from nuclear facilities comprises the steps of: mixing the spent ion exchange resins with water to form a reaction mixture; setting and controlling the pH of the reaction mixture in a range from 1.0 to 3.5, preferably in a range from 2.0 to 3.0; adding an oxidant to the reaction mixture, with the temperature of the reaction mixture maintained at 90°° C. or less so that the spent ion exchange resin and the oxidant react with each other to form an aqueous reaction solution comprising the organic reaction products of the spent ion exchange resin; and electrochemically oxidizing the organic reaction products, wherein carbon dioxide is produced and a carbon-depleted aqueous reaction solution having a TOC (total organic carbon) value of less than 50 ppm is obtained. Furthermore, an apparatus for the conditioning of spent ion exchange resins from nuclear facilities is described.

Abstract: A nuclear fuel assembly for a boiling water reactor extends along a fuel assembly axis and includes a base including a lower tie plate, a head including an upper tie plate and a lift handle, a bundle of fuel rods extending axially between the lower tie plate and the upper tie plate, and a water channel extending within the bundle of fuel rods with axially connecting the base to the head such that the load of the base is transferred to the head via the water channel. The fuel assembly further comprises a tie rod extending between the base and the head. The tie rod is axially fixed to the base and connected to the head via a connection assembly comprising a stopping member configured to abut an abutting surface of the head for limiting a downward movement of the base relative to the head during lifting of the fuel assembly, in case of a breakage of the water channel.