Abstract: An emission monitoring system for a venting system of a nuclear power plant is configured for low consumption of energy and high reliability. The emission monitoring system has a pressure relief line connected to a containment and contains a high-pressure section, a low-pressure section, and a sampling line. The sampling line opens into the low-pressure section of the pressure relief line and is guided from there to a functional path and through the sampling line steam flows. A jet pump containing a pump fluid connector, a suction connector and an outlet connector is provided. A pump fluid feed line has an inlet side opening into the high-pressure section of the pressure relief line and is guided from there to the jet pump and connected to the pump fluid connector. A sample return line is guided from the functional path to the jet pump and connected to the suction connector.

Abstract: The invention relates to a method and a device for repairing damage to a container wall located under water. A repair device has a carrier having means for fastening the carrier to the container wall, a repair die that has at least one injection opening and at least one outlet opening and can be pressed against the container wall in a remotely controlled manner in order to cover the damage to the container wall and to seal off a chamber formed by the repair die and the container wall, a reservoir for an adhesive, a remotely controllable injection device is arranged on the carrier, which remotely controllable injection device is provided in order to inject adhesive from the reservoir into the injection opening of the repair die so that the chamber formed by the repair die and the container wall is filled by means of the adhesive.

Abstract: A method for the quantitative analysis of the composition of a gas mixture is specifically suited for the analysis of the atmosphere of a containment in a nuclear installation. Even under difficult conditions, this method enables the simple, reliable and as direct as possible determination of physical parameters, thereby permitting the characteristic definition of the ignition field and the combustion behavior of the gas mixture. A measuring device is provided with a thermal conductivity detector with a first measuring bridge, a thermal tonality detector with a second measuring bridge, and a common evaluation unit. With reference to bridge voltages present on the two measuring bridges, the hydrogen content and the oxygen content are simultaneously determined in the evaluation unit.

Abstract: A method involving periodic implementation, during the same operating cycle of the reactor, the method including following steps of: a) calculating, from measurements provided by sensors present in the reactor, the three-dimensional distribution of the local power in the core, b) simulating at least one accidental transient occurrence of power applied to the calculated three-dimensional distribution of local power, c) identifying, using thermomechanical calculations, at least one rod which is the most likely to be subject to a fracture of the cladding thereof during the simulated transient occurrence of power, and d) determining, using thermomechanical calculations on the rod identified, the value of the parameter which is representative of the operability of the reactor.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: The present invention concerns a mosaic module for a control panel of a power plant, the control panel comprising a plurality of mosaic modules being arranged adjacent to each other, each mosaic module being adapted to be fixed to a mechanical support structure, the mosaic module comprising: at least one display device for displaying a function of the power plant and/or an input device, at least one optical power supply input for receiving a power supply for the mosaic module via one or more first optical links, and at least one optical data input for receiving data to be displayed on the display device and/or at least one optical data output to transmit an input command received from the input device.

Abstract: The strip is of the type comprising a wall portion for delimiting a cell with interlaced strips, a spring formed in the strip and provided on the wall portion for biasing a fuel rod extending through the cell away from the wall portion, the spring comprising a cantilevered tab formed in the strip and a contact portion formed at least partially in the tab and protruding from the tab for contacting a fuel rod received in the cell.

Abstract: The invention relates to a method for reducing the radioactive contamination of the surface of a component used in a nuclear reactor, which component is in contact with radioactively contaminated water, in which method a hydrophobic film is produced on the surface of a component by virtue of the surface being wetted with an aqueous solution which contains a film-forming amphiphilic substance.

Abstract: A method for cleaning and conditioning the water-steam circuit of a power plant, especially of a nuclear power plant, uses an amine as a film-forming agent. The amine is metered into the working medium circulating in the water-steam circuit. The film-forming agent forms a hydrophobic film on the surfaces of the circuit. During the process, the concentration of at least one impurity contained in the water and the concentration of the film-forming agent in at least in the feed water of the steam generator are measured and monitored. The concentration of the film-forming agent is varied, as needed, subject to the concentration of the impurity.

Abstract: This invention relates to a control method for a pressurized water nuclear reactor, which comprises a core generating thermal power and means of acquiring magnitudes representative of core operating conditions. The method comprises a step to regulate the temperature of the primary coolant, if the temperature of the primary coolant for a given thermal power is outside a predefined set temperature interval (?TREF) depending on the reactor power. The set temperature interval (?TREF) is characterized by variable amplitude (?T) on a thermal power range between N % and 100% nominal power, where N is between 0 and 100 and comprises a zero amplitude at 100% nominal power, a zero amplitude at N % nominal power.

Abstract: A fuel rod support insert for a nuclear fuel assembly spacer grid, a spacer grid and a nuclear fuel assembly are provided. The support insert is for a nuclear fuel assembly spacer grid comprising interlaced straps defining a lattice of cells for receiving fuel rods, the insert being adapted to be secured to the straps for extending in at least one cell. The insert extends along an axis intended to be parallel to that of a cell and comprises two axially spaced end portions for connecting the insert to the straps and at least one elongated blade-like spring extending axially between the end portions for supporting a fuel rod. The spring has a non-rectilinear cross-section in each plane perpendicular to the insert axis.

Abstract: The invention relates to a device for closing a pipe opening, opening upwards, of a drive casing pipe accommodating the control rod drive of a boiling water reactor, said device comprising a closure plug with a plug housing, opening downwards, a sealing element guided axially so as to be movable within the plug housing, which element has a sealing face providing a sealing contact with an annular counter face defining the pipe opening, a locking element provided to fix the plug housing on the drive casing pipe, and a spring, which is supported by the upper end thereof on the plug housing and by the lower end thereof on the sealing element and holds said element in a lower end position.

Abstract: A method for producing a wear-resistant and corrosion-resistant stainless steel part for a nuclear reactor is provided. This method includes steps of providing a tubular blank in austenitic stainless steel whose carbon content is equal to or lower than 0.03% by weight; shaping the blank; finishing the blank to form the cladding; hardening the outer surface of the cladding by diffusing one or more atomic species; the blank, before the providing step or during the shaping or finishing step, being subjected to at least one hyper quenching with sub-steps of: heating the blank to a sufficient temperature and for a sufficient time to solubilize any precipitates present; quenching the blank at a rate allowing the austenitic structure to be maintained in a metastable state at ambient temperature and free of precipitates.

Abstract: A method for encapsulating a fuel rod or a fuel rod section in a container includes inserting the fuel rod or fuel rod section into the container. One of the ends of the container is connected to a purging-gas line. The container is dehydrated and purged by use of a purging gas. The ends of the container are connected to a bypass line in such a way that a closed gas circuit is produced and a hot gas is circulated in the gas circuit until the absolute moisture content reaches an end value at which the absolute moisture content no longer rises. The container is disconnected from the gas circuit and subsequently the container is closed in a fluid-tight manner at both ends.

Abstract: Position-indexed storage device comprising a support and a basket installed on the support, the basket comprising a plurality of compartments distributed around a longitudinal axis, the basket comprising a central passage with a longitudinal axis, the support comprising a shaft with a longitudinal axis, the shaft being housed in the passage such that relative displacement is possible between the basket and the support along the longitudinal axis, and mechanical means for displacement of the basket relative to the support by a given angle about the longitudinal axis are supported by the internal surface of the passage and the external surface of the shaft, said mechanical displacement means being actuated in each cycle, each cycle comprising a relative separation or approach displacement along the longitudinal axis of the basket and the support.

Abstract: The present invention relates to a method for detecting temporally varying thermomechanical stresses and/or stress gradients over the wall thickness of metal bodies, in particular pipelines. In the method, the temperature on the outer surface of the body is measured in order to determine a temperature progression and stress progression therefrom. In addition, electromagnetic ultrasonic transducers are used at at least one measuring point on the outer surface in order to determine the progression of the stresses and/or stress gradients over time over the wall thickness of the body in conjunction with the result of the temperature measurement. The method allows the fatigue monitoring of pipelines even in the event of rapid stress changes.

Abstract: The present invention relates to a method for controlling a pressurized water reactor (100) comprising the steps that involve measuring the effective power (Pe) of the nuclear reactor; acquiring a reference value for the desired power (Pc); acquiring an estimated duration (DURATION) for the increase in power in order to achieve said reference value of the target power (Pc) desired, said estimated duration (DURATION) corresponding to the time taken for the power to increase from said effective power (Pe) to said reference value for the target power (Pc); determining the reference position (Z) of at least one control rod cluster among said plurality of control rod clusters (40) in order to achieve said reference value for said target power (Pc) desired as a function of said estimated duration (DURATION), of said measured effective power (Pe) and of said reference value for said target power (Pc); monitoring the position of said at least one control rod cluster so as to position it in its reference position (Z).

Abstract: A method for carrying out emergent shaft sealing for a reactor coolant pump and a shaft sealing assembly is provided. The method comprises the steps that a passive stopping sealing member is arranged in a first sealing assembly and forms an opening under the normal operation condition, without affecting the first sealing assembly using a liquid film formed by fluid static pressure to prevent reactor coolant from leaking along a pump shaft; under the station blackout condition, after sensing high-temperature fluid flowing into the first sealing assembly, the passive stopping sealing member is closed and holds the pump shaft of the reactor coolant pump tightly, the gap between the first sealing assembly and the pump shaft of the reactor coolant pump is blocked, and therefore the reactor coolant is prevented from leaking along the pump shaft.

Abstract: A photothermal examination method and corresponding examination unit are provided. In the method first and second zones of the surface to be characterized are scanned simultaneously, and first and second photosensitive surfaces separate from one another acquire images of the infrared radiation emitted by these two zones.

Abstract: A method for thermochemically treating a part while masking a portion and corresponding mask are provided. The method includes the steps of providing a mask comprising a body with a seat, at least a deformable sealing washer located in the seat, and a tightening bushing, the body having a cavity, placing a first portion of the part in the cavity, a second portion of the part being located in a passage in the sealing washer, and a third portion of the part being located outside the mask, moving the tightening bushing to its tightened position so that the sealing washer is deformed and applied against the second portion of the part, and applying a thermochemical treatment to the third portion of the part.