Abstract: The disclosure relates to a device for checking a fuel rod comprising a testing container having first and second chambers, a first checking device arranged in the testing container, wherein the testing container has at least one inlet opening, at least one outlet opening, and an insertion opening for inserting the fuel rod into the second chamber, and wherein a valve is arranged in a connecting channel connecting the first and the second chamber. A method is disclosed for checking a fuel rod in a water-filled basin of a submerged nuclear plant having such a device, wherein the fuel rod is inserted into the second chamber of the testing container through the insertion opening while the valve is closed, wherein a fluid is fed in via the at least one inlet opening, and wherein the valve is opened in order to check the fuel rod with the first checking device.

Abstract: A clamp sealing apparatus configured for preventing leakage from a defective region of a tube is provided. The clamp sealing apparatus includes a first clamping section and second clamping section tangentially connected to each other; and a seal held by the first and second clamping sections. Each of the first and second clamping sections includes a segment of a clamp body and a segment of a compression section. The clamp body and the compression section is configured for axially compressing the seal between the clamp body and the compression section such that the seal is pressed radially against an outer diameter surface of the tube onto the defective region by the axial compression to seal the defective region. A method of sealing a defective region of a tube to prevent fluid leakage through the defective region is also provided.

Abstract: A probabilistic method for determining an operability interval for fasteners in a nuclear power plant assembly is provided.

Abstract: A pressure-relief system for a containment of a nuclear plant has a pressure-relief line which is led through the containment and is closed by a shutoff device, and a wet scrubber being switched into the pressure-relief line lying outside the containment, for the pressure-relief gas flow developing in the pressure-relief operating mode with the shutoff device being open. An effective, reliable operation of the wet scrubber with a compact structural configuration is made possible. This is achieved by a reservoir, arranged in the containment or fluidically connected therewith such that an overpressure, as compared with the outer environment, present in the containment, is transferred to the reservoir, and a feeding line which is led from the reservoir to the wet scrubber and can be closed by a shutoff device, for feeding a liquid active as a scrubbing liquid from the reservoir to the wet scrubber.

Abstract: A sealing device for a jet pump of a boiling water reactor is provided. The jet pump includes an inlet mixer and a diffuser receiving the inlet mixer at a slip joint such that an outer circumferential surface of the inlet mixer is received in an inner circumferential surface of the diffuser at the slip joint. The diffuser includes a plurality of guiding fins, each guiding fin including a radially inner surface, a radially outer surface and lateral surfaces extending radially between the inner and outer surfaces. The sealing device includes a seal configured for sealingly contacting the outer circumferential surface of the inlet mixer and a collar configured for holding the seal against the outer circumferential surface of the inlet mixer. The collar includes portions configured for being received radially between the radially inner surfaces of the guiding fins and the outer circumferential surface of the inlet mixer.

Abstract: The invention is a process of repairing cracked or microstructurally damaged portions of irradiated materials, such as nuclear reactor pressure vessels and shrouds. A damaged portion of the irradiated substrate is first removed, such as by electrical discharge machining (EDM). After removing the damaged portion, the recast layer inherent in the EDM process is then removed. Once the repair area substrate material has been removed to a calculated depth, the created cavity is then filled without releasing transmutated elements within the irradiated material. A chamber may be placed on the irradiated material surrounding the repair area to create an isolated work space.

Abstract: A method for calculating a PCI margin associated with a loading pattern of a nuclear reactor including a core into which fuel assemblies are loaded according to the loading pattern is implemented by an electronic system. The fuel assemblies include fuel rods each including fuel pellets of nuclear fuel and a cladding surrounding the pellets. This method includes calculating (100) a reference principal PCI margin for a reference loading pattern of the fuel assemblies in the core; calculating (110) a reference secondary PCI margin for the reference pattern; calculating (120) a modified secondary PCI margin for a modified loading pattern of the fuel assemblies in the core, and calculating (130) a modified principal PCI margin for the modified pattern, depending on a comparison of the modified secondary PCI margin with the reference secondary PCI margin.

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: A method for conditioning the circulatory system (1) of a power plant is especially suited for a nuclear power plant. An amine, which is a film-forming agent, is metered into the working medium circulating in the circulatory system. The film-forming agent forms a hydrophobic film on the surfaces of the circulatory system. During the process, the concentration of the film-forming agent is monitored in at least one measuring point by way of measurement, and metering of the film-forming agent is stopped once its concentration in the working medium has reached a value of 1 ppm to 2 ppm in at least one measuring point M1.

Abstract: A method and a corresponding device for the pressure relief of a nuclear power plant having an outlet for a relief flow. The relief flow is guided out of a containment into the atmosphere via a relief line provided with a filter system. The filter system has a filter chamber with a filter-chamber inlet and outlet and a sorbent filter arranged therebetween. The relief flow is guided in a high-pressure section of the relief line past the filter chamber, with the latter being heated, and the relief flow is expanded at the end of the high-pressure section and dried. In order for efficient retention of iodine-containing organic compounds, the relief flow is guided through a bed filter, guided in a superheating section past the high-pressure section of the relief line and in the process is heated, guided in this state directly thereafter through the filter chamber having the sorbent filter.

Abstract: A nuclear plant has a containment shell and a pressure relief line passing out of the containment shell and sealed by a shut-off valve, and through which a pressure relief flow can flow during relief operation, such that it is configured for particularly reliable management of critical scenarios where there is a considerable pressure increase within the containment shell at the same time as the release of hydrogen and/or carbon monoxide. A gas flow treatment device is provided upstream from the respective pressure relief line, and contains a flow duct and has a lower inflow opening and an upper inflow/outflow opening. Catalytic elements for eliminating hydrogen and/or carbon monoxide are arranged in the flow duct above the lower inflow opening. During a critical fault, the flow duct is flowed through from bottom to top by a gas mixture present in the containment shell by the principle of natural convection.

Abstract: A pressure-relief system for the containment of a nuclear power facility allows reliable operation of a wet scrubber for the pressure relief flow with a simultaneously compact structural design. The pressure relief system has a pressure relief line guided through the containment and can be closed by a shut-off valve, a wet scrubber arranged in a portion of the pressure relief line located inside the containment, for the pressure relief flow which forms in the pressure-relief mode when the shut-off valve is open, a reservoir arranged inside the containment and is fluidically connected to the remaining inner space of the containment such that any overpressure, with respect to the surroundings outside the containment, prevailing in the containment is transferred at least in part to the reservoir, and a supply line leading from the reservoir to the wet scrubber for supplying the wet scrubber with fluid from the reservoir.

Abstract: A method for treating waste water from the decontamination of a metal surface in a primary coolant circuit of a nuclear reactor comprises discharging a predetermined amount of an oxidation solution from the primary coolant circuit into a reduction zone connected to the primary coolant circuit and reacting the oxidation solution with a reducing agent to form a reaction solution that is freed of oxidizing agent, and passing the reaction solution over an ion-exchange resin in order to form a desalinated solution, and returning the desalinated solution to the primary coolant and/or disposing of the desalinated solution. A waste water treatment apparatus for carrying out the method is also provided.

Abstract: A control rod drive replacement device includes a channel for mounting on a support under a nuclear reactor pressure vessel and a control rod drive extractor removably received in the channel. The control rod drive extractor includes a base and a housing pivotably connected to the base. The housing is configured for receiving a control rod drive. The base includes a base axis. The housing is pivotably attached to the base at the base axis for pivoting between a horizontal orientation in which the housing is aligned within the channel and a vertical orientation in which the housing is aligned for receiving the control rod drive from the nuclear reactor pressure vessel. The housing is vertically movable with respect to the base in the vertical orientation.

Abstract: A process for the recovery of a radioisotope from a waste resin of a nuclear power plant comprises the steps of: a) treating a waste resin loaded with at least one radioisotope with an organic acid or alkaline compound to release the at least one radioisotope and to obtain a process solution containing the at least one radioisotope; b) separating the at least one radioisotope from the process solution through a reaction specific to the radioisotope so as to obtain a treated process solution depleted of the at least one radioisotope, wherein said depleted process solution comprises the organic acid or alkaline compound and optionally a non-reacted radioisotope; c) reacting the organic acid or alkaline compound in the depleted process solution from step b) by thermal and/or photochemical oxidation to form gaseous reaction products; and d) reloading the waste resin with the reacted process solution from step c) to bind the non-reacted radioisotope on the waste resin.

Abstract: A sealing device is provided to form a sealed region about one or more surfaces to be treated. The sealing device has an open end with a rim configured to matingly engage a treatment surface. The sealing device is braced both vertically and laterally, and the sealed region is flooded and pressurized. A peening nozzle and manipulating tooling are positioned within an interior volume of the sealing device. Pressurized fluid is ejected from the nozzle causing the formation of cavitation bubbles. The nozzle flow causes the cavitation bubbles to settle on the surfaces to be treated. The collapsing impact of the cavitation bubbles imparts compressive stress in the materials of the treatment surfaces.

Abstract: A nuclear fuel cladding is provided. The nuclear fuel cladding includes a base cladding; and at least one nanomaterial layer deposited on a surface of the base cladding, the nanomaterial layer having an average grain size of between 5 to 400 nanometers. A method of manufacturing nuclear fuel cladding is also provided. The method includes depositing nanoparticles on a base cladding to form at least one nanomaterial layer, the nanoparticles having an average grain size of between 5 to 400 nanometers.

Abstract: A sealing member is provided to create a sealed region about an annulus formed between an inner body, such as a thermal sleeve, and an outer body, such as a control rod drive housing nozzle. Liquid is introduced into the sealed region to create a flooded region, which is pressurized to a desired level. A nozzle is provided into the flooded region, the nozzle being configured to fit within the annulus. Pressurized fluid is ejected from the nozzle, causing the formation of cavitation bubbles. The nozzle flow causes the cavitation bubbles to settle on the surfaces forming the annulus. The collapsing impact of the cavitation bubbles imparts compressive stress in the materials of the surfaces forming the annulus.

Abstract: A fuel rod for a nuclear fission reactor is disclosed and claimed. The fuel rod includes an elongate hollow cladding configured to retain a nuclear fuel therein. The cladding includes an elongate hollow tube. Fiber layers are positioned around the outside surface of the tube or within the tube forming an integral part thereof. Both the tube and the fibers are formed of a ceramic material. A fuel assembly including a plurality of such fuel rods is also disclosed and claimed.

Abstract: The invention is an innovative design/repair methodology for PWR piping nozzles and vessel nozzles that are attached to the piping/vessel base material with a full penetration weld joint geometry. The development of a robust repair methodology for nozzles of this configuration is necessary due to plant aging, potential material degradation in the original materials of construction, potential increased nondestructive examination requirements, and PWSCC phenomena in the susceptible original materials of construction. The purpose/objective of the repair methodology is to provide a means of partially replacing the existing pressure boundary susceptible materials with PWSCC-resistant materials to facilitate the long-term repair life of the plant. The invention may be applied to a plurality of nozzle, piping, and vessel sizes with a full penetration weld joint.