A method for checking the functional equivalence of two pieces of software for control systems, in particular control systems based on programmable logic controllers, engineered from function block diagrams (120, 196) with a plurality of interconnected function blocks, the method comprising the following steps: a) generating a first instance of compilable source code from a first plurality of function block diagrams (70, 74, 78) by virtue of a first code generator obeying to a first set of sequentialization rules; b) parsing said first instance of source code and identifying the set of all function block I/O ports and junction points (nodes) and the set of all connections between them (arcs) of the data flow; c) using the information from step b) to reconstruct a first data flow graph (86) of said first instance of source code in the upstream direction; d) generating a second instance of compilable source code from a second plurality of function block diagrams (90, 94, 98) by virtue of the same or a second code generator obeying a second set of sequentialization rules; e) parsing said second instance of source code and identifying the set of all function block I/O ports and junction points (nodes) and the set of all connections between them (arcs) of the data flow; f) using the information from step e) to reconstruct a second data flow graph (106) of said second instance of source code in the upstream direction; g) comparing said first identified data flow graph (86) and said second identified data flow graph (106) with each other, node by node, and thereby checking if the transitive closure of input up to that node in said second instance of source code is the same as in said first instance of source code.
The invention relates to a fan arrangement (2) with a plurality of ventilators (4) which are arranged parallel to one another and are designed to generate an air flow along a common main flow direction (6), wherein • at least one flow channel (10) is paired with each ventilator (4), • each flow channel (10) is provided with a flap (14) which can be pivoted between an open position and a closed position, and • each flap (14) is constructed and arranged in the flow channel (10) so as to be supported by the air flow in the open position and so as to be brought into the closed position by a reverse flow directed opposite the air flow. The aim of the invention is to reliably prevent a flow short-circuit in the event that individual ventilators (4) malfunction and to ensure a natural draft convection in the event that all ventilators (4) malfunction. According to the invention, this is achieved in that each flap (14) is constructed so as to be brought into the open position by virtue of its inherent weight in the absence of a flow.
The invention relates to a seismic sensor unit (4) for an emergency shutdown system (2) of a nuclear power plant, comprising - at least one accelerometer (6) for at least one axis, - a trigger module (8) which processes a signal derived from a measured signal of said accelerometer (6) and generates an output signal from which a trip signal is derived. It is an objective of the present invention to provide a reliable, robust, and non- complex seismic sensor unit (4) of such a type. The proposed solution according to the invention is characterized in that said trigger module (8) comprises an analog peak-hold-circuit (18) processing an input signal (X) and generating an output signal (Y) with the following properties: - if the input signal (X) is rising and the level of the output signal (Y) is below the level of the input signal (X), the output signal (Y) increases almost in- stantly with the input signal (X), - if the level of the input signal (X) is below the level of the output signal (Y), the output signal (Y) essentially keeps the current level unless a reset condition is reached.
A circuit arrangement, in particular for a safety I&C system (4) of a nuclear power plant (6), is to keep the proven diagram-centric project-specific engineering approach known from CPU-based systems while reaping the benefits of FPGA technology. To this end, the circuit arrangement according to the invention comprises: • a generic FPGA (18) with a number of logic blocks (20), and • at least one dedicated PLD (22) which operates as an application-specific switch-matrix for said logic blocks (20).
H03K 19/177 - Logic circuits, i.e. having at least two inputs acting on one outputInverting circuits using specified components using elementary logic circuits as components arranged in matrix form
A process of incorporating technetium into an electroless deposit, forming an alloy that is extremely resistant to corrosion and reduces the mobility of technetium on a geologic time scale is disclosed and claimed. The process includes providing a liquid containing technetium, such as an aqueous waste stream generated during the used nuclear fuel reprocessing activities. The technetium is collected and concentrated, and provided into an electroless deposition bath. A substrate, such as suitably prepared zero valent iron or stainless steel, is introduced into the bath to initiate autocatalytic electroless deposition of the technetium onto the substrate due to the difference in electrochemical potential between the plating bath and the metals in solution. This causes a layer of technetium metal to form on the substrate. The electroless deposition is continued until virtually all of the technetium has been removed from the bath, and then continues to build a layer of technetium-free material on the outermost surface of the substrate. One or more additional deposition steps may be performed to armor the plated substrate against leaching/corrosion in a nuclear waste disposal facility.
C23C 18/16 - Chemical coating by decomposition of either liquid compounds or solutions of the coating forming compounds, without leaving reaction products of surface material in the coatingContact plating by reduction or substitution, i.e. electroless plating
C23C 18/34 - Coating with one of iron, cobalt or nickelCoating with mixtures of phosphorus or boron with one of these metals using reducing agents
This patent discloses and claims a system for providing in-plane stabilization to the tubes of a heat exchanger. The system increases the friction between the tubes and the existing anti-vibration bars by pressing the tube bundle together in the out-of-plane direction. The invention involves a structure that develops forces by reacting either off the inner diameter of the tube bundle wrapper, or off an assembled stressor hoop, which is self-supporting on top of the tube bundle, or involves purposely deforming connecting hoops to bias the anti-vibration bars in a way that the tube U-bend is no longer planar. It imposes a controlled clamping and/or deforming force on the tube bundle, either by direct contact with the outer tubes or indirectly via the existing anti -vibration bar structure.
The inventive tool is designed to ensure proper plant operation position of the core shroud head bolt assemblies of a boiling water nuclear reactor. The tool exerts an upward pulling force on stuck retainers to return them to the locking position around the nut. The tool includes a frame, an engagement member, and a collar. The engagement member is lowered toward and end of the bolt until it comes into contact therewith. Further movement of the engagement member causes the frame and collar to move upward, lifting the retainer and extending the spring.
The invention provides at least two electrochemical sensors positioned in a nuclear reactor or in a system adjacent to the nuclear reactor, wherein at least one of the at least two electrochemical sensors has a heated zirconium electrode, and the at least two electrochemical sensors produce voltages proportional to an electrochemical corrosion potential for a surface that each of the at least two electrochemical sensors are installed upon. The invention also provides an arrangement configured to accept the voltages produced by the at least two electrochemical sensors, wherein the arrangement is configured to determine an electrochemical corrosion potential of a zirconium fuel rod in the nuclear reactor based upon the voltages of the at least two electrochemical sensors.
A method for retrofitting a boiling water reactor is provided. The method includes removing a mixing chamber from a slip joint defined by a diffuser and the mixing chamber, the mixing chamber having an inner surface and a bottom edge directing flow to the diffuser such that a recirculation zone at an entrance to the slip joint creates a diverging effective path for the leakage flow entering the slip joint. The method also includes providing a new inner surface and new bottom edge, the new inner surface and the new bottom edge being reshaped to decrease the size of the recirculation zone. A jet pump is also provided.
The invention relates to a position measuring system (1) for detecting an extremal position (Xmin, Xmax) of a guide rod (3) which extends along a path (x) that is rectilinear relative to the guide system (2) and can be moved along the rectilinear path (x), wherein said system operates particularly reliably, and requires a minimum of electrical lines. The position measuring system (1) is equipped with a plurality of sensor elements (5) and at least one magnetic element (7), wherein the magnetic element (7) is designed to form a magnetic field (H), the magnetic element (7) is connected to the guide rod (3), the or each sensor element (5) is designed to detect a magnetic field (H) within a detection region, the field strength of said magnetic field being greater at the location of the sensor element (5) than a predefined threshold value, and at least one sensor element (5) is connected to the guide system (2) and is arranged in an environment (Ux) of the path (x). The invention further relates to a corresponding measuring method.
G01D 5/20 - Mechanical means for transferring the output of a sensing memberMeans for converting the output of a sensing member to another variable where the form or nature of the sensing member does not constrain the means for convertingTransducers not specially adapted for a specific variable using electric or magnetic means influencing the magnitude of a current or voltage by varying inductance, e.g. by a movable armature
G01F 23/74 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by floats of the free float type using magnetically actuated indicating means for sensing changes in level only at discrete points
The present invention provides a method and an apparatus for calibrating a first self-powered neutron detector for long term use in a nuclear reactor core with a second self-powered neutron detector, where the emitter material of the second self-powered neutron detector has a neutron absorption cross-section that is greater than the neutron absorption cross-section of the first emitter material for the first self-powered neutron detector.
The present invention provides a method and an apparatus for calibrating a first self-powered neutron detector for long term use in a nuclear reactor core with a second self-powered neutron detector, where the emitter material of the second self-powered neutron detector has a neutron absorption cross-section that is greater than the neutron absorption cross-section of the first emitter material for the first self-powered neutron detector.
The invention relates to a concrete component (50) having a front face (52), in particular of an installation found in power plant and industrial construction. The component should have a versatile arrangement for load attachment, which can be set up with minimal effort. According to the invention said aim is achieved by an arrangement for load attachment comprising a plurality of anchors (2), wherein the anchors (2) are arranged on the front face (52) in a regular grid covering substantially the whole area and are embedded in the concrete component (50).
The invention relates to a method and corresponding device for depressurizing a nuclear power plant (2) having a containment shell (4) for containing activity carriers and having an outlet (10, 10') for a depressurization flow, wherein the depressurization flow is conducted out of the containment shell (4) into the atmosphere via a depressurization line (12, 12') provided with a filter system, wherein the filter system comprises a filter chamber (16) having a filter chamber inlet (124), a filter chamber outlet (128), and a sorbent filter (18) lying therebetween, and wherein the depressurization flow is first conducted in a high-pressure section (70), then is depressurized by means of expansion at a throttle device (72), then conducted at least partially through the filter chamber (16) having the sorbent filter (18), and finally blown out into the atmosphere. In order to enable an especially efficient and effective retention of activity carriers contained in the depressurization flow, in particular organic compounds containing iodine, the depressurization flow depressurized by the throttle device (72) is conducted through a superheating section (80) immediately before the depressurization flow enters the filter chamber (16), in which superheating section the depressurization flow is heated by direct or indirect heat transfer from the depressurization flow not yet depressurized in the high-pressure section (70) to a temperature that is at least 10 °C, preferably 20 °C to 50 °C, above the dew point temperature present there.
The invention relates to a component (20a-e) for conducting or receiving a fluid, in particular a component (20a-e) of a fluid-conducting line system (20) of an industrial facility, in particular of a line system of a tertiary cooling circuit of a nuclear power plant, the wall of said component comprising a supporting structure (2) made of a glass-fiber-reinforced plastic. The component is provided with an electrically insulating inner and outer protective layer (4 and 6) on the inner and outer component surface, respectively. An electrically conductive inner intermediate layer (10) lies between the inner protective layer (4) and the supporting structure (2), said layer being provided with an electrical terminal (8), and an electrically conductive outer intermediate layer (14) lies between the outer protective layer (6) and the supporting structure (2), said layer being provided with an electrical terminal (12) and being electrically insulated from the inner intermediate layer (10).
G01M 3/18 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for pipes, cables, or tubesInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for pipe joints or sealsInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for valves
16.
METHOD AND DEVICE FOR TAKING A SAMPLE IN A STEAM GENERATOR
The invention relates to a method for taking a sample (50) of a deposit (28) on a secondary side (26) of a pipe base plate (10) of a steam generator (4) of a nuclear power plant (2). Said method consists of the following: a steam generator pipe (22) is removed from the pipe base plate (10) in order to expose a hot pipe bore (20) penetrating said pipe base plate (10), a removal tool (38) of a device (36) for taking the sample (50) is introduced into the hot pipe bore (20) by the primary side (24) of the pipe base plate (10) which is opposite the secondary side (26), one part of the deposit (28) is mechanically removed by the removal tool (38) as a sample (50), the sample (50) is transported through the hot pipe bore (20) to the primary side and removed from the steam generator (4), the removal tool (38) is removed from the steam generator (4). Said device for (36) taking a sample (50) contains a removal tool (38) which can be introduced by a primary side (24) of the pipe base plate (10) which is opposite the secondary side (26) in an exposed hot pipe bore (20) penetrating the pipe base plate (10) for mechanically removing one part of the deposit (28) as a sample (50).
The invention relates to a filter (28) for a cooling water (12)-conducting primary circuit (14) of a nuclear power station (2) comprising a reactor container (4) with a discharge opening (20 a, b) for the cooling water (12) that opens into the interior (22) of the reactor container and that can be accessed when the nuclear power station (2) is brought to a stop for revision. Said filter comprises a base carrier (29) with an entrance (30), at least one filter cartridge (34) held on the base carrier (29) such that it can be interchanged and which is fluidically coupled to said entrance (30), and a fixing device (36) that can be fixed to the discharge end (20 a, b) such that the entrance (30) couples fluidically to the discharge opening (20, a b). In a method for filtering cooling water (12) during the revision stop: a filter (28) according to one of claims 1 to 7 is fixed, by means of its fixing device (36), with its entrance (30) in fluid connection with the discharge opening (20 a, b), the cooling water (12) is circulated through said discharge opening (20 a, b), and the filter (28) is removed from the discharge opening (20 a, b).
G21C 19/30 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products
G21C 19/307 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for liquids
B01D 27/00 - Cartridge filters of the throw-away type
B01D 35/00 - Filtering devices having features not specifically covered by groups , or for applications not specifically covered by groups Auxiliary devices for filtrationFilter housing constructions
The invention relates to a method for chemical decontamination of an oxide-coated surface of a metal structural part or of a system in a nuclear power plant with several cleaning cycles, each involving an oxidation step, in which the oxide layer is treated with an aqueous solution containing an oxidation agent, and a subsequent decontamination step, in which the oxide layer is treated with an aqueous solution of an acid. At least one oxidation step is carried out in an acid solution and at least one oxidation step in an alkaline solution.
A fuel assembly loading machine (2) comprises a guide mast (10) and telescopic elements which can be telescopically lowered out of the guide mast (10) and raised again and are in the form of a centering bell (12), a fuel assembly gripper (14) and a control assembly gripper (16), wherein the telescopic elements are coupled to one another, and are operatively connected to a common main lifting mechanism (18), in such a way that, when the telescopic elements are raised out of the lowered position, initially the control assembly gripper (16) carries along the fuel assembly gripper (14), and then the fuel assembly gripper (14) carries along the centering bell (12), and wherein the guide mast (10) has a fixed stop (22), which fixes the end position of the centering bell (12) which can be reached during lowering. Such a fuel assembly loading machine (2) is intended to be developed such that the amount of time required for the fuel assembly change is reduced in comparison with the conventional technology, and such that a quicker and more effective response to any centering problems which may occur is possible than has hitherto been the case. According to the invention, a stop (40) is fitted on the guide mast (10) for this purpose, said stop being capable of being moved by means of a separate special lifting mechanism (42) in the lifting or lowering direction (38) and interacting with an associated supporting flange (20) of the centering bell (12), with the result that the centering bell (12) can be raised out of the end position of said centering bell into an intermediate position lying above this by the movable stop (40) being raised.
G21C 19/10 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements
G21C 19/105 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with grasping or spreading coupling elements
G21C 19/16 - Articulated or telescopic chutes or tubes for connection to channels in the reactor core
20.
DEVICE AND METHOD FOR REMOVING FOREIGN BODIES FROM THE BOTTOM OF A REACTOR PRESSURE VESSEL
The invention relates to a device (24) for removing foreign bodies (22) from the bottom (8) of a nuclear reactor pressure vessel (2) that is flooded with water (4) and comprises coupling connectors (10) projecting into the interior (20), which can be placed on one of the coupling connectors (10) and contains a suction device (30) which comprises a suction tube (32) having an intake opening (36) that can be moved relative to the remaining device (24), a filter (38), an outlet (42) and a suction pump (40) delivering water (4) from the intake opening (36) via the suction tube (32) and the filter (38) to the outlet (42). In a corresponding method, the device (24) is placed on one of the coupling connectors (10) and a suction device (30) is operated, wherein the suction pump (40) delivers water (4) from the intake opening (36) through the suction tube (32) via the filter (38) to the outlet (42), the intake opening (36) for taking in the foreign bodies (22) from the bottom (8) and/or the coupling connectors (10) is moved relative to the remaining device (24), and the foreign bodies (22) are collected in the filter (38).
G21C 13/036 - Joints between tubes and vessel walls, e.g. taking into account thermal stresses the tube passing through the vessel wall, i.e. continuing on both sides of the wall
G21C 17/01 - Inspection of the inner surfaces of vessels
21.
METHOD FOR DISMANTLING A LOWER CORE SUPPORT STRUCTURE INSTALLED IN A REACTOR PRESSURE VESSEL OF A NUCLEAR POWER PLANT
The invention relates to a method for dismantling a lower core support structure (12) installed in a reactor pressure vessel (8) of a nuclear power plant, the installed height thereof being greater than the distance between the top edge of the opened reactor pressure vessel (8) and the water level (4) of a reactor pit (2) flooded up to a walkable working plane (23), wherein the core support structure (12) is raised to an intermediate position, in which the core support structure is completely below the water level (4) and only a top part (22) thereof protrudes past the top edge of the opened reactor pressure vessel (8). In said intermediate position, the core support structure (12) is at least indirectly placed on the top edge of the opened reactor pressure vessel (8) and the top part (22) is then cut off from the core support structure (12). Said methodology prevents the core support structure (12) from protruding above the surface (4) of the water in the flooded reactor pit (21) during transport.
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
22.
MONITORING SYSTEM FOR AN INNER CHAMBER OF A MACHINE
The invention relates to a monitoring system (20) for an inner chamber of a machine (2) containing a radar source (26) radiating radar radiation (28) into the inner chamber (6), a radar receiver (34) receiving radar radiation (32) reflected in the inner chamber (6) and emitting same as a received signal (36), a control and evaluation unit (38) for determining an actual signature (40a) from the received signal (36) and for comparing said signature to a stored target signature (40b) representing a fault-free machine (2), and an output unit (42) for outputting an error signal (46, 52a, b) when a deviation of the actual signature (40a) from the target signature (40b) exceeds a tolerance value (44).
G01S 7/41 - Details of systems according to groups , , of systems according to group using analysis of echo signal for target characterisationTarget signatureTarget cross-section
G01S 13/04 - Systems determining presence of a target
G01S 13/88 - Radar or analogous systems, specially adapted for specific applications
F02C 7/05 - Air intakes for gas-turbine plants or jet-propulsion plants having provisions for obviating the penetration of damaging objects or particles
The invention relates to a method for chemically decontaminating the surface of a metal component of the primary circuit of a pressurized-water reactor, said surface comprising an oxide layer. The method is divided into two process stages and is further implemented as follows: - in the first process stage, at least one treatment cycle is performed, comprising an oxidation step, a reduction step, and a subsequent first decontamination step, wherein the component is treated in the oxidation step with an aqueous solution comprising an oxidizing agent, which converts trivalent chrome present in the oxide layer into hexavalent chrome; the component is treated in the reduction step with an aqueous solution comprising a reducing agent for reducing excess oxidation agent from the oxidation step; the component is treated in the first decontamination step with an aqueous solution comprising exclusively or predominantly at least one decontamination acid that forms no antisoluble deposits with metal ions present in the solution, in particular bivalent metal ions; and the solution is fed through an ion exchanger for removing metal ions present in the solution, which metal ions originate from the oxide layer and/or the base metal of the component, - in the second process stage, at least one treatment cycle is performed, comprising a second decontamination step, wherein the component is treated with an aqueous solution comprising exclusively or predominantly oxalic acid as a decontamination acid.
A method for removing corrosion products from a system, the method including: adjusting the system temperature to between 115° F. to 212° F.; injecting a cleaning dissolution solvent into the system; injecting a gas into the system after the system is filled with the cleaning dissolution solvent; the gas mixing with the solvent in the system; draining the solvent from the system after a predetermined period of time of dissolution; injecting a passivation composition into the system; injecting a gas into the system, the gas mixing the passivation composition; draining the system of the composition after a predetermined period of time of passivation; rinsing the system with a low volume solution; and rinsing the system at with a full volume solution.
A dry conversion reactor for converting uranium hexafluoride to uranium dioxide, the dry conversion reactor including a gas-phase reaction segment and a fluidized bed segment, wherein at least one of the gas-phase reaction segment and the fluidized bed segment is a replaceable segment. A method for operating a dry conversion reactor utilizing a uranium hexafluoride to uranium dioxide conversion process, the method including replacing at least one conversion reactor segment.
The invention relates to a method and a device for measuring the deformation of a fuel element of a pressurized water reactor, comprising a plurality of control rod guide tubes (2), wherein a probe (4) displaceable in the direction of the longitudinal axis of the control rod guide tube (2) is inserted in the interior of at least one control rod guide tube (2), comprising at least one reference axis, and supported in at least one section of the control rod guide tube (2) on the interior wall thereof, such that the orientation of the at least one reference axis is unambiguously a function of the orientation of the longitudinal axis (7) of the control rod guide tube (2) in said section, and the orientation of the at least one reference axis in said section of the control rod guide tube (2) is determined relative to the orientation of said or a further reference axis of the probe (4) in a different section.
G21C 17/06 - Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination
G01B 7/28 - Measuring arrangements characterised by the use of electric or magnetic techniques for measuring contours or curvatures
G01B 5/25 - Measuring arrangements characterised by the use of mechanical techniques for measuring angles or tapersMeasuring arrangements characterised by the use of mechanical techniques for testing the alignment of axes for testing the alignment of axes
G01B 7/16 - Measuring arrangements characterised by the use of electric or magnetic techniques for measuring the deformation in a solid, e.g. by resistance strain gauge
27.
METHOD AND DEVICE FOR MACHINING A SEALING SEAT OF A SHUT-OFF VALVE
The invention relates to a method for machining a sealing seat (24a, b), arranged at the end (26) of a connection pipe (8a, b), of a shut-off valve (6) fitted in a power generating or industrial plant (2), which method comprises the following steps: a) an upper valve part (16) and the internal housing fittings (32) are removed from the housing (10) of the shut-off valve (6), whereby a housing opening (14) is exposed, b) a clamping device (40), having a counter-bearing (56), is introduced through the housing opening (14) into the or a further connection pipe (8a, b) and fastened to the inner wall (50) thereof, c) a machine tool (58), having a bearing (64), is introduced into the housing (10) through the housing opening (14) and mounted with its bearing (64) on the counter-bearing (56), d) a machining step (B1-4) is carried out with the machine tool (58) on the sealing seat (24a, b), e) the machine tool (58) is detached from the counter-bearing (56) and removed through the housing opening (14), f) if need be, steps c) to e) are repeated with another or the same machine tool (58), g) the clamping device (40) is detached from the connection pipe (8a, b) and removed through the housing opening (14), and h) the upper valve part (16) and the internal fittings (32) are attached to the housing (10). A device for machining a sealing seat (24a, b), arranged at the end (26) of a connection pipe (8a, b), of a shut-off valve (6) fitted in a power-generating or industrial plant (2) comprises a clamping device (40), which can be introduced through a housing opening (14) of the shut-off valve (6) into the or a further connection pipe (8a, b) and includes a counter-bearing (56) and a fastening element (47) interacting with the inner wall (50) of the connection pipe (8a, b), and at least one machine tool (58), which can be introduced through the housing opening (14) into the housing, for carrying out a machining step (B1-4) on the sealing seat (24a, b), wherein the machine tool (58) has a bearing (64), which can be mounted in the counter-bearing (56).
B24B 15/03 - Machines or devices designed for grinding seat surfacesAccessories therefor in valve housings using portable or mobile machines
28.
CYCLONE SEPARATOR FOR THE PHASE SEPARATION OF A MULTIPHASE FLUID STREAM, STEAM TURBINE SYSTEM HAVING A CYCLONE SEPARATOR AND ASSOCIATED OPERATING METHOD
The invention relates to a cyclone separator (1) for the phase separation of a multiphase fluid stream, having a housing (2) which is configured substantially rotationally symmetrically about a mid-axis (M) and encloses a hollow chamber (3), having at least one feed line (6) for the fluid stream, which is designed for an inflow of the fluid stream that is directed substantially tangentially to the interior (11) of the housing, and having at least one discharge line (24) for the separated gaseous component of the fluid stream. Such a cyclone separator is to be improved to the effect that it is suitable for heating the gaseous content of the fluid stream and makes low demands on material and required space. In addition, the aim is to ensure a uniform and preferably homogeneous flow distribution of the steam to be heated as it enters the heating phase. To this end, the invention provides for the hollow chamber (3), when viewed in the radial direction from the mid-axis (M), to comprise an outflow chamber (16) having a substantially circular cross section and, following same in the aforesaid order, a heating chamber (14), an intermediate chamber (15), a dryer chamber (13) and an inflow chamber (12) having a substantially circularly annular cross section in each case, wherein the inflow chamber (12) is delimited on the outside by means of the housing (2), wherein the heating chamber (14) contains heating elements designed to heat the gaseous content, wherein at least one fine separator (28) and at least one associated condensate-collecting trough (32) are arranged in the dryer chamber (13), and wherein the at least one condensate-collecting trough (32) is connected to at least one condensate discharge pipe (34) which is arranged in the intermediate chamber (15) and through which the condensate forming in the at least one fine separator (28) during operation is led out of the hollow chamber (3).
A method for retrofitting a boiling water reactor slip joint of a jet pump to reduce vibrations is provided. The method includes removing a mixing chamber from an existing slip joint defined by a diffuser and the mixing chamber, the existing slip joint defining an existing annular gap, and providing a new slip joint defining a new annular gap, the new annular gap being reshaped to permit reduced vibration. A jet pump and a method of operating a jet pump are also provided.
A method to perform an analysis of two types of CRUD on a nuclear fuel rod, including providing a nuclear fuel rod with first and second layers of CRUD on an exterior of the fuel rod; brushing the first layer of CRUD from a selected area on the fuel rod with a CRUD tool having a brushing device; a force applied to the brushing device to remove the first layer of CRUD; collecting the first layer of CRUD from the brushing device, scraping the second layer of CRUD from the fuel rod in the selected area with the tool, wherein the tool has a scraping device and a second force is applied to the tool for scraping, collecting the second layer of CRUD from the scraping device, and analyzing the first and second layers of CRUD separately with a scanning electron microscope.
End piece (2) for a corrugated hose (6), comprising: a bushing (8) having a front region (12) and a rear region (20) opposite of a longitudinal direction (L), said rear region being located inside the corrugated hose (6) in the mounted state of the bushing (8). The front region (12) of the bushing (8) consists of a collar (14), which forms a groove (16) along the circumference of the bushing (8) for receiving the free end (4) of the corrugated hose (6). The rear region (20) of the bushing (8) consists of a plurality of flexible tongues (22), which have a relief direction that is oriented perpendicular to the longitudinal direction (L) of the bushing (8) and directed toward the outer chamber (AN) of the bushing (8). At least a portion of the flexible tongues (22) comprises a radially projecting detent protrusion (26) for implementing an engagement active in the longitudinal direction (L) of the bushing (8). A retaining sleeve (10), which can be received by the bushing (8) with precise fit and, in the mounted state of the end piece (2), extends in an inner chamber (IN) enclosed by the bushing (8), limits the radial play of the flexible tongue (22).
F16L 25/00 - Construction or details of pipe joints not provided for in, or of interest apart from, groups
F16L 33/26 - Arrangements for connecting hoses to rigid membersRigid hose-connectors, i.e. single members engaging both hoses specially adapted for hoses made of metal
F16L 33/30 - Arrangements for connecting hoses to rigid membersRigid hose-connectors, i.e. single members engaging both hoses comprising parts inside the hoses only
32.
METHOD AND APPARATUS FOR INSPECTING AN ANNULAR WELD ON A MAIN COOLANT PIPE CONNECTED TO A REACTOR PRESSURE VESSEL IN A NUCLEAR POWER PLANT
In a method and an apparatus for inspecting an annular weld (10) on a main coolant pipe (4) connected to a reactor pressure vessel (2) in a nuclear power plant, a self-propelled submersible vehicle (12) is introduced into the open, flooded reactor pressure vessel (2), said vehicle (12) comprising at least one inspection device (20) that can be placed against the internal peripheral surface of the main coolant pipe (4) and can be moved in the peripheral direction of the pipe (4). The submersible vehicle (12) is advanced into the main coolant pipe (4), which is also flooded, and is fixed in the region of the weld (10) by means of radially deployable extension arms (16). The inspection device (20) is then placed against the weld (10) and is moved therealong in the peripheral direction.
The invention relates to a device for phase separating a multi-phase fluid flow, having a housing (2) designed substantially rotationally symmetrically about a center axis (M) and enclosing a hollow space (3), having at least one infeed line (6) for the fluid flow designed for inflow of the fluid flow directed substantially tangentially to the interior (11) of the housing, and having at least one outlet line (24) for the separated gaseous portion of the fluid flow, wherein said device heats the gaseous portion of the fluid flow, such as steam, and requires little material and space. To this end, heating elements designed for heating the gaseous portion are disposed in the hollow space (3) in an annular chamber (14) placed concentrically about the center axis (M).
Sealing device (2) for a device for measuring the fill level (4) in a fluid container, particularly in a pressure vessel (6) of a nuclear facility, comprising at least one thermal element (8) which is connected to the pressure vessel (6) by means of a pressure pipe (10) and to an evaluation unit (14) by means of a pipeline (12) connected to the pressure pipe (10), wherein the pressure pipe (10) and the pipeline (12) are connected to each other by means of a threaded pipe connection (16) comprising a center part (31) having two threaded sections (20, 22), wherein one threaded section (20) is connected to a pipe sleeve (26) by means of a screw connection (34), and the other threaded section (22) is connected to the pressure pipe (10) by means of a further screw connection (36).
G01F 23/22 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water
F16L 19/065 - Joints in which sealing surfaces are pressed together by means of a member, e.g. a swivel nut, screwed on, or into, one of the joint parts in which radial clamping is obtained by wedging action on non-deformed pipe ends the wedging action being effected by means of a ring
G21C 17/035 - Moderator- or coolant-level detecting devices
35.
OPTICAL CABLE, HAVING ENCODED INFORMATION ON LENGTH AND TYPE OF CABLE
The present invention has the object to create a simple possibility for tracing back and identifying cables (4) in a cable installation. In particular, a fast laying of factory-delivered cables (4) at construction sites is to be enabled. Furthermore, a cable piece randomly picked from an installation is to be identifiable as simply as possible. For this purpose, a cable (4) is provisioned according to the invention, having a cable sheath (8), a length marker (10) arranged on the cable sheath (8) in a continuous manner, and a plurality of cable markers (16), which are arranged on the cable sheath (8) at specified distances from each other as viewed in the longitudinal direction of the cable (4).
A leak detecting apparatus for detecting flaws in a surface submerged in liquid is provided that includes a vacuum head for contacting the surface and forming a vacuum chamber filled with liquid between the surface and the vacuum head, the vacuum head including a multi-surface elastomer seal for contacting the surface; and a vacuum generator for pulling gas through a flaw in the surface into the vacuum head and to the gas separation chamber so as to displace the liquid from the gas separation chamber. A method for detecting flaws in a surface submerged in liquid is also provided.
A motor operated valve diagnostic data acquisition system including: a motor operated valve assembly; a transmitter, the transmitter being permanently connected in a housing of a motor operated valve assembly; a receiver, the transmitter transmitting signals to the receiver via motor operated valve cables, the cables being existing motor operated valve cables; and electronic circuitry connected to the receiver, the electronic circuitry processing the signals received by the receiver.
F16K 37/00 - Special means in or on valves or other cut-off apparatus for indicating or recording operation thereof, or for enabling an alarm to be given
G01L 5/00 - Apparatus for, or methods of, measuring force, work, mechanical power, or torque, specially adapted for specific purposes
G01L 5/12 - Apparatus for, or methods of, measuring force, work, mechanical power, or torque, specially adapted for specific purposes for measuring axial thrust in a rotary shaft, e.g. of propulsion plants
38.
DEVICE AND METHOD FOR DETECTING A MODIFICATION OF THE AERODYNAMIC INFLOW CONDITIONS OF ROTATING TURBINE BLADES
The invention relates to a method for detecting a modification of the aerodynamic inflow conditions of the turbine blades (6) rotating when the turbine (2) is in operation. Said method consists of the following steps: the operating noise (x(t)) of the turbine (2) is detected (16) and filtered (20) for determining a rotating sound (d(t)) caused by the rotating turbine blades (6), a value of the amplitude and/or phase of the rotating sound (d(t)) is recorded (22) for a plurality of successive times, the recorded values are analysed (22) and a modification of the aerodynamic inflow conditions of the rotating turbine blades (6) is detected prior to or after inputting the modification using a difference in the values of the amplitude and/or phase of the rotating sound (d(t)).
The invention relates to a device for releasing an explosive gas via a chimney arranged on a roof of a building. The upper free end of the chimney is provided with a release head consisting of a plurality of distribution pipes projecting from the chimney in the form of a star. Each pipe has an outlet on the free end thereof. The aim of the invention is to be able to reduce the pressure wave load exerted during an explosion even during adverse wind conditions. To this end, the end regions of the distribution pipes are embodied as jet pumps in which the explosive gas is the propellant sucking the ambient air.
The invention relates to a method for chemically decontaminating the surface of a metal component, wherein, in a first treatment step, an oxide layer formed on the component by corrosion of the material of said component is removed from the surface of the component by means of a first aqueous treatment solution containing an organic decontamination acid and in a subsequent second treatment step, the surface at least partially free of the oxide layer is treated with an aqueous solution containing an active component for removing particles which adhere to the surface. The active component consists of at least one anionic surfactant from the group consisting of sulphonic acids, phosphonic acids, carboxylic acids and salts of said acids.
A piping assembly for directing fluid and mitigating acoustic and vortex coupled resonance is provided that includes a main pipe delivering fluid in a first direction; a standpipe coupled to the main pipe at an intersection; and a scoop positioned at the intersection directing the fluid towards the standpipe. A scooping insert and a method for disrupting vortex shedding in a piping assembly are also provided.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
A nickel, chromium, iron alloy and method for use in producing weld deposits and weldments formed therefrom. The alloy comprises, in weight percent, about 28.5 to 31.0% chromium; about 0 to 16% iron; less than about 1.0% manganese; about 2.1 to 4.0% niobium plus tantalum; 1.0 to 6.5% molybdenum; less than 0.50% silicon; 0.01 to 0.35% titanium; 0 to 0.25% aluminum; less than 1.0% copper; less than 1.0% tungsten; less than 0.5% cobalt; less than about 0.10% zirconium; less than about 0.01% sulfur; less than 0.01% boron; less than 0.03% carbon; less than about 0.02% phosphorous; 0.002 to 0.015% magnesium plus calcium; and balance nickel and incidental impurities. The method includes the steps of forming a welding electrode from the above alloy composition and melting the electrode to form a weld deposit. A preferred weldment may be in the form of a tubesheet of a nuclear reactor.
B32B 15/04 - Layered products essentially comprising metal comprising metal as the main or only constituent of a layer, next to another layer of a specific substance
C22C 19/05 - Alloys based on nickel or cobalt based on nickel with chromium
B23K 35/22 - Rods, electrodes, materials, or media, for use in soldering, welding, or cutting characterised by the composition or nature of the material
44.
FUEL ELEMENT FOR A PRESSURIZED-WATER NUCLEAR REACTOR
In a fuel element for a pressurized-water reactor, in addition to spacers (6), flow-guiding structural parts (8) are arranged which comprise four outer webs (20) which, in a plane oriented perpendicularly to the central longitudinal axis (4), surround a square inner region (24) of which the centre point (M) lies on the central longitudinal axis (4). At their lower longitudinal side facing the flowing cooling water in the operating state, the outer webs are provided with deflection lugs (30) pointing towards the inner region and are structurally identical, wherein mutually opposite outer webs (20) are arranged mirror-symmetrically with respect to a centre plane (21) extending in the axial direction. Such a structural part (8) forms, at most for a number of fuel rods (2) which is smaller than their total number in the fuel element, cells (12) through which a respective fuel rod (2) is guided, wherein the number of these cells (27), which are situated in a row (14) or column (16), is smaller than the number of the fuel rods (2) respectively situated in this row (14) or column (16).
G21C 3/322 - Means to influence the coolant flow through or around the bundles
G21C 3/326 - Bundles of parallel pin-, rod-, or tube-shaped fuel elements comprising fuel elements of different compositionComprising, in addition to the fuel elements, other pin-, rod-, or tube-shaped elements, e.g. control rods, grid support rods, fertile rods, poison rods or dummy rods
G21C 3/352 - Spacer grids formed of assembled intersecting strips
45.
Boiling water reactor nuclear power plant with alcohol injection
A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
46.
METHOD FOR CONDITIONING A PRECIPITATING SOLUTION THAT ARISES DURING THE WET-CHEMICAL CLEANING OF CONVENTIONAL OR NUCLEAR PLANTS, SAID SOLUTION CONTAINING ORGANIC SUBSTANCES AND METALS IN IONIC FORM
The invention relates to a method for conditioning a precipitating solution that arises during the wet-chemical cleaning of conventional or nuclear plants, said solution containing organic substances and metals in ionic form, in which at least a portion of the organic substance is decomposed through electrochemical treatment or UV radiation of the precipitating solution and wherein at least one metal precipitates with the addition of phosphoric acid, and the resultant phosphate precipitate is removed from the precipitating solution.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
A method for monitoring the fill level of a liquid (F) in a liquid container, in particular for monitoring the coolant fill level in a reactor pressure vessel (4) of a pressurized water reactor system, wherein a determination is made about the drop in the liquid level (10) below the installed height of a heated thermal element (HT) by way of the temperature difference (DT) measured between said heated thermal element (HT) and an unheated thermal element (UHT) associated therewith, both of said elements being disposed in the liquid container, said method comprising the especially reliable triggering of an alarm with a short response time while maintaining a low level of physical and controls engineering complexity when the liquid level to be monitored falls below a critical value. To this end, according to the invention the temporal progression of the temperature difference (DT) is continuously monitored for any significant increase, in particular sudden increase, within a time interval of pre-determined duration (Δt) prior to the respective evaluation time (t0), wherein an alarm signal is issued as soon as the change (ΔDT) in the temperature difference (DT) within said time interval reaches or exceeds a pre-determined limit value.
G01F 23/22 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water
G01F 23/24 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water by measuring variations of resistance of resistors due to contact with conductor fluid
49.
METHOD FOR CONDITIONING A CLEANING SOLUTION RESULTING FROM THE WET CHEMICAL CLEANING OF A NUCLEAR STEAM GENERATOR
In a method for conditioning a cleaning solution resulting from the wet chemical cleaning of a nuclear steam generator, the cleaning solution is electrolytically treated and radioactive metal nuclides contained in the cleaning solution are precipitated at a cathode.
A device for repairing a damaged location (21) in an underwater region of a wall of a vessel or pool, in particular in the region of a wall of a pool of a nuclear reactor plant, comprises a guiding system (6), which can be attached along a side wall (12) at a distance therefrom and can be fixed thereto, for guiding a carriage (20) which can be made to move in the longitudinal direction of the guiding system (6). Arranged on the carriage (20) is a displaceably mounted holder (30) for a repair support (22), which can be placed with an adhesive surface against the wall region containing the damaged location (21). With such a device, wall regions that are at a great depth and can only be accessed through narrow gaps can be repaired quickly and reliably.
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
The invention relates to a pencil-shaped electrical heating element, in particular for application in a device for measuring filling level in a liquid container, in particular in a reactor chamber of a nuclear plant, comprising an electrically-conducting sleeve and at least one electric line embedded therein which is in conducting contact with the sleeve. The electrical line has at least two line sections with differing electrical properties and/or thermal conductivities.
H05B 3/48 - Heating elements having the shape of rods or tubes non-flexible heating conductor embedded in insulating material
G01F 23/22 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water
The invention relates to a physico-chemical method for cleaning the secondary chamber of a heat exchanger in a nuclear plant, wherein the secondary chamber (6) is dried. A cleaning solution is introduced into the chamber to treat the deposits present in the secondary chamber (6).
The invention relates to a method for conditioning a contaminated ion exchange resin, by mixing the same with water and at least partly breaking up the same into water-soluble components by means of an oxidising agent added to the water, wherein the resulting aqueous solution is immobilised with a binder, optionally after concentration by evaporation of water.
A nuclear power plant (10) is provided including a BWR (20), a reactor cooling system cooling the BWR (20), an HWC hydrogen injection system (60) connected to the reactor cooling system and an alcohol injection system (100) connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
The invention relates to a recombiner element (4) comprising a plurality of catalyst elements (8) that are arranged in a common housing (6) and that trigger a recombination reaction with oxygen when hydrogen is carried along in a feed gas flow, the housing (6) surrounding the catalyst elements (8), which are arranged therein, in a funnel-type way in such a manner that the heat released by the recombination reaction supports the gas flow inside the housing (6) by a convection effect, which recombiner element safeguards a reliable removal of the hydrogen from the gas mixture with an especially high degree of operational safety even under comparatively extreme conditions or scenarios of said type. According to the invention, at least one of the catalyst elements (8) arranged inside the housing (6) has a predetermined ignition zone (20) in which a surface temperature of more than 560°C is produced in the convective operation at ambient conditions of approximately 1 bar and 100°C at a hydrogen concentration in the feed gas flow of more than 5% by volume.
The invention relates to a safety system (1) for a nuclear plant comprising a plurality of catalytic recombiner elements (40, 44) that trigger a recombination reaction with oxygen when hydrogen is carried along in a feed gas flow, which system safeguards a reliable removal of the hydrogen from the gas mixture with an especially high degree of operational safety even under comparatively extreme conditions or scenarios of said type. According to the invention, the recombiner elements and/or the flow paths interconnecting two recombiner elements each on the gas side are designed in such a manner that a pressure pulse (40) produced in the gaseous medium by an ignition during the recombination reaction in a first recombiner element (40) triggers a gas displacement process in the feed section of a second, adjacent recombiner element (44), said process having a flow rate of at least 5 m/s.
The invention relates to a fuel element (100) for a boiling water reactor, having a multiplicity of fuel rods (102) which are combined to form a bundle and which stand with the lower end thereof on a lower holding plate (311), and having a fuel element base (101), the upper outlet opening of which faces toward the holding plate (311) and the lower inlet opening of which faces toward a lower core grate (201) of the boiling water reactor, wherein a flow duct (301) extends through the fuel element base (101) between the upper outlet opening and a lower inlet opening along a central longitudinal axis (A), which flow duct (301) is traversed by cooling water during operation, wherein the fuel element base (101) has a throttle element (302) which is spaced apart from the inlet and outlet openings in the direction of the central longitudinal axis (A) and which comprises an impact body (303) which is held centrally in the flow duct (301) by a plurality of webs (401) which extend transversely with respect to the central longitudinal axis (A), which impact body (303) occupies a part of the cross-sectional area of the flow duct (301).
An improved manipulator system having a remotely actuated tube walker assembly for reciprocally moving the manipulator along a tube bundle for inspection and repair of the tubes of the tube bundle wherein the tube walker has a movable inner section and an outboard section with both sections having groups of hydraulically actuated locking fingers which are selectively engageable with the tubes of the tube bundle so that when the fingers of the outboard section are locked the fingers of the inner section are disengaged and the inner portion can be moved axially to an aligned position with new tube locations.
Methods for removing nuclear waste from a component are provided. In one embodiment, the method includes the steps of supplying oxalic acid to the nuclear waste on the component to form an oxalic acid/waste solution, feeding the oxalic acid/waste solution from the component, feeding an oxidant to the oxalic acid/waste solution to form carbon-dioxide, water and a precipitate, and separating the precipitate from the water. A system for removing nuclear waste is also provided.
A62D 3/30 - Processes for making harmful chemical substances harmless, or less harmful, by effecting a chemical change in the substances by reacting with chemical agents
60.
Boiling water reactor nuclear power plant with alcohol injection
A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 19/307 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for liquids
61.
Reduced order stress model for online maneuvering, diagnostics of fuel failure and design of core loading patterns of light water reactors
The invention is principally directed to a reduced order model, XEDOR, facilitating the prediction of and the diagnostics of pellet-clad interaction stress-corrosion-cracking failure of nuclear fuel rods. The invention more particularly relates to assessment of susceptibility to PCI failure for guidance in the design of fuel loading in nuclear reactors. The invention additionally relates to the protection against PCI failure by providing operational information to operators of a nuclear reactor during power maneuvering, including predictive calculations prior to executing power maneuvers. Additionally, the invention relates to the diagnostics of an event suggesting a possible PCI cladding failure.
A collecting line for leakage monitoring in an installation is permeable to a substance (S) to be monitored in at least one section extending along the longitudinal direction. The collecting line consists of at least two subsections (10, 14) that are interspaced from a first end (12) of the collecting line. The distance (a2) of the second subsection (14) is larger than that of the first subsection (10). The invention is characterized in that the permeability to the substance (S) to be monitored is higher in the first subsection (10) than in the second subsection (14).
The invention relates to a method for decontaminating surfaces, which have been contaminated with alpha emitters, of nuclear plants, which method is carried out subsequently to a decontamination process which is aimed at the removal of oxide layers. The surfaces are treated with an aqueous solution which contains a cationic or zwitterionic surfactant and oxalic acid, wherein at least a part of the solution, after having acted on a surface, is conducted across an ion exchanger.
The invention relates to a method for the detection of a leakage (30a, b) in a double pipe (2), wherein a medium (22) located in an intermediate chamber (10) between the interior (6) and exterior pipe (8) is moved toward a first end (16a) of the double pipe (2), whereupon medium (22) flows in from the second end (16b). On the first end (16a) the medium (22) is examined for any leakage (30a, b) of a characteristic material (7, 36), upon the detection (34) of which a leakage signal (34) is generated and a location (L1, 2) of the leakage (30a, b) is calculated based on the transport time (Tx) of the characteristic material (7, 36) from the leakage (30a, b) to the first end (16a) and based on the mass flow (Avi, Evi) of the medium (22). A corresponding device (4) contains a conveyor unit (18, 36) for moving the medium (22) through the intermediate chamber (10), a material sensor (28a, b) disposed at the first end (16a) for examining the medium (22) for the characteristic material (7, 36), a mass flow sensor (24a, b) at the first end (16a) for the medium (22), a control and analysis unit (32) for generating a leakage signal (34) and calculating of the location (L1, 2) of the leakage (30a, b) based on the transport time (Tx) and the mass flow sensor (24a, b).
G01M 3/22 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for pipes, cables, or tubesInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for pipe joints or sealsInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for valves
65.
ELECTROCHEMICAL CORROSION POTENTIAL DEVICE AND METHOD
The invention provides at least two electrochemical sensors positioned in a nuclear reactor or in a system adjacent to the nuclear reactor, wherein at least one of the at least two electrochemical sensors has a heated zirconium electrode, and the at least two electrochemical sensors produce voltages proportional to an electrochemical corrosion potential for a surface that each of the at least two electrochemical sensors are installed upon. The invention also provides an arrangement configured to accept the voltages produced by the at least two electrochemical sensors, wherein the arrangement is configured to determine an electrochemical corrosion potential of a zirconium fuel rod in the nuclear reactor based upon the voltages of the at least two electrochemical sensors.
The invention relates to an apparatus for detecting slow movements or changes in the length of a test piece (16), particularly by applying a tensile force that affects the test piece (16). Said apparatus comprises a measuring brace (26) that has two spaced-apart legs (28) and is to be attached to the test piece (16) as well as an extension-sensitive sensor (38) which is disposed on the measuring brace (26). The invention further relates to a method for measuring slow movements or changes in the length of a test piece (16), especially by applying a tensile force that affects the test piece (16). In said method, the changes in the length of the test piece (16) are detected by means of a measuring brace (26) that is attached thereto and at least one extension-sensitive sensor (38) which is disposed on the measuring brace (26).
G01B 7/16 - Measuring arrangements characterised by the use of electric or magnetic techniques for measuring the deformation in a solid, e.g. by resistance strain gauge
67.
METHOD FOR PREDICTION OF LIGHT WATER REACTOR FUEL DEFECTS USING A FUEL CONDITION INDEX
A method to assess light water reactor fuel integrity is presented having the steps of granting access in a nuclear reactor fuel pool to at least one of a discharged fuel rod and a nuclear fuel assembly, calculating an operating flux for the fuel rod, measuring a thickness of CRUD on the fuel rod, measuring a thickness of oxide on the fuel rod, calculating a maximized flux for the at least one fuel rod for a position of the one fuel rod in a nuclear reactor, calculating a maximized deposit for the fuel rod, calculating a maximized oxide thickness for the fuel rod, calculating a fuel condition index of the fuel rod, comparing the fuel condition index to an index constant, and removing the fuel rod from operation when the fuel condition index is greater than the index constant.
The invention relates to a safety device (2) for protecting against excess pressure in a pressure system, in particular a pressure container (4) or a nuclear plant. Said safety device comprises a safety armature (6) that opens with high reliability when a critical system pressure is reached and allows the pressure system to decompress, and also close again in a particularly reliable manner after successful decompression to a determined value. According to the invention, the safety armature (6) comprises two safety valves (12, 48) that are arranged behind one another in a flow channel (10), each safety valve (12, 48) being connected on the control side to the control device (40, 72) via a control line (38, 70), said control line (40, 72) being associated therewith and being connected on the pulse side to the pressure system via a pressure relief line (42, 74), said control device operating according to the closed circuit principle.
F16K 17/10 - Safety valvesEqualising valves opening on surplus pressure on one sideSafety valvesEqualising valves closing on insufficient pressure on one side spring-loaded with auxiliary valve for fluid operation of the main valve
69.
METHOD FOR REMOVING DEPOSITS CONTAINING MAGNETITE AND COPPER FROM CONTAINERS IN INDUSTRIAL AND POWER PLANTS
The invention relates to a method for removing deposits containing magnetite and copper from a container, particularly from a steam generator of a nuclear power plant, wherein - in a first step the container is treated using an alkaline cleaning solution comprising a complexing agent forming a complex soluble with iron ions, a reducing agent, and an alkalizing agent, and - in a second step a further complexing agent forming a more stable complex with iron III ions than the complexing agent used in the first step and an oxidant are added to the cleaning solution of the first step present in the container.
A method to characterize the power transfer of a nuclear component is provided including the steps of obtaining a sample of a deposit layer on a side of a nuclear component, obtaining a scanning electron microscope image of an outside surface of the sample, obtaining a scanning electron microscope image of an inside surface of the sample, analyzing the scanning electron microscope images of the outside and inside surfaces of the sample for a presence of capillaries and steam chimneys, and calculating the power transfer of the component based on a number of steam chimneys in the deposit layer.
G01N 23/225 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or by measuring secondary emission from the material using electron or ion microprobes
A method to determine deposits in a steam generator having the steps of creating a calibration standard having at least two rings of deposit material, subjecting the calibration standard to an eddy current signal, wherein an amplitude of the signal reflected from the calibration standard is used to obtain a polynomial equation fit of the reflected eddy current signals to actual thickness of the rings, obtaining a steam generator with tubes, initiating an eddy current signal into the tubes of the steam generator, detecting and recording reflections of the eddy current signal initiated into the tubes of the steam generator, and determining a thickness of the deposits in the steam generator from the recorded reflections of the eddy current signal and the polynomial equation.
G01B 7/06 - Measuring arrangements characterised by the use of electric or magnetic techniques for measuring length, width, or thickness for measuring thickness
72.
METHOD FOR THE PRODUCTION OF A COLLECTING LINE FOR DETECTING AND LOCATING AN AGENT DISCHARGING INTO THE ENVIRONMENT OF THE COLLECTING LINE IN CASE OF LEAKAGE, AND COLLECTING LINE PRODUCED BY MEANS OF THIS METHOD
The invention relates to a method for the production of a collecting line, and to a collecting line produced according to this method for detecting and locating an agent discharging into the environment of the collecting line in case of leakage. Such a collecting line comprises a carrier pipe (18), the wall of which is equipped with openings (16) that are sealed by means of an elastic material that is permeable for the agent. According to the invention, a hose (4) made of the permeable elastic material, the inside diameter of which is smaller than the outside diameter of the carrier pipe (18), is elastically expanded by creating a pressure difference between the interior and the exterior of the hose (4), the carrier pipe (18) being inserted into the radially expanded hose (4), and the radially expanded hose (4) being attached to the outside surface of the carrier pipe (18) by reducing the pressure difference.
The invention provides a method by means of which a fuel rod cladding tube (4) can be treated or prepared such that the influence of iron oxide deposits on the surface thereof can be studied and analyzed as safely as possible, and at the same time precisely, under near-operational conditions. For this purpose, the invention provides that the fuel rod cladding tube (4) is coated at least partially with an iron oxide layer (6) by means of immersing the same in an aqueous electrolyte medium (8) comprising iron oxide particles, wherein the iron oxide particles are produced by means of anodic oxidation of a ferrous working electrode (10).
A method and apparatus for removing uranium (IV) and uranium (VI) from sands and soils. The method and device assays a volume of soil and determines a presence and position of uranium enrichment. A concentration of uranium enrichment is determined, and compared to a threshold concentration. The volume of soil is processed and transported to a coarse screen, deck screen and classified in a classifier. In the classifier, the volume of soil is separated into a washed and fines fraction, assayed and transported to a uranium recovery facility, then transported to a silicon removal process and subsequent ammonia and metals removal processes.
The invention relates to a method for determining the filling level of a liquid in a liquid container by comparing a temperature T which has been determined using an unheated thermocouple with a temperature THT which has been determined using a heated thermocouple, wherein the unheated thermocouple and the heated thermocouple can be arranged in the liquid container. The invention also relates to an apparatus for determining the filling level of a liquid in a liquid container, said apparatus having means for comparing a temperature which has been determined using an unheated thermocouple with a temperature which has been determined using a heated thermocouple, wherein the unheated thermocouple and the heated thermocouple are arranged in the liquid container. As a result of the thermal compensation for the temperature dependence of the heat transfer coefficient α(T) of the liquid, it is possible to determine the temperature difference THT-T in a manner independent of temperature and thus to reliably measure the liquid level.
G01F 23/24 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water by measuring variations of resistance of resistors due to contact with conductor fluid
76.
METHOD FOR AIDING THE LOADING OR UNLOADING OF A CORE OF A PRESSURISED WATER REACTOR
The invention relates to a method for loading and unloading a core of a pressurised water reactor. According to said method, a force (F) acting transversally to the axial direction (57) of a fuel element (50) of the core is exerted against said fuel element (50), which is adjacent to an unloading or loading fuel element position (52), in an upper region of the fuel element (50), by means of a device. The device used is a submersible vessel (2) that can be freely manoeuvred under water, and on which an extensible carrier arm (22, 24) is arranged, said arm being at least indirectly positioned against the fuel element (50).
A degassing plant (2) for reactor coolants (R) has a degasser column (6) which is coupled to the primary cooling circuit of a pressurized water reactor and through which reactor coolant (R) flows. The degassing plant (2) furthermore comprises a coolant evaporator (54) with a first heat exchanger (52) and a strip steam condenser (34) with a second heat exchanger (44), wherein a partial flow of the reactor coolant (R) flows through the heat exchanger (52) of the coolant evaporator (54) on the secondary side, and wherein the heat exchanger (44) of the strip steam condenser (34) is switched, on the primary side, in a vapour and gas outlet line (32) which is connected to the degasser column (6). The degassing plant (2) is intended to be configured such that, with as simple a design as possible and taking into consideration relevant safety procedures, a particularly effective and at the same time energy-efficient separation of gasses, which are dissolved in the reactor coolant (R) and cannot be condensed, is made possible, wherein the thermal load of the assigned nuclear intermediate cooling system is furthermore intended to be kept as low as possible. To this end it is provided that the heat exchanger (52) of the coolant evaporator (54) is switched in a heat-pump circuit (58) on the primary side, which heat-pump circuit is coupled to the heat exchanger (44) of the strip steam condenser (34) with respect to the heat flux, which is established during plant operation, such that the heat liberated in the strip steam condensation is transferred at least partially to the reactor coolant (R), which flows through the coolant evaporator (54), and thus leads to its evaporation.
G21C 15/16 - Cooling arrangements within the pressure vessel containing the coreSelection of specific coolants comprising means for separating liquid and steam
G21C 19/307 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for liquids
A method to characterize the power transfer of a nuclear component is provided including the steps of obtaining a sample of a deposit layer on a side of a nuclear component, obtaining a scanning electron microscope image of an outside surface of the sample, obtaining a scanning electron microscope image of an inside surface of the sample, analyzing the scanning electron microscope images of the outside and inside surfaces of the sample for a presence of capillaries and steam chimneys, and calculating the power transfer of the component based on a number of steam chimneys in the deposit layer.
G01N 23/225 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or by measuring secondary emission from the material using electron or ion microprobes
The invention relates to a line assembly for transporting a liquid (0), in particular crude oil. Said line assembly comprises an over-ground pipeline (2) below which a sensor line (6) is arranged in order to detect a leak, said sensor line extending along the pipeline (2) in the longitudinal direction thereof. According to the invention, said sensor line (6) is arranged in a collecting container (8), in the deepest point thereof, that also extends along said pipeline and is fixed to the pipeline (2), and the lateral edges (10) of said collecting container extending in the longitudinal direction are at a distance to the outer surface (12) of the pipeline (2). Means for reducing an air flow (A) passing transversally to the longitudinal direction of the pipeline, between the collecting container (8) and the outer surface (12) of the pipeline (2), are arranged inside the collecting container (8).
G01M 3/04 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point
G01M 3/18 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for pipes, cables, or tubesInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for pipe joints or sealsInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using electric detection means for valves
80.
DEVICE FOR REMOVING SOLID PARTICLES FROM THE COOLING MEDIUM CIRCULATED IN THE PRIMARY CIRCUIT OF A NUCLEAR REACTOR
The geometric dimensions and shapes of a device for removing solid particles from the cooling medium (K) that is circulated in the primary circuit of a nuclear reactor, in particular a boiling water nuclear reactor, are such that the device can be used as a fuel element or fuel assembly configured for this nuclear reactor and can be inserted in lieu of such a fuel element or fuel assembly into an empty fuel element or assembly position of the reactor core (40) of the nuclear reactor.
To permit the particularly safe, easy handling of fuel elements (10), the invention discloses a fuel element loading machine (2) with a lifting apparatus (8) which comprises a guide mast (12) and a fuel element gripper (16) which can be extended telescopically from the guide mast (12). According to the invention, a video camera (32) is for this purpose arranged on the lifting apparatus (8), is connected on the data side to a control station (34) and its viewing direction pointed downwards.
G21C 19/105 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with grasping or spreading coupling elements
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
A method to analyze crystals in a deposit on a surface of a nuclear generating station heating surface, wherein the method extracts a sample of material from the surface of the nuclear generating station heating surface and also includes conducting at least one of a high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample and a scanning transmission electron microscope/selected area electron diffraction/spot and elemental mapping analysis of the sample.
A method to determine deposits in a steam generator having the steps of creating a calibration standard having at least two rings of deposit material, subjecting the calibration standard to an eddy current signal, wherein an amplitude of the signal reflected from the calibration standard is used to obtain a polynomial equation fit of the reflected eddy current signals to actual thickness of the rings, obtaining a steam generator with tubes, initiating an eddy current signal into the tubes of the steam generator, detecting and recording reflections of the eddy current signal initiated into the tubes of the steam generator, and determining a thickness of the deposits in the steam generator from the recorded reflections of the eddy current signal and the polynomial equation.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
G05B 19/418 - Total factory control, i.e. centrally controlling a plurality of machines, e.g. direct or distributed numerical control [DNC], flexible manufacturing systems [FMS], integrated manufacturing systems [IMS] or computer integrated manufacturing [CIM]
A method for determining the inner surface and an outer surface of a sample flake through the use of digital image photography wherein digital images of a sample flake first surface and second surface are compared to a known inner surface and known outer surface of a reference flake.
The invention (2) relates to a device (2) that is simply constructed and displays a high fail safeness for using heated thermoelements (HT) and unheated thermoelements (UHT) as transmitters for measuring a liquid in a liquid container, in particular in the reactor container (4) of a nuclear installation, makes an especially precise and reliable measurement of the level height (H) possible. In addition, the invention proposes a plurality of elongated measuring tubes (6), at a distance from one another, wherein each measuring tube (6) exhibits a number of thermoelements (HT, UHT) arranged such that they are distributed in a longitudinal direction, and wherein an unheated thermoelement (UHT), acting as a reference transmitter, is allocated to a thermoelement (HT), acting as a primary transmitter, arranged at a first measuring tube (6) and heated by a heating element (HE) and which is arranged at a measuring tube (6) that is different from the first measuring tube (6).
G01F 23/22 - Indicating or measuring liquid level or level of fluent solid material, e.g. indicating in terms of volume or indicating by means of an alarm by measuring physical variables, other than linear dimensions, pressure or weight, dependent on the level to be measured, e.g. by difference of heat transfer of steam or water
G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
87.
NUCLEAR ENGINEERING PLANT AND CLOSURE APPARATUS FOR ITS CONTAINMENT
A nuclear engineering plant (4) with a containment (2), whose interior chamber (14) is subdivided by means of a gas-tight partition wall (16) into a systems chamber (18) containing the reactor pressure vessel (8) and the primary coolant circuit (10) and into an operating chamber (20) which is accessible during normal operation, is intended to ensure a particularly high operational reliability, in particular also in incident situations, in which hydrogen is released in the systems chamber (18), with production and operational outlay being kept low. For this purpose, the invention provides a number of overflow openings (22a, 22b, 22c) in the partition wall (16), wherein the respective overflow opening (22a, 22b, 22c) is closed by means of a closure element (24) of a closure apparatus (26) which opens automatically when a trigger condition associated with the respective overflow opening (22a, 22b, 22c) is reached, and wherein closure apparatuses (26) which open both as a function of pressure and independently of pressure are provided. A closure apparatus (26) which is particularly expedient in this connection furthermore has a closure element (24) comprising a bursting film (50) or a bursting diaphragm, wherein the closure apparatus (26) is designed such that it frees the overflow opening (22a, 22b, 22c) automatically when a predetermined environment-side trigger temperature is reached.
F16K 17/14 - Safety valvesEqualising valves opening on surplus pressure on one sideSafety valvesEqualising valves closing on insufficient pressure on one side with fracturing member
F16K 17/38 - Safety valvesEqualising valves actuated in consequence of extraneous circumstances, e.g. shock, change of position of excessive temperature
In a combustion element for a light-water reactor, having a large number of fuel rods which are guided in a plurality of axially spaced-apart spacers (4), which each form a grid comprising grid webs (14, 16) with a large number of cells (6), which are arranged in rows (10) and columns (8), a peripheral web (16) of a spacer (4) has a cutout (40) introduced after damage in the region of one cell (6). A repair sleeve (20) with an outer face protruding into the cutout (40) is inserted into the cell (6) and runs flush with the undamaged outer face of the spacer (4). A combustion element damaged in the region of the spacer can thus be repaired and reused.
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
A fuel assembly for a pressurized water nuclear reactor contains a multiplicity of fuel rods which are guided in a plurality of axially spaced spacers (4) which in each case form a square grid, composed of grid webs (141-17, 161-17), with a multiplicity of cells (6) arranged in rows (10) and columns (8). In each case one control rod guide tube (12) is guided through a number of these cells (6). At least one spacer (4) is designed to be mechanically stronger in a first partial region (18) than in a second partial region (19). In this second partial region (19), the spacer (4) is provided with at least one resisting element (20) which protrudes into a flow sub-channel (17) formed between the fuel rods (2) and increases the flow resistance, the resisting element counteracting a reduction, associated with the mechanically weaker design, in the flow resistance in the second partial region (19) and in this manner effecting a homogenization of the hydraulic behaviour of a spacer which is mechanically inhomogeneous on account of the varying mechanical design.
The invention relates to a method and device for detecting the position of a pulsed mechanical effect on a plant component (2), whereby an operating noise in the plant component (2) is continuously recorded by means of a plurality of sensors (4s) which are arranged on the plant component (2) and converted by the above into a measured signal (Ms), said measured signals of the sensors undergoing a transformation in a first time window (ﶴt1). A first evaluation function (K1,s) is derived from a plurality of first transformations determined in said manner, said evaluation functions displaying the appearance of the pulsed mechanical effect on a plant component (2). According to the invention, when detecting an effect following the second shorter time window (ﶴt2) having the same algorithms, second transformed and respectively, second evaluation functions (K2,s) are derived, from which, respectively, one time point (ts) is determined, wherein the sound signal produced by the effect impacts upon the sensor (4s). From there, running time differences produced between the sensors (4s) can be exactly reconstructed on the position of the effect.
The invention relates to a method and device for detecting a pulsed mechanical effect on a plant component (2), whereby an operating noise in the plant component (2) is continuously recorded by means of a sensor (4) which is arranged on the plant component (2) and converted by the above into a measured signal (M) which is subjected to a mathematical transformation. According to the invention, a sliding mean is calculated by determining the quantiles from the determined transformations and the evaluation function (K(t)), which displays the application of a pulsed mechanical effect on the plant component (2), is derived from said sliding mean.
The device (2) according to the invention contains a feed line (4) for cleaning agent under high pressure, a nozzle carrier (6) which is connected to said feed line in a rotationally fixed manner and on which at least one cleaning nozzle (44b-e) and at least one counter nozzle are arranged, wherein the counter nozzles (46) are oriented (62a) and dimensioned in such a manner that recoil forces (Fr, Fc) accumulate to form a transverse force (Fges) in the direction of the cleaning nozzles (44b-e), and a conically tapering, eccentric point (58) at the opposite peripheral position of the transverse force (Fges). In the method according to the invention, the nozzle carrier (6) is rotated (88), the angle of rotation of the nozzle carrier (6) is detected, the retraction length of the nozzle carrier (6) is detected, the nozzle carrier (6) is introduced into the pipe system (80) as far as the junction (96), and the nozzle carrier rotated (88) in such a manner that the transverse force (Fges) points in the direction of the partial pipe (98a, b) to be approached, and the nozzle carrier (6) is advanced (82) further without rotating (88) until the nozzle carrier (6) is located in the partial pipe (98a, b).
B08B 9/043 - Cleaning the internal surfacesRemoval of blockages using cleaning devices introduced into and moved along the pipes moved by externally powered mechanical linkage, e.g. pushed or drawn through the pipes
The invention relates to a tool for manually treating a damaged point in the sealing region (8) between a reactor pressure vessel (1) and its cover (3), having the following configuration: it has a basic body (10) provided with at least one handle (12); on one side of the basic body (10) there are two sliding elements (14) which leave a space (19) free between them and serve for supporting on a surface of the sealing region (8) to be treated; the end faces (32) of the sliding elements (14) are of planar design and run in a common flat plane; a cutting means for machining is arranged in the space (19) in such a way that it essentially does not project beyond the end faces (32) of the sliding elements (14).
The invention relates to a device for carrying out a tightness test on at least one fuel rod container containing a fuel rod, comprising the following steps: a) a testing container in which a fuel rod container (5) is arranged in a substantially vertical alignment is provided within the fuel cooling pond (1) of a nuclear power station, b) a cushion of gas (18) is produced above the fuel rod container in the testing container and is connected to a negative pressure source, c) in a region arranged between the cushion of gas (18) and the upper end of the fuel rod container there is a viewing window (12), d) an underwater camera (30) is directed at the viewing window (12) and, in the event of a leak of the fuel rod capsules, the rising gas bubbles are observed.
Disclosed are a method and a device for calibrating a humidity sensor (18), in which a carrier gas having a predefined input steam concentration and a predefined rate of flow (Qv) flows into a mixing chamber (2) communicating with a water reservoir (6) via a steam-permeable partition (4), the temperature (Tw) of the water reservoir (6) being set to a constant value, and steam is admixed to the carrier gas in the mixing chamber (2). A measured value (CM) for the output steam concentration in the steam/carrier gas mixture discharged from the mixing chamber is determined by means of the humidity sensor (18) and is compared to a real value (Cout) of the output steam concentration in the mixture discharged from the mixing chamber (6), said real value (Cout) being defined by a relation depending on the temperature (Tw) of the water reservoir (6), the rate of flow (Qv), and the input steam concentration into the mixing chamber (6).
The invention relates to a method for decontaminating an oxide layer-containing surface of a component or a system of a nuclear facility. According to said method, the oxide layer is treated with a gaseous oxidant.
A feeding system (2) for an absorber liquid (4) containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, having a storage container (6) for the absorber liquid (4) is to be designed for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid (4) or corrosion of the container wall (10) of the storage container (6) is to be excluded. For this purpose, the invention provides for the storage container (6) to be connected to a pressure container (16) via an overflow line, wherein the pressure container (16) is filled with a motive fluid (14).
An apparatus and method to remove uranium from a body of material wherein the method includes the steps of depositing the body of solid material in an ultrasonic extractor and depositing an amount of acid in the ultrasonic extractor. The method also provides for the steps of heating a jacket of the ultrasonic extractor, transporting the body of solid material in the ultrasonic extractor and the amount of acid such that the body of solid material and the acid contact each other inside the heated ultrasonic extractor while the ultrasonic extractor provides ultrasonic energy to both the body of solid material and the amount of acid, wherein the amount of acid strips uranium from the body of solid material. The method further provides for collecting the amount of acid and the body of solid material in the ultrasonic extractor in different positions, transporting the amount of acid with the stripped uranium to an extraction mixer settler, and settling uranium product from the extraction mixer settler.
A method for checking the functional equivalence of two pieces of software for control systems, in particular control systems based on programmable logic controllers, engineered from function block diagrams (120, 196) with a plurality of interconnected function blocks, the method comprising the following steps: a) generating a first instance of compilable source code from a first plurality of function block diagrams (70, 74, 78) by virtue of a first code generator obeying to a first set of sequentialization rules; b) parsing said first instance of source code and identifying the set of all function block I/O ports and junction points (nodes) and the set of all connections between them (arcs) of the data flow; c) using the information from step b) to reconstruct a first data flow graph (86) of said first instance of source code in the upstream direction; d) generating a second instance of compilable source code from a second plurality of function block diagrams (90, 94, 98) by virtue of the same or a second code generator obeying a second set of sequentialization rules; e) parsing said second instance of source code and identifying the set of all function block I/O ports and junction points (nodes) and the set of all connections between them (arcs) of the data flow; f) using the information from step e) to reconstruct a second data flow graph (106) of said second instance of source code in the upstream direction; g) comparing said first identified data flow graph (86) and said second identified data flow graph (106) with each other, node by node, and thereby checking if the transitive closure of input up to that node in said second instance of source code is the same as in said first instance of source code.
A nickel, chromium, iron alloy and method for use in producing weld deposits and weldments formed therefrom. The alloy comprises, in weight percent, about 28.5 to 31.0% chromium; about 0 to 16% iron, preferably 7.0 to 10.5% iron, less than about 1.0% manganese, preferably 0.05 to 0.35% manganese; about 2.1 to 4.0% niobium plus tantalum, preferably 2.1 to 3.5% niobium plus tantalum, and more preferably 2.2 to 2.8% niobium plus tantalum; 0 to 7.0% molybdenum, preferably 1.0 to 6.5%, and more preferably 3.0 to 5.0% molybdenum; less than 0.50% silicon, preferably 0.05 to 0.30% silicon; 0.01 to 0.35% titanium; 0 to 0.25% aluminum; less than 1.0% copper; less than 1.0% tungsten; less than 0.5% cobalt; less than about 0.10% zirconium; less than about 0.01% sulfur; less than 0.01% boron, preferably less than 0.0015% boron, and more preferably less than 0.001% boron; less than 0.03% carbon; less than about 0.02% phosphorous; 0.002 to 0.015% magnesium plus calcium; and balance nickel and incidental impurities. The method includes the steps of forming a welding electrode from the above alloy composition and melting the electrode to form a weld deposit. A preferred weldment made from the alloy and method includes an alloy substrate in the form of a tubesheet of a nuclear reactor.
patents
