Abstract: A measuring tool (10) comprising two jaws (15b, 15c) which can be moved one with respect to the other in a first horizontal direction and moved back towards each other by an elastic means, is set up level with the spacer-grid (3) of the fuel assembly. The measuring tool (10) is displaced in a second horizontal direction perpendicular to the first horizontal direction and to a front face (6a) of the spacer-grid (3) such that the jaws (15b, 15c) each come into contact with a lateral face (6b, 6c) of the spacer-grid (3) and the distance between the jaws (15b, 15c) is measured in the first horizontal direction. The tool (10) is mounted so as to rotate freely about an axis parallel to the vertical longitudinal direction of the fuel assembly (1) such that the measuring tool (10) is oriented freely with respect to the front face (6a) of the spacer-grid (3), when the measuring tool is displaced in the second horizontal direction.

Abstract: A method and an arrangement for sensing the clearance between an object and an object-adjacent surface, which is particularly for use in difficult ambient conditions, such as in dirty, corrosive environments and under varying temperature and pressure conditions, e.g. to measure and hold constant the clearance between the sector plates (3, 4) of a regenerative rotary air preheater and the end surfaces of the rotor (2, 8). The sensing device (7) includes a compressed-air operated pipe (9) which is mounted on the object (3, 4) and which includes a sound-emitting opening (14) located adjacent the surface (8), wherein changes in the distance (S) between the object (3, 4) and the surface (8) are represented by changes that occur in the resonant frequency of the pipe (9).

Abstract: The invention relates to a sensor for a measuring an electrochemical corrosion potential comprising a sensor tip, a conductor electrically connected to the sensor tip, an insulating member which surrounds the conductor, a connecting member which surrounds the conductor; and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member.

Abstract: A reactor stud cleaning apparatus is presented for cleaning studs, nuts, and washers used to secure reactor heads on nuclear reactors. The cleaning apparatus is a self-contained unit having a housing that includes two independent sealable compartments allowing cleaning of two studs simultaneously. Each sealable compartment has a topside covering with a port that allows reactor studs to be lowered into the apparatus keeping the longitudinal axis of the reactor stud maintained in a substantially vertical position. Inside the sealable compartments is a turn table for vertically mounting a reactor stud and for rotating the stud about its longitudinal axis. Brushes, rotatably mounted inside the compartment, are used to clean the reactor stud. The apparatus also has a cleaning fluid circulation system for circulating solvents or other cleaning agents or fluids over the reactor stud during cleaning.

Abstract: Compliance with administrative limits on cumulative exposure of control rod groups in the reactor core, is monitored by computing the incremental effective exposure for each group commensurate with core power, for each time increment at which each group is within the position range where an administrative limit is imposed. The increments of effective exposures for each group are accumulated, and the accumulated effective exposure for each group is compared with the administrative limit to each group. This comparison is then displayed to the reactor operator, preferably using either a “rolling wheel” or “sliding bar” format.

Abstract: A nuclear reactor instrumentation system including a plurality of in core nuclear instrumentation assemblies arranged in a gap between a number of fuel assemblies charged in a reactor core. The income nuclear instrumentation assemblies each include a fixed type neutron detector assembly having a plurality of fixed type neutron detectors dispersively arranged in a core axial direction and a fixed type gamma thermometer assembly having a plurality of fixed type gamma ray heat detectors arranged at least in a same core axial direction as the fixed type neutron detectors. A power range detector signal processing device is operatively connected to the fixed type neutron detector assemblies through signal cables. A gamma thermometer signal processing device is operatively connected to the fixed type gamma thermometer assembly of the in core nuclear instrumentation assembly through a signal cable.

Abstract: Scan methods of inspecting piping and welds of a pipe elbow in a reactor pressure vessel of a boiling water reactor are disclosed. The scan apparatus includes a motor connected to a scan head having two ultrasonic transducer probes each including an ultrasonic transducer, a transducer arm connected to each transducer probe, a scan platform having an arcuate cutout, and a connector connecting each transducer arm to the scan platform. The scan head allows the transducer probes to remain substantially in contact with the pipe elbow while the scan head traverses the pipe elbow. The scan apparatus further includes a pivot arm connecting the scan platform to a pivot pin. The method includes the steps of positioning the scan head and the transducer probe such that the ultrasonic transducer is substantially in contact with the pipe elbow surface. Then moving the scan head axially along the pipe elbow in a first direction while maintaining the ultrasonic transducer in substantial contact with the pipe elbow surface.

Abstract: A maintenance of a suppression pool or spent fuel storage pool in a reactor primary containment vessel of a nuclear power plant includes a cleaning step of a water in the suppression pool and a decontamination step of an interior of the suppression pool, and in the maintenance, the coated film applied on an inner surface of a wall portion of the suppression pool is inspected. These cleaning, decontaminating and inspecting steps are performed by divers while maintaining a water level in the suppression pool.

Abstract: An improved heated junction thermocouple cable arrangement for use with a nuclear reactor heated junction thermocouple having sensors with an unheated thermocouple, a heated thermocouple, and a heater coil, shares power conductors between multiple heater coils. The thermocouple conductors are separated from the power conductors to eliminate the possibility of interference with the thermocouple conductor signals. The heater coils are ganged in parallel and if one heater coil fails, power is still supplied to the remaining heater coils connected to a pair of power conductors.

Abstract: A valve control apparatus for generating an anticipatory output response for controlling a fluid flow in a fluid flow line includes a valve provided inside the fluid flow line, at least one sensor for providing at least one input signal, said at least one input signal responsive to properties of said fluid flow in advance of said valve, and a letdown back-pressure controller receiving said at least one input signal and providing a tunable output signal for selectively opening and closing said valve to vary flow in the flow line responsive to said at least one input signal.

Abstract: An apparatus for ultrasonically examining a weld in a nuclear steam supply system includes an elongated guide rod for positioning within a pipe of the nuclear steam supply system. An ultrasonic transducer is positioned at an end of the elongated guide rod. A collapsible shoe encloses the ultrasonic transducer. The collapsible shoe includes a biasing mechanism to allow the collapsible shoe to pass through the pipe while the pipe is at a first circumference and while the pipe is at a second circumference. The collapsible shoe continuously contacts the pipe to establish ultrasonic coupling for the ultrasonic transducer.

Abstract: A method of controlling the system pressure in a power generating system, having a turbine-generator and a BWR, that modulates the core thermal power of the reactor while maintaining the main turbine control valves in a constant steady position is described. The core thermal power may be adjusted by adjusting the control rod density within the reactor core or by adjusting the flow rate through the reactor which may be accomplished by modulating the speed of variable frequency recirculation pumps or by modulating recirculation flow control valves. The method includes transferring the power generation system from normal turbine control valve modulation pressure control to core thermal power modulation pressure control. Additionally the method includes modifying the bypass valve closure bias and the power control bias to accommodate the variances from core power modulation pressure control over normal pressure control.

Abstract: A main steam pressure disturbance preventing apparatus of a nuclear power plant including, one of a pressure detector and a water level detector provided in one of a steam system from a reactor of the nuclear power plant and a drain system connected to the steam system, an instrumentation pipe connected between one of the pressure detector and the water level detector and one of the steam system and the drain system, and catalyst installed in one of the pressure detector, the water level detector and the instrumentation pipe for recombining hydrogen and oxygen.

Abstract: The invention relates to a sensor for a measuring an electrochemical corrosion potential comprising a sensor tip, a conductor electrically connected to the sensor tip, an insulating member which surrounds the conductor, a connecting member which surrounds the conductor; and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member.

Abstract: A method is provided for determining the closeness to criticality of a nuclear reactor during start-up, comprising the steps of completing a control rod withdrawal step, thereby generating a change in an output signal of a neutron detector; measuring the output signal after the completion of the control rod withdrawal step and during a transient portion of the output signal; calculating the effective neutron multiplication factor (Keff) based upon the measured output signal and elapsed time between output signal measurements; and determining the closeness to criticality of the nuclear reactor based upon the calculated value of the effective neutron multiplication factor (Keff). The invention also encompasses apparatus used to perform the above method.

Abstract: A method is provided for performing nondestructive examination of a metal specimen using neutron activated positron annihilation wherein the positron emitter source is formed within the metal specimen. The method permits in situ nondestructive examination and has the advantage of being capable of performing bulk analysis to determine embrittlement, fatigue and dislocation within a metal specimen.

Abstract: A radiation detecting apparatus comprises a container having a vacuum space, a radiation detecting element disposed in the vacuum space for detecting radiation, and a cooling element disposed in the container for cooling the radiation detecting element to a preselected cooling temperature during operation of the radiation detecting apparatus. A physical absorbing material is disposed in the container for absorbing gases from the vacuum space. A chemical absorbing material capable of activation by heat is disposed in the container for absorbing gases from the vacuum space. A heating device is provided for heating the chemical absorbing material. A measurement and control circuit measures a temperature of the radiation detecting element or a degree of vacuum in the vacuum space to determine whether or not the chemical absorbing material should be activated by the heating device.

Abstract: A device for inspecting the interior of steam generators capable of visually inspecting interior of tubes in steam generators, including upper portions steam generator tubes, tops and bottoms of support plates, wrapper-to-support plate welds, and other internal structures.

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

Abstract: A method for detecting the dropping of at least one control element in a reactor core of a power station by delaying signals output by detectors disposed along a fall path of the control element in such a way that they are approximately simultaneous. This improves a registration of the dropping of the control element by improving a ratio between a useful signal and a noise signal.

Abstract: A method and an apparatus introduce a self-propelled in-pipe manipulator into the interior of a pipeline. A hollow body receives the in-pipe manipulator, is open at least at one end surface and is positioned at an opening of the pipeline, from where the in-pipe manipulator can drive on its own into the pipeline.

Abstract: The spring force measurement system includes a draw rod standard, an insertion tube and an extension tube, together with a cable received through the tubes connecting the standard and a load cell, in turn connected by the cable to a winch. The draw rod standard is located in the fuel bundle below the spacer having a spring whose spring force is to be measured. By drawing the standard through the spacer cell, the spring force is computed using the measured tension in the connecting cable as the standard is withdrawn and the known values of the coefficient of friction between the spacer spring projection and the draw rod standard and the combined weight of the cable and draw rod standard. The measuring process is repeated for next higher spacers.

Abstract: An underwater inspection/repair apparatus comprises a sealing device provided around an opening portion of a watertight vessel, a pushing mechanism provided to the watertight vessel, a water discharge pump provided to the watertight vessel for discharging water in an inside of the watertight vessel, and a compressed air supplying device for supplying a compressed air into the inside of the watertight vessel. A top end of the sealing device is pushed against an inner wall surface of the reactor vessel by a reaction force generated when the pushing member of the pushing mechanism is pushed against the reactor internal structure, so that the inside of the watertight vessel can be isolated in a watertight manner. The pneumatic water discharge pump includes a pneumatic pressure cylinder driven by a pneumatic pressure, and a water discharge cylinder cooperated with the pneumatic pressure cylinder.

Abstract: A temperature measurement system includes, for each of two nuclear steam supply system (NSSS) loops, four temperature sensors each of which provides a value of the temperature of reactor coolant flowing in the NSSS hot leg of the loop; a mechanism for determining, for the hot leg, a first bias value associated with a first pair of the temperature sensors and a second bias value associated with a second pair of the temperature sensors from the value of the temperature of the reactor coolant of each of the temperature sensors; a mechanism for averaging the first bias value and the values of the first pair of the temperature sensors to determine a first average temperature therefrom; and a mechanism for averaging the second bias value and the values of the second pair of the temperature sensors to determine a second average temperature therefrom, whereby two separate average temperatures of the temperature of the reactor coolant are determined for each of the two loops.

Abstract: A preventive maintenance apparatus for structural members inside a nuclear reactor pressure vessel includes a ring-shaped guide rail having a plurality of lugs which is placed on an upper flange of a core shroud provided inside a reactor pressure vessel. At least some of lugs separately engage with a plurality of guide rods provided on an inner surface of the reactor pressure vessel. A turntable is rotated on the guide rail. A first discharging nozzle moving apparatus placed on the turntable moves a first discharging nozzle for adding compressive remaining stress to an outer surface of the core shroud in a radial direction of the core shroud and in an axial direction of the core shroud. A second discharging nozzle moving apparatus placed on the turntable moves a second discharging nozzle for adding compressive remaining stress to an inner surface of the core shroud in a radial direction of the core shroud and in an axial direction of the core shroud.

Abstract: The present invention concerns a monitor for measuring both the gamma spectrum and neutrons emitted by an object such as a spent nuclear fuel pin or pin assembly or nuclear waste material. According to the invention, it comprises a lead block (1) presenting a front face (8) intended to be brought close to said pin (7) or assembly to be measured and incorporatinga gamma detector (2) for gamma spectroscopy located close to a rear face of said block and associated with a collimator (4) extending from said front face to said gamma detector,two neutron detectors (3) which extend parallelly to each other and to said front face and are disposed symmetrically on either side of the axis of said collimator (4) close to said front face (8).

Abstract: The electromagnetic drive mechanism for the control rods of a nuclear reactor has a stationary gripper and coil, a movable gripper and coil, and a lifting armature and coil for moving the movable gripper to advance or retract the control rods in steps. A coil current driver responsive to a controller provides currents to the coils individually and in combinations. During the operations, coil current signals are sensed, sampled, digitized and processed to generate coil current data such as amplitude as a function of time. The measured coil current data is compared to nominal current data as well as historical current data, and the historical data is updated or appended to include the measured data. The comparison includes a check for correct timing relationships, such as the timing of the current notch occurring due to increased inductance upon pull-in of a gripper.

Abstract: Automated inspection assemblies for scanning pipes of a nuclear reactor are described. In one embodiment, the assembly includes a mounting subassembly and a scanning subassembly. The mounting subassembly includes a clamp configured to be mounted to selected pipes in a nuclear reactor pressure vessel, and the scanning subassembly is movably coupled to the mounting subassembly. The scanning assembly includes a scanning head configured to scan at least a portion of the circumference of the pipe to be inspected, and the scanning head includes a substantially "U" shaped transducer support assembly sized to receive the pipe. Transducer elements are coupled to the transducer support assembly legs, and the transducer support assembly is configured to rotate about the pipe to inspect the pipe.

Abstract: An improved test for determining densification characteristics of nuclear fuel including a fissionable ceramic material is disclosed. The test includes controlling the testing atmosphere to impede loss through vaporization of silica and in turn density reduction due to compositional changes.

Abstract: The present invention, in one aspect, is a method for identifying an optimum core loading arrangement. The method generally has an initialization phase and a running, or search, phase. In the initialization phase, an initial core loading arrangement is identified based on the relative reactivity levels of the bundles to be loaded and the reactor core locations. Once the initial core loading arrangement is identified, such arrangement is then optimized, within the defined constraints, in the running phase. More specifically, in the running phase, each core location is analyzed to determine whether such core location reactivity level can be changed from the initial reactivity level to either satisfy a constraint or optimize cycle energy, or both.

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

Abstract: Determined results of which measured data of a plant is normal/abnormal are mapped to a plant model in an abstract function level hierarchically represented with flow structures of mass/energy and goals to be accomplished by the respective flow structures so as to detect a failure propagation network. In the failure propagation network, a flow structure at the lowest hierarchical level of a network with a top goal of the highest priority is selected as an object to be diagnosed first. In other words, flow structures that are functionally important and close to the origin of anomaly are preferentially diagnosed. The diagnosis for flow structures detects abnormal elements through hypothesis and test. The abnormal element are evaluated with respective mass/energy balances or the like. With respect to the evaluated result, the origin of failure is inferred based on the rule of experience. When the abnormal element has a detailed flow structure, it is diagnosed.

Abstract: An automated system for identifying nuclear power plant components includes a camera which provides an input signal from an image of the nuclear power plant components. A digitizer generates a digitized image of the nuclear power plant components from the input signal. A workstation locator routine locates the component identifier of the nuclear power plant components from pixel elements of the digitized image and a workstation determining routine determines the component identifier. The determining routine includes plural recognizer routines each having an output providing an intermediate recognition of the component identifier, and a combining routine combining the outputs of the recognizer routines to recognize the component identifier. Each of the outputs includes an intermediate identifier and a corresponding confidence value.

Abstract: In a method of measuring the contamination profile of the inner surfaces of a radioactively contaminated pipe embedded within a concrete structure in a nuclear reactor facility, the detecting surface of a detector to be inserted into the contaminated pipe is fitted with a movable shield plate or cover, which is moved by a specified distance and the radioactive contamination profile within a limited area of the inner surfaces of the pipe is measured in terms of the difference in radiation intensity measured in two states, one where the shield plate or cover is in contact with the detecting surface and the other where it is spaced from the latter.

Abstract: The apparatus is for internal inspection of a guide tube of a nuclear fuel assembly and measurement of deformations of a zone of smaller diameter in a terminal portion of the tube. The apparatus has a stick whose distal portion presents, in succession, starting from a distal end of the stick-shaped means, a guiding swelling having an easy fit in the zone of smaller diameter and two projections. The first projection carries two sets of diametrically opposite strain sensors provided with conductors linking them to a proximal end of the stick. A sleeve slides in a running portion of the guide tube above the zone of smaller diameter and slides over the stick. The sleeve and said stick-shaped means have mutually cooperating abutments for limiting displacement of the sleeve towards the distal portion to a position where the sleeve surrounds the stick.

Abstract: An electrochemical corrosion potential sensor includes a sensor tip electrically joined to a conductor, and a ceramic insulator joined to the tip around the conductor. A sleeve is joined to the insulator around the conductor, and is electrically insulated from the tip by the ceramic insulator. The insulator has an exposed surface axially separating the tip and sleeve, and a ceramic coating is bonded thereto for preventing dissolution of the insulator by reactor water. In a preferred embodiment, the ceramic insulator is sapphire, and the ceramic coating is yttria-stabilized-zirconia or magnesia-stabilized-zirconia which may be plasma sprayed over the insulator.

Abstract: The apparatus for inspecting the rods of a cluster insertable in the core of a nuclear reactor comprises a body provided with sliding guide means for rods in m different orientations of the cluster, m being a submultiple of the number n of rods in the cluster, and a measurement jig carrying n/m heads (16) for eddy current examination and n/m probes (18) for ultrasound scanning placed on different guide paths. Each ultrasound probe (18) is constituted by an annular support surrounding a rod path and carrying a ring of individual transducers. The transducers are connected to a circuit for sequentially exciting the transducers in successive groups, causing the tube to be scanned circumferentially, and for electronic focusing towards the surface of a tube travelling on the path.

Abstract: Fluid flow rates are determined with a nonradioactive tracer agent in the operation of a water cooled nuclear fission reactor employed for driving electrical power generators. The technique is proposed for use to routinely calibrate other flow measuring means such as restricted flow nozzles.

Abstract: A sealing device comprising a supporting assembly (9) containing sealing members (16) supporting the instrument column (5), and an assembly (17, 18) for lifting the instrument column (5) in sealing engagement with the supporting members (16). The sealing members (16) supporting the instrument column (5) include a clamping ring (28) and at least one O-ring seal (29, 30). Axial movement of the instrument column (5) enables clamping of the inner seal (30) via the clamping ring (28). When axial movement of the instrument column (5) is restricted, the seals may be clamped by means of a suitably sized clamping ring (28).

Abstract: The measurement device includes a means (18) for leaktight passage of a measurement conduit (20), produced in the form of a thimble, and an element (25) for connecting the end of the measurement conduit (20) to a smooth cylindrical extension element. The connection element (25) is produced in the form of a sleeve (26), the internal bore of which includes a first part (35) for receiving the measurement conduit (20) and a second part (36) for receiving the smooth cylindrical extension element. The connection element (25) includes, at the ends of the sleeve (26), a first removable means (28) for fastening and leaktight passage of the measurement conduit (20), and a second removable means (30) for fastening and leaktight passage of the smooth cylindrical extension element. The invention applies both to mobile-probe devices for measuring in the core of the reactor and to fixed-probe measurement devices.

Abstract: An in-core monitor protector intended for use during in-vessel services while work is being performed in the reactor core region of a BWR. The purpose of the protector is to prevent damage to or displacement of the in-core monitors, which could be caused by the installation or removal of various types of equipment through the top guide of the core during an outage. This protection is accomplished by partly surrounding the in-core monitor with the protector. The primary component is an aluminum channel of such size as to surround the monitor on three sides and of such length as to extend from the bottom of the top guide to the top of the fuel support castings. The closed sides of the channel prevent direct contact between servicing tools and the monitor. A top subassembly is used to secure the protector to the top guide. A bottom subassembly is used to secure the protector to the fuel support castings.

Abstract: A system diagnoses the operability of a rotating electrical apparatus such as a reactor coolant pump (RCP) motor. The RCP motor includes a rotatable shaft, a stator, a bearing assembly and a bearing insulator. The bearing assembly electrically contacts the rotatable shaft and the bearing insulator at least partially electrically insulates the rotatable shaft and the bearing assembly from the stator. The system includes a shaft transformer and voltage source for inducing a voltage in the rotatable shaft; a variable resistor in parallel with the bearing insulator having first and second resistances; an analog-to-digital converter for sensing a first voltage associated with the first resistance and a second voltage associated with the second resistance; and means for determining the operability or the impedance of the bearing insulator from the first and second voltages and the second resistance.

Abstract: A control rod mechanism for transmitting the rotation of an electric motor via a drive shaft to an elevator apparatus for raising and lowering a hollow piston on an upper end thereof, which is provided a control rod that controls the output of a nuclear reactor. The control rod mechanism is used to insert the control rod into a reactor core or remove it therefrom, and also to rapidly insert the control rod into the reactor core by the injection of hot water to force the hollow piston upward. The control rod mechanism includes a magnetic coupling having an inner magnet, which is divided into a plurality of parts and is disposed at a lower portion of the drive shaft in order to transmit the rotational power of the electric motor to the drive shaft. The control rod mechanism also includes a drive-side outer magnet, which is divided into a plurality of parts and is provided on a rotary shaft of the electric motor in an arrangement on an outer side of the inner magnet.

Abstract: A device for carrying out a surveillance program to monitor the environmental effect of radiation and high-temperature oxygenated water on the shroud material in an operating reactor. The device is a holder containing shroud material specimens and encapsulated neutron flux monitors. The holder is installed on the shroud wall and left in place for a predetermined period of reactor operation time for the purpose of monitoring the neutron fluence effect on the shroud material specimens housed in the holder. The specimen holder is later removed from the reactor. The specimens are removed from the holder and examined in a laboratory to determine any changes in material properties. The specimen holder comprises a bolt-like device having a hollow head and a threaded shaft. The hollow head provides space for material specimens and neutron flux monitors. The specimen holder is attached to the shroud wall by screwing the threaded shaft into a threaded hole in shroud wall.

Abstract: A system and method of the present invention measures the relative stress in the cladding of a nuclear fuel rod caused by the internal gas pressure of the rod. This is done by determining the attenuation ratio of two specific frequencies from a broadband low frequency acoustic energy spectrum. Relative pressure differentials are measurable and therefore any loss of fuel rod gas pressure is a direct indication of fuel rod cladding failure.

Abstract: The present invention, in one aspect, is a method for identifying an optimum control blade positioning arrangement. The method generally has an initialization phase and a running, or search, phase. In the initialization phase, an initial control blade positioning arrangement is identified. Once the initial control blade positioning arrangement is identified, such arrangement is then optimized, within the defined constraints, in the running phase. More specifically, in the running phase, each control blade position is analyzed to determine whether the control blade position can be changed from the initial position to either satisfy a constraint or optimize cycle energy, or both.

Abstract: In an apparatus for operating and monitoring a plant, a manipulating device provides the plant with manipulation signals for operating the plant. In response to information supplied from the plant and a manipulating signal of the manipulating device, a first display data producing device produces first display data needed for the operation of the plant changing every moment. On the basis of data supplied from the first display data producing device, a first display device displays information relating to the operation of the plant. In response to information supplied from the plant and a manipulating signal of the manipulating device, a second display data producing device produces data needed for monitoring the plant. On the basis of data supplied from the second display data producing device, a second display device displays information relating to the monitoring of the plant on a larger display screen than that of the first display device.

Abstract: A probe inspects steam generator tubing for defects. The probe includes a transducer which generates a localized ultrasonic Lamb wave. The ultrasonic wave is transferred to the tubing by a coupling medium, such as water, that physically couples the transducer and the tubing. Defects in the tubing reflect the ultrasonic wave to the probe which detects the reflections. The results are then used to determine the length and depth of such defects as cracks, pitting, and thinning. The localized ultrasonic wave performs an inspection sensitive enough to detect ligaments between crack segments. This allows highly accurate predictions of tubing integrity and rupture strength.

Abstract: This invention provides a device for capacitively measuring the distance to an object, which comprises an electrode that will couple capacitively with the object, a shield that surrounds the electrode and is electrically isolated from the electrode by means of insulation, and a layer of insulation that surrounds the shield, wherein the insulation between the electrode and the shield, and the insulation that surrounds the shield, are formed by deposition.This invention further provides a method for measuring the distance to a solid object, which comprises forming an electrode that will capacitively couple with the object, depositing a layer of insulation over at least part of the electrode so that the insulation surrounds the electrode, depositing a layer of metal over the insulation to form a shield, and depositing a second layer of insulation over the shield.

Abstract: In an apparatus for operating and monitoring a plant, a manipulating device provides the plant with manipulation signals for operating the plant. In response to information supplied from the plant and a manipulating signal of the manipulating device, a first display data producing device produces first display data needed for the operation of the plant changing every moment. On the basis of data supplied from the first display data producing device, a first display device displays information relating to the operation of the plant. In response to information supplied from the plant and a manipulating signal of the manipulating device, a second display data producing device produces data needed for monitoring the plant. On the basis of data supplied from the second display data producing device, a second display device displays information relating to the monitoring of the plant on a larger display screen than that of the first display device.