Abstract: The present invention provides a nuclear fuel assembly, where a boron-containing compound is used as a burnable poison and is distributed in a majority of the rods in the assembly. The assembly comprises a plurality of fuel rods, each fuel rod containing a plurality of nuclear fuel pellets, wherein at least one fuel pellet in more than 50% of the fuel rods in the fuel assembly comprises a sintered admixture of a metal oxide, metal carbide or metal nitride and a boron-containing compound.

Abstract: A method of operating a nuclear power plant includes determining and licensing a maximum power level at which the power plant can be operated subsequent to the beginning of a fuel cycle, and with the power plant being operated at less than its maximum power rating at certain times such as at the beginning of a fuel cycle. The maximum power level is greater than the power level that would be calculated based upon an assumption that the heat flux peaking factor (FQ) and enthalpy rise peaking factor (F?H) remain at their maximum level throughout an entire fuel cycle. The maximum power rating takes advantage of factors such as known reductions in FQ and F?H at certain points in the fuel cycle, the marginal additional capacity of the Balance Of Plant, and the occasional optimization of process parameters such as ultimate heat sink temperature and atmospheric conditions.

Abstract: A method is provided for determining and controlling feed water level in a steam generator of a nuclear power plant, whereby the water level sensor is calibrated in terms of the pressure drop across structural components in the feed water path. The water level differential pressure sensor is calibrated so that maximum water level is indicated as the level corresponding to the upper tap level plus the pressure drop across the foregoing structural components at maximum power plant power.

Abstract: A reactor pressure vessel supports the flanges of a core barrel assembly and an upper support plate. These flanges have aligned holes for diverting inlet coolant water to the vessel head for cooling the head. The aligned holes have different cross-sectional areas for controlling unexpected backflow of coolant water from the vessel head around the periphery of the upper support plate.

Abstract: A reactor service manipulator delivery system supported on a remotely operated trolley that travels on top of a track attached to the steam dam of a BWR shroud. The trolley has a mast and pulley system designed to manipulate a TV camera for inspection of various reactor components inboard and outboard of the shroud at the same time that fuel is being moved within the core during a refueling operation so that refueling can be performed in parallel with the inspection.

Abstract: A single page enterprise wide level display provides a comprehensive readily understood representation of the overall health status of a complex plant. Color coded failure domains allow rapid intuitive recognition of component failure status. A three-tier hierarchy of displays provide details on the health status of the components and systems displayed on the enterprise wide level display in a manner that supports a logical drill down to the health status of sub-components on Tier 1 to expected faults of the sub-components on Tier 2 to specific information relative to expected sub-component failures on Tier 3.

Abstract: A system for controlling a complex process in which a supervisory sequential controller interface provides automatic sequencing of a plurality of sequential steps of a procedure for operating the complex process, and a display in dynamic communication with the supervisory sequential controller interface and responsive thereto, displays an overview of the status of progression through the sequential steps relative to the entire process. The display provides a global view of the status of the procedure identifying the state of substantially each criterion that needs to be satisfied to advance to the next sequential step and providing verification that substantially each command to be given upon satisfaction of the criterion, has been executed.

Abstract: A nuclear fuel assembly having a parallel array of elongated fuel elements supported between an upper and lower nozzle. Main support grids are substantially evenly spaced along the elongated dimension of the assembly to maintain the spacing between fuel elements. A plurality of auxiliary vibration-resistant grids respectively positioned between the main support grids in the middle third elongated dimension of the fuel elements, that are smaller in height and provide a larger fuel element contact area than the main support grids.

Abstract: A method of inspecting a heat exchanger including a plurality of tubes having ends fixed in a tubesheet includes the steps of: establishing a baseline for each tube of its location in the tubesheet and its unique eddy current pattern resulting, at least in part, from fixing the tube end in the tubesheet; inserting an eddy current probe into a first identified tube; obtaining eddy current data for the first identified tube in the region of the tubesheet; comparing the eddy current data obtained for the first identified tube with the baseline eddy current data for the tube, in order to verify the correctness of the tube identification; and accepting the obtained eddy current data if the correctness of the tube identification has been verified by the comparison. A computer program product for automatically carrying out the steps of the foregoing method is also disclosed.

Abstract: A reactor coolant system of a pressurized water nuclear reactor with fuel assemblies in the core is simultaneously shutdown and decontaminated in less than 90 hours after the shutdown is initiated. A LOMI or Can-Derem decontamination reagent is added to reactor coolant to reduce iron and nickel in oxidized coatings before removing dissolved hydrogen and reducing the gamma emitting activity of the reactor coolant.

Abstract: A display system in which subsequent failures of plant equipment and plant systems are predicted to occur and in which the probability of failure before a specified date and the probability of failure after a specified date is determined and displayed and in which the calendar date is determined and displayed when the desired probability that the equipment not fail prior to the calendar date is specified. The system includes an Equipment Failure And Degradation Module that determines the remaining equipment/system life; a Probability-of-Failure Predictor Module that determines the probability of the equipment/system failing prior to a specified date and the probability of the equipment/system failing after a specified date; and a Date-of-Failure Predictor Module that determines the calendar date that corresponds to a specified probability that equipment not fail prior to the date.

Abstract: A reduced pressure drop debris filter bottom nozzle for use in a fuel assembly of a nuclear reactor includes a plate member formed with a plurality of first flow holes and a plurality of second flow holes. The first flow holes are oblong in cross section and may be of an oval configuration. The first flow holes each have in cross section a major axis and a minor axis, the major axis being longer than the minor axis. The second flow holes are circular in cross section and may have a diameter substantially equal to the minor axis of the first flow holes. The bottom nozzle also features support ligaments that extend between the first flow holes and that are axially aligned with fuel rods within the fuel assembly.

Abstract: Phenomena in water systems caused by hydraulic transients known as “waterhammer” are suppressed by surge vessels containing crushable, cellular, polymeric materials such as polyethylene, polyurethane and polystyrene foam. The foam may be one piece or in the form of several separate particles contained in nylon netting. A water permeable screen is located between a nozzle connection and the cellular polymeric materials for retaining the materials in the surge vessels as the water flows into and out of the vessels. The foam occupies much of the volume in the surge vessels not occupied by the water and the gas contained in the foam can not readily dissolve into the water.

Abstract: A solid state semiconductor neutron detector that automatically varies its sensitivity to provide a pulsed output over the entire range of operation of a nuclear reactor. The sensitivity is varied by changing the thickness or makeup of a converter layer that emits charged particles to the active region of the semiconductor surface.

Abstract: A method and apparatus for controlling components in a component control system in a nuclear power plant includes an image processing system and display processor (IPS-DP) for issuing an encrypted command for a selected component in a component command system (CCS) in the nuclear power plant; a soft controller for receiving the encrypted command and matching the command with the selected component in the CCS for issuing a control command to the selected component; and control channel gateway for receiving the encrypted command from the IPS-DP and soft controller, deciphering the encrypted command, and, if authenticated, issuing a control command for said selected component in the CCS. Security and isolation, while complying with applicable codes and regulations, are assured.

Abstract: A flexible penetration attachment for BWR wet-well strainers is provided that causes a suction pipe to communicate with liquid in the wet-well through an aperture in a wet-well wall. The penetration includes a boot for defining an outer circumference of the aperture. A suction pipe having an outer surface extends through the aperture such that an annular space is defined between the aperture outer circumference and the suction pipe outer surface. A resilient seal having a generally U-shaped cross section with first and second ends is placed within the annular space. The first end is fixedly secured to the outer surface of the suction pipe and the second end is fixedly secured to the aperture outer circumference such that the suction pipe may move in response to a force applied to the suction pipe.

Abstract: A sleeve assembly for refurbishing a fuel rack having cells in which fresh or spent nuclear fuel assemblies may be stored, in which the cells have elongate rack walls extending from a rack base plate and the rack base plate has flow holes communicating with the cells. The sleeve has at least one elongate wall extending from the topside of a sleeve base having an opposed bottom side. The sleeve base has a flow hole extending therethrough that communicates with one of the rack base plate flow holes. A pin assembly disposed in the sleeve base flow hole has resilient tabs extending beyond the bottom side of the sleeve base for extending into a rack base plate flow hole and resiliently engaging the rack base plate when the sleeve assembly is installed in one of the cells. The pin assembly resists horizontal and vertical movements of the sleeve assembly, permits water flow into the cell and permits sleeve assembly removal tools and inspection devices to access the pin assembly.

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 seal assembly for sealing a space between an annular flange on a nuclear reactor vessel and a surrounding ledge of a refueling canal to provide a temporary water barrier during refueling operations. The seal assembly includes an annular closure plate having an outer portion secured to the surrounding ledge and an inner portion supporting a first sealing surface. A second sealing surface opposing the first sealing surface is formed on or secured to the annular flange of the reactor vessel. An annular space between the first and second sealing surfaces provides a ventilation path from the reactor cavity during normal plant operation. The annular space is sealed by an inflatable seal during refueling operations to provide a water barrier between the refueling canal and the reactor vessel. The inflatable seal is secured to and supported by an annular support structure that straddles the annular space.