Abstract: A tool to slide a channel on a nuclear reactor fuel bundle assembly, the tool includes: a plate having a slot to receive a handle of the fuel bundle and a lower surface that engages an upper edge of the channel; at least one post extends up from the plate, and an arm is attached to a pivot on the post and includes a first end to receive a downward force and a second end adapted to engage the handle of the fuel bundle to apply an upward force to the handle and push down on the channel.

Abstract: A nuclear reactor fuel bundle assembly including: a fuel bundle including an array of fuel rods and water rods mounted in an upper tie plate and housed in walls of a channel, and a pore type debris shield mounted at least partially in the channel, above or below the upper tie plate, the shield extending to or over the walls of the channel, whereby deflecting and/or capturing falling debris from entering the fuel assembly, wherein the shield is design to be durable, yet flexible, and porous.

Abstract: The apparatus for creating and editing a nuclear reactor core template includes a graphical user interface and a processor controlling the graphical user interface to display a graphical representation of a nuclear reactor core. Via the graphical user interface, the processor provides a user with graphical tools for at least one of assigning fuel bundle categories to fuel bundle positions in the graphical representation and editing assigned fuel bundle categories to the fuel bundle positions in the graphical representation.

Abstract: A nuclear reactor fuel bundle assembly has been developed that includes: a fuel bundle including an array of fuel rods attached to a lower tie plate, an upper tie plate and housed in walls of a channel, and a water rod, having an upper discharge end below and unattached to the upper tie plate, and the upper discharge end having an unobstructed opening.

Abstract: The method and apparatus adaptively determine weighting factors within the context of an objective function for handling optimality conditions and constraints within an optimization search. The objective function is defined as a sum of credit and penalty components. The credit components represent the optimality conditions for the problem. The penalty components represent the constraint violations for the problem. Initially, each component is made up of a weight multiplied by a mathematical expression, called a term, that quantifies either an optimality condition or a constraint violation. The set of credit and penalty weights are adaptively determined based on the progress of an optimization search. Both static and dynamic representations of the modified objective function are used to perform the adaption.

Abstract: A method of evaluating constraint functions, the evaluation being based at least in part on a fuel bundle movement criteria.

Abstract: A system and computer implemented method for automatically performing reload licensing analysis of a nuclear reactor and preparing and generating a reload licensing report. The system and method include collecting input data and processing a plurality of nuclear reactor analysis modules. Each of the nuclear reactor analysis modules receives analysis module input data and generates analysis output data. The analysis module input data is based at least in part on the collected input data. The method also includes verifying at least one analysis output data from one nuclear reactor analysis module by comparing the analysis output data to predetermined analysis output data for the one nuclear reactor analysis module. The method further includes generating a reload licensing report for the nuclear reactor as a function of the analysis output data from two or more nuclear reactor analysis modules.

Abstract: In the method, a reference core design is generated based on a defined set of limits. The set of limits may include a target hot excess reactivity constraint to be satisfied over a given core energy cycle and a given desired control blade definition that is set for the cycle. A reference core design is generated based on the defined limits. A unique subset of fresh fuel bundles is subject to an iterative improvement process including replacing, at each fuel location, at least one of the current fresh fuel bundles with at least one of the selected fresh fuel bundles, and simulating reactor operation on the reference core design to obtain a plurality of outputs to be ranked based on the defined set of limits. The highest ranked output may represent an accepted core design with set control blade definition that satisfies the target hot excess reactivity constraint.

Abstract: A method of fuel bundle consideration in a reactor. The method includes creating or editing bundle groups including fuel bundles.

Abstract: A method, arrangement and computer program is described for generating a database of selectable fresh fuel bundle designs usable in one or more nuclear reactors. In an example, an initial population of candidate fresh fuel bundle designs may be generated and a set of rod-type changes to make to a given candidate in the initial population may be established. A given candidate fresh fuel bundle may be modified by making at least one rod-type change from the set therein. A core loaded with the modified bundle design may be simulated to generate bundle performance outputs. The bundle performance outputs may be ranked based on a plurality of user-input limits. The modified candidate fresh fuel bundle design may be stored, if the bundle performance outputs meet, or are within an accepted margin to, the user-input limits, so as to generate the database.

Abstract: A method and arrangement of determining pin enrichments for a fuel bundle of a nuclear reactor, where a plurality of input parameters and target conditions may be input and enrichment changes, to be made across the fuel bundle, may be calculated using response matrix technology. Fuel bundle pin enrichment data may be output that satisfies the target conditions. The method and arrangement may enable production of fuel bundles having a desired local peaking, exposure peaking and R-factor performance. Consequently, given fuel cycles typically may be loaded and operated such that less fuel may be needed for identical cycle lengths, potentially resulting in improved fuel cycle economics. Additionally, because fuel bundle development may require fewer iterations, there may be a substantial cycle time reduction in the bundle design process, potentially reducing cost and enhancing profitability.

Abstract: A method, arrangement and computer program is described for determining a set of standardized rod types for use in a nuclear reactor core. The method may include defining a set of rod type-related limits, and determining, based on the limits, an initial population of rod types to be evaluated for use in cores of a selected number of nuclear reactor plants. Based on the initial population of rod types a database of selectable fresh fuel bundle designs applicable to the one or more cores may be generated. Bundle data related to at least a subset of the selectable fresh fuel bundle designs may be retrieved from the database, and a target number of rod types may be selected from the initial population based on the retrieved bundle data as the set of standardized rod types.

Abstract: In the method, a set of limits applicable to a core may be defined, and a test core loading pattern design, to be used for loading the core, may be determined based on the limits. Reactor operation on at least a subset of the core may be simulated to produce a plurality of simulated results. The simulated results may be compared against the limits, and data from the comparison may indicate whether any of the limits were violated by the core during the simulation. A designer or engineer may use the data to modify the test core loading pattern, creating one or more derivative core loading pattern design(s) for simulation and eventual perfection as an acceptable core loading pattern design for the core.

Abstract: In the method, a set of limits applicable to a test rod pattern design are defined, and a sequence strategy for positioning one or more subsets of the test rod pattern design is established. Reactor operation on a subset of the test rod pattern design, which may be a subset of fuel bundles in a reactor core for example, is simulated to produce a plurality of simulated results. The simulated results are compared against the limits, and data from the comparison is provided to indicate whether any of the limits were violated by the test rod pattern design during the simulation. A designer or engineer may use the data to determine which operator parameters need to be adjusted (e.g., control blade notch positions for example) in order to create a derivative rod pattern design for simulation, and eventually perfect a rod pattern design for a particular core.

Abstract: In the method, a set of limits are defined and a reference core design is generated based on the limits, and includes an initial loading pattern of current fresh fuel bundles arranged in a plurality of fuel locations. A unique subset of fresh fuel bundles is selected for evaluation as the reference core design is subjected to an iterative improvement process. The iterative process includes replacing, at each fuel location, at least one of the current fresh fuel bundles with at least one of the selected fresh fuel bundles, and simulating reactor operation on the reference core design to obtain a plurality of outputs. The outputs may be ranked based on the defined set of limits, and the highest ranked output may be selected as an accepted core design for the nuclear reactor.

Abstract: A spacer for a nuclear fuel assembly includes dividers orthogonally interdigitated relative to one another defining fuel and water rod cells. The dividers defining fuel rod cells have openings carrying an integrally formed spring which projects into the fuel cell to bear the fuel rod in the cell against an opposite stop carried by an opposing divider. The dividers are arranged in a 10×10 array with diagonally opposing quadrants being mirror images of one another. The dividers are surrounded by a peripheral band with integrally formed stops projecting into the peripheral cells enclosing the spacer.

Abstract: A new type of coated component for use in boiling water reactor (“BWR”) fuel assemblies, particularly Zircaloy spacers, having protective coatings applied to selected surfaces of the spacers in order to prevent the formation and propagation of “shadow corrosion” on adjacent zirconium alloy structures. In its broader aspects, the coating material is applied to those surfaces of the BWR components having electro-chemical characteristics that differ from zirconium alloys, such as Inconel spacers or springs. The new coatings impart an electro-chemical potential to the surfaces of the components that is substantially similar to the adjacent zirconium alloy, thereby preventing or significantly inhibiting shadow corrosion.

Abstract: According to the method, a reactor core is shut down during the operation cycle. One or more fuel bundles of the reactor core are then moved to new positions within the reactor core to increase a total energy, output of the reactor core as compared to continuing operation of the reactor core without the shutting down and moving steps.

Abstract: A nuclear fuel assembly includes a fuel bundle resting on a lower tie plate. The lower tie plate includes a grid having raised bosses disposed in a rectilinear array for receiving the end plugs of the fuel rods and webs interconnecting the bosses. Between the raised bosses and the end plugs, there is disposed a filter plate having holes registering with the holes through the bosses for receiving the end plugs and apertures providing an approximate 40% open area through the filter plate. The webs of the tie plate grid are recessed from the upper edges of the bosses, facilitating flow through the apertures of the filter plate.

Abstract: A nuclear fuel assembly includes a fuel bundle resting on a lower tie plate. The lower tie plate includes a grid having raised bosses disposed in a rectilinear array for receiving the end plugs of the fuel rods and webs interconnecting the bosses. Between the raised bosses and the end plugs, there is disposed a filter plate having holes registering with the holes through the bosses for receiving the end plugs and apertures providing an approximate 40% open area through the filter plate. The webs of the tie plate grid are recessed from the upper edges of the bosses, facilitating flow through the apertures of the filter plate.