Global Nuclear Fuel - Americas, LLC

General Electric Company

Steven Barry Sutton

David Joseph Kropaczek

Christian Carlos Oyarzun

William Earl Russell, Ii

Mehdi Asgari

William Charles Cline

Carey Reid Merritt

Glen Alan Watford

John Dean Fuller

Margaret Estelle Harding

Russell M Fawcett

Lukas Trosman

Gerald Dean Kvaall, Jr

Roland Otto Jackson

William Earl Ii Russell

Carey Reid Merrit

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 allow a user to selectively populate a loading map with fuel bundles residing in at least one fuel pool.

Method and apparatus for moving nuclear fuel

A method of evaluating constraint functions, the evaluation being based at least in part on a control blade movement criteria.

Method and apparatus for evaluating a proposed solution to a constraint problem

Method and apparatus for facilitating recovery of nuclear fuel from a fuel pool

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

Methods of editing fuel bundle groups

Methods of using fuel bundle groups as evaluation constraints

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.

Method and apparatus for creating and editing a nuclear reactor core loading template

In an embodiment of a method of evaluating robustness of a proposed solution to a constraint problem, operational output data for at least first and second modified versions of the proposed solution is generated. The first modified version has at least one control variable of the proposed solution perturbed in a first direction and the second modified version has the at least one control variable of the proposed solution perturbed in a second direction. At least a portion of the generated operation output data is then presented such as on a graphical user interface.

Method and apparatus for evaluating robustness of proposed solution to constraint problem and considering robustness in developing a constraint problem solution

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

Method of fuel bundle consideration in a reactor

In a method of determining an operating margin to a given operating limit in a nuclear reactor, operational plant data from an on-line nuclear reactor plant is accessed, and reactor operation is simulated off-line using the operational plant data to generate predicted dependent variable data representative of the given operating limit. The predicted dependent variable data is normalized for evaluation with normalized historical dependent variable data from stored operating cycles of plants having a similar plant configuration to the on-line plant. A time-dependent average bias and a time-dependent uncertainty value for the predicted dependent variable data are determined using the normalized historical dependent variable data, and a risk-tolerance level for the on-line plant is obtained. An operating margin to the given operating limit is determined based on the determined time time-dependent average bias value and time-dependent uncertainty value so as to satisfy the risk-tolerance level of the evaluated plant.

Method of determining margins to operating limits for nuclear reactor operation

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.

Method, arrangement and computer program for generating database of fuel bundle designs for nuclear reactors

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.

Method, arrangement and computer program for determining standardized rod types for nuclear reactors

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.

Method and arrangement for developing core loading patterns in nuclear reactors

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.

Method and arrangement for developing rod patterns in nuclear reactors

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.

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Wilmington, NC, United States of America

Steven Barry Sutton