Global Nuclear Fuel - Americas, LLC

Jens Georg Munthe Andersen

Charles Lee Heck

James Courtney Shaug

Francis Thomas Bolger

A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. Based on a predetermined generic transient bias and uncertainty distribution in the change in critical power ratio (&Dgr;CPR/ICPR), shifted histograms of fuel rod critical power ratio (CPR) are generated. Ultimately, the operating limit minimum critical power ratio (OLMCPR) of the reactor is directly evaluated from a shifted histogram of fuel rod CPRs for a particular set of initial conditions which result in a probability calculation representing the number of fuel rods subject to a boiling transition (NRSBT) during the transient condition being equal to a predetermined value.

Determination of operating limit minimum critical power ratio

A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. A data processing system is used to execute specific program routines that simultaneously simulate the thermal operating characteristics of fuel rods within the reactor during a transient operational condition. The method employs a multi-dimensional approach for the simulation of postulated operational events or an anticipated operational occurrence (AOO) which produces a transient condition in the reactor&mdash;such as might be caused by single operator error or equipment malfunction. Based on a generic transient bias and uncertainty in the change in critical power ratio (&Dgr;CPR/ICPR), histograms of fuel rod critical power ratio (CPR) are generated. Ultimately, the operating limit minimum critical power ratio (OLMCPR) of the reactor is evaluated from a histogram of probability calculations representing the number of fuel rods subject to a boiling transition (NRSBT) during the transient condition. The histogram may be readily displayed by the data processing system and used to statistically demonstrate an OLMCPR compliance of the reactor core design with USNRC regulations.

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Wilmington, NC, United States of America

Jens Georg Munthe Andersen