Induced nuclear reactions: processes – systems – and elements – Testing – sensing – measuring – or detecting a fission reactor...
Reexamination Certificate
2000-10-04
2003-03-18
Hoff, Marc S. (Department: 2857)
Induced nuclear reactions: processes, systems, and elements
Testing, sensing, measuring, or detecting a fission reactor...
C376S255000
Reexamination Certificate
active
06535568
ABSTRACT:
TECHNICAL FIELD
This invention relates generally to methods for demonstrating compliance of a nuclear reactor with fundamental licensing criteria for nuclear fuel rod internal pressure, and more particularly to a method for establishing thermal-mechanical limits for the use and operation of nuclear fuel rods.
BACKGROUND
Most countries have some sort of regulatory body or commission for administering the safe use and distribution of nuclear fuels and power generation within their jurisdictions. In the United States of America, for example, the Nuclear Regulatory Commission (USNRC) is the primary governmental regulatory body that is responsible for licensing the construction and operation of nuclear power plants. In its regulatory capacity, the USNRC is responsible for setting up and reviewing fundamental safety criteria for granting a license for the operation of a nuclear reactor. Under such review, an applicant for a license must be able to demonstrate that the operation of a particular reactor complies with fundamental safety criteria set forth by the USNRC. Reactor design and operation evaluation procedures are proposed and, once approved by the USNRC, may form the basis for the granting of an operating license.
Procedures for demonstration of compliance typically include (among other things) providing statistical/empirical evidence showing that the fuel rods within a reactor function within a given margin of safety at or below some predetermined power level that will ensure that the thermal and mechanical stresses on the fuel rod cladding for all rods in the reactor core are kept at a safe level during the life and use of the fuel—for example, to prevent any cracking or ruptures of fuel rod cladding and subsequent leaking of contaminates. Reactor operating limits are established to ensure that reactor operation is maintained within a fuel rod thermal-mechanical design and safety analysis basis. These operating limits may be defined, for example, by the maximum allowable fuel pellet operating power as a function of fuel pellet exposure level—usually expressed in terms of the maximum linear heat generation rate (MLHGR) (i.e., the maximum heat generated by a fuel rod per unit length of the rod (Kw/ft) verses exposure (Gwd/mt)).
As an example, under GE Company's USNRC approved “GESTAR” licensing basis, it is required that fuel design Thermal-Mechanical (T-M) analysis for a reactor be performed using the following conditions: (i) either worst tolerance assumptions are applied or probabilistic analysis are performed to determine statistically bounding results (i.e., upper 95% confidence), and (ii) operating conditions are taken to bound the conditions anticipated during normal steady-state operation and anticipated operational occurrences (AOOs). Based on the T-M analysis performed, operating limits are established to ensure that actual fuel operation is maintained within the fuel rod thermal-mechanical design and safety basis. These operating limits define the maximum allowable fuel pellet operating power level as a function of fuel pellet exposure.
The conventional methodology for designing thermal-mechanical (T-M) operational limits for fuel rods is to apply a bounding power history analysis on a single hypothetical fuel rod. For example, fuel rods may be evaluated to ensure that the effects of fuel rod internal pressure during normal steady-state operation will not result in fuel failure due to an excessive fuel rod cladding pressure loading. Such an evaluation is based, for example, on the amount of fission gas released by the uranium fuel pellets in a fuel rod and the resultant pressure within the rod to determine the cladding creep-out rate due to internal gas pressure during normal steady-state operation. (For example, a T-M performance analysis program, such as GE Company's GESTR-MECHNICAL (GSTRM) performance code program, may be used to evaluate a fuel rod.) A generic power limit curve is developed from the T-M evaluation which provides an operating envelope that is valid for the operation of fuel rods in every fuel cycle for all reactors. An example of such a fuel T-M limit envelope is illustrated by curve
10
in the graph of
FIG. 1
which shows the maximum linear heat generation rate (LHGR) versus pellet exposure.
In developing such a generic operating limit envelope, typically only the fuel rod or rods within a reactor core that experience the maximum expected power and exposure conditions are evaluated. However, it is known that no single fuel rod in a reactor core actually operates with a power history equivalent to its limiting power history, but rather operates substantially below the limiting power for a majority of its operational duty. Prior to the present invention, it has been unfeasible to identify which fuel rods or rod within the core may be operating in the most limiting manner with regard to internal fission gas release. In this regard, the conventional approach of developing a bounding power history for generating thermal-mechanical limits for fuel rods is overly conservative and unnecessarily restricts the operation of nuclear reactors below an optimum level.
DISCLOSURE OF THE INVENTION
The method of the present invention overcomes the above limitations by constructing actual power history profiles for every fuel rod in the reactor core and evaluating the internal pressure for each fuel rod individually. Conventional computer programs for generating a power history for an individual fuel rod based on reactor design or actual reactor operations are known. For example, around about the year 1987, General Electric Company developed a computer program called “LERN”. The LERN program is capable of generating a power history profile for a single fuel rod and is typically used, for example, to perform special studies for particular fuel rods (one at a time). In the present invention, fuel rod power histories are constructed for each rod in the reactor core using both pre-acquired historical operating data of each fuel rod during past fuel cycles and a projection of reactor operations into an upcoming fuel-cycle.
In implementing the method of the present invention, a computer program may employ the above described LERN-type technology or equivalent software to evaluate individual fuel rod power histories for each nuclear fuel rod in the reactor core based both on empirical information acquired during previous fuel cycles and a projected operation of the reactor in an upcoming fuel cycle. Using the constructed power histories for each fuel rod, the program then performs a rod-by-rod internal pressure analysis wherein it computes thermal and mechanical overpower limits and a maximum internal pressure for each rod in the upcoming fuel cycle to determine which rod or rods are operating with the most limiting power history—i.e., closest to its internal pressure, and thermal and mechanical overpower limits. The power history of this fuel rod or rods is then used as the basis for setting safe operational limits, and determining compliance with U.S. NRC requirements, rather than relying on an overly conservative conventional bounding power history analysis (which is typically based solely on a single hypothetical fuel rod). Compliance of reactor operation with U.S. NRC licensing requirements is demonstrated by confirming that the fuel rod identified as having the maximum fuel rod internal pressure or thermal and mechanical overpower stresses for the upcoming fuel cycle does not violate the particular fundamental licensing criteria.
Preferably, the above fuel rod evaluation process is performed for each operating fuel-cycle of a reactor. By using this T-M limit setting approach, higher peak operating powers for each rod may be used without violating fundamental safety and governmental imposed criteria for reactor operation. Moreover, implementation of the disclosed method of the present invention allows operation of a reactor using a much less restrictive fuel rod thermal-mechanical limit and, consequently, results in substantial improvements in operatio
Global Nuclear Fuel - Americas, LLC
Hoff Marc S.
Miller Craig Steven
Nixon & Vanderhye P.C.
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