Determination of operating limit minimum critical power ratio

Details Determination of operating limit minimum critical power ratio Determination of operating limit minimum critical power ratio

2002-03-27

2004-01-06

Carone, Michael J. (Department: 3641)

Induced nuclear reactions: processes, systems, and elements

With control of reactor

By electronic signal processing circuitry

C376S259000, C376S277000

Reexamination Certificate

active

06674825

ABSTRACT:

FIELD OF THE INVENTION
The present invention relates generally to methods for evaluating nuclear power core operation, and more particularly to an improved method and apparatus for determining an operating limit minimum critical power ratio (OLMCPR) so as to effectuate increased efficiency and operation of Boiling Water Reactors (BWR).
BACKGROUND OF THE INVENTION
A Boiling Water Nuclear Reactor (BWR) generates power from a controlled nuclear fission reaction. As shown in
FIG. 1
, a simplified BWR includes a reactor chamber
101
that contains a nuclear fuel core and water. Generated steam is transferred through pipe
102
to turbine
103
, where electric power is generated, then water returns to the core through pipe
104
. As shown in
FIG. 2
, the core
201
is made of approximately five hundred (500) bundles
202
of fuel rods arranged in a specific manner within the reactor core. As shown in
FIG. 3
, each bundle
301
contains roughly one hundred (100) fuel rods
302
. Water in the core surrounds the rods. Heat generated by a nuclear reaction is transferred from the rods to the water circulating through the core, boiling some of the water. The heat generated in the core is carefully controlled to maintain safe and efficient reactor operations.
In a Boiling Water Nuclear Reactor there are basically three modes of heat transfer that must be considered in defining thermal limits for the reactor: (i) N ideate boiling, (ii) transition boiling and (iii) film boiling. “Nucleate boiling” is the preferred efficient mode of heat transfer in which the BWR is designed to operate. “Transition boiling” is manifested by an unstable fuel rod cladding surface temperature which rises suddenly as steam blanketing of the heat transfer surface on the rod occurs, then PS to the nucleate boiling temperance as the steam blanket is swept away by the coolant flow, and then rises again. At still higher fuel rod/bundle operating powers, “film boiling” occurs which results in higher fuel rod cladding temperatures. The cladding temperature in film boiling, and possibly the temperature peaks in transition boiling, may reach values which could cause weakening of the rod cladding and accelerated corrosion. Fuel rod overheating is conservatively defined as the onset of the transition from nucleate boiling to film boiling. The conventional basis for reactor core and fuel rod design is defined such that some “margin,” accommodating various design and operational “uncertainties,” is maintained between the most limiting operating condition and the transition boiling condition at all times for the life of the core.
The onset of transition boiling can be predicted by a correlation to the steam quality at which boiling transition occurs-called the “critical quality.” Steam quality can be readily measured and is generally a function of measuring distance above the boiling boundary (boiling length) for any given mass flow rate, power level, pressure and bundle flow geometry among other factors. A “critical power” is defined as that bundle power which would produce the critical quality of steam. Accordingly, a “critical power ratio” (CPR) is defined as the ratio of the critical power to the bundle operating power at the reactor condition of interest. CPR is descriptive of the relationship between normal operating conditions and conditions which produce a boiling transition. CPR is conventionally used to rate reactor design and operation. To assure a safe and efficient operation of the reactor, the CPR is kept above a prescribed value for all of fuel assemblies in the core. Reactor operating limits are conventionally defined in terms of the most limiting fuel bundle assembly in the core—defined as the “minimum critical power ratio” (MCPR). Reactor operating limits are often stated in terms of MCPR.
In nuclear power generation engineering, it is widely recognized that there is a possibility, however small, that the occurrence of a reactor transient event combined with the various “uncertainties” and tolerances inherent in reactor design and operation may cause transition boiling to exist locally at a fuel rod for some period of time. Accordingly, MCPR operating limits are conventionally set in accordance with the United States Nuclear Regulatory Commission (USNRC) design basis requirement that transients caused by a single operator error or a single equipment malfunction shall be limited such that, taking into consideration uncertainties In the core operating state, more than 99.9% of the fuel rods are expected to avoid boiling transition during that error or malfunction. A safety limit minimum critical power ratio (SLMCPR) defined under current USNRC requirements as the MCPR where no more than 0.1% of the fuel rods are subject to boiling transition (also known as NRSBT for Number of Rods Subject to Boiling Transition). The corresponding OLMCPR describes the core operating conditions such that the MCPR is not lower than the SLMCPR to a cc in statistical confidence.
i. An Ideal Approach
In principle, the OLMCPR could be calculated directly such that for the limiting anticipated operational occurrence (AOO), less than 0.1% of the rods in the core would be expected to experience boiling transition. This approach is described in U.S. Pat. No. 5,912,933, by Shaug et at. The process involved is shown in
FIG. 4
, which depicts histogram
40
of rod CPR values
401
versus number of rods
402
at the specific CFR value. While the CPR value is usually associated with a fuel bundle, it actually refers to the limiting rod in a bundle. Each rod in the bundle has a CPR value at is determined by the local power distribution and relative position of the rod within e bundle (R-factor). The lowest CPR of any one rod in the bundle Is used to characterize the CPR for the entire bundle.
The CPR
401
for a given rod has an associated probability distribution function (PDF) which reflects the uncertainties in its determination. The PDF may be determined experimentally and is shown as an Experimental Critical Power Ratio (ECPR) distribution
410
. Thus, if a nominal CPR value (
411
) is 1.0, then the PDF
410
of probable actual CPR values range from 0.90 to 1.10. The variability in the rod CPR values is due to uncertainties in the initial rod condition, i.e., uncertainties in the measurements of parameters at the reactor operating state (core power) and in the modeling of derived parameters (power distribution).
To take the effect of a transient event on the CPR values into account, a safety margin is introduced to CPR values by shifting the acceptable nominal CPR value
405
for the lowest rod CPR to a larger CPR value, i.e., 1.25. The ECPR histogram distribution
403
for the lowest CPR rod is thus shifted such that the entire CPR histogram is above a CPR value of 1.20, and well above a CPR value of 1.0. Moreover, the nominal CPR values
407
for rods other than the lowest CPR rod are above the nominal CPR value, e.g., 1.25, of the lowest CPR rod.
During a transient in rod operation, the histogram
407
of rod CPRs shifts to the left to lower CPR values, resulting in the histogram
408
. With this shift, the “nominal” CPR value
406
during the transient is at the point, e.g., 1.05, where the minimum CPR value is reached during the transient. The limiting rod will have an associated PDF
404
, which includes both the uncertainties in the initial rod conditions and “transient uncertainties.” The maximum change in critical power ratio during the transient (&Dgr;CPR
409
) includes uncertainties in the modeling of the transient, both the physical models and plant parameters.
Ideally, this transient &Dgr;CPR
409
and associated OLMCPR would be generated as shown in
FIG. 5
, and described as follows:
Step
1
: Assume a set of base core operating conditions using the parameters to run the plant that generates a core MCPR equal to the OLMCPR, as shown by block 501.
Step
2
: Using the parameters, such as core power, core flow, core pressure and others, that predict a general bundle CPR set forth in block
506
, determine the MCPR for eac

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