Patent Number: 055552790
Section: summary

BACKGROUND OF THE INVENTION In presently known nuclear reactors, particularly in boiling water reactors, the reactor core typically contains a plurality of vertically oriented fuel assemblies arranged in an array such that a self-sustaining nuclear fission reaction can be controlled. The typical core is generally contained in a pressure vessel and submerged in a liquid such as water. The water may serve as both a core coolant and a neutron moderator. A series of vertically oriented, moveable control rods, composed of neutron absorbing material, are insertable between the fuel assemblies such that core reactivity control may be accomplished by adjusting the control rods within the core. In the usual case, water flows through channels located amongst the fuel rods and control rods in the core and is pumped upwardly from a lower plenum below the core to an upper plenum above the core. To monitor the power distribution within the core, it is common practice to place in-core neutron detectors both radially and axially throughout the core. The signals from these neutron detectors are used to monitor core conditions and to initiate action in the event of a detected abnormality in the reactor. Core monitoring may include providing a reactor operator with analog or digital indications of the monitored conditions and providing an alarm when an abnormal condition is detected. Action, such as shutdown of the reactor, may also be automatically initiated when an abnormal core condition is detected. A boiling water reactor is capable of entering a state called thermal-hydraulic instability that can challenge established fuel thermal and mechanical safety limits. Examination of the thermal-hydraulic stability of a reactor must be performed to prevent potential damage to the core. Thermal-hydraulic instability may be described as follows. Pressure perturbations at the core inlet cause flow disturbances that travel up the fuel channels as time-varying density waves. These waves result in local deviations from the steady-state axial pressure drop distribution. The local pressure drop in a fuel assembly is highly dependent on void fraction. Since the coolant voiding increases axially with greater core elevation, the highest void fraction is found at the channel outlet. The effect of density waves on the total channel pressure drop is therefore effectively delayed in time--the void sweeping time--until the perturbation is felt at the channel exit. When the channel pressure drop time delay (phase lag) nears 180 degrees out of phase with the channel inlet flow variations, the fuel assembly can become thermal-hydraulically unstable. Thus, the thermal-hydraulic stability margin of a fuel channel is dependent on the phase lag caused by void sweeping time, and the gain which is dependent on the channel void distribution. An additional complexity is introduced in boiling water reactor stability because of the reactor power dependency on coolant density. Local void reactivity responds to the time-varying density wave described above. The reactivity change affects local neutron flux and is manifested after a time delay as changes in fuel cladding surface heat flux and ultimately in local coolant voiding. This mechanism can also provide positive feedback to density wave oscillations. The neutronic feedback gain is dependent on how closely the fuel thermal time constant approximates the void sweeping time, and on the local void fraction. The two feedback mechanisms, thermal hydraulic and neutronic, are coupled in a boiling water reactor core and produce oscillations in both core flow and thermal power. These oscillations can affect margins to fuel safety limits. In addition, core instabilities can occur even when neither feedback mechanism alone is sufficient to generate power oscillations. Generally, existing thermal-hydraulic instability detection systems do not have the means to rapidly and accurately notify an operator of the core's stability margin. In the current invention, a simulated decay ratio signal is generated that relates to the thermal-hydraulic stability of a nuclear reactor which is input into means to take corrective action, notify an operator, or initiate an automatic suppression function upon reaching a minimum specified stability margin. A computer-based system is used, utilizing algorithms that provide a fast system response time such that a reliable indication of reductions in thermal-hydraulic stability of a core may be obtained prior to the core actually becoming unstable, and so that reactor operators may have sufficient time to take compensatory measures, or an automatic suppression function may be initiated. SUMMARY OF THE INVENTION A power oscillation detection system and method for monitoring thermal hydraulic stability in a nuclear reactor core which indicates to nuclear plant operators the reactor stability margin, or generates automatic initiation of corrective action when instabilities occur, is provided. In a preferred embodiment, the system contains a plurality of in-core neutron flux detectors spatially distributed throughout the reactor core. Each flux detector provides an output signal. A band-pass filter removes high frequency components of the output signals that are in excess of the characteristic frequency range for thermal-hydraulic instability. A computer-based signal processor system employing a period based algorithm utilizes a counting means that evaluates the output signal of each flux detector to determine a time-dependent count for each flux detector output signal corresponding to consecutive flux detector signal oscillations with periods that are within a given time tolerance and range. This count correlates to the thermal-hydraulic stability of the reactor. A representative maximum oscillation period count for a given reactor stability state is desired to be used to determine a decay ratio, however, that count generally occurs only once every several minutes for each individual flux detector signal. To create a faster response time, the present invention utilizes a computer-based signal processor system using a simulated decay ratio algorithm that employs a count combining means for determining the maximum oscillation period count of a designated group of flux detectors that are spatially distributed throughout the reactor core, over a substantially shorter evaluation time. This evaluation time may be dynamically established utilizing real-time processed data from the in-core flux detectors. Utilizing this computer-based system, the decay ratio may be updated, for example, approximately every five seconds. To overcome expected statistical variations in the maximum oscillation period count between the consecutive evaluation periods, the signal processor that processes the simulated decay ratio signal may employ several techniques to provide improved performance. First, a spike rejection function may be used to reduce anomalous changes in the combined oscillation period count value that are not related to actual core performance. Second, the maximum combined oscillation period count for several evaluation periods may be processed to develop a consistent and responsive measure of the core stability performance. Finally, the processed maximum oscillation period count is used to establish the simulated decay ratio signal. This simulated decay ratio signal provides an estimate of the nuclear power reactor core decay ratio that is representative of the reactor's thermal-hydraulic stability margin. This representative decay ratio signal may then be used as input to an indicating device such that a reactor operator can monitor core conditions, to generate an alarm at a desired simulated decay ratio level, or, to generate an automatic reactor trip signal at a desired simulated decay ratio level. It is accordingly an object of the present invention to provide a computer-based oscillation detection system for monitoring and indicating thermal-hydraulic stability margin in a nuclear reactor. A further object of the present invention is to provide a computer-based oscillation detection system for monitoring and indicating unacceptable losses in thermal-hydraulic stability margin that, if unmitigated, are capable of inducing sustained oscillations of a character potentially damaging to the reactor core. Another object of the current invention is to provide a computer-based oscillation detection system for monitoring and indicating thermal-hydraulic stability margin wherein neutron flux is monitored at axial and radially distributed locations throughout the reactor core to detect and indicate changes in thermal-hydraulic stability margin. A further object of the present invention is to provide a computer-based oscillation detection system for monitoring and indicating thermal-hydraulic stability margin, wherein reactor personnel are notified in the event of the transition to thermal-hydraulic instability induced flux oscillations. Another object of the present invention is to provide a computer-based oscillation detection system for monitoring and indicating thermal-hydraulic stability margin, having a fast response time, wherein corrective controls are instituted in the event of changes in stability margin. Another object is to provide a computer-based oscillation detection system for monitoring and indicating of thermal-hydraulic stability margin, having a fast response time, wherein an automatic suppression function is initiated in response to the detection of a specified reactor stability state. Other objects and advantages of the present invention will become apparent from the following description taken in conjunction with the accompanying drawings.