阅读说明：本技术 亚临界堆芯反应性偏差预测技术 (Subcritical reactor core reactivity deviation prediction technology ) 是由 P·J·萨巴斯蒂安尼 M·W·迪克斯 L·R·格罗布迈尔 于 2018-12-12 设计创作，主要内容包括：本申请公开了一种用于确定在达到反应堆临界状态之前核反应堆堆芯的整体堆芯反应性偏差和相应估计临界条件的方法。该方法首先需要收集和评估计数率反比(ICRR)数据；具体来说,将测量的ICRR数据与预测的ICRR数据进行拟合。然后,将整体堆芯反应性偏差确定为对预测值进行均匀反应性调整的量,其使得在测量值和预测值之间产生理想比较。(A method for determining an overall core reactivity deviation and corresponding estimated critical conditions of a nuclear reactor core prior to reaching a reactor critical state is disclosed. The method first requires collecting and evaluating Inverse Count Rate Ratio (ICRR) data; specifically, the measured ICRR data is fitted to the predicted ICRR data. The overall core reactivity deviation is then determined as an amount of uniform reactivity adjustment to the predicted value that results in a desired comparison between the measured and predicted values.)

1. For determining K_effA method of overall core reactivity deviation for a nuclear reactor core of less than 1, the method comprising the steps of:

measuring a subcritical neutron flux for one or more conditions of a nuclear reactor core;

calculating a predicted value of spatially corrected subcritical neutron flux for the one or more conditions of the nuclear reactor core;

determining a difference between the measured subcritical neutron flux and the predicted spatially corrected subcritical neutron flux; and

recording the difference as a deviation in overall core reactivity.

2. The method of claim 1, wherein the measuring step is performed using the output of a source sector detector.

3. The method of claim 1, further comprising the steps of:

applying regression statistics of the measured values of subcritical neutron flux to corresponding calculations of spatially corrected predicted values of subcritical neutron flux; and is

Applying quantitative measure-and-predict criteria on regression statistics to detect various core abnormalities while the nuclear reactor core is in a subcritical state and before the nuclear reactor core reaches a critical state.

4. The method of claim 1, further comprising the steps of: the deviation in reactivity between the predicted core and the actual core is determined by determining a unified analytical reactivity adjustment, which is the systematic global reactivity deviation required to bring the measured neutron flux data into agreement with the predicted spatially corrected predicted values.

5. The method of claim 1, wherein the measuring, calculating, and determining steps are performed under a plurality of steady state subcritical conditions.

6. The method of claim 5, wherein the plurality of steady state subcritical conditions comprise a plurality of state points.

7. The method of claim 5, wherein the plurality of steady state subcritical conditions are obtained by the step of repositioning control rods while maintaining other reactor conditions at steady state.

8. A processing device programmed to perform the method of claim 1.

9. A machine-readable medium comprising instructions for performing the method of claim 1.

Technical Field

The disclosed concept relates generally to methods for predicting when a nuclear reactor core becomes critical, and more particularly, to a method for determining an overall core reactivity deviation and corresponding estimated critical conditions of a nuclear reactor core before reactor critical conditions are reached.

Background

In a pressurized water reactor power generation system, heat is generated in the core of a pressure vessel by a fission chain reaction occurring in a plurality of fuel rods supported in the core. The fuel rods are held in spaced relation within the fuel assembly with the spaces between the fuel rods forming coolant channels through which the boronized water flows. The hydrogen in the coolant water slows down neutrons emitted by the enriched uranium in the fuel rod to increase the number of nuclear reactions, thereby improving the efficiency of the process. Control rod guide thimbles are interspersed within the fuel assembly in place of the locations of the fuel rods and serve to guide the control rods, which are operable to be inserted into and withdrawn from the core. When inserted, the control rods absorb neutrons, thereby reducing the number of nuclear reactions and reducing the amount of heat generated within the core. The coolant flows from the reactor, through the components and to the steam generator tube side, where heat is transferred to water in the steam generator shell side under lower pressure conditions, which results in the generation of steam for driving the turbine. Coolant exiting the tube side of the steam generator is returned to the reactor in a closed loop cycle driven by the main coolant pump to restart the process.

The power level of a nuclear reactor is generally divided into three sections: a source or start-up section, an intermediate section, and a power section. The power level of the reactor is continuously monitored to ensure safe operation. Such monitoring is typically performed by neutron detectors placed outside and inside the nuclear reactor core for measuring the neutron flux of the reactor. Since the neutron flux at any point in the reactor is proportional to the fission rate, the neutron flux is also proportional to the power level.

Fission and ionization chambers have been used to measure flux within the source, intermediate and power sections of reactors. Typical fission and ionization chambers are capable of operating at all normal power levels; however, they are generally not sensitive enough to accurately detect the low level of neutron flux emitted within the source section. Thus, when the power level of the reactor is within the source section, a separate low level source section detector is typically used to monitor the neutron flux.

Fission reactions occur within the core when free neutrons at the proper energy level strike the atoms of fissionable materials contained within the fuel rods. These reactions result in the release of a large amount of thermal energy that is extracted from the core in the reactor coolant, and these reactions release additional free neutrons that can be used to generate more fission reactions. Some of these released neutrons escape the core or are absorbed by a neutron absorber (e.g., control rods) and therefore do not cause conventional fission reactions. By controlling the amount of neutron absorbing material present in the core, the rate of fission can be controlled. Random fission reactions will always occur in fissionable material, but when the core is deactivated, the released neutrons will be absorbed at such a high rate that a sustained series of reactions will not occur. By reducing the neutron absorber material until the number of neutrons in a given generation equals the number of neutrons in the previous generation, the process becomes a self-sustaining chain reaction, and the reactor is said to be "critical". When the reactor is critical, the neutron flux is about six orders of magnitude higher than when the reactor is out of service. In some reactors, in order to accelerate the increase in neutron flux in a deactivated core to achieve a practical transition interval, an artificial neutron source is implanted in the nuclear reactor core among the fuel rods containing fissionable material. Such artificial neutron sources provide a local increase in neutron flux and thus contribute to the power of the reactor.

Without a neutron source, the ratio of the number of free neutrons in one generation to the number of free neutrons in the previous generation is called the "neutron multiplication factor" (K)_eff) And used as a measure of reactor reactivity. In other words, the critical metric for a nuclear reactor core is K_effI.e., the ratio of neutron production to total neutron loss (due to both destruction and loss). When K is_effAbove 1, more neutrons are produced than are destroyed. Similarly, when K_effLess than 1, more neutrons are destroyed than are produced. When K is_effBelow 1, the reactor is said to be "subcritical". Until recently, there has been no direct method for source zone excore detectors to measure when critical conditions have occurred. Nuclear power plant operators typically estimate when a critical condition occurs by a variety of methods. A method of estimating when a critical condition occurs is estimated by plotting the inverse of the count rates obtained from the source zone detectors as a function of a change in conditions (e.g., control rod withdrawal) used to bring the nuclear power plant to the critical condition. When the nuclear power plant is in a critical state, the count rate of the source section approaches infinity, and therefore, the Inverse Count Rate Ratio (ICRR) becomes zero. The ICRR curve is almost never linear due to the physical causes of the reactions occurring within the nuclear reactor core. Changes in control stick position have a significant effect on the shape of the ICRR curve. Therefore, there is great uncertainty in estimating the critical condition of a nuclear power plant from the ICRR curve, and it is also under strict scrutiny by the american nuclear regulatory commission and nuclear power operations institute.

Recently, a method has been devised to directly predict when the reactor reaches a critical state. This process is described in U.S. Pat. No.6,801,593. According to the method, the reactivity of the core is improved while monitoring the output of the source zone detector. The correction factor linearizes the ICRR, enabling a predictable extrapolation of the curve. Thus, the method describes a core reactivity measurement procedure in which the count rates are inversely proportional to the spatial correction. However, this method does not address the accuracy of the core reactivity measurement, which depends on the accuracy of the measured neutron radiation level. In particular, it is very important to accurately determine the incremental change in the measured neutron level. In a normally operating neutron radiation detector, the largest neutron measurement error component is typically caused by a quantity commonly referred to as the "background signal". The response of the background signal in the detector measurement is not due to source neutrons. This results in errors in the measured core reactivity changes. In order to improve the accuracy of the neutron population measurements and obtain a corresponding improvement in accuracy during the ICRR reactivity measurement, it is necessary to remove any significant background signal component from the measurements before using them to calculate a change in reactivity. Prior to US patent 7,894,565, there was no direct method to determine the content of background signals in neutron signal measurements from typical neutron detectors used in commercial nuclear power facilities. One such method is provided by US 7,894,565, but there is still room for improvement in estimating when the core reaches a critical state. In addition, there is a need for a method that can determine whether the core is operating as designed and whether there is an anomaly before the core reaches a critical state. Currently, this analysis can only be performed after the core reaches a critical state as part of a low power physical test procedure that must be successfully concluded before the reactor reaches full power.

Disclosure of Invention

The disclosed concept provides a method for determining K_effA method for total core reactivity deviation of a nuclear reactor core of less than 1. The method comprises the following steps: a measurement step of measuring a subcritical neutron flux (i.e., a measured neutron detector response) for one or more conditions of a nuclear reactor core. The method further includes a calculating step of calculating a predicted value of the spatially corrected subcritical neutron flux (i.e., a predicted neutron detector response) for one or more conditions of the nuclear reactor core. The method then determines the difference between the measured neutron detector response and the predicted neutron detector response and records the difference as the overall core reactivity deviation. In one implementation of the methodIn an example, the measuring step is performed from the output of the source segment detector, and preferably, the measuring step, the calculating step, and the determining step are performed under a plurality of steady state subcritical conditions (i.e., state points). Desirably, multiple steady state subcritical conditions are achieved by repositioning the control rods while maintaining other core conditions at steady state.

The method may further comprise the steps of: regression statistics of the measured and predicted values of the neutron detector responses are used and quantitative measure-predict criteria are applied on the regression statistics to detect various core anomalies while the nuclear plant is in a subcritical state and before the nuclear plant reaches a critical state. The method may further comprise the steps of: the deviation in reactivity between the predicted core and the actual core (i.e., the assembled core after initial construction or refueling) is determined by determining a unified analytical reactivity adjustment, which is the systematic overall reactivity deviation required to bring the measured neutron flux data into agreement with the predicted neutron detector response.

The method may be performed by a processing device programmed to perform the method. Instructions for performing the method may be obtained on a machine-readable medium for use by a processing device in performing the method.

Drawings

A further understanding of the disclosed concept can be obtained from the following description of the preferred embodiments when read in conjunction with the following drawings, in which:

fig. 1 is a schematic diagram of the primary side of a nuclear power generation system.

Detailed Description

FIG. 1 illustrates the primary side of a nuclear power plant 10 in which a nuclear steam supply system 12 supplies steam for driving a turbine generator (not shown) to produce electricity. The nuclear steam supply system 12 has a pressurized water reactor 14 that includes a reactor core 16 housed within a pressure vessel 18. The fission reactions within the reactor core 16 generate heat that is absorbed by the reactor coolant, light water, passing through the core. The heated coolant is circulated through the hot leg tubes 20 to the steam generator 22. Reactor coolant is returned from the steam generator 22 to the reactor 14 by a reactor coolant pump 24 through cold leg tubes 26. Typically, a pressurized water reactor has at least two steam generators and usually three or four steam generators 22, each of which is fed with heated coolant through a hot leg tube 20, which hot leg tube 20 together with a cold leg tube 26 and reactor coolant pumps 2 make up a primary loop. Each primary loop supplies steam to a turbine generator. One such circuit is shown in figure 1.

The coolant returning to the reactor 14 flows downwardly through the annular downcomer and then upwardly through the core 16. The reactivity of the core, and therefore the power output of the reactor 14, is controlled in a short period of time by control rods that can be selectively inserted into the core. The long-term reactivity is regulated by controlling the concentration of neutron moderators, such as boron dissolved in the coolant. As the coolant circulates through the entire core, boron concentration modulation affects the reactivity of the entire core uniformly. On the other hand, the control rods affect the local reactivity, thus resulting in asymmetry of the axial and radial power distributions within the core 16.

The conditions within the core 16 are monitored by several different sensor systems. These systems include an excore detector system 28 that measures the neutron flux escaping from the reactor 14. The excore detector system 28 includes a source zone detector used when the reactor is shutdown, a mid-zone detector used during startup and shutdown, and a power zone detector used when the power of the reactor is above about 5%. In-core detectors are also commonly used during power operations; however, they are not relevant to the present application.

It is often desirable to have an Estimated Critical Condition (ECC) as part of any reactor startup process. ECC is a combination of control rods and primary system conditions (e.g., soluble boron concentration, coolant temperature) that are expected to produce critical reactor conditions. From a reactivity management perspective, it is valuable for the ECC to closely match the actual critical conditions of the core (i.e., the true combination of control rod positions and primary system conditions that produce critical reactor states). Further, the plant specifications include operational constraints (also referred to as LCOs), i.e., the core reactivity is measured within a specified amount of predicted core reactivity. The relevant monitoring is done after each core refueling and before starting power operation (typically > 5% of rated thermal power) and typically once after each month.

Various ECC combinations may be determined by core design predictions prior to reactor core operation. However, prior to reactor criticality, more accurate ECC speculation can be obtained through ICRR monitoring and evaluation, which can identify the presence or absence of any overall core reactivity deviations. The overall core reactivity deviation is defined as the difference between the predicted reactivity state of the core and the actual reactivity state of the core as determined by the measurements. The overall core reactivity deviation may then be incorporated into the updated ECC projections prior to the reactor criticality.

ICRR monitoring is a conventional practice during a shutdown/startup condition, requiring power from a neutron detector (M)_R) Is measured. After a reactive operation (e.g., control rod withdrawal) is performed and a new steady state condition (state point) is reached, another measurement (M) is collected_i)。M_R/M_iThe ratio is defined as the ICRR of state point i. As other reactive operations occur, the ICRR may be updated and monitored according to changes from the reference measurements and then how the reactor progresses toward (or away from) the reactor criticality. If the reactor is intended to be started (i.e., brought into a critical state), a positive response is added to the core (e.g., control rod withdrawal, primary system soluble boron dilution) and the ICRR is expected to be near zero.

As described in U.S. patent No.6,801,593, ICRR is not linear unless the reactor is very close to critical due to the physical nature of the reactions occurring within the reactor core; control rod position changes and proximity to critical conditions as part of a pre-critical test can have a significant effect on the shape of the ICRR curve. Thus, U.S. Pat. No.6,801,593 provides a method for linearizing the measured ICRR as control rod position changes or core conditions change.

The method described in US patent US6,801,593 relies on the use of spatially corrected ICRR (ICRR)_sc) As a measurement parameter, the measurement parameter is a neutron detector measurement value (M)_R/M_i) But depends on the core design by means of Spatial Correction Factors (SCFs). U.S. Pat. No.6,801,593 defines SCF as a function of a static spatial factor and a predicted eigenvalue obtained from subcritical, static calculations with or without a fixed neutron source.

Due to ICRR_scDepending in part on design predictions, ICRR is therefore applied_scThe use as a primary measurement parameter may be inherently affected by masking effects in which errors or deviations in the design predictions also affect the measurement values. Therefore, from the perspective of reactor physics measurements, it is desirable to eliminate the predictive component from the measurements to eliminate potential masking effects. Thus, the disclosed concept first defines the measured ICRR (a "pure" measurement without a predictive component, M)_R/M_i) And predicted ICRR (a "pure" prediction without measurement component but taking into account any spatial effects, at measurement M_RAnd M_iMay be caused by changes in nuclear plant configuration or core conditions) may result in the spatial effect.

After collecting the plurality of ICRR measurements, the measured ICRR may be compared to the predicted ICRR at each state point. It is then possible to quantify the overall reactivity deviation, defined as the linear fit and y-intercept (zero when performing a linear fit of the measured ICRR to the predicted ICRR), by determining a uniform reactivity adjustment to the predicted ICRR at each state point that results in ideal behavior. Fundamentally, the predicted values are adjusted to match the measured values, and the adjustment is used to correct the prediction for future evolution (e.g., eventually approaching a critical state).

Recognizing that (I/M) theory is actually realized by monitoring the measured neutron detector response for changes from a baseline or reference condition, equation (1) is a familiar relationship to nuclear reactor operators.

M_R*(1-k_R)∝M_i*(1-k_i) (1)

Wherein M is_RAnd M_iNeutron detector response, k, at a reference state point condition and a subsequent state point condition, i, respectively_RAnd k_iK under the reference state point condition and the subsequent state point condition i, respectively_effThe value is obtained.

Rearranging the terms yields a new equation (2).

In this form, the left side of the equation is now only the measured count rate ("raw" or measured ICRR, I without spatial correction)_MAnd i) ratio. The right side of the equation consists of core eigenvalues, which can be calculated by nuclear design (predicted ICRR, I)_PI) making a prediction that takes into account spatial effects caused by changes in control stick position or changes in primary system conditions during the measurement. Such separation of the measured value from the predicted value is desirable to eliminate potential masking effects. The simplified form is:

I_M，i∝I_P，i(3)

the true regression of equation (3) can be written as:

I_M＝β₁*I_P+β₀(4)

the final estimate of the true regression, equation (5), may be used as the basis for core design validation prior to performing power plant operations; specifically, the increase in ICRR and the total measured change may be compared to design predictions when the reactor is shutdown. The resulting assessment is not affected by masking effects, and the agreement of the measured values with the predicted values (within a predetermined tolerance range) indicates that the core is in the designed behavior.

Ideally, the constructed measured core is identical to the designed predicted core, such that in the equations(4) β₁Equal to one and β₀Equal to zero. However, in practice, this is not the case; there may be some minor differences in the linear fit between the measured ICRR response and the predicted ICRR response. Regardless of the cause, it is particularly useful to quantify the systematic reactivity deviations so that they can be used for critical state prediction and monitoring purposes.

Returning to equation (2), redefining the reference neutron detector measurements as the normalization constant (C) and rearranging the terms yields the following results:

by combining the normalization constants and the prediction term into the predicted detector response at state points i (pi) that also take into account spatial effects as explained before, equation (6) can be simplified and expressed as a true regression:

M_i＝β₁*P_i+β₀(7)

to quantify the global deviation, the measurements of a set of neutron detectors are fitted to their corresponding predicted values. The final estimate of the true regression is defined in equation (8).

Ideally, the y-intercept of the measured neutron detector response and the predicted neutron detector response is zero. Assuming that the regression estimate is linear and the data points are closely fitted, the overall measurement and predicted reactivity deviation can be estimated by determining the amount of reactivity adjustment required to force the y-intercept (b) to zero for the linear fit defined in equation (8). All state points (by P)_iChange in value) is an estimated core reactivity deviation, which yields a linear fit with a y-intercept (b) of zero.

Thus, the disclosed concept utilizes a direct comparison of the raw subcritical neutron flux measurement at each state point condition with the corresponding predicted value. This is in contrast to existing power reactor physical testing methods, which require calibration of the measurement data prior to evaluation of the results; in case a complete separation of measured and predicted values is used, the benefit of this approach is to prevent masking effects (i.e. to eliminate interdependencies between measured and predicted values).

In addition, the disclosed concept utilizes regression statistics of raw neutron detector measurements to corresponding predicted values, as well as quantitative measure-predict criteria based thereon, to detect various core anomalies when the nuclear power plant is in subcritical conditions and before the nuclear power plant achieves a critical state. The benefits of this approach are: it may provide an additional safety measure because abnormal core conditions may be detected during hot standby testing and anticipated during the eventual approach to critical conditions.

Furthermore, the disclosed concept utilizes a method to determine reactivity deviation between the predicted and actual cores by determining a unified analytical reactivity adjustment (system global reactivity deviation) required to reconcile the measured neutron flux data with the predicted values. This is in contrast to previous power reactor physical test methods that determine reactivity differences based on measured reactivity at critical reactor conditions. The benefits of this approach are: it provides a way to identify abnormal reactivity indications/behaviors in subcritical conditions as a means to provide reactivity management guidance and/or accident prevention. Moreover, the method provides a reactive bias offset directly to a predictive model used in a nuclear power plant safety analysis.

Application of the method requires neutron detector measurements and corresponding core state predictions, which are provided by existing core design specifications and take into account subcritical neutron flux distributions. The basic use of the method is to monitor and infer the subcritical state of the core. Related applications include monitoring negative reactivity conditions or outage margins, and predicting estimated critical conditions prior to plant startup. The method is equivalent to a subcritical physics test, which integrates monitoring and prediction functions to finally perform a series of comparisons of measured values with predicted values to confirm the operational consistency of the constructed core with the design after refueling; this result was previously only achievable in low power tests after the reactor became critical.

The key information required for safe and effective operation of a subcritical reactor core is the negative reactivity of the core; that is, the amount that the core is in a subcritical state, also referred to as the shutdown margin. This information was only inferred, and not directly measured, prior to the development of the methods described herein.

The basic use of the method is to infer and monitor the negative reactivity of the subcritical core for any static configuration of interest (i.e., for steady state combinations of control rod position and primary system conditions) by using neutron detector signal measurements and advanced subcritical core predictions. The comparison of a series of subcritical measured values with predicted values during the startup of the nuclear power plant lays the foundation for an integrated application of the method, i.e. the comparison of measured values with predicted values is performed under a plurality of steady-state subcritical conditions, wherein each of said steady-state subcritical conditions is referred to as a state point.

The method is performed under static conditions and sub-critical conditions (dynamic and critical conditions relative to conventional low power physical testing). The revolutionary nature of this approach is that it is not merely an extension of the steps performed during low power physical testing. Rather, this approach achieves the same goals as low power physical testing; after refueling and before returning to normal operation, tests are performed to determine if the operating characteristics of the core are consistent with the design predictions to ensure that the core can operate as designed.

Performing this method while achieving the same goals as low power physical testing yields inherent safety, human performance, and test performance advantages over low power physical testing. Performing measurements under static and subcritical conditions may substantially enhance safety and reactivity management of the nuclear power plant. The method is seamlessly integrated into the routine startup activities of the nuclear power plant without the need for infrequently performed tests and upgrades and special tests that are exceptional to the nuclear power plant operation, which improves the reliability and human performance of the tests. Thus, core design based on this approach is validated to provide broad benefits for almost any type of nuclear power plant.

It should be appreciated that the methods described herein may be performed by a processor or processing device of a computer system or by other means of performing the function. Thus, a processor with the necessary instructions, which may be programmed directly in the processor or on a machine-readable medium accessed by it for performing such a method or method element, forms a means for performing the method or element of the method. Further, the elements of an apparatus embodiment described herein are examples of means for performing the functions performed by the elements to carry out the invention.

While specific embodiments of the disclosed concept have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the disclosed concept which is to be given the full breadth of the appended claims and any and all equivalents thereof.