JPH0249476B2 - - Google Patents

Description

[Detailed Description of the Invention] [Technical Field of the Invention] The present invention relates to a method and apparatus for monitoring the stability of thermo-hydraulic vibrations of a fuel bundle in a boiling water nuclear reactor, and in particular, to a method and apparatus for monitoring the stability of thermal-hydraulic vibrations of a fuel bundle in a boiling water nuclear reactor, and in particular, to The stability is monitored by focusing on the peaking coefficient, the peaking coefficient of the axial power distribution, and the axial peak position, and if the stability deteriorates, control rods around the fuel bundle are inserted to reduce the output of the fuel bundle. The present invention relates to a nuclear reactor output control method and device for lowering and stabilizing the output of a nuclear reactor.

[Technical background of the invention]

Cooling water that passes through fuel bundles in the core of a boiling water reactor generally forms a two-phase flow of water and steam to remove heat generated within the core, but this two-phase flow includes voids, It is known that there is a possibility of thermal-hydraulic oscillations based on feedback between pressure drop and flow rate.

There are many fuel channels in the reactor core, and since they form parallel flow paths, each fuel bundle undergoes thermal-hydraulic oscillations under the condition of a constant differential pressure between the inlet and the outlet. It can be considered that this occurs independently of the channel.

Furthermore, due to the locality of vibration, feedback via nuclear properties is considered to be secondary, so the first
This can be ignored as an approximation.

Such a mode of vibration is called channel stability, while a mode of vibration in which the feedback phenomenon of the two-phase flow of the fuel bundle, the core characteristics of the core, and the dynamic characteristics of the recirculation flow path are related is called core stability. There is.

Core stability during actual nuclear power plant operation can be monitored by global signals such as core average neutron flux signal, recirculation flow rate, etc., but channel stability can be monitored locally. Because it is a phenomenon, it is not expected that it can always be monitored only by the readings from the in-core neutron detector.

Therefore, at the design stage, channel stability is thoroughly analyzed to determine the range in which stable operation can be achieved.

[Problems with background technology]

The operating conditions of a nuclear power plant that has started operation may change over a long period of time due to changes in the type of fuel, operational constraints, etc. In that case, it is natural to conduct stability analysis to determine the operating range. However, it is not wrong to say that the stability limit of channel stability in actual nuclear reactors changes from moment to moment, including small changes that do not lead to this.

Furthermore, since the operating range at the above-mentioned design stage is determined based on the most severe conditions, it is thought that there is a considerable stability margin in normal cases.

Therefore, if the stability of such channel stability can be evaluated for all fuel bundles under actual operating conditions, channel stability can be monitored thereby.

[Purpose of the invention]

The present invention has been made based on such conventional circumstances, and is capable of evaluating and monitoring the thermal-hydraulic stability of a large number of fuel bundles in a reactor core in accordance with each operating state, and is capable of suppressing local vibrations. It is an object of the present invention to provide a method for controlling the output of a nuclear reactor and a device for controlling the output thereof, which can prevent such occurrence.

[Summary of the invention]

That is, the nuclear reactor output control method of the present invention is as follows:
In a nuclear reactor in a power operation state, a first coefficient representing the power of the fuel bundle in the reactor core relative to the core average, a second coefficient representing the peak height of the axial power distribution shape of the fuel bundle, and a second coefficient representing the peak height of the axial power distribution shape of the fuel bundle; A third coefficient representing the peak position is determined, and the thermal-hydraulic stability of the fuel bundle is determined based on these first to third coefficients. The reduction ratio, which is a parameter representing stability, is expected to increase based on the stability analysis results based on the reference fuel bundle output and axial output distribution. If the stability margin is below a predetermined value, the stability is recalculated using the fuel bundle output and axial output distribution shape that corresponds to the actual operating conditions. It is characterized by inserting control rods into the reactor core to reduce output and ensure sufficient stability. The power control device for a nuclear reactor of the present invention also includes a core performance calculation device that performs nuclear thermal and hydraulic calculations in the core of a nuclear reactor, and a nuclear thermal and hydraulic power calculation device based on the nuclear thermal and hydraulic calculation values calculated by this core performance calculation device. The first coefficient represents the power of the fuel bundle in the core relative to the core average, which is the value of the stability evaluation parameter, the second coefficient represents the peak height of the axial power distribution shape of the fuel bundle, and A stability parameter calculation device that calculates the third coefficient representing the peak position of the power distribution, a stability analysis of the reference fuel bundle output and axial power distribution shape, and a stability parameter calculation device that is used for this stability analysis. a stability calculation device that calculates the stability evaluation parameters for the output and axial output distribution shape of the fuel bundle that serve as a reference, and the stability calculated by the stability calculation device and the stability parameter calculation device, respectively; The value of the stability evaluation parameter calculated by the stability parameter calculation device is greater than the value of the reference stability evaluation parameter calculated by the stability parameter calculation device. and a comparison device that causes the stability calculation device to recalculate the stability when the stability evaluation parameter does not satisfy predetermined criteria regarding the magnitude relationship of these stability evaluation parameters; The fuel bundle is characterized by a control rod insertion device that inserts control rods around the fuel bundle into the reactor core when the stability value is close to unstable.

[Embodiments of the invention]

The details of the present invention will be explained below.

In general, there are many parameters that can affect channel stability.
Within the normal operating range of a nuclear power plant, the issues are the power and flow rate of each fuel bundle, as well as the axial power distribution.

Regarding the output and flow rate, the higher the output and the lower the flow rate, the worse the channel stability becomes.

Therefore, in general, the higher the output and the lower the flow rate of operation, the worse the stability becomes.

In addition to the output and flow rate, another factor that has a large effect on channel stability is the output distribution in the axial direction.

In other words, this axial power distribution shape not only has a great influence on channel stability, but also changes considerably in actual operating conditions due to control rod insertion and withdrawal, Zenon effects associated with combustion, etc. Since each fuel channel has a different shape, it is necessary to take into account the effect of the axial power distribution shape in order to evaluate the stability in accordance with actual operating conditions.

By the way, in general, the relationship between the axial power distribution shape and the stability of the channel stability is that the higher the peak height of the axial power distribution shape, and the closer the peak position is to the core (fuel It can be said that the lower the bundle (bundle), the worse the stability.

In other words, if the peak height is high and the peak position is at the bottom, it means that the amount of heat generated is large at the bottom, and if the amount of heat generated at the bottom is large, the boiling point of the coolant will be at the bottom. , it can be considered that stability deteriorates because the region of two-phase flow, which is a destabilizing factor, becomes longer.

In this embodiment, taking the above points into consideration, channel stability is monitored by focusing on the output of the fuel bundle and the axial output distribution shape.

That is, the output coefficient P ^r _j of the bundle is used as a parameter representing the magnitude of the output of the j-th fuel bundle among the large number of fuel bundles accommodated in the reactor core.
Use.

This P ^r _j can be expressed by the following formula.

P ^r _j =Q _j /(Q/N) where Q _j is the output of the j-th fuel bundle, Q
is the power of the entire reactor core, and N is the number of fuel bundles in the reactor core.

Furthermore, the peaking coefficient and peak position of the axial output are used to characterize the axial output distribution.

Generally, the axial power distribution shape is, for example, the first
As shown in the figure, the fuel bundle is calculated by dividing it into a large number of nodes, for example 1 to 24, along the axial direction, and this will be followed here as well.

In Fig. 1, the output is plotted on the horizontal axis, and the axial nodes are plotted on the vertical axis, and each curve is
Each shows the power distribution of the fuel bundle.

Now, if the axial output coefficient at the i-th node in the axial direction of the j-th fuel bundle is P ⁱ _j , P ⁱ _j can be expressed by the following equation.

P ⁱ _j =Q ⁱ / _j /(Q _j /N _ax ) where Q ⁱ _j is the output at the i-th node of the j-th fuel bundle, and N _ax is the number of nodes in the axial direction.

The largest one among these P ⁱ _j is defined as the axial peaking coefficient P ^ax _j .

Therefore, it can be expressed as P ^ax _j = ^max _i (P ⁱ _j ).
Using this P ^ax _j , the height of the peak of the axial output distribution type can be evaluated.

That is, the larger P ^ax _j becomes, the higher the peak height becomes.

Next, the following quantities are defined as parameters for expressing the position of the peak of the axial output distribution.

P ^l _j = (N _ax − N ^p / _j ) / N _ax where N ^p _j is the node number where the peak of the j-numbered fuel bundle exists, and the node numbers are numbered sequentially upward from the bottom of the fuel bundle. shall be provided.

According to this P ^l _j , it can be seen that the larger the value of P ^l _j, the lower the peak position is in the fuel bundle.

The axial power distribution shape ^of the j-th fuel bundle is the product of the two parameters P ^ax _j and P ^l _j defined above, and is determined by P p j defined as P ^p _j = P ^ax _j・P ^l _{j .} can be evaluated based on

In addition, the following parameter P _j including the output size
Define.

P _j = P ^r _j・P ^p _j = P ^r _j・P ^ax _j・P ^l _jThese values are combined with the stability analysis performed in advance, that is, the channel stability to determine the possibility of thermo-hydraulic oscillations. By comparing the values of the above-mentioned various parameters determined from the combustion bundle output and axial output distribution used in the analysis, it is possible to evaluate the stability of the channel stability in accordance with the actual operating point.

FIG. 2 shows a nuclear reactor power control device according to an embodiment of the present invention. The reactor power control device shown in the figure includes a core performance calculation device 1, a stability calculation device 2, a stability parameter calculation device It consists of a device 3, a stability parameter comparison device 4, and a control rod insertion device 5. Note that the dotted line indicates the case where the criteria are not satisfied when comparing the stability parameters.

That is, the core performance calculation device 1 inputs process quantities such as core flow rate, core pressure, core power, average neutron flux, and control rod pattern from the reactor 6, and calculates the output of each fuel bundle based on these process quantities. , calculate the flow rate and axial power distribution.

The stability parameter calculation device 3 inputs the output and axial power distribution of each fuel bundle calculated by the core performance calculation device 1, and based on these, calculates the above-mentioned parameters P _j , P ^r _j , P ^ax _j , Calculate the value of P ^l _j .

The stability calculation device 2 determines the reference axial distribution, analyzes the channel stability for the most unstable fuel bundle, and calculates the operating point and width reduction ratio (D,
The relationship with R) is determined in advance as an equal reduction width ratio line as shown in FIG.

In other words, normally, in stability analysis, the standard axial output distribution is set, and the stability is evaluated based on it using the width reduction ratio shown in Figure 4. (D, R)
When D exceeds 1, the vibration diverges, and the smaller D and R are, the more stable it is.

Then, the stability calculation device 2 determines the values of each of the above-mentioned parameters based on the channel stability analysis result for the most unstable fuel bundle analyzed by the stability calculation device 2.

Note that the values of each parameter are denoted by P ^r _c , P ^ax _c , P ^l _c , and P _c with a subscript c.

Here, P _c =P ^r _c・P ^ax _c・P ^l _c .

The stability parameter comparison device 4 compares the values of the parameters P ^r _c , P ^ax _c , P ^l _c , P _c calculated by the stability calculation device 2 and the parameters P _j , P calculated by the stability parameter calculation device 3. Compare the values of ^r _j , P ^ax _j and P ^l _j .

However, this comparison is performed only in the area in FIG. 3 where the width reduction ratio exceeds a predetermined value smaller than 1 and is smaller than 1.

When the attenuation ratio reaches a region exceeding 1, a neutron absorbing material is immediately inserted to lower the output to a region where the attenuation ratio becomes less than 1.

The comparison of these various parameters will be performed in the following steps.

First, the stability of various parameters is quantitatively evaluated in advance, and the relationship between the range of change of the various parameters and the range of change of the width reduction ratio is investigated.

Next, each parameter is compared, and if there is a fuel bundle for which the value of the parameter calculated by the stability parameter calculation device 3 is larger than the value of the parameter calculated by the stability calculation device 2, that is, the actual stability is If it is predicted that the stability will be worse than the results of the stability analysis performed in advance, the stability calculation device 2 will calculate the stability of the fuel bundle using realistic analysis conditions (fuel bundle output and axial output distribution). We will recalculate the stability.

However, by setting a certain width for each parameter,
Calculation is performed when one of the parameters satisfies one of the following formulas.

That is, P ^r _j > P ^r _c + ΔP ^r P ^ax _j > P ^ax _c + ΔP ^ax P ^l _j > P ^l _c + ΔP ^l P _j > P _c + ΔP...() Calculate if any of the following formulas is true.

Also, even if the parameter becomes large, if the difference falls within one of the following, P ^r _c + ΔP ^r ≧P ^r _j ＞P ^r _c P ^ax _c + ΔP ^ax ≧P ^ax _j ＞P ^ax _c P ^l _c + ΔP ^l ≧P ^l _j ＞P ^l _c P _c +ΔP ≧P _j ＞P _c ... () If at least two or more of the four parameters apply to the above equation, the calculation will be recalculated.

If the result calculated by the stability calculation device 2 is close to instability, that is, if the width reduction ratio is less than or equal to 1 and close to 1, the control rod insertion device 5 inserts the control rods around the fuel bundle. Insert it to reduce the output of the fuel bundle and maintain sufficient stability.

〔Effect of the invention〕

As described above, according to the present invention, it is possible to evaluate and monitor the thermal-hydraulic stability of a large number of fuel bundles in a reactor core in accordance with each operating state, and to prevent the occurrence of local vibrations. becomes possible.

[Brief explanation of the drawing]

FIG. 1 is a graph showing the axial power distribution of a fuel bundle, FIG. 2 is a block diagram showing a nuclear reactor power control device according to an embodiment of the present invention, and FIG.
Graph showing the relationship between the operating point and the constant reduction width ratio line, 4th
The figure is a graph for explaining the width reduction ratio. 1... Core performance calculation device, 2... Stability calculation device, 3... Stability parameter calculation device, 4... Stability parameter comparison device, 5... Control rod insertion device.

Claims (1)

[Scope of Claims] 1. In a nuclear reactor in a power operation state, a first coefficient representing the power of the fuel bundle in the reactor core relative to the core average, and a second coefficient representing the peak height of the axial power distribution shape of the fuel bundle. and a third coefficient representing the peak position of the power distribution, and based on these first to third coefficients, the thermal-hydraulic stability of the fuel bundle is calculated based on the coefficients used in the stability analysis performed in advance. In addition, it is monitored by comparison with the reference fuel bundle output and axial output distribution shape, and the stability analysis results using the reference fuel bundle output and axial output distribution shape are used to determine the width reduction, which is a parameter representing stability. If the ratio is expected to increase, and the stability is recalculated using the fuel bundle output and axial output distribution shape that match the actual operating conditions, and the stability margin falls below the predetermined value. A method for controlling the output of a nuclear reactor, which comprises: inserting control rods around the fuel bundle into the reactor core to lower the output and ensure sufficient stability. 2. A core performance calculation device that performs nuclear thermal-hydraulic calculations in the core of a nuclear reactor, and a core performance calculation device that calculates the stability evaluation parameters based on the nuclear thermal-hydraulic calculation values calculated by this core performance calculation device. a first coefficient representing the power of the fuel bundle relative to the core average, a second coefficient representing the peak height of the axial power distribution shape of the fuel bundle, and a third coefficient representing the peak position of the power distribution. A stability parameter calculation device is used to perform stability analysis on the reference fuel bundle output and axial output distribution shape in advance, and the reference fuel bundle output and axial output used in this stability analysis are A stability calculation device that calculates the value of the stability evaluation parameter for the distribution shape and the stability evaluation parameter values calculated by this stability calculation device and the stability parameter calculation device, respectively, are compared to calculate the stability. The value of the stability evaluation parameter calculated by the stability parameter calculation device is larger than the value of the reference stability evaluation parameter calculated by the stability parameter calculation device, and these stability evaluation parameters a comparison device that causes the stability calculation device to recalculate the stability when a predetermined judgment criterion regarding the magnitude relationship of is not satisfied, and a stability value recalculated by the stability calculation device is close to unstable 1. A power control device for a nuclear reactor, comprising a control rod insertion device for inserting control rods around a fuel bundle into a reactor core.