JPH0531755B2 - - Google Patents

Description

[Detailed Description of the Invention] [Technical Field of the Invention] The present invention relates to a nuclear instrumentation device for a nuclear reactor, and particularly to a method for accurately calibrating the gain between neutron monitors of a nuclear instrumentation system during startup or shutdown of a nuclear reactor. Regarding nuclear instrumentation equipment for nuclear reactors.

[Technical Background of the Invention] Conventionally, various neutron monitors have been used during startup, operation, or shutdown of a nuclear reactor.
The gains between these neutron monitors are calibrated in accordance with the plant conditions of the reactor.

That is, for example, in a boiling water reactor, calibration between each neutron monitor is performed as follows according to each state of the startup curve shown in FIG.

First, we will explain the startup of a boiling water reactor.

After the control rods are withdrawn from the fully inserted state, which is the reactor's shutdown state, to increase the neutron flux and reach a critical state, the neutron flux and core thermal output are further increased by withdrawing the control rods, and the reactor water is Transition to nuclear heating that increases temperature and furnace pressure. After that, after the core pressure reaches approximately 65 kg/cm ² and the core thermal output is around 10% of the rated value, the reactor mode switch is switched from "startup" to "operation" and the generator is attached. Furthermore, by withdrawing the control rod, the rated value
The output is increased to 60% and the target control rod pattern is created.

During this period, the nuclear instrumentation system ^is operated by the ^startup region ^neutron monitor ( (hereinafter referred to as SRM) is used, and the process from criticality through temperature and pressure increases to mode switch switching is 10 ⁸ to ^{10 13} n・cm ² /sec.
In the neutron flux region, an intermediate output neutron monitor (hereinafter referred to as IRM) is used.
After that, 10 ¹² to 10 ¹⁴ n after switching the mode switch.
In the cm ² /sec neutron flux region, the neutron flux is monitored by an average power neutron flux monitor (hereinafter referred to as APRM) and a local power system neutron monitor (hereinafter referred to as LPRM).

Here, APRM is divided into channels
The LPRM readings are averaged and calibrated periodically during reactor operation by a probe in the movable core (hereinafter referred to as TIP).

On the other hand, SRM and IRM calibration is performed during the reactor start-up test as follows.

That is, it is determined by comparing the SRM gain with the count value for the neutron source and satisfying the SRM lower limit condition.

Next, when the count rate of SRM reaches the level of 10 ⁵ cps, the respective average values of SRM and IRM are set so that they overlap by more than 1 decade.
Calibrate the IRM gain. At this time, it is determined that IRM turns on scale when the SRM counting rate is 10 ⁵ or more.

Furthermore, when switching between IRM and APRM,
It is determined that the average values of IRM and APRM overlap by one decade or more, and APRM turns on scale when the IRM reading value is in the highest range.

At this time, the gains of APRM and LPRM are used as nominal values or calibrated values from the previous test.

[Problems with Background Art] However, the method of calibrating the gain between each neutron monitor as described above has the following problems.

In other words, in general, the gain of SRM and IRM is that only the core average value of neutron flux is the gain of IRM, respectively.
The axial and radial positions of each neutron flux detector are different, as they are calibrated to overlap according to and APRM.
This creates the problem of having to recalibrate each time the control rod pattern changes.

Figure 3 shows the position of each neutron flux detector in a cross section within the core. In this example, marked with an Indicate with △ marks at four locations in D.
IRMs are located at 8 locations E, F, G, H in the center of the core and I, J, K, L on the outer periphery, and are also marked with ●.
LPRMs are placed in 43 locations every 4 bundles.

Figure 4 shows the axial arrangement of each of these neutron flux detectors, where 1 is the fuel assembly and 2 is the fuel assembly.
SRM, 3 IRM, 4 LPRM, 5 fixed neutron source and 6 TIP. Similar to the fixed neutron source 5, the SRM2 and IRM3 are located at the lower part of the core and about 2/3 of the height of the fuel assembly 1;
LPRM4 is arranged at equal intervals at four locations in the axial direction, and is calibrated by TIP6.

As described above, since the spatial arrangement of each neutron flux detector is different, each reading value will be different depending on a change in the control rod pattern. In other words, if the control rod pattern in the cross section in the x direction of Figure 3 is created as shown in Figures 5a and 5b for equivalent core thermal output, the neutron flux distribution in this case will be as shown in Figure 6. .

Therefore, if the control rod patterns are different for the same core thermal power with APRM readings,
IRM readings will show different values,
This creates an IRM channel that cannot overlap with APRM, resulting in an unfavorable situation for the safe operation of the reactor.

[Object of the invention] The present invention provides a method for determining thermal output, or APRM readings.
This was done with the focus on the fact that by accurately understanding the relationship with IRM readings, it is possible to accurately know the state of the reactor at startup and shutdown, regardless of changes in control rod patterns. An atomic technology that improves operational safety by determining the calibration coefficient of the nuclear instrumentation system of a nuclear reactor from various data and easily and accurately switching the nuclear instrumentation system during startup and shutdown. The purpose is to provide reactor nuclear instrumentation equipment.

[Summary of the Invention] That is, the present invention provides intermediate power system neutron flux detectors that are distributed in the core of a nuclear reactor, and a system that calculates a neutron flux distribution based on the plant state of the reactor, or A neutron flux distribution output device that inputs actual measured values from a movable in-core probe located symmetrically with the neutron flux detector and estimates the reading of the intermediate power system neutron flux detector, and this neutron flux distribution output device The estimated reading value of each intermediate output system neutron flux detector is compared with the actually measured reading value of each intermediate output system neutron flux detector, and the absolute value of each deviation is below a predetermined tolerance value. and a neutron flux gain setting device for setting the gain of each intermediate power system neutron flux detector.

[Embodiment of the Invention] The present invention will be described in detail below with reference to an embodiment shown in the drawings.

FIG. 1 shows the configuration of a nuclear reactor instrumentation system according to an embodiment of the present invention. In the figure, numeral 7 is a plant status input device, 8 is a neutron flux distribution calculation device, and 9 is a neutron flux detector gain setting. equipment, 10 is a nuclear reactor, 1
1 indicates a neutron flux detector.

Various signals M indicating the plant status at the time of starting or stopping the nuclear reactor 10 are sent to the plant status input device 7.
is input. This signal M is data regarding reactor pressure, core inlet enthalpy, core flow rate, and thermal power.The thermal power data includes the average value of APRM readings after APRM calibration, and the average value of APRM readings before APRM calibration or when APRM cannot be used. In such cases, values calculated based on the thermal balance of the reactor are used.

The various signals inputted to the plant status input device 7 are then inputted as a signal N to the neutron flux distribution calculation device 8, where these data are used to solve the core three-dimensional diffusion equation. The neutron flux at the location is determined.

The neutron diffusion equation is given by the following equation.

−D ₁ D ² φ+Σ ₁ φ ₁ = 1−β/λνΣ〜fφ ₁ ... Here, D ₁ , Σ ₁ , φ ₁ , λ, νΣ〜f are respectively,
D ₁ , Σ ₁ , νΣ~f are functions of reactor pressure, core inlet enthalpy, core flow rate, and core thermal power. It is.

Thermal neutron (φ ₂ ) at neutron flux detector position rj
is determined by the following equation using the fast neutron flux φ ₁ which is the solution of the equation.

φ ₂ (rj)=gj・f _∞ j ・φ ₁ (rj) ... j=1, 2, ...-s Here, gj and f _∞ j are the neutron swelling at the neutron flux detector position rj, respectively. Factors to consider and infinite neutron spectrum. Also, Ns indicates the number of neutron flux detectors.

The thermal neutron flux φ ₂ at the neutron flux detector position rj determined in the neutron flux distribution calculation device 8 in this way is inputted to the neutron flux detector gain setting device 9 as a signal O, and the reading of the neutron flux detector is The gain of the neutron flux detector is set so that the reactor operation restriction conditions are satisfied and the gain is sufficiently close to the target gain value.

This gain setting is performed by the method described below.

Detector reading at neutron flux detector position rj
Rj is expressed as Rj=Sj·φ ₂ (rj) . . . j=1, 2, . . . , Ns, where Sj is the gain of the j-th detector. Here, to adjust Sj, if φ゜ and R゜ are respectively the core average value of neutron flux and the average value of neutron flux detector readings, then the reading value Rj゜ of each detector is Rj゜=R゜/ φ゜φ ₂ (rj) ... j=1, 2, ... , Ns.

The absolute value of the difference between the reading value Rj゜ of the neutron flux detector obtained in this way and the actual reading value R _* j of the neutron flux detector obtained as the output signal P of the neutron flux detector 11 is the allowable value. A gain adjustment signal Q of the neutron flux detector is sent from the neutron flux detector gain setting device 9 to the neutron flux detector 11 so that the gain is smaller than ε.

In the above embodiment, the neutron flux detector position
Thermal neutron flux φ ₂ at rj was calculated by the neutron flux distribution calculation device 8, but the present invention is not limited thereto. The gain of the neutron flux detector can be set by directly determining the flux.

In other words, if the reactor core satisfies 1/4 symmetry as shown in Figure 3, the LPRM string will be divided into bundles of each cell due to symmetry, considering the four bundles around the control rods as one cell. The value of each corner adjacent to is determined.

For example, IRMI is at the same position as the LPRM string L ₁ due to symmetry in the plane of symmetry that includes the Y direction. Furthermore, since each LPRM is equipped with a TIP, it is possible to measure the relative power distribution along the core axis and calibrate the LPRM and other neutron flux detectors.

[Effects of the Invention] As described above, according to the reactor nuclear instrumentation device of the present invention, the gain calibration of the neutron flux detector is appropriately set by the setting signal from the neutron flux detector gain setting device. Therefore, the dependence of the neutron flux detector gain on the control rod pattern is reduced, and the absolute value of the neutron flux detector reading can be more accurately grasped, which improves the nuclear instrumentation system during reactor startup and shutdown. Switching becomes easier. Furthermore, since the readings of each neutron flux detector overlap during this switching, the number of bypassed detectors is also reduced, and the safety of reactor operation can be improved.

[Brief explanation of drawings]

Fig. 1 is a block diagram showing a reactor nuclear instrumentation system according to an embodiment of the present invention, Fig. 2 is a graph showing a startup curve of a boiling water reactor, and Fig. 3 is a neutron flux detection of a boiling water reactor. 4th diagram of the radial arrangement of the reactor in the core
The figure is a layout diagram of the neutron flux detector in the axial direction within the core, and Figures 5a and 5b are explanatory diagrams showing examples of control rod patterns.
FIG. 6 is a graph showing the neutron flux distribution in the radial direction. 1...Fuel assembly, 2...SRM, 3...IRM, 4...
LPRM, 5... Fixed neutron source, 6... TIP, 7... Plant status input device, 8... Neutron flux distribution calculation device, 9... Neutron flux detector gain setting device, 10...
Nuclear reactor, 11...neutron detector.

Claims (1)

[Scope of Claims] 1. Intermediate power system neutron flux detectors distributed within the core of a nuclear reactor; a neutron flux distribution output device that inputs actual measured values by a movable in-core probe arranged at a position symmetrical to the flux detector and estimates the reading value of each of the intermediate power system neutron flux detectors, and this neutron flux distribution output Compare the readings of each intermediate output system neutron flux detector estimated by the device with the actually measured readings of each intermediate output system neutron flux detector, and check that the absolute value of each deviation is less than or equal to a predetermined tolerance value. A nuclear reactor nuclear instrumentation system characterized by comprising a neutron flux gain setting device for setting the gain of each of the intermediate power system neutron flux detectors.