JPS60213887A - Nuclear reactor pipe rupture detection method and device - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to the detection of broken pipes and the fracture area during a loss of coolant accident in a nuclear reactor, and the present invention relates to the detection of broken piping and the fracture area during a loss of coolant accident in a nuclear reactor. The present invention relates to a method and apparatus for detecting a pipe break in a nuclear reactor, which is suitable for detecting a pipe break and a break area.

[Background of the invention]

The configuration of a conventional boiling water reactor emergency core cooling system is shown in FIG. 1, and the operating conditions are shown in FIG. 2. When a rupture occurs in the coolant pipe 3 that transports the heat generated in the reactor core 1 to the turbine 5 (
Loss of coolant accident), coolant leaked and reactor pressure vessel 2
The water level in the reactor 1 falls, the reactor 1 scrams, and the uranium fission reaction stops. However, heat generation (decay heat) of the fission products continues to occur. The pressure and water level within the reactor pressure vessel 2 further decrease, and the pressure within the containment vessel 4 that houses the reactor pressure vessel 2 increases. As shown in FIG. 2, when the water level in the reactor pressure vessel 2 falls from the normal water level through the scram level to the low water level, radioactive materials in the coolant are released into the coolant pipes that penetrate the containment vessel 4 ( In order to prevent leakage to the outside of the containment vessel 4 through the main steam piping) 3, the main steam isolation valve 6 is closed.

At the same time, for the purpose of depressurizing the reactor pressure vessel 2 and mitigating the rise in fuel rod temperature of the reactor core 1 due to decay heat generation and coolant reduction, the emergency diesel start-up time (approximately 30 seconds), water injection into the core 1 is started. Furthermore, when the water level in the reactor pressure vessel 2 falls to the low water level, the valve 8 of the automatic pressure reduction system is activated to reduce the pressure in the reactor pressure vessel 2 based on the product logic with the high pressure signal of the containment vessel 4. Then, the steam in the reactor pressure vessel 2 is released to the sub-reduction/pool 9 to lower the pressure. When the pressure in the reactor pressure vessel 1 further decreases, the low-pressure core spray 10 and the low-pressure core water injection system 11 are activated to start injecting coolant into the reactor core 1 for the purpose of spray cooling and submergence cooling of the fuel rods. .

As described above, the operating conditions of the current emergency core cooling system shown in FIG. 2 are constant regardless of the type of broken pipe or the size of the broken area. Therefore, current operating conditions are not necessarily optimal for all failure modes. For example, in the case of high-pressure core sprain operation, when a small rupture occurs in a pipe occupied by liquid phase coolant, the temperature of the fuel tube becomes the highest, although it is below the safety standard. The temperature can also be lowered by changing the operating conditions of the automatic pressure reduction system.

Therefore, if the rupture mode can be detected, the emergency core cooling system can be activated under the optimal operating conditions for each rupture, and the reactor can be safely brought to a cold shutdown state.

Conventional methods for detecting pipe breakage in a nuclear reactor include detection based on reactor pressure or containment vessel pressure, as well as measurement signals of both.

There is no rupture detection method based on the signal from the water level needle that measures the residual water in the reactor pressure vessel, which changes as the rupture occurs. Experiments simulating loss of coolant accidents have shown that in the event of a large rupture, the water level drops instantly, which is not optimal for rupture detection.
Compared to water level changes, gradual reactor pressure or containment vessel pressure is more effective for rupture detection. However, in the case of small fractures, which occur more frequently than large fractures, pressure changes are generally slow, and the detection sensitivity is not necessarily good.Contrary to large fractures, it is more reliable to rely on water level changes.

[Object of the Invention] The object of the present invention is to provide an atomic system for detecting fractured piping and fracture area with high reliability based on the signal of the reactor pressure vessel water level gauge during a loss of coolant accident in a nuclear reactor, especially in the case of a small fracture. An object of the present invention is to provide a method and apparatus for detecting a pipe break in a furnace.

[Summary of the invention]

The following describes the results of a simulation experiment of small breaks in various pipes of a boiling water reactor, simulating high-pressure core sprain operation, which is the most severe safety issue. The piping that is assumed to be ruptured is the main steam, water supply, high-pressure core slag, and recirculation system piping, as shown in Figure 3. Figure 3 shows the connection height of each piping to the reactor pressure vessel. is shown.

Figure 4 shows the reactor pressure changes at the time of small pipe breaks. When there is a small break in the main steam and water supply piping, where the connection to the pressure vessel is high and the gas-phase coolant mainly flows out from the break, the reactor pressure gradually decreases.

On the other hand, in the case of high-pressure core spray or recirculation system piping rupture that connects to the reactor pressure vessel at a low location, the reactor pressure increases after the main steam isolation valve closes, but is maintained within a predetermined range by the operation of the safety relief valve. . Regardless of the type of pipe rupture, a small rupture is characterized by the fact that the reactor pressure remains almost constant until the automatic depressurization system is activated.

Figure 5 shows changes in the water level in the downcomer when each pipe breaks small. The outflow mass flow rate W of the coolant from the fracture opening is given by the product of the critical mass velocity Gc and the fracture opening area A1. , In the experiment shown in Fig. 5, since Aa is the same for each pipe break, G
The outflow flow rate W changes depending on the size of c, and the water level drop in the downcomer varies. In other words, if the high-pressure core spray or recirculation system piping connected to the lower part of the reactor pressure vessel ruptures, there will be no liquid-phase coolant outflow, and the critical mass velocity Gcl for liquid-phase outflow is the same as the critical mass velocity for gas-phase outflow at the same pressure. Get
Since the pipe is also large, the water level in the downcomer will fall more quickly in the event of a large outflow or a rupture of the main steam supply pipe.

FIG. 6 shows the change in the maximum value (maximum cladding tube temperature) of the surface temperature of an electric heater simulating a nuclear reactor fuel rod at each time. In the case of main steam and feed water piping rupture (mainly when gas-phase coolant flows out), the maximum cladding temperature is only slightly higher than the initial temperature, but liquid-phase coolant flows out and the high-pressure reactor core with a large flow rate In the event of a rupture of the spray or recirculation system piping, there is sufficient margin to meet safety standards, but the safety standards are also high in the event of a major rupture. This increase in maximum cladding temperature can be prevented by changing the operating conditions of the non-commercial core cooling system.

To summarize the experimental results above, in the case of a small rupture, which corresponds to a crack in a pipe or a malfunction of a valve, and which occurs more frequently than a large rupture, (1) the reactor pressure remains in the ruptured pipe until the automatic decompression device operates; It remains constant regardless.

(2) When a pipe ruptures at the bottom of the pressure vessel where the liquid coolant flows out from the rupture port, the maximum cladding tube temperature increases even in the case of a major rupture, but when the pipe ruptures at the top of the pressure vessel, it hardly rises.

It can be said.

FIG. 7 shows the change in the remaining water volume in the downcomer, which is obtained by multiplying the downcomer flow path area at each water level based on the downcomer water level changes shown in FIG. 5. The following can be clearly seen from Figure 5 or more clearly from Figure 7.

(1) Until the water level in the downcomer falls and the broken pipe is exposed, the liquid phase coolant continues to flow out and the reactor pressure is almost constant, so the critical flow rate Gc remains constant and the remaining water volume (2) decreases linearly according to dV=Gatedt ρ6 (constant). Note that ρt indicates the density of residual water.

(cutting) When the broken pipe is exposed, the outflow coolant becomes a gas phase and p1
The critical mass velocity Gc suddenly changes, and the volume reduction rate O changes. Thereafter, the critical mass velocity Gc becomes constant due to the outflow of gas lees, so the mass of the remaining water decreases linearly with a gentler slope than the outflow of the liquid phase coolant.

From the above study, if the change in the occupied volume reduction rate of the remaining cooling water (exposure of the broken pipe) is detected from the water level gauge reading and the pre-known downcomer flow path cross-sectional area, the following method can be used to detect the broken pipe and It becomes possible to detect the fracture area.

As shown in FIG. 8, since the volume occupied by the downcomer of the remaining coolant at the time of exposure takes a value specific to each broken pipe, a broken pipe can be detected from the remaining water volume at the time of change in the volume reduction rate. The swimming method is particularly effective in small breaks where the water level changes relatively slowly.

On the other hand, in the event of a major rupture, the depressurization rate of the reactor pressure vessel is large;
There is uncertainty in the converted residual water volume value due to the drift of the water level needle or fluctuations in the indicated value.

However, even in this case, the break exposure time can be detected from the change in the average slope of the time change of the water level gauge indication value. Furthermore, as shown in Fig. 8, for example, the water level drop time ΔTA from the scram water level to the exposed water level at the fracture opening and the drop time °ΔTs from the exposed water level to the reference water level set below the exposed water level are measured, and the ratio ΔT^/ If we take ΔTi, we get ρtΔV^, and ΔVA/ΔVm takes a value specific to each pipe,
Since G-/G and t take a single value when the furnace pressure is constant at the beginning of a fracture or a small fracture, ΔTA/ΔTi is equal to the fracture i.
It is constant regardless of ij and am, and can detect broken piping.

In addition, if the rupture pipe is known, the outflow volume ΔV and outflow time ΔTA of the coolant are known, so in the case of a small rupture where the reactor pressure p is constant, A, = ρtΔV a / G
a (P) −・= − According to (2), when the reactor pressure p changes greatly such as in the case of a major rupture, AI=ρtΔv h /G c ・・・−・−(a
), the fracture area As can be detected.

[Embodiments of the invention]

An embodiment of the present invention will be described below with reference to FIG. 149 and FIG. 10.

A measurement signal from a water level gauge 12 that measures the water level in the reactor pressure vessel 2 is input to a function generator 22 and a multiplier 23 through a filter 20 to remove fluctuating components. The function generator 22 generates a signal proportional to the flow cross-sectional area of the downcomer, which corresponds to the human water level. The multiplier 23 multiplies the water level signal and the channel cross-sectional area signal, and outputs a signal proportional to the amount of volume reduction of the coolant per unit time. The same signal is branched, one of which is input to a subtracter 25 through a delay circuit 24, and the difference between it and the non-delayed signal is calculated by the subtracter 25. The signal generator 26 generates a signal proportional to the delay time of the delay circuit 24, which is input to the divider 27, which divides the signal from the subtracter 25 to generate a signal proportional to the volume reduction rate of the downcomer coolant. Output. The same signal is branched, one of which passes through a delay circuit 28 and is input to a comparator 29 together with the other.

Comparator 29 generates a signal when a difference occurs between the two signals.

When the water level gauge 12 issues a scram signal, this signal is input as a start signal to the timer 21, and the timer 21 is activated. The timer 21 inputs the signal from the comparator 29 as a stop signal, measures the time ΔTA until the volume reduction rate of the downcomer changes, and generates a signal proportional to Δl11A at the same time as the stop signal is input.

The signal from the multiplier 23, which is proportional to the amount of volume reduction of the downcomer coolant at each time, is constantly integrated by an integrator 30, and is input to the gate element 31 as an integral value.

The gate element 31 opens in response to an output signal from the comparator 29 indicating the time when the volume reduction rate changes, and outputs the volume reduction amount ΔV of the coolant in the downcomer.

FIG. 1θ shows the latter half of this embodiment. A and B in Figure 10
, C, D, and E indicate that they are connected to the same symbols in Figure 89, respectively. The signal from the water level gauge 12 from which fluctuating components have been removed by the filter 20 is sent to a comparator 33 together with the output of a signal generator 32 that generates a signal proportional to an arbitrarily provided reference water level.
is input.

The comparator 33 issues a signal at the time when the water level in the reactor pressure vessel 2 drops below the reference water level.

The timer 34 measures and outputs ΔT3, using the signal from the comparator 29 that outputs a signal when the volume reduction rate changes as a start signal, and the signal from the comparator 33 that outputs a signal when the reference water level drops as a stop signal. In the divider 35, ΔTh
and ΔTB, the comparator 40 calculates ΔT1/ΔTm.
A signal proportional to is input. The signal generators 36 to 39 generate ΔTA/ΔTs specific to each pipe connected to the reactor pressure vessel.
is output to the comparator 40. In the comparator 40, the comparator 35
The signals from the signal generators 36 to 39 are compared, and an identification signal of the matched pipe, that is, the broken pipe is outputted to the terminal F.

The measurement signal from the pressure gauge 13 of the reactor pressure vessel is input to the function generator 41, and the critical mass velocity G6. and the critical mass velocity Get at the time of liquid phase outflow are output from the respective terminals.

Get and Get are integrated by integrators 42 and 43, respectively, and output the time-integrated average value Gc of the critical mass velocity since the start of the scram up to each time. In other words. The selector 44 selects gas phase or liquid phase outflow based on the broken pipe signal from the comparator 40, and inputs Gav or G, t to the multiplier 45. The multiplier 45 multiplies ΔT^ from the timer 21 by the unit, and inputs the result to the divider 46. In the divider 46, the signal ΔV from the gate element 31
Divide A (outflow coolant volume) by Gc·ΔTA to obtain the fracture area Am and output it to terminal G.

As described above, according to this embodiment, a broken pipe and a broken area can be detected with high sensitivity based on the water level gauge signal of the reactor pressure vessel at the time of a loss of coolant accident, regardless of the size of the broken area.

FIG. 11 shows a second embodiment of the invention. Measurement signals (analog quantities) from the water level gauge 12 and pressure gauge 13 of the reactor pressure vessel 2 are converted into digital quantities by the AD converter 47 and input into the computer 48 . Computer 48 is the ninth
The program executes the processing shown in the block diagram of FIG.

A third embodiment of the present invention will be described with reference to FIGS. 12 and 13. FIG. 12 showing the first half is almost the same as the first embodiment. In this embodiment, as shown in Fig. 13, a signal is generated in which the remaining coolant volume when the volume reduction rate changes is proportional to the remaining coolant volume when each pipe is exposed (takes a value specific to each pipe). A comparator 40 compares the signals from the signal generators 56 to 59 to detect a broken pipe. Fracture area detection is the same as in the first embodiment.

〔Effect of the invention〕

According to the present invention, in the event of a loss of coolant accident in a nuclear reactor, it is possible to reliably detect the fractured piping and fracture area, especially in the case of a small fracture, based on the signal from the reactor pressure vessel water level gauge. Operating conditions can be optimized depending on the type of fracture, and even in the event of a small fracture in the liquid phase pipe, the reactor can be safely brought to a cold shutdown state without increasing the fuel rod temperature.

[Brief explanation of drawings]

Figure 1 is a piping system diagram showing the boiling water reactor emergency core cooling system, Figure 2 is a block diagram showing its operating conditions, and Figure 3 is the connection height of each piping to the reactor pressure vessel. Pattern diagram,
Figure 4 shows changes in reactor pressure at the time of small breaks in each pipe, Figure 5 shows changes in water level in the downcomer at the time of small breaks in each pipe, and Figure 6 shows a simulated change in maximum debt pipe temperature. Figure shown, M7
Figure 8 shows the change in the remaining coolant volume, Figure 8 shows the relationship between the change time of the volume reduction rate and the remaining coolant volume, and Figure 9 shows the relationship between the change time of the volume reduction rate and the remaining coolant volume.
and FIG. 10 are block diagrams showing the first embodiment of the present invention,
FIG. 11 is a block diagram showing a second embodiment of the present invention;
2 and 13 are block diagrams showing a third embodiment of the present invention. 1... Reactor core, 2... Reactor pressure vessel, 3...
・Coolant piping, 4... Containment vessel, 5... Turbine,
6...Main steam isolation valve, 7...High pressure furnace! 6 spray,
8...Automatic decompression system, 9...Suppression pool,
10...Low pressure core spray, 11...Low pressure core water injection system, 12...Reactor water level gauge, 13...Reactor pressure gauge, 20...Filter, 21...Timer, 22...・
Function generator, 23... Multiplier, 24... Delay circuit,
25... Subtractor, 26... Signal generator, 27...
Divider, 28... Delay circuit, 29... Comparator, 30
... Integrator, 31... Gate element, 32... Signal generator, 33... Comparator, 34... Timer, 35...
...Divider, 36-39...Signal generator, 40...
Comparator, 41... Function generator, 42.43... Integrator, 44... Selector, 45... Multiplier, 46...
Divider, 47... AD converter, 48... Computer, 49... Display. Agent Patent Attorney Tatsuyuki Unuma rl '5 # To i 7th session time

Claims (1)

[Claims] 1. In a nuclear reactor in which primary system pipes such as main steam pipes are connected at different heights, when any of the secondary system pipes breaks, the volume of residual water in the flash reactor pressure vessel or A method for detecting a pipe break in a nuclear reactor, characterized by detecting a pipe break based on a change in mass reduction rate. 2. A nuclear reactor according to claim 1, characterized in that a broken pipe is detected based on the volume or mass of residual water at a time when the volume or mass reduction rate of water remaining in the pressure vessel changes. Piping break detection method. & In claim 2, a broken pipe is detected using the relationship between the known water level and the residual water volume or mass for the reactor pressure vessel, and a signal from a water level system installed in the reactor pressure vessel. A method for detecting a rupture in a nuclear reactor pipe, the method comprising: detecting a rupture in a nuclear reactor pipe; 4. In claim 1, the time from the time when the water level drops to an arbitrarily set level after the rupture occurs until the volume or mass reduction rate changes, and the time set arbitrarily from the time when the volume or mass reduction rate changes. 1. A method for detecting a pipe break in a nuclear reactor, characterized by detecting a pipe break based on the ratio of the time it takes for the water level to drop to a certain level. 5. In claim 1, the time from the closing of the main steam isolation valve until the volume or mass reduction rate changes, and the time from the time when the volume or mass reduction rate changes until the water level drops to an arbitrarily set level. A method for detecting a pipe break in a nuclear reactor, characterized by detecting a pipe break based on a ratio to time. 6. In claim 1, the outflow mass velocity from the fracture port is determined based on the time until the volume or mass reduction rate changes and the change in reactor pressure up to the time of this change,
A nuclear reactor pipe fracture detection method characterized by also detecting a fracture area. 7. In a reactor pipe rupture detection device that is equipped with a pressure gauge and a water level gauge and connects primary system piping such as main steam piping at different heights, a rupture occurs when any of the secondary system piping ruptures. A broken pipe is detected from the ratio of the time from the time the water drops to an arbitrarily set water level until the volume or mass reduction rate changes, and the time from the time the reduction rate changes until the water drops to the arbitrarily set water level. Nuclear reactor piping, characterized in that the outflow mass velocity from the rupture port is calculated based on the circuit, the time until the rate of decrease changes, and the change in reactor pressure up to the time of this change, and the rupture area is calculated. Break detection device.