JPH04104090A - Pressure releasing apparatus for nuclear reactor - Google Patents

Abstract

PURPOSE:To enhance safety by providing a control means forcibly opening an opening means when a cooling material abnormally flows out of a reactor container. CONSTITUTION:When a dry well high pressure signal 10 and a low pressure ECCS pump operating signal 15 generated by the outflow of a cooling material are automatically detected, an ADS is operated and an escape safety valve 1 is forcibly opened and the steam in a reactor pressure vessel 2 is discharged into the cooling water in a pressure suppression chamber 7 through an exhaust pipe 5 and the pressure in the reactor pressure vessel 2 is rapidly lowered. As a result, the cooling water in the pressure suppression chamber 7 is injected in the reaction pressure vessel 2 by the pump 14 of a low pressure ECCS 11 and a core can be cooled within a short time.

Description

[Detailed Description of the Invention] [Industrial Application Field] The present invention relates to an emergency depressurization device for a nuclear reactor in a 1M nuclear power plant, and in particular, a system for forcibly closing a safety valve when an abnormality occurs in the reactor. This invention relates to a pressure release device that opens the reactor to reduce the pressure in the reactor.

[Prior art] In nuclear power plants, even if an accident occurs, the core of the reactor must be cooled. ECCS
)have. With the current settings, loss of coolant accidents (L
Assuming that the reactor water level is low and the dry well pressure is high, the ADS (automatic depressurization function) is activated.
By releasing the pressure inside the reactor and bringing the reactor to a low pressure state, the reactor core can be cooled by the low pressure core cooling system in addition to the high pressure core cooling system.

On the other hand, if the reactor is maintained at high pressure in a transient event where the dry well pressure high signal does not occur and the reactor water level drops, in addition to the high pressure emergency core cooling system, the low pressure emergency A special core cooling system cannot be used.

For this reason, as described in Japanese Patent Application Laid-Open No. 57-113395, a startup logic for automatically reducing the pressure of the nuclear reactor shown in FIG. 4 was devised. This is an ADS automatic startup logic based on a combination of a reactor water level low signal and a high-pressure emergency core cooling system inactivation signal, and a reactor water level low signal and a short time (5 minutes).
This consists of ADS automatic activation logic based on the continuation signal.

[Problems to be Solved by the Invention] The above-mentioned conventional technology takes into account that the progression of an event, that is, an event in which a low reactor water level signal occurs without a high dry well pressure signal, progresses relatively slowly. However, there was a problem in that it was difficult to put it into practical use.

The present invention clarifies the events that require automatic ADS activation, accurately controls the signals corresponding to those events, and improves the ADS automatic activation logic to improve the reliability of the atomic couplant during transients and accidents. The aim is to improve

Means for Solving the Problem] The configuration of the pressure relief device for a nuclear reactor according to the present invention for solving the above problem is to reduce the pressure of the reactor by guiding water vapor in the reactor vessel to the condensing means via the relief means. The opening device is equipped with a control means that can forcibly open the opening means when coolant abnormally flows out from the reactor vessel and the high-pressure water injection system is determined to be inactive. .

[Function] In response to an event in which coolant abnormally flows out from the reactor pressure vessel, the reactor pressure relief device of the present invention releases the relief safety valve while there is sufficient water in the reactor pressure vessel. can be operated automatically. In particular, the safety relief valve can be automatically operated for all events of coolant flowing into the drywell and coolant flowing into the water within the suppression chamber.

Thereby, the reactor pressure vessel is depressurized so that the low pressure ECC 5 is activated with sufficient reliability and the reactor core is sufficiently cooled for the above event.

Conventionally, in cases such as LOCA, a measure was taken to restore the cooling water level in the reactor pressure vessel during a short period of continuous signal from the scram level, but with the present invention,
During the 30-minute continuous abnormality signal, take careful measures1.
This will give them room to operate.

[Example code] Hereinafter, one embodiment of the present invention will be described using FIG. 1.

FIG. 1 is a schematic system diagram of a pressure relief device for a nuclear reactor according to the present invention.

The configuration of FIG. 1 is as follows: 1 is a safety valve, 2 is a reactor pressure vessel, 3 is a dry well, 4 is a main steam pipe, 5 is an exhaust pipe, 7 is a pressure suppression chamber, and 8 is an atom Reactor water level low signal detector,
10 is a dry well pressure high signal determiner, 15 is a low pressure spray operation signal determiner, 14.19 is a pump, 31
is a signal judger.

The present invention relates to a start control mechanism for a safety relief valve 1, which includes a safety relief valve 1 attached to a main steam pipe 4 in a dry well 3, and an exhaust pipe 5 connecting a reactor pressure vessel 2 and a pressure suppression chamber 7. It is.

Its start-up control is based on conventional LOCA (loss of coolant accident) signals, namely, reactor water level low signal 8, dry well pressure high signal 10. In addition to the ADS activation signal consisting of the low-pressure ECC5 pump activation signal 15, the reactor water level low signal (in this case refers to the reactor water level that activates the high-pressure ECC5)
A circuit is added to automatically start the ADS when a signal 31 continues for 30 minutes or more.

The operation of this startup control will be explained below.

Reactor pressure vessel 2 as in loss of coolant accident (LOCA)
In the event that coolant flows out from the reactor, in addition to the operation of the high-pressure ECC5 pump, the reactor water level low signal 8 (in this case, refers to the reactor water level that starts the low-pressure EC: C5) is activated by the flowing coolant. Resulting dry well pressure high signal 10. When the low-pressure ECC5 pump operation signal 15 is automatically detected, the ADS is activated after a certain time delay, and the safety relief valve 1 is forcibly opened.
Steam in the reactor pressure vessel 2 is released through the exhaust pipe 5 into the cooling water in the pressure suppression chamber 7, and the pressure in the reactor pressure vessel 2 is quickly reduced, allowing the pump 14 of the low-pressure ECC 5l 1 to operate. As a result, the pump 14 of the low pressure ECC 311
By injecting the cooling water in the pressure suppression chamber 7 into the reactor pressure vessel 2, the reactor core can be cooled in a short time.

On the other hand, during reactor isolation, the reactor isolation cooling system (RCI C
) or high-pressure ECC5, etc., there is a possibility that the dry well pressure high signal 10 will not occur and the ADS will not start automatically, so the low-pressure ECC5 will be injected into the reactor pressure vessel 2. Water may not be injected using the pump 14. In response to such an event, if the reactor water level low signal (in this case refers to the reactor water level that activates the high-pressure water injection system) signal 31 has been occurring continuously for more than 30 minutes, the ADS will automatically When activated, the relief safety valve 1 is forcibly opened, the pressure inside the reactor pressure vessel 2 is quickly lowered, and water is injected into the reactor pressure vessel 2 by the low pressure ECC5 pump 14 to cool the reactor core.

That is, the water level in the reactor pressure vessel 2 is measured with a water level gauge,
When the detector 31 detects that the water level in the reactor pressure vessel 2 has remained below the ADS startup water level for more than 30 minutes, the ADS
A feature of the present invention is to automatically start the DS.

Here, the reason why the reactor water level low signal continues for 30 minutes will be explained using a 1100 MWe class BWR-5 plant (standard type plant) as an example.

FIG. 3 is a schematic diagram showing the structure of the reactor pressure vessel and the coolant water level.

FIG. 3 shows the structure and water level inside the reactor pressure vessel 2. Each water level generally triggers the activation of engineering safety equipment. In the case of BWR-5, high-pressure water injection systems such as the reactor isolation cooling system (RCIC) and high-pressure core spray system (HPC8) are activated by the reactor water level signal (level 3), and the reactor water level (level 1) signal is activated. Low pressure core spray system (LPC8),
Low-pressure ECC5 systems such as the low-pressure core water injection system (LPCI) are activated. Here, assuming an event in which the high-pressure water injection system during reactor isolation does not operate as described above, a scram (level 3
) Steam is released from the safety valve 1 into the cooling water of the pressure suppression chamber 7 due to the decay heat of the reactor core, so the water level in the reactor pressure vessel 2 gradually decreases.

Now, the amount of steam CM) that should flow out is calculated using a linear equation.

Here, M: Amount of steam flowing out per second (tons) Po: Reactor thermal output (MWT) P(t) Reactor thermal output after the scram (decay heat after scram) Hg: Steam enthalpy ( ca 41 /I [g]
t Time after Niscrum: 1m (seconds) f: Conversion factor (Kca Q /MvT/5ec) Insert the coefficient into the above formula and calculate the standard BWR-5 plant's 30
A rough evaluation of the amount of steam flowing out (M) per minute (1800 seconds) is as follows.

M = 3440 (MIIT) x 239 (Kca
It /NVT/5ee) Valve 62 tons In Figure 3, the amount of cooling water held from the start of scram when the reactor water level is low (level 3) to the upper end of the effective fuel rod (level) is approximately 80 tons or more. The cooling water level from the upper end to the lower end (level) is approximately 100 tons.
It's on.

Therefore, even if the amount of water equivalent to 30 minutes (approximately 62 tons) flows out of the reactor 3 and pressure vessel 2 from the low reactor water level (level 3) as described above, the reactor water level will be lower than the upper end of the effective fuel rod ( Since there is sufficient cooling water in the reactor core, cooling of the reactor core will be ensured sufficiently and safely.

As described above, even if the high-pressure water injection system does not operate effectively in the event that coolant abnormally flows out from the 1M slave reactor pressure vessel 2, the ADS can be automatically operated to reduce the pressure inside the reactor pressure vessel. Accordingly, it is possible to provide a start-up control mechanism for the safety relief valve 1 that can operate the low-pressure ECC 5 system and sufficiently cool the reactor core.

Next, another embodiment of the present invention will be described with reference to FIG.

FIG. 2 is a system diagram of another embodiment of the present invention. In FIG. 2, 16 is a high-pressure core spray system, 24 is
Reactor isolation cooling system, 2:2.26 is 1, signal judge, and other symbols are the same as in FIG. 1.

The difference between the system shown in FIG. 2 shown in this example and that shown in FIG. 1 is that
In the automatic start-up control mechanism of the ADS, simultaneous signals for inactivation of the high-pressure core spray system 16 and inactivation of the reactor isolation cooling system 24 are combined. ' That is, when the inoperation of the high-pressure core spray system 16' and the isolation cooling system 17 is detected at the same time, the iADs are activated after a certain time delay, and the safety relief valve 1 is forcibly opened, thereby reducing the 1M slave reactor pressure. The steam in the container 2 is introduced into the cooling water in the pressure suppression chamber 7 through the exhaust pipe 5 and condensed.

As a result, the pressure inside the reactor pressure vessel 2 quickly decreases,
Since water can be quickly injected from the low-pressure core spray system 11 into the reactor pressure vessel 2, the core is cooled in a short time and safety is ensured.

[Effect of the invention] According to the present invention, the following effects can be expected.

(1) Improving the reliability of the emergency core cooling system Assuming that the high-pressure water injection system does not operate effectively in the event of a loss of coolant accident in which coolant abnormally flows out of the reactor pressure vessel or during reactor isolation. By observing the automatic start-up of the reactor pressure reduction operation instead of manual operation, it is possible to reliably cool the reactor core.

As a result, the emergency core cooling system can be used in response to any event such as abnormal outflow of coolant from the reactor pressure vessel, and the reliability of the nuclear couplant can be sufficiently improved. (2) Improve safety margin It is possible to prevent erroneous operations, and the safety margin is improved.(There is time to consider core cooling measures.) On the other hand, the present invention provides a new high-pressure water injection system (for example, a high-pressure water injection system using an accumulator, a high-pressure water injection system using a high-pressure
It has the same effect as adding CC8''), and a sufficient safety margin can be achieved with simple and inexpensive equipment.

The economic effect is large.

[Brief explanation of the drawing]

Fig. 1 is a schematic system diagram of the first embodiment of the present invention, Fig. 2 is a schematic system diagram of another embodiment of the present invention, and Fig. 3 is the structure and cooling water level of the reactor pressure vessel. The schematic diagram, Figure 4', is
- Conventional example/me system schematic diagram. -Explanation of the code'
〉- 1...Safety valve, 2...Reactor pressure vessel, 3...Dry well, 4...Main steam pipe 55pa...Exhaust pipe, 7...
Pressure suppression chamber, 11...Low pressure ECC8, 16...High pressure EC
C5, 14', 1'9... Pump, 17... Isolation cooling system, 8°10.15, 22, 26, 31... Signal determiner.

Claims (1)

[Scope of Claims] 1. In a pressure relief device for a nuclear reactor that guides water vapor in a reactor vessel to a condensing means through an opening means, when coolant abnormally flows out from the reactor vessel and the high-pressure water injection system is inoperable. A pressure relief device for a nuclear reactor according to claim 1, characterized in that the pressure relief device for a nuclear reactor is equipped with a control means capable of forcibly opening the opening means when it is determined that the condition is as follows. In,
means for detecting the operating state of each of a reactor isolation cooling system and a high-pressure core spray system; and inputting a signal from the detection means to bring the isolation cooling system and high-pressure core spray system into an inoperable state. A nuclear reactor characterized by comprising a control means capable of forcibly opening the opening means while the reactor water level continues to be below the set water level for starting the high-pressure water injection system for 30 minutes or more when it is determined that the opening means is present. pressure relief device.