JPS58173492A - Method of conquering design standard accident and virtual accident of atomic power plant - 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 The present invention relates to a gas-cooled high-temperature nuclear reactor housed in a prestressed concrete pressure vessel, the spherical fuel elements of which are composed of fissile material particles embedded in a graphite matrix, and a prestressed concrete pressure vessel. a reactor switching building enclosing a reactor switching building, at least one isolation system and residual heat exhaust system, and a plurality of barriers for retention of fissile material, the coating of fissile material particles being a first barrier, a graphite The present invention relates to a method for overcoming design basis and hypothetical accidents in nuclear power plants, where the substrate constitutes the second barrier, the third barrier is realized by a prestressed concrete pressure vessel, and the reactor protection building constitutes the fourth barrier.

Nuclear power plants are designed according to current standards and regulations that require overcoming design basis accidents (10/a to greater than 10/a probability of occurrence). Accidents rarer than this are classified as hypothetical accidents.

Nuclear power plants are usually not designed with this hypothetical accident as a target, and damage scale or risk (certain occurrence x damage scale)
can only be confirmed.

As a primary measure for reducing or eliminating risks during the operation of nuclear power plants, it is conceivable to guarantee a reduction in radioactivity in the event of an accident, including a hypothetical accident. In high-temperature nuclear reactors of the nine structures mentioned at the outset, the aforementioned barriers are used to accomplish this task. However, the effectiveness of the barrier, apart from any external effects, depends substantially on operating and accident temperatures and operating and accident pressures. Along with this, chemical processes due to high temperatures and the presence of foreign substances (corrosion of graphite,
Decomposition of concrete) is also important in ensuring the containment of fission products.

Therefore, the safety of nuclear power plants depends first of all on overcoming temperature problems. In other words, is it sufficiently probable that the failure temperature of all barriers in 411[ is exceeded?
or must be excluded altogether. Pressure protection of equipment is also extremely important.

The purpose of the present invention is to reduce the risk of radioactivity leakage from the nuclear power plant and its damage to the environment in both design basis accidents and hypothetical accidents at the nine nuclear power plants mentioned at the beginning. The objective is to provide a method that can be virtually eliminated at low cost.

According to the invention, this object is achieved by a plurality of the following steps, which follow one after the other.

a) The first step is through automatically triggered protective measures;
Also, emergency measures carried out manually when this protective measure fails or as an adjunct to this protective measure
This is a stage in which the functions of the first and second barriers are guaranteed by utilizing the negative temperature coefficient unique to high-temperature nuclear reactors to prevent an unreasonable temperature rise.

b) The second step is the failure of @1 and the second barrier, on the one hand by measures for temperature limitation of the prestressed concrete pressure vessel and its seals, and on the other hand by measures for pressure limitation; stage to ensure the functionality of the barrier, in which automatically activated protective measures are first used and, if this protective measure fails or as an aid, emergency measures are carried out manually. Ru.

C) The third stage consists of the prevention of the primary gas leakage from the prestressed concrete pressure vessel by, on the one hand, the containment of the reactor protection building and, on the other hand, by the pressure relief process via the fission product retention system. This is a step to ensure the functionality of the fourth barrier during the event, in which the automatically triggered protective measures are first used, and if this protective measure fails or as an adjunct to this protective measure, emergency measures are activated manually. pressure relief can be provided through passive protection devices.

On the one hand, a small number of auxiliary devices and measures according to the system and methods are integrated into a complete safety design, and on the other hand, this safety design is adapted to specific site conditions (for example in urban areas) or to changing safety requirements (due to scientific and technological developments). It has enough flexibility so that it can be adapted at any time.

In the proposed method, the unique properties of high temperature nuclear reactors are optimally utilized. individual protective measures that are automatically triggered;
The availability of sufficient time for the staggered development of manual protective measures (repair and emergency measures) ensures reliable containment of the fission products even in the event of a hypothetical accident.

As mentioned above, the effect of the barrier to retain fission products is
In particular, it depends on the temperature inside the high-temperature reactor. Temperature increases typically result from abnormalities in the relationship between heat production and heat removal. Failure to maintain the fissile material particle and fuel element temperatures within the reactor below failure limits can result in failure of these two fission product barriers and radioactivity leakage into the next circuit. do.

The guarantee of radioactive confinement in the primary circuit is determined by the seals of the vessel penetrations and the temperature of the pressure vessel itself, as well as the pressure created in the secondary circuit. In case of a major failure of the next circuit cooling system, which cannot be removed by repair or emergency measures,
An increase in cold gas temperature and pressure in the primary circuit occurs.

- A pressure increase in the next circuit can lead to damage due to overpressure in the pressure vessel if protective measures are not activated. The above-mentioned protective measure is a planned relaxation of the pressure in the pressure vessel. It also prevents unexpected failure of container seals that are exposed to superheated temperatures as a result of an increase in cold gas temperature.

Relief of pressure results in more gas escaping into the reactor protection building. The reactor protection building, by its design and function, guarantees radiation furrow containment as a fourth barrier. The accident behavior of reactor support buildings is also determined by the temperature and pressure that occurs. If sufficient exhaust heat from the reactor cannot be activated, the reactor and metal fittings will become even hotter and more gas will escape from the pressure vessel. In this case, the increase in radioactivity associated with the high core temperature must be taken into account. Manually initiated auxiliary emergency measures maintain containment of the reactor support building.

A situation arises in nuclear reactor saline structures that renders radiation ridge containment ineffective only when radiation reaches the interior of the reactor structure and damage to the building due to overpressure must be considered.

In that case, a distinction is made between pressure loads caused by the combustion of combustible materials in the reactor SS material and pressure loads caused by the accumulation of large amounts of non-condensable gases. The sources of flammable or non-condensable gases are the thermal decomposition of the reactor pressure vessel concrete and the corrosion of graphite that occurs when air ingress into the secondary circuit occurs simultaneously with water ingress.

In order to avoid overpressure failure of the reactor protection building itself in the above cases, the reactor protection is Since the pressure in the building is relieved, it is unlikely that radioactivity exceeding the permissible level will reach the environment in this case as well. In addition, passive protection devices that operate auxiliarily can be provided. This device is also connected to the nine fission product retention devices.

The effectiveness of the method according to the invention is based, on the one hand, on the use of a plurality of graded barriers, and on the other hand, on the preservation of the functionality of the individual barriers by means of various protective measures carried out in a staggered manner. be. In another embodiment of the invention, at least some of the automatically or manually initiated protective measures are provided by means of devices designed to increase the redundancy of different structures.

The first step of the method according to the invention, i.e. the implementation of the prevention of an undue temperature rise, involves, on the one hand, the shutdown of the high-temperature reactor and, on the other hand, the cooling of the reactor, activating the residual heat evacuation device for this purpose. .

The possibility of interrupting the operation of the core is ensured by a first isolation system comprising absorption rods inserted into the core and side reflectors. In the event of an accident, the @1 shutoff system is automatically activated, or if the automatic system malfunctions, an absorber rod is manually inserted. If it is not possible to insert an absorbent rod, a second shutoff system comprising a large number of small absorbent bulbs can be used by manual operation.

The function of small absorption spheres is sufficient to keep the reactor subcritically cold. The period of feeding the small absorbing spheres into the core or reacting the first shutoff system can extend to more than 10 hours due to the negative temperature coefficient of high temperature reactors as well as the high failure temperature of the coated fissile material particles.

The second Imperial Palace system is also in poor condition, and the first #! If repair of the broken line is not possible, a shutdown of the high-temperature reactor takes place via a negative temperature coefficient.

- The cooling properties of the secondary circuit are determined by two factors: the circulation of gas and the presence of heat absorbers in the primary circuit. There, the residual heat removal device is equipped with a heat exchanger and a circulation blower.

The circulation blower is automatically started or manually activated as required. If it cannot be started, several hours (5 to 10 hours) are used for repair processing and emergency procedures.

In the case of high-temperature nuclear reactors, where sufficient natural circulation flow is available due to the special arrangement and construction of the primary circuit components, this natural convection may last for an extended period of time during an unsuccessful attempt to start the circulation blower. Reliably exhaust heat from the reactor core. However, as a result of the core flow reversal, an increase in cold gas temperature occurs.

Exhaust of residual heat from the core is usually carried out via a main heat exchanger. When this heat exchanger stops, it is automatically switched to the auxiliary heat exchanger. If the auxiliary heat exchanger also malfunctions, manual emergency measures are still required for several hours.

In particular, the high heat capacity of the reactor core and reactor accessories is effectively used for this rounding.

A complete shutdown of the residual heat extraction system will result in a temperature rise - a cessation of residual heat extraction from the circuit will damage the first and second barriers, i.e. the coating of fissile material particles and the graphite matrix, over a long period of time (several days). The third barrier, the prestressed concrete pressure vessel, must be impeccable as this would result in a large amount of fissile material leaking from the next circuit.

The second stage of the method according to the invention therefore includes a pressure device. First, the concrete in the pressure vessel decomposes, producing a large amount of H.
2. Release of CO and other gases into the reactor protection building must be prevented by sufficient heat exhaustion.

In the case of high temperature nuclear reactors where type lined liner cooling systems are available, the above objectives are served by the liner cooling system. The liner cooling system limits the temperature of the core and primary circuit. This is because over a long period of time (approximately 10 days) all residual heat can be removed from this system. Manually performed auxiliary emergency procedures ensure the safety function of the liner cooling system closure.

Even if the liner cooling system is defect-free, pressure will still be generated inside the pressure vessel, so when overheating temperatures appear after a long accident period, pressure relief must be planned in order to avoid failure due to overpressure in the pressure vessel. must be carried out. -Limiting of the pressure in the next circuit is carried out by a safety valve. If the failure of the safety valve on reaching the response pressure cannot be ruled out (resulting in the leakage of the entire contents of the next circuit into the reactor protection building), emergency procedures may be carried out, e.g. by manually operated auxiliary shutoff. By means of a valve, the blockage of the secondary circuit can be restored; further measures for pressure limitation of the primary circuit can be initiated manually via the gas purification equipment located in the secondary circuit.

After the highly radioactive primary gas has leaked into the reactor protection building due to the high core temperature, it is the reactor protection building that remains as the last barrier to the environmental leakage of fission products. Reactor protection buildings, by their design and function, ensure the containment of radioactivity. In addition, manually operated shutoff flags can ensure the closure of the reactor protection building with a high degree of redundancy. In order to maintain a permissible pressure for the reactor protection building, a pinox valve connected to the filter section is operated. The fission products are retained in the filter section, thereby substantially preventing any undue radioactivity leakage into the environment. The pinox valve can be operated automatically or manually in a staggered manner.

If, as a result of the shutdown of the liner cooling system, the concrete of the pressure vessel decomposes and with it a large amount of gas leaks into the reactor protection building, the above-mentioned device is activated. If even this equipment becomes inoperable (hypothetical accident), flammable or non-condensable gases will accumulate within the reactor protection building, and the containment of radioactivity will become uncertain due to overpressure in the combustion chamber.

In order to overcome this accident (probability of occurrence 10/I),
Further pressure relief can be achieved in the event of an accident via a safety valve using a rupture disc.

In this case, the discharge channel is also connected to a radioactive storage device, for example a filter section or a water reservoir. Therefore, it can be said that no environmental damage will occur in this case either.

Claims (1)

[Claims] 1) A gas-cooled high-temperature nuclear reactor housed in an f-rested concrete pressure vessel, the spherical fuel elements of which are composed of fissile material particles embedded and coated in a graphite matrix; an enclosing reactor protection building, at least one interconnection system and a residual heat evacuation device;
a plurality of barriers for retention of fissile material, the coating of fissile material particles being the first barrier, the graphite matrix being the second barrier, and the third barrier being realized by a breast-stressed concrete pressure vessel. , a method for overcoming design basis accidents and hypothetical accidents in nuclear power plants in which the reactor defense building constitutes the fourth barrier involves several successive steps, namely: a) first;
The second stage is achieved by automatically activated protective measures, and when this protective measure is not activated, by manually implemented emergency measures as an adjunct to these protective measures, and by the negative temperatures characteristic of high-temperature reactors. b) the second step involves ensuring the functionality of the @1 and second barriers by preventing unreasonable temperature rises by utilizing the coefficients; In the event of failure of the second barrier, measures to limit the temperature of the breast-stressed concrete pressure vessel and its seals, on the one hand, and pressure limits, on the other hand, ensure the effectiveness of the third barrier. A protective measure that includes and is automatically triggered is first used, and when this protective measure fails, or as an adjunct to the protective measure, emergency measures are manually activated.
C) The third step is to remove the pressure from the breast-stressed concrete pressure vessel, on the one hand by sealing off the reactor shield and, on the other hand, by relieving the pressure via the fission product retention device. A protective measure which includes preserving the function of the fourth barrier during the primary gas flow and which is automatically activated is first used, and when this protective measure is not activated or A method characterized in that, as an auxiliary, emergency measures are carried out manually and pressure relief can be auxiliary carried out via passive protective devices. 2) Automatically and manually triggered protective measures. 3) For carrying out the first step 1. The method according to claim 1, characterized in that on the one hand the high-temperature nuclear reactor is shut down and on the other hand a residual heat evacuation device is activated. 4) For the shut-off of the high-temperature reactor, the first Absorber rods that form the isolation system are inserted into the reactor core and side reflectors, and the insertion is automatically initiated, or manually when the automatic device is malfunctioning.
When the first fast-acting system is malfunctioning, a manually operated second shut-off system, which includes the supply of small absorption bulbs, is used and the 117
3.) The method according to claim 3, characterized in that measures are taken to cause the cut-off system to react. 5) The method according to claim 3 or 4, characterized in that the high-temperature reactor is refreshed by a negative temperature coefficient when both the first and second shutdown systems are stopped. . 6) - In a high-temperature nuclear reactor where sufficient natural circulation flow can be utilized due to the special arrangement and structure of the next circuit components, if all circulation blowers reach +II at the end of forging, residual heat will be removed from the core by this natural circulation flow. A method according to claim 3 or 5, characterized in that the method comprises discharging. 7) To remove residual heat, activate the device for removing residual heat, consisting of a heat exchanger and a circulation blower, with automatic startup or, if this device stops, repair measures and emergency 4. A method according to claim 3, characterized in that it is carried out manually after the procedure has been carried out. - 8) for the performance of the second stage in high-temperature reactors where a cape-lining liner cooling system is available, this liner cooling system is used for the removal of heat from the primary circuit;
2. A method as claimed in claim 1, characterized in that the pressure in the primary circuit is limited via a safety valve. 9) A method according to claim 8, characterized in that the auxiliary measures for limiting the pressure in the secondary circuit are carried out manually via a gas cleaning installation installed in the secondary circuit. 10) By manipulating the shut-off flag for the implementation of the third stage, the nine points connected with the filter section are 2. A method according to claim 1, characterized in that the method comprises operating an analog valve. 11) The method according to item 1 or 10 of the scope of Patent d, characterized in that the auxiliary pressure relief of the reactor protection building is carried out via a safety valve equipped with a rupture disc and communicating with the filter section. .