CN215988119U

CN215988119U - Steam generator cooling system used after emergency shutdown of high-temperature gas cooled reactor

Info

Publication number: CN215988119U
Application number: CN202121688835.2U
Authority: CN
Inventors: 沈洵; 骞宏伟; 李彦峰; 王凯
Original assignee: State Nuclear Electric Power Planning Design and Research Institute Co Ltd
Current assignee: State Nuclear Electric Power Planning Design and Research Institute Co Ltd
Priority date: 2021-07-23
Filing date: 2021-07-23
Publication date: 2022-03-08
Anticipated expiration: 2031-07-23

Abstract

The utility model provides a steam generator cooling system used after a high-temperature gas cooled reactor is in emergency shutdown, which comprises a cooling steam inlet pipeline, a steam generator and a cooling steam outlet atmosphere discharge pipeline, wherein the cooling steam inlet pipeline is communicated with an inlet of the steam generator, an outlet of the steam generator is communicated with the cooling steam outlet atmosphere discharge pipeline, and the cooling steam outlet atmosphere discharge pipeline is provided with a flow regulating valve. The utility model realizes the accurate control of the cooling steam parameters after the emergency shutdown, so that the steam generator can be cooled quickly and safely.

Description

Steam generator cooling system used after emergency shutdown of high-temperature gas cooled reactor

Technical Field

The utility model belongs to the technical field of nuclear power, and particularly relates to a cooling system for a steam generator after emergency shutdown of a high-temperature gas cooled reactor.

Background

When a high temperature gas cooled reactor nuclear power plant (HTR-PM) normally operates, a main helium fan drives helium coolant to pass through a reactor core to take heat generated by nuclear fuel fission of the reactor core, the average temperature of helium at an inlet and an outlet of the reactor is respectively 250 ℃ and 750 ℃, high-temperature helium at 750 ℃ enters a steam generator through an inner pipe of a concentric pipe of a hot gas guide pipe, the heat is transferred to two loops of feed water through a heat transfer pipe of the steam generator, and superheated steam which heats the feed water at 205 ℃ to about 570 ℃ is supplied to a steam turbine to generate electricity. The high temperature superheated steam generated by the HTR-PM effectively increases the efficiency of the power generation, but its high temperature operation mode also presents some new challenges.

When an accident happens to the reactor and the emergency shutdown is triggered by the protection system, after the accident reason is found out and the corresponding fault is eliminated, the reactor should be provided with a restarting condition as soon as possible so as to improve the operability and the economy of the power station. When the high-temperature gas-cooled reactor is normally stopped, the first loop and the second loop can undergo a sufficient power reduction and cooling process, and the temperature of the reactor and related system equipment can be smoothly reduced to the temperature required for restarting. When the high-temperature gas cooled reactor is in emergency shutdown, the reactor can be rapidly shut down, and a fan baffle and an inlet and outlet isolation valve of an evaporator are rapidly closed. Because the reactor has large heat capacity, the first loop and the second loop are not fully cooled, and the reactor, the evaporator and other equipment are continuously in a high-temperature state and cannot be restarted quickly. If only relying on natural cooling, the cooling speed of the first loop and the second loop is very slow, and the temperature required by restarting needs to be reduced for several months, which is not acceptable in engineering. Meanwhile, after the reactor is stopped urgently, the main feed water loses a heating source, the temperature of the main feed water is far lower than that of the evaporator, so that the main feed water cannot be cooled by injecting the main feed water, once the main feed water is injected, the evaporator is subjected to large cold impact, the safety of evaporator equipment is influenced, and the method is not feasible.

Based on the design characteristics, equipment performance and operation characteristics of the high-temperature reactor, in order to shorten the cooling time of the first loop and the second loop of the reactor after the emergency shutdown as much as possible, ensure the safety of equipment such as an evaporator and the like, and further improve the operability of the HTR-PM set, a small-flow cooling technical scheme after the emergency shutdown needs to be formulated. After the reactor is triggered to be shut down emergently, an auxiliary steam system, a start-up shutdown system, an evaporator accident discharge system, a fuel loading and unloading system, a helium purification accident cooling dehumidification train and other systems are comprehensively utilized, a reactor core, an evaporator and related pipeline equipment of the reactor are cooled simultaneously through the coordination work of primary loop small-flow helium and secondary loop small-flow steam until the temperature of hot helium at the outlet of the reactor is reduced to about 200 ℃, and the unit has a restart condition.

The current small-flow cooling technology after emergency shutdown adopts the general scheme that: after the reactor triggers the emergency shutdown, the fan baffle is closed, the steam generator inlet and outlet isolation valves are closed in an interlocking manner, and the steam generator is isolated from other equipment and pipelines of the two loops. The operator eliminates accidents such as cold loss, pressure loss, pipe breakage of a steam generator and the like, confirms that the necessary conditions of the small-flow cooling operation after the emergency shutdown are met, and starts the small-flow cooling operation within 200 minutes of the shutdown. This low flow cooling operation is divided into 4 stages of steam generator blowdown, steam generator pre-cool and core cooling. The time when the reactor triggers the scram is defined as zero time, the time for starting all subsequent operations is counted from the zero time, and a schematic diagram of a general scheme of the small-flow cooling after the scram is given in figure 1. The duration of the two phases of steam generator pre-cooling and core cooling is the longest in the 4 phases of the low flow cooling operation, which directly determines the cooling effect and cooling time of the reactor. The key to determining the cooling effect and cooling time is the need to continuously provide cooling steam at a suitable, steady pressure and flow rate.

The connection mode of a heat transfer pipe and a tube plate of the Steam Generator (SG) is expansion joint and seal welding (non-strength welding), the stability and the precision of the parameter control of the scram cooling steam are poor, and the safety and the reliability of the SG are greatly influenced. When the steam flow is too large, large local temperature gradient and thermal stress are easily caused, and the fatigue life and local deformation of the control equipment are not facilitated; when the steam flow is too small, the operation control difficulty can be improved, and the cooling efficiency is reduced. At present, a small-flow cooling steam gas source is provided by an auxiliary boiler room, and the current engineering system does not relate to how to accurately control parameters such as pressure, flow and the like of cooling steam so as to meet the cooling requirement. For example, patent document CN106782720A discloses a system and method for heating and cooling a start-stop reactor plant with auxiliary steam: the outlet of the reactor is communicated with the primary side inlet of the steam generator, the inlet of the reactor is communicated with the primary side outlet of the steam generator, the secondary side outlet of the steam generator is communicated with the inlet of the steam-water separator, the outlet of the steam-water separator is communicated with the inlet of the steam bypass, the outlet of the steam bypass is communicated with the inlet of the steam condenser and the inlet of the deaerator, the high-temperature steam outlet of the auxiliary electric boiler is communicated with the inlet of the steam-water separator, the low-temperature steam outlet of the auxiliary electric boiler is communicated with the secondary side inlet of the steam generator, the system and the method can realize the quick response of the start-stop reactor equipment in the normal start-stop process of the unit, can realize the quick cooling of the reactor under the accident condition of the unit, and have higher reliability and safety in the start-stop process of the high-temperature gas cooled reactor unit. Patent document CN106782720A also cools SG with a small flow rate of cooling steam, but the entire system is complicated, and the flow rate of cooling steam is greatly affected by other systems, and cannot be accurately adjusted.

For another example, patent document CN108278590A discloses a shutdown cooling system and method for a high temperature gas cooled reactor nuclear power plant, where the system includes an evaporator, an evaporator outlet valve group, a steam-water separator, an evaporator inlet valve group, a bypass valve group, a high pressure heater, an auxiliary steam header, a condenser, a feed pump, and a deaerator. The method comprises the following steps: 1) introducing steam in the auxiliary steam header into the steam side of the high-pressure heater and the deaerator to increase the water supply temperature at the inlet of the evaporator; 2) adjusting the water supply pressure at the inlet of the evaporator by using a water supply pump, controlling the water injection flow of the evaporator by using an evaporator inlet adjusting valve, further controlling the temperature reduction speed of the evaporator, and controlling the pressure of the evaporator by using an evaporator outlet adjusting valve; 3) discharging the working medium at the outlet of the evaporator into a steam-water separator; 4) controlling the pressure of the steam-water separator by using a bypass regulating valve and auxiliary steam entering the steam-water separator; 5) and recovering water in the steam-water separator to a deaerator, and establishing a two-loop water circulation to cool the evaporator. Patent document CN108278590A discloses heating feedwater with auxiliary steam and cooling the feedwater with high-temperature water, and the cooling media are different between the two. The temperature of the high-temperature water is still very different from the temperature of the evaporator, and once the high-temperature water is injected, the high-temperature water brings great cold impact to the evaporator, so that the safety of the evaporator equipment is influenced, and the high-temperature water is not feasible.

Therefore, the existing engineering system cannot meet the use requirement of accurate control of cooling steam parameters.

SUMMERY OF THE UTILITY MODEL

Aiming at the problems, the utility model provides a cooling system for a steam generator after a high-temperature gas cooled reactor is in emergency shutdown, which realizes the accurate control of the cooling steam parameters after the high-temperature gas cooled reactor is in emergency shutdown.

In order to achieve the purpose, the utility model adopts the following technical scheme:

a cooling system for a steam generator after a high-temperature gas cooled reactor is in emergency shutdown comprises a cooling steam inlet pipeline, a steam generator and a cooling steam outlet atmosphere exhaust pipeline, wherein the cooling steam inlet pipeline is communicated with an inlet of the steam generator, an outlet of the steam generator is communicated with the cooling steam outlet atmosphere exhaust pipeline, and a flow regulating valve is arranged on the cooling steam outlet atmosphere exhaust pipeline.

Preferably, the steam generator inlet is communicated with a main water supply pipeline, the main water supply pipeline is communicated with the cooling steam inlet pipeline, and the cooling steam inlet pipeline is provided with a cooling steam inlet isolation valve group.

Preferably, one end of the cooling steam inlet pipeline is communicated with the steam generator inlet, and the other end of the cooling steam inlet pipeline is communicated with the auxiliary electric boiler.

Preferably, the cooling steam outlet atmospheric discharge pipeline is further provided with a cooling steam atmospheric discharge front isolation valve and a silencer.

Preferably, the outlet of the steam generator is communicated with a main steam pipeline, and the main steam pipeline is communicated with the cooling steam outlet atmospheric air discharge pipeline.

The utility model has the beneficial effects that: steam that produces the auxiliary electric boiler is as cooling steam, and through cooling steam inlet pipe access main water supply pipe back, get into SG and cool off SG, through the cooling steam export atmospheric discharge pipeline that newly increases on SG export main steam pipeline, behind isolation valve, flow control valve, the muffler before the atmospheric discharge of cooling steam, discharge to the atmosphere, realize the accurate control to the post cooling steam parameter of promptly stopping piling.

Additional features and advantages of the utility model will be set forth in the description which follows, and in part will be obvious from the description, or may be learned by practice of the utility model. The objectives and other advantages of the utility model will be realized and attained by the structure particularly pointed out in the written description and claims hereof as well as the appended drawings.

Drawings

In order to more clearly illustrate the embodiments of the present invention or the technical solutions in the prior art, the drawings used in the description of the embodiments or the prior art will be briefly described below, and it is obvious that the drawings in the following description are some embodiments of the present invention, and those skilled in the art can also obtain other drawings according to the drawings without creative efforts.

FIG. 1 shows a prior art post scram low flow cooling overview;

FIG. 2 illustrates a diagram of a conventional post island scram low flow cooling system of the present invention;

fig. 3 shows a schematic diagram of the overall scheme of the post scram low flow cooling in example 2.

Detailed Description

In order to make the objects, technical solutions and advantages of the embodiments of the present invention clearer, the technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the drawings in the embodiments of the present invention, and it is obvious that the described embodiments are some, but not all, embodiments of the present invention. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.

Example 1:

a cooling system for a steam generator after emergency shutdown of a high-temperature gas cooled reactor comprises a cooling steam inlet pipeline, a steam generator and a cooling steam outlet atmosphere exhaust pipeline, wherein the cooling steam inlet pipeline is communicated with an inlet of the steam generator; the steam generator export with cooling steam outlet atmospheric discharge pipeline intercommunication, steam generator export intercommunication have main steam conduit, main steam conduit with cooling steam outlet atmospheric discharge pipeline intercommunication, cooling steam outlet atmospheric discharge pipeline is equipped with isolation valve, flow control valve and muffler before the cooling steam atmospheric discharge in proper order, cooling steam inlet pipeline one end with steam generator entry intercommunication, the other end and supplementary electric boiler intercommunication.

On the conventional island side, as shown in fig. 2, steam generated by the auxiliary electric boiler is used as cooling steam and is connected to a water supply valve group on a main water supply pipeline through a cooling steam inlet pipeline, then the cooling steam enters #1SG and #2SG respectively to be cooled, the cooled steam passes through a cooling steam outlet atmospheric air exhaust pipeline newly arranged on an SG outlet main steam pipeline, is exhausted to the atmosphere after passing through a cooling steam pre-atmospheric air exhaust isolation valve, a flow regulating valve and a silencer, and the cooling steam outlet atmospheric air exhaust pipeline is provided with the cooling steam pre-atmospheric air exhaust isolation valve and the flow regulating valve (namely, the cooling steam regulating valve, see fig. 2) to accurately control the flow of the cooling steam after emergency shutdown, so that the cooling speed of the reactor is controlled. Specifically, the method comprises the following steps: after the reactor triggers the scram, the operator confirms that the necessary conditions for carrying out the post-scram small-flow cooling operation are met, the small-flow cooling operation is started within 200 minutes of the scram, and the small-flow cooling operation is divided into 4 stages (as shown in FIG. 1): in the stage of water discharging of the steam generator, steam discharging of the steam generator, precooling and core cooling of the steam generator, after the emergency shutdown of the #1 reactor, the nuclear island confirms that the necessary conditions for small-flow operation after the emergency shutdown are met, the conventional island confirms that auxiliary steam of the auxiliary electric boiler is available and finishes heating of a cooling steam pipeline during the emergency shutdown, the hot standby preheating of a steam-water separator of the #1 startup and shutdown system is stopped, after the third stage (precooling SG) starts, namely the emergency shutdown of the reactor starts to precool the steam generator, the front isolation valve of cooling steam exhaust atmosphere is opened, the main steam isolation valve on the main steam pipeline is closed, then the air exhaust pipeline of a cooling steam outlet newly added on the main steam pipeline of the #1SG outlet is opened, and the air exhaust pipeline passes through the front isolation valve of cooling steam exhaust atmosphere and the flow regulating valve. Controlling the flow of the cooling steam to be 0.5-0.55t/h through a #1SG flow regulating valve, and maintaining the absolute pressure of an outlet of the #1SG to be more than or equal to 1.1MPa for 60 min; then, the reactor core enters a fourth stage (cooling reactor core), and the flow of the cooling steam is controlled to be in gradient arrangement along with the time change through a #1SG flow regulating valve, and the method specifically comprises the following steps: controlling the flow rate to be 0.9-1.1t/h within 60-72min, and keeping for 12 min; controlling the flow rate to be 1.8-2.2t/h within 72-84 min; controlling the flow rate to be 2.7-3.3t/h within 84-96 min; controlling the flow rate to be 3.6-4.4t/h within 96-108 min; controlling the flow rate to be 4.5-5.5t/h within 120min 108-; controlling the flow rate to be 5.7-6t/h after 120min until the cooling working condition is finished, and installing a silencer on an air exhaust pipeline at a cooling steam outlet, wherein the noise behind the silencer is less than or equal to 85 decibels; wherein, the pressure at the outlet of the steam generator is controlled to be more than or equal to 1.1MPa, and the fluctuation range is controlled to be less than 0.5 percent; controlling the absolute pressure of the junction of the nuclear island and the conventional island of the main water supply pipeline of the steam generator to be more than or equal to 1.2MPa, the temperature to be 188-. Steam parameters of the pre-cooling steam generator on the nuclear island side and the cooling core stage are shown in table 1, steam flow requirements from the pre-cooling steam generator are shown in table 2, and steam pressure control requirements in the small-flow cooling process are shown in table 3;

TABLE 1 precooling steam generator and core cooling phase steam parameters

TABLE 2 steam flow requirements from Pre-chilled steam Generator

TABLE 3 steam pressure control requirements during low flow cooling

When the #2 reactor is in emergency shutdown, the cooling system operation flow is consistent with the #1 reactor. The utility model does not consider the working condition that two stacks are cooled simultaneously in emergency shutdown.

Example 2:

the steam generator cooling system used after the scram of the high temperature gas-cooled reactor in the embodiment 1 is applied to a start-stop reactor system in a nuclear power plant demonstration project of the high temperature gas-cooled reactor in a gulf of stone island in Shandong, Huaneng, after the reactor triggers the scram, an operator confirms that the necessary conditions for carrying out the small-flow cooling operation after the scram are met, and the small-flow cooling operation is started within 200 minutes of the scram. The small flow cooling operation is divided into 4 stages as shown in fig. 3: the method comprises the steps of draining water of a steam generator (through an SG accident discharge system), discharging steam of the steam generator (through a main water supply pipeline low-flow discharge valve), pre-cooling the steam generator (by adopting a start-stop reactor system, the low-flow steam parameters are that the temperature is 188 ℃, the pressure is 1.1MPa, and the flow is 0.5t/h), and cooling a reactor core stage (the flow of helium in a primary loop is 1kg/s, the low-flow steam parameters are that the temperature is 188 ℃, the pressure is 1.1MPa, and the flow is linearly increased to 6t/h from 0.5t/h within 1h and then is maintained).

Example 3:

the difference from the steam generator cooling system for the high temperature gas cooled reactor after the scram in example 1 is that after the third stage (pre-cooled SG) is started, the cooling steam inlet isolation valve set is opened, and cooling steam flows into SG (existing system). The flow of the cooling steam is controlled to be 0.5t/h through a flow regulating valve (namely, a cooling steam regulating valve, see figure 2) on the atmosphere discharging pipeline of the #1SG cooling steam outlet, and the pressure of the #1SG outlet is maintained to be more than or equal to 1.1 MPa. After the fourth stage (cooling the core) is started, the flow control valve on the air exhaust pipeline is discharged through a #1SG cooling steam outlet, and the flow of the cooling steam is controlled to meet the flow requirement in the table 2. And (3) maintaining the SG outlet pressure to be more than or equal to 1.1MPa until the whole emergency shutdown cooling working condition is finished, and closing a flow regulating valve on an atmospheric discharge pipeline of the #1SG cooling steam outlet. The operation of cooling #2SG is the same as #1 SG.

The utility model is applied to the stage of precooling and cooling the reactor core of the steam generator, accurately controls the parameters of the cooling steam after emergency shutdown and ensures the smooth implementation of the small-flow cooling technical scheme after emergency shutdown.

The utility model can be applied to emergency shutdown systems of other high-temperature gas cooled reactor nuclear power plant demonstration projects. For the existing high-temperature gas cooled reactor nuclear power station, if the aim of rapidly cooling after emergency shutdown is realized, the technical scheme of the utility model has the advantages of minimum influence on the existing system and minimum reconstruction cost. There are no other alternatives at all. For high-temperature gas cooled reactor nuclear power projects which are not built yet, the requirement of rapid cooling after emergency shutdown can be taken into consideration in the early design stage, and at the moment, the technical scheme that cooling steam is discharged to a two-loop drainage radioactivity monitoring pool or other discharge end points with stable pressure environments can be selected.

In summary, in the technical scheme of the utility model, the temperature, pressure and flow control of the cooling steam are key technologies, the discharge end point of the cooling steam is atmosphere, the backpressure of the atmospheric environment is stable, the setting mode is favorable for accurately controlling the parameters of the cooling steam in emergency shutdown, and the fluctuation of the parameters of the cooling steam passing through the SG is small; the system is simple to operate and control, cannot be influenced by a plurality of systems such as a circulating water system, a vacuumizing system, a condensed water system and the like, and meets the requirement of scram cooling on key operation time nodes; the transformation cost is low, and the operation economy is high. According to the technical scheme, the existing auxiliary steam system is fully utilized, and only one pipeline is additionally arranged on the outlet pipeline of the evaporator of the original system, so that the accurate control of the flow of cooling steam can be realized, the influence on the matching design of a solidified piling machine is small, and the equipment and the pipeline which are installed and constructed are not influenced; and for the solidified system process design, only the working condition operation control design of the start-stop system needs to be modified.

Although the present invention has been described in detail with reference to the foregoing embodiments, it will be understood by those of ordinary skill in the art that: the technical solutions described in the foregoing embodiments may still be modified, or some technical features may be equivalently replaced; and such modifications or substitutions do not depart from the spirit and scope of the corresponding technical solutions of the embodiments of the present invention.

Claims

1. The cooling system for the steam generator after the high-temperature gas cooled reactor is in emergency shutdown is characterized by comprising a cooling steam inlet pipeline, a steam generator and a cooling steam outlet atmosphere discharge pipeline, wherein the cooling steam inlet pipeline is communicated with an inlet of the steam generator, an outlet of the steam generator is communicated with the cooling steam outlet atmosphere discharge pipeline, and a flow regulating valve is arranged on the cooling steam outlet atmosphere discharge pipeline.

2. The cooling system for the steam generator after the high temperature gas cooled reactor emergency shutdown as claimed in claim 1, wherein the steam generator inlet is communicated with a main water supply pipeline, the main water supply pipeline is communicated with the cooling steam inlet pipeline, and a cooling steam inlet isolation valve group is arranged on the cooling steam inlet pipeline.

3. The steam generator cooling system for the high temperature gas cooled reactor after emergency shutdown as claimed in claim 1, wherein the cooling steam inlet pipe is communicated with the steam generator inlet at one end and is communicated with the auxiliary electric boiler at the other end.

4. The system as claimed in claim 1, wherein the cooling steam outlet exhaust pipe is further provided with a cooling steam pre-venting isolation valve and a silencer.

5. The system as claimed in claim 1, wherein the steam generator outlet is communicated with a main steam pipeline, and the main steam pipeline is communicated with the cooling steam outlet and the atmosphere discharging pipeline.