KR20210098111A - Analysis Methodologies of Severe Accident Sensitivity Assesment for Station Black Out in CANDU - Google Patents

Abstract

The present invention relates to a method for severe accident sensitivity assessment on a power loss accident of a heavy-water reactor, to assess sensitivity due to a power loss accident with a high frequency of core damage. The present invention provides a method for severe accident sensitivity assessment on a power loss accident of a heavy-water reactor, to provide a method for assessing the effect of safety system operation and non-operation on the development of a serious accident caused by a power loss accident with a high frequency of core damage. The method for severe accident sensitivity assessment on a power loss accident of a heavy-water reactor includes the steps of: (a) setting initial condition for a safety system; (b) checking a pressure signal; (c) losing coolant due to increased pressure; (d) outputting the state of a moderator cooling system; and (e) outputting the state of damage progress.

Description

Analysis Methodologies of Severe Accident Sensitivity Assesment for Station Black Out in CANDU}

The present invention relates to a method for evaluating the impact of a serious accident of a heavy water reactor power loss accident for evaluating the effect of a power loss accident with a high frequency of heavy water core damage.

Conventionally, according to Korean Patent Application Laid-Open No. 2017-0016588, a radioactive information confirmation unit that checks whether radioactive material is leaked, a main area in which any one or more systems that can leak radioactive material exist, and a nuclear power plant shutdown It includes a zone classification part that classifies it as a hazardous zone where there is one or more systems that must be defended so that the initial event, which is all abnormal events that causes, can be resolved, and a major zone analysis part that analyzes through the faulty tree.

The heavy water reactor serious accident evaluation code is implemented by ISAAC (Integrated Severe Accident Analysis code for CANDU plants), which was developed by KAERI/FAI in 1995, and was modeled by reflecting the CANDU characteristics based on the severe accident analysis code (MAAP4). Although there is a right to use, there is a problem in that there are no safety system operation or non-operational impact evaluation results for the development of serious accidents caused by power loss accidents with a high frequency of damage to the core of the heavy water reactor.

This impact evaluation is an additional evaluation item for continuous operation of Wolsong Unit 1, which was evaluated to supplement Probabilistic Safety Assessment (PSA). It is necessary to research and develop a method for evaluating the impact of a heavy water power loss accident in order to evaluate the impact of the loss accident.

The present invention is intended to solve the above problems, and the method for evaluating the impact of a heavy water reactor power loss accident to provide a safety system operation and non-operation impact evaluation method for the development of a serious accident caused by a power loss accident with a high frequency of core damage The purpose is to provide

The present invention provides a method for evaluating the impact of a serious accident in a heavy water power loss accident, comprising the steps of: a) setting initial conditions for a safety system in the case of the heavy water power loss accident; b) when the pressure on the primary side of the steam generator increases due to the power loss accident in the heavy water reactor, generating a signal for generating a damage to the reactor building for each set time or generating a decrease signal for the cooling water on the secondary side of the steam generator due to interruption of water supply to the steam generator; c) exposing the core to loss of coolant through a liquid discharge valve and a degassing condenser due to an increase in the pressure on the primary side of the steam generator; d) outputting the moderator cooling system status as the moderator cooling system is not operating when the temperature of the moderator increases due to heat transfer from the horizontal fuel pipe at the primary side of the steam generator; And e) when the overpressure threshold of the primary system of the steam generator is reached, the horizontal fuel pipe is damaged, the calandria rupture disk is damaged, and the calandria rupture disk is damaged, and the damage progress state is output as the inner core damage progresses.

Preferably, the step of confirming the pressure signal in step b) is outputting a state signal of a breakage of the water supply pipe due to steam overpressure due to the breakage of the water supply pipe located inside the reactor building when the nuclear reactor building damage occurrence signal is generated at a first set time; include

Preferably, the step of confirming the pressure signal of step b) is a step of outputting a state signal for breaking the main steam pipe by steam overpressure by breaking the main steam pipe located inside or outside the nuclear reactor building when the nuclear reactor building damage occurrence signal is generated at the second set time includes ;

Preferably, in step a), when the initial condition of the safety system is set, the water spray system among the safety system is set to non-operation, and the reactor building damage pressure is set higher than the initial condition of the safety system, so that the reactor building pressure behavior in step b) is Calculating; includes.

Preferably, when the secondary side cooling water reduction signal in step b) is generated, in the safety system, when rapid cooling using a steam generator is performed and the spray tank cooling water is used as the steam generator coolant, the damage progressed in step e) is delayed for at least a certain period of time.

Preferably, when setting the initial condition of the safety system, when the terminal shielding cooling system is available, the step of outputting in an undamaged state at least the effect on the reactor building damage or calandria damage is included.

Preferably, when setting the initial condition of the safety system, assuming that the bellows breaks when the pressure tube is damaged and high-pressure safety injection and medium-pressure safety injection are used, the damage progress state in step e) is delayed for at least 3 to 5 hours; includes; .

Preferably, step a) includes setting the initial condition of the safety system to an initial condition in which the cooling system of the main system, including the steam generator water supply system and calandria, which are safety systems, does not operate.

According to the present invention, after a heavy water power loss accident, the safety system operation and non-operation in the event of a power loss accident are identified by understanding the effects of reactor building pressure, hydrogen concentration, fuel pipe melting, calandria pipe damage, etc. It can be reflected, and it has the effect of contributing to the approval and approval by submitting the evaluation when the Wolseong Unit 1 continuous operation permit is granted.

1 is an exemplary diagram illustrating a heavy water reactor type nuclear power plant for evaluating the impact of a serious accident of a heavy water reactor power loss accident according to an embodiment of the present invention.
2 is an exemplary diagram illustrating a loop of a coolant system for evaluation of the impact of a serious accident in a heavy water power loss accident according to an embodiment of the present invention.
Figure 3 is an exemplary view showing a serious accident impact evaluation system of the power loss accident in heavy water according to an embodiment of the present invention.
4 is a flowchart illustrating a method for evaluating a serious accident impact of a heavy water power loss accident according to an embodiment of the present invention.
5 is an exemplary view illustrating a pressure behavior of the reactor coolant system according to the progress of an accident in the reactor coolant system.
6 is an exemplary view showing a change in the amount of cooling water in the reactor coolant system.
7 is an exemplary view showing the secondary side pressure of the steam generator of Loop 1 and Loop 2, which are reactor coolant systems.
8 is an exemplary view showing the water level of the steam generator.
9 is an exemplary view showing the maximum fuel rod temperature in the core.
10 is an exemplary view showing the transfer of the melt from the fuel pipe to the calandria.
11 and 12 are exemplary views showing the amount of moderator in calandria and the amount of cooling water in the reactor compartment.
13 is an exemplary view illustrating a pressure change in a nuclear reactor building.
14 and 15 show the emission rates of the inert gas and CsI and CsOH outside the reactor building.
16 is an exemplary view showing the amount of hydrogen generated by the inside of calandria and MCCI (Molten Core Concrete Interaction).
17 is an exemplary view showing the amount of hydrogen removed by PAR (Passive Autocatalytic Recombiner).
18 is an exemplary view showing a hydrogen mole ratio in a boiler room of a nuclear reactor building.

Hereinafter, the present invention will be described in more detail with reference to the drawings. It should be noted that the same components in the drawings are denoted by the same reference numerals wherever possible. In addition, descriptions of well-known functions and configurations that may unnecessarily obscure the gist of the present invention will be omitted.

As shown in FIG. 1 , the heavy water reactor has 380 Pressure Tubes in a Large calandria vessel surrounds fuel channels 20 and an End Shield 30 ( 10) is installed horizontally like the Calandria Tube, and fueling machines attach to individual fuel channels (40) if necessary. could develop into an accident. As shown in FIG. 2 , the coolant system of a heavy water reactor type nuclear power plant consists of two loops, and it is possible to isolate in case of an accident, thereby minimizing an accident.

3 is a block diagram illustrating a system for evaluating the impact of a serious accident of a heavy water power loss accident according to an embodiment of the present invention. 3, the serious accident impact evaluation system 1 of the heavy water power loss accident includes a safety system setting unit 110, a pressure signal confirmation unit 120, a processing unit 130, a moderator cooling system state output unit ( 140 ), a damage progress state output unit 150 , and a control unit 160 .

The safety system setting unit 110 is configured to set the initial condition of the safety system in case of a power loss accident in heavy water. Here, the initial condition of the safety system is that all power-related systems are lost in a power plant blackout accident, so the cooling system of the main systems including the steam generator water supply system and calandria, that is, the emergency core cooling system, the moderator cooling system, the terminal shielding cooling system, and the local air It is set as an accident condition in which the core is eventually damaged because the cooler does not work. That is, it refers to the initial condition in which the cooling devices of the main systems including the safety system, the steam generator water supply system and the calandria, that is, the emergency core cooling system, the moderator cooling system, the end-shielding cooling system, the local air cooler, etc. do not operate. Under these initial conditions, the liquid relief valve (LRV) of the reactor coolant system is a fail-open type valve, so it does not close once opened. assumed. In the basic accident, recovery measures for the operator were not considered, however, it is assumed that the reactor shutdown is successful, the valve of the reactor building water supply system is operated by the battery, so it operates despite a power outage, and the passive hydrogen recombiner, which is a passive system, also operates. did.

The pressure signal confirmation unit 120 generates a signal for generating a signal for generating damage to the reactor building for each set time when the pressure on the primary side of the steam generator is increased due to a power loss accident with heavy water or a decrease signal for cooling water on the secondary side of the steam generator due to interruption of water supply to the steam generator. .

The processing unit 130 exposes the core to the loss of coolant through the liquid discharge valve and the degassing condenser due to an increase in the pressure on the primary side of the steam generator.

The moderator cooling system status output unit 140 is configured to output the moderator cooling system status as the moderator cooling system is not operated when the moderator temperature increases due to heat transfer from the horizontal fuel pipe at the primary side of the steam generator.

When the damage progress state output unit 150 reaches the overpressure threshold of the primary system of the steam generator, the horizontal fuel pipe is damaged and the calandria rupture disk is damaged, and the core damage proceeds due to the decrease in the thermal margin and water level of the moderator in the calandria. This is a configuration that outputs the damage progress status.

The control unit 160 is a component for controlling the above-mentioned components, and a method for evaluating the impact of a serious accident of a heavy water power loss accident with this control unit will be described as follows.

4 is a flowchart illustrating a method for evaluating a serious accident impact of a heavy water power loss accident according to an embodiment of the present invention.

First, the control unit sets the initial condition of the safety system in case of a power loss accident in heavy water (a). Next, when the pressure on the primary side of the steam generator increases due to power loss accident with heavy water, it generates a signal to generate a signal for damage to the reactor building for each set time or a decrease signal for the cooling water on the secondary side of the steam generator due to interruption of water supply to the steam generator. Step (b), Next, the core is exposed due to the loss of coolant through the liquid discharge valve and the degassing condenser due to the increase in the pressure on the primary side of the steam generator (c), and then the moderator temperature by heat transfer from the horizontal fuel pipe at the primary side of the steam generator is Step (d) of outputting the moderator cooling system status as the moderator cooling system is not working if it increases. Next, when the overpressure threshold of the primary system of the steam generator is reached, the horizontal fuel pipe is damaged, the calandria rupture disk is damaged, and the calandria As the inner core damage progresses, the step (e) of outputting the damage progress status is reached.

Here, in the pressure signal confirmation step of step b), when the nuclear reactor building damage occurrence signal is generated at the first set time, the control unit outputs a water supply pipe breakage state signal due to steam overpressure due to the breakage of the water supply pipe located inside the reactor building.

In addition, in the pressure signal confirmation step of step b), when generating the reactor building damage occurrence signal at the second set time, the control unit generates the main steam pipe breakage status signal due to steam overpressure caused by the breakage of the main steam pipe located inside or outside the reactor building. print out

In addition, in step a), when the initial condition of the safety system is set, the control unit sets the spray system in the safety system as non-operational and sets the reactor building damage pressure higher than the initial condition of the safety system to set the reactor building pressure in step b). Calculate the behavior

In addition, the control unit performs rapid cooling using a steam generator in the safety system at the time when the secondary-side coolant reduction signal in step b) is generated, and when the water spray tank coolant is used as the steam generator coolant, damage proceeds in step e) The condition appears to be delayed for at least about 3 days or more.

In addition, when the initial condition of the safety system is set, the control unit outputs at least the effect of damage to the reactor building or calandria in an undamaged state when the terminal shielding cooling system is available.

In addition, when the control unit sets the initial conditions for the safety system, assuming that the bellows is broken when the pressure tube is damaged, and when high-pressure safety injection and medium-pressure safety injection are used, the damage progress state in step e) is delayed for at least about 3 to 5 hours.

Hereinafter, a detailed description of a method for evaluating the impact of a heavy water power loss accident will be described with reference to the accompanying drawings.

A serious accident analysis code used in the present invention will be described as follows. The ISAAC code was used to simulate a serious accident at Wolsong Unit 1. The ISAAC code uses MAAO4/PWR developed for comprehensive serious accident simulation in light water reactors as a reference code, and is a figure 8 shape consisting of 380 horizontal fuel pipes and two independent closed circuits to simulate the characteristics of serious accidents unique to heavy water reactors. of the primary system (PHTS) of In addition, various safety systems were added for accident mitigation. Therefore, the ISAAC code includes temperature rise in fuel rods and horizontal fuel lines, deformation and reaction with moderators, oxidation reaction between zircaloy and water vapor, damage to horizontal fuel lines, core relocation, suspended debris bed formation and heat transfer, It is possible to predict the accident progression according to the characteristics of heavy water channels such as the behavior of fuel materials at the bottom of the calandria and damage to the calandria. In addition, when the bellows of the end joint are broken at the same time due to the breakage of the pressure pipe in a high-pressure environment, the accident process in which the bellows is broken in the case of a supply pipe breakage accident or a power plant blackout accident that proceeds due to a high-pressure accident, and the stopping cooling system in the stationary condition The ISAAC was improved to simulate an accident in which coolant is lost due to leakage in the cooling pump bearings. In the present invention, accident analysis was performed using ISAAC version 4.03.

In one embodiment of the present invention, the initial event was reviewed by applying the analysis methodology from the PL (Point Lepreau, Canada) nuclear power plant to the latest results of the Wolsong Unit 1 PSA (Probabilistic Safety Assessment), and the frequency of core damage at the Wolsong Unit 1 Terminal shield cooling system failure (LOS), which accounts for about 15% of the total, was added as the sixth initial accident. If the six initial accidents are listed, supply pipe stagnation rupture accident (SFB), power plant blackout accident (SBO), accident at standstill (SSA), steam generator customs rupture accident (SGTR), small coolant loss accident (SLOCA), and termination It is a loss of shield cooling system accident (LOS). Here, the selection of an end-shielding cooling system failure (LOS) accident scenario will be described as follows. End-shielding cooling system failure (LOS) accident is an initial accident that was not included in the existing PL, and was classified as P-6 in this analysis. In the case of Wolsong Unit 1, the failure rate of the terminal shielding cooling system accounts for about 15% of the total core damage frequency, and it has the fourth highest contribution after class 4 power loss accident, cooling water system malfunction, and coolant distribution pipe breakage.

The assumptions for accident analysis are as follows. Since the assumptions for each scenario in the six initial events are slightly different, the assumptions for each case are added along with the assumptions that can be commonly applied to most accidents.

- The reactor is at full power (2157.5 MW(th)) except for the shutdown accident.

- For reactor building leakage, only leakage due to leakage rate is considered.

- For small LOCA (Loss of Coolant Accident), the fracture area is assumed to be RIH 2.5% (5.325x10-3 m2), and the fracture height is assumed to be 10.696 m, which is the RIH floor height.

- Except for SBO (Station Blackout) accidents, grade 4 and 3 power is available.

- Moderator cooling system and main water supply system are not available in most cases.

- ECCS operated in three stages of high pressure, medium pressure and low pressure is not available in most accidents, but in some accidents high pressure and medium pressure injection are available, or low pressure injection where the heat exchanger fails is available in some cases.

- In the case of a LOCA (Loss of Coolant Accident) accident, the two closed circuits of the PHTS (Primary Heat Transport System) are isolated from each other through the closed circuit isolation valve when the pressure is lower than 5.516 MPa(a). The pressure is controlled at the liquid release valve set point until the pressure tube is broken.

- In the steam generator, the secondary side pressure is controlled at the set pressure of the main steam safety valve (MSSV), and when the steam generator rapid decompression (SGCC) is available, the MSSV is automatically opened 30 seconds after the LOCA signal.

- The PHTS pump is simulated to automatically stop after 2 minutes when the pressure is lower than 2.5 MPa(a) to prevent damage by air bubbles at low pressure.

- When the coolant in the primary system is lost, the horizontal fuel pipe is heated, but decay heat is removed by the external moderator surrounding the fuel pipe. However, in some high-pressure accidents, the pressure tube swells and deforms, and high-pressure coolant and water vapor are released from the fuel tube to the calandria, causing the rupture disk to burst and the moderator to be released into the boiler room of the nuclear reactor building.

- When the coolant inside the pressure tube is depleted and the moderator water level in the calandria is lowered and the fuel tube is exposed inside and out, the fuel pipe begins to deform, moves away from its original position, and relocates around a healthy horizontal fuel pipe in the lower space to solidify and SDB (Suspended Debris Bed) is formed. When the moderator water level is continuously lowered, the exposed fuel pipe increases and the fuel material transferred to the Suspended Debris Bed (SDB) increases. All materials are assumed to be transported to the bottom of the calandria.

- The fuel material at the bottom of the calandria depletes the moderator and heats the calandria wall, but the outer wall is cooled naturally through the shield coolant in the reactor compartment outside the calandria to maintain the soundness of the calandria. However, in this process, the shielding coolant is evaporated and , it increases the pressure of the reactor building and threatens the integrity of the reactor building.

- When the level of the shielding coolant decreases and the fuel material on the bottom of the calandria is exposed, the external wall cooling is no longer performed, and thus the calandria is expected to be damaged.

- When the calandria is damaged, the fuel material inside the calandria is relocated to the reactor compartment and reacts with the concrete on the floor. Initially, the shielding coolant is evaporated, followed by concrete erosion until the floor is punctured.

- The reactor building sprinkler system is a passive system and operates when the reactor building pressure rises above 14 kPa(g) and stops when the reactor building pressure falls below 7 kPa(g). The amount of cooling water that can be sprayed is 1,559 m3.

- While the sprinkler system is a short-term reactor building heat removal system, the Local Air Cooler (LAC) is responsible for the long-term heat removal function of the reactor building as long as power is supplied. There are 35 LACs in all, but 16 of them with large capacity are safe and can be operated manually in the main control room, and 12 out of 16 are simulated as success criteria.

- A total of 27 passive hydrogen couplers (PARs) (10 small, 15 medium, 2 large) are installed in Wolsong Unit 1 for hydrogen control in the nuclear reactor building, and the installation location is a small 10 in the upper dome. There are 5 medium-sized dogs, 5 medium-sized ones in the boiler room, 3 medium-sized ones in the moderator room, and 1 medium-sized and 1 large-sized one in each fuel change room. The number and method of operation can be adjusted by input, and in this analysis, when the hydrogen concentration reaches 2%, the operation is delayed for 30 minutes, and when the concentration is lower than 0.2%, it stops. In order for PAR to work, adequate oxygen must be supplied.

- The reactor building was simulated to fail at 426 kPa(a), and 358 kPa(a) was also used for sensitivity analysis.

- Liquid release valve (LRV) opening pressure: 10.24 MPa(g)

The ISAAC model added for the new accident analysis analyzes the scenarios in six initial events, and the moderator leakage due to the bellows breakage of the end joint pipe, which was not considered in the existing PSA analysis, and the cooling pump bearing of the static cooling system under the stationary condition. The following new models have been added to ISAAC to simulate accidents in which coolant is lost due to leakage in

① Bellows break model

If the pressure pipe is broken in an accident where the Primary Heat Transport System (PHTS) is maintained at high pressure, the bellows of the end joint pipe may be simultaneously damaged due to pressure propagation in the fuel pipe. This bellows breakage may accompany supply pipeline stagnation breakage (SFB) and power plant blackout (SBO).

In ISAAC, the following input variables were introduced to simulate the failure of the pressure tube and the leakage of the moderator from Calandria to the reactor building.

- AFAILBELLOW: Bellows fracture area. If this value is greater than 0.0, the moderator leakage from Calandria to the reactor building is calculated.

- ZFAILBELLO: Breaking height from the bottom of the calandria

- JNFAILBELLOW: The number of the reactor building compartment where the moderator leaks through the bellows.

- ZNFAILBELLOW: Breaking height at the bottom of the reactor building compartment (JNFAILBELLOW)

- ISTAGRUP(ICH,IL): Control variable of fracture area of horizontal fuel pipe pressure pipe after SFB

If the ISTAGRUP variable value is 1, the pressure pipe breaking area is limited to the maximum flow path area of the inlet/outlet supply pipe. If it is any other value, the pressure pipe breaking area is twice the pressure pipe flow path and the sum of the inlet/outlet supply pipe flow path. Calculated by multiplying by the actual number of fuel lines.

②Static state initial model

Since the PHTS (Primary Heat Transport System) condition in the stationary state is different from the full power, it is necessary to introduce the PHTS pressure, water level, and the temperature in the PHTS, steam generator and pressurizer, and variables indicating that the PHTS coolant is separated from the steam. . Based on these input parameters, ISAAC defines the PHTS initial conditions. Under these conditions, the PHTS water level is located at the inlet and outlet capillary height, and decay heat is removed by the static cooling system. Similar initial conditions are defined for steam generators and pressurizers. The parameters used are:

- HALFLP: half-loop operation control variable

= 0 for normal operation

= 1 for half-loop operation

- TIHALF: Elapsed time (s) from reactor shutdown

- TGHL(IL): gas temperature in closed loop IL (K)

- XWHLB(IL): Water level in broken loop of closed loop IL (based on calandria floor) (m)

- TWHLB(IL): Coolant temperature in broken loop of closed loop IL

- XWPBHLB(IL): Water level in broken loop pump bowl of closed loop IL

- TWPBHLB(IL): coolant temperature in broken loop pump bowl of closed loop IL

- XWHLU(IL): unbroken loop level of closed loop IL

- TWHLU(IL): Coolant temperature in the unbroken loop of the closed loop IL

- XWPBHLU(IL): water level in unbroken loop pump bowl of closed loop IL

- TWPBHLU(IL): coolant temperature in unbroken loop pump bowl of closed loop IL

- PPS0HL (IL): PHTS pressure (Pa)

- PPZ0HL: Pressurizer pressure

- TGPZ0HL: pressurizer initial gas pressure

- TWPZ0HL: Pressurizer initial coolant temperature

- TSG0HL(1): Coolant and gas temperature in closed circuit-1 broken SG

- TSG0HL(2): Cooling water and gas temperature in closed circuit-2 broken SG

- TSG0HL(3): closed loop-1 coolant and gas temperature in unbroken SG

- TSG0HL(4): closed loop-2 coolant and gas temperature in unbroken SG

- MWSGHL(1): closed loop-1 broken loop SG coolant mass

- MWSGHL(2): closed loop-2 broken loop SG coolant mass

- MWSGHL(3): closed loop-1 unbroken loop SG coolant mass

- MWSGHL(4): closed loop-2 unbroken loop SG coolant mass

- PSG0HL(1): closed loop-1 broken loop SG pressure

- PSG0HL(2): closed loop-2 broken loop SG pressure

- PSG0HL(3): closed loop-1 unbroken loop SG pressure

- PSG0HL(4): closed loop-2 unbroken loop SG pressure

The power plant blackout accident according to this embodiment is an accident in which even the emergency diesel generator fails after the loss of Class IV power, which corresponds to a representative high-voltage accident. In a power plant blackout accident, all power-related systems are lost, so the cooling systems of the main systems including the steam generator water supply system and calandria, that is, the emergency core cooling system, the moderator cooling system, the terminal shield cooling system, and the local air cooler, do not work. Ultimately, it leads to an accident in which the core is damaged. Since the liquid discharge valve (LRV) of the reactor coolant system is a fail-open type valve, it does not close once opened. In the basic accident, recovery measures for the operator were not considered, however, it is assumed that the reactor shutdown is successful, the valve of the reactor building water supply system is operated by the battery, so it operates despite a power outage, and the passive hydrogen recombiner, which is a passive system, also operates. did. In the sensitivity calculation, analysis was performed when it was assumed that the main water supply pipe or the main steam pipe broke at the same time or the functions of each system were partially maintained. The calculation time for the analysis was set to 500,000 seconds (about 139 hours) after the start of the accident. In order to determine the impact of a serious accident according to the progress of such an accident, it is evaluated whether the functions of each system such as bellows breakage, steam generator rapid cooling, emergency core injection, sprinkling, firing igniter, and emergency power supply time are maintained or not.

Power loss accident (SBO, Station Blackout)-A The basic accident background analysis is summarized as follows. When the accident started and the cooling water on the secondary side of the steam generator was depleted due to the interruption of water supply to the steam generator, the pressure on the primary side increased, and the core was exposed due to the loss of coolant through the liquid discharge valve and the degassing condenser. Since the moderator cooling system is not working, the moderator temperature increases due to heat transfer from the horizontal fuel pipe. As the horizontal fuel pipe is damaged by the overpressure of the primary system, the calandria rupture disk is damaged, and the core damage proceeds due to the decrease in the thermal margin and water level of the moderator in the calandria. In a basic accident, since there is no recovery action by the operator, the depletion of coolant in Calandria, damage to the reactor building, and damage to the container of Calandria occur sequentially.

Next, the flow of the evaluation of the impact of a serious accident on the power loss accident in the heavy water reactor will be described in detail.

5 is an exemplary diagram illustrating a pressure behavior of a reactor coolant system according to an accident progression of the reactor coolant system. 5 shows the pressure behavior of the reactor coolant system according to the accident progression after a power outage accident occurred in the power plant and the reactor was immediately stopped. The pressure decrease immediately after the accident is due to a decrease in the core output, and the amount of heat transfer to the secondary side decreases due to a decrease in the water level due to the interruption of water supply to the secondary system, and the reactor coolant system pressure rises again. phosphorus reaches 10.24 MPa (g). After maintaining this pressure for a while, when about 13,000 seconds have elapsed, the horizontal pressure tube creeps and the reactor coolant system pressure decreases to the calandria pressure and thereafter becomes the same as the calandria pressure. The pressurizer pressure shows the same behavior as the reactor coolant system because it is not isolated from the reactor coolant system.

The initial amount of coolant in the reactor coolant system, which was about 41 tons per closed circuit, increased to about 46 tons due to the pressure drop caused by the shutdown of the reactor, cooling water was introduced from the pressurizer. begins to be emitted. 6 is an exemplary view showing a change in the amount of cooling water in the reactor coolant system. As shown in FIG. 6 , after about 10,000 seconds, about 40 tons of cooling water remains in each closed circuit and is reduced by discharge through the degassing condenser. The amount of cooling water in the pressurizer decreases rapidly in the beginning, and it maintains an upward trend due to the increase in the reactor coolant system pressure, and then the coolant decreases through the pressurizer pressure pipe along with the decrease in the reactor coolant system pressure and is depleted after about 14,000 seconds.

7 is an exemplary view showing the secondary side pressure of the steam generator of Loop 1 and Loop 2, which are reactor coolant systems. The pressure on the secondary side of the steam generator of Loop 1 and Loop 2 is shown in FIG. 7 . Before the pressure tube is broken, the secondary system pressure increases and maintains to 5.11 MPa(a), which is the opening/closing set pressure of the main steam safety valve (MSSV) due to heat transfer from the reactor coolant system. Although the four steam generators are modeled individually, they are all connected to each other by a steam header to show similar pressure changes as shown in FIG. 7 . 8 is an exemplary view showing the water level of the steam generator. The water level in the steam generator is shown in FIG. 8 . The steam generator is completely depleted about 10,000 seconds after the start of the accident.

9 is an exemplary view showing the maximum fuel rod temperature in the core. 9 shows the maximum fuel rod temperature in the core. The maximum temperature shows the highest value in the core regardless of the location, and the temperature distribution from about 21,000 seconds to 2,000 K or more indicates that the fuel rod is melting. The behavior of these fuel rods is directly related to the transport of the melt. 10 is an exemplary view showing the transfer of the melt from the fuel pipe to the calandria. As shown in FIG. 10 , the first melt transfer from the fuel pipe to the calandria starts after about 23,000 seconds, and about 123 tons of nuclear fuel material is collected at the bottom of the calandria. During the transport to the calandria bottom, melt cooling and steam generation, zircaloy oxidation, etc. are taken into account. The melt discharged from the fuel rods into the calandria first depletes the moderator and evaporates the coolant in the reactor compartment surrounding the calandria, ultimately damaging the calandria (at about 156,800 seconds), which in turn causes the It can be seen that the melt rapidly decreases.

11 and 12 are exemplary views showing the amount of moderator in calandria and the amount of cooling water in the reactor compartment. 11 (Water mass in Calandria for Case SBO-A, (kg)) and FIG. 12 (Water mass in Reactor Vault for Case SBO-A, (kg)) show the amount of moderator in Calandria and the amount of cooling water in the reactor compartment each is shown. In the moderator system, which has lost its cooling function, the temperature and pressure increase due to heat transfer from the fuel pipe, and around 14,300 seconds when the pressure difference with the reactor building becomes 20 psid, the rupture disk explodes and the moderator begins to leak into the reactor building, causing an accident. After 43,800 seconds, it can be seen that 217 tons of moderators are completely depleted. When the outside of the fuel pipe is completely exposed from the fuel pipe located in the upper part of the calandria due to the decrease in the moderator level, the fuel pipe is rapidly heated, and as a result, the fuel material as shown in FIG. It begins to be transported to the bottom of this calandria. About 156,800 seconds after the start of the accident, the calandria is destroyed by the core melt transferred to the calandria floor, and the calandria melt reacts with the water in the reactor compartment, and the coolant stock in the reactor compartment is also drastically reduced. After about 434,000 seconds, the reactor compartment floor is penetrated and the melt is placed into the reactor building floor. Figure 13 (see Figure 7-35).

13 is an exemplary view illustrating a pressure change in a nuclear reactor building. The reactor building pressure change is shown in FIG. 13 (Boiler room pressures for Case SBO-A, (Pa)). The increase in pressure up to 43,800 s from the onset of the accident is due to steam inflow from the degassing condenser liquid discharge valve due to overpressure in the reactor coolant system and evaporation of the moderator by the core melt after fuel line failure. After the moderator of Calandria is depleted, the pressure decreases, then the pressure rises again as the coolant in the reactor compartment is evaporated by heat transfer through the wall of the Calandria, and after about 74,300 seconds, the reactor building breakage pressure (426 kPa(a)) to reach

14 (Noble gas release fraction into environment for Case SBO-A (fraction)) and 15 (CsI / CsOH release fraction into environment for Case SBO-A (fraction)) are inert gas and CsI and CsOH outside the reactor building. indicates It shows that 100% of inert gas and about 2.7% of CsI and CsOH are released after the reactor building is damaged. FIG. 16 (Hydrogen mass generated from calandria and MCCI for Case SBO-A, (kg)) shows the amount of hydrogen generated inside calandria and by Molten Core Concrete Interaction (MCCI). About 642 kg of hydrogen is generated inside Calandria, and about 2,880 kg of hydrogen is generated by MCCI (Molten Core Concrete Interaction). 17 (Total mass of hydrogen removed by PAR for Case SBO-A, (kg)) shows the amount of hydrogen removed by PAR (Passive Autocatalytic Recombiner). About 355 kg of hydrogen is removed by PAR (Passive Autocatalytic Recombiner). As shown in FIG. 18 (Hydrogen and oxygen mole fractions in boiler room for Case SBO-A), the hydrogen mole ratio of the reactor building boiler room is maintained at a maximum of 3% at the initial stage of the accident, and then increases to a maximum of 54% as oxygen is depleted. .

Next, the analysis of the sensitivity accident process will be described as follows.

Sensitivity calculation results analyzed for cases where the main water supply pipe or main steam pipe breakage is simulated at the same time or the functions of each system are partially maintained can be summarized as follows (refer to Table 1)

1 or more SG coolant depleted Broken one or more pressure tubes CV coolant exhaustion nuclear pile
building damage callandria
damage nuclear pile
compartment
floor breakage SBO-A 10,080 14,290 43,800 74,290 156,840 434,380 SBO-A1 1,400 4,380 32,070 26,130 142,100 412,160 SBO-A2 1,160 7,950 36,860 31,790 149,270 424,130 SBO-A3 1,050 6,190 32,330 66,920 140,470 406,390 SBO-A4 10,080 14,280 43,950 no failure set point 165,140 No Fail SBO-A5 10,080 14,280 287,110 361,930 479,770 No Fail SBO-A6 10,080 14,290 47,550 No Fail No Fail No Fail SBO-B 10,080 13,800 50,380 92,280 168,240 No Fail SBO-B1 10,080 14,290 53,320 86,830 171,450 No Fail SBO-C 10,080 13,800 349,560 No Fail No Fail No Fail SBO-C1 10,080 13,800 355,710 No Fail No Fail No Fail SBO-C2 10,070 14,300 410,400 No Fail No Fail No Fail SBO-C3 10,070 14,300 409,160 No Fail No Fail No Fail SBO-D 10,080 13,800 321,670 205,090 No Fail No Fail

Table 1 shows a summary of the major accident progress by power plant outage accident scenario, and the calculation time is 500,000 seconds.

- SBO-A1: The case where the water supply pipe located inside the reactor building is broken do.

- SBO-A2: This is a case where the main steam pipe located inside the reactor building is broken, and the steam generator hardly performs a cooling function, so the overall accident progress is accelerated. This happens.

- SBO-A3: This is a case where the main steam pipe located outside the reactor building is broken, and the steam generator hardly performs the cooling function, so the overall accident progresses faster.

- SBO-A4: The same as the basic accident process, but the spraying system did not work and the reactor building damage pressure was artificially set high to observe the reactor building pressure behavior. The calculation results show that the maximum pressure increased to about 1 MPa(a) at calandria failure.

- SBO-A5: This is a case where rapid cooling using a steam generator is performed at the time the emergency core cooling signal is generated and the spray tank cooling water is used as the steam generator cooling water. Even if it is carried out, it shows the result that the accident progress is delayed by about 3 days or more after that.

- SBO-A6: When a terminal shielding cooling system is available, it can at least prevent damage to the reactor building and damage to the calandria.

- SBO-B: Assuming that the bellows is broken when the pressure tube is broken, high-pressure safety injection and medium-pressure safety injection are used, which can bring about 3-5 hours of delay.

- SBO-B1: Similar to SBO-B, but without bellows failure. Calandria coolant depletion and failure are slightly delayed, but reactor building failure is accelerated.

- SBO-C, SBO-C1, SBO-C2, SBO-C3: This is the case in which the emergency core cooling system is operated by emergency power for 72 hours after core damage. . Loop isolation has little effect, and bellows breakage affects the calandria coolant depletion time for about 60,000 seconds. don't give

- SBO-D: It is similar to SBO-C, but the heat exchanger of the emergency core cooling system does not work, and the calculation result shows that the reactor building is damaged in about 205,000 seconds.

The present invention is not limited by the embodiments of the present invention and the accompanying drawings in the above description, and it will be apparent to those skilled in the art that various substitutions, modifications and changes are possible without departing from the technical spirit of the present invention. will be.

110: safety system setting unit
120: pressure signal confirmation unit
130: processing unit
140: moderator cooling system status output unit
150: damage progress output unit
160: control unit

Claims (8)

In the method of evaluating the impact of a serious accident of a power loss accident in heavy water,
a) setting an initial condition of the safety system in case of power loss accident with the heavy water;
b) when the pressure on the primary side of the steam generator due to the power loss accident in the heavy water reactor increases, generating a signal for generating a damage to the reactor building for each set time or generating a signal for reducing the cooling water on the secondary side of the steam generator due to interruption of water supply to the steam generator;
c) exposing the core to loss of coolant through a liquid discharge valve and a degassing condenser due to an increase in the pressure on the primary side of the steam generator;
d) outputting the moderator cooling system status as the moderator cooling system is not operating when the moderator temperature increases due to heat transfer from the horizontal fuel pipe at the primary side of the steam generator; and
e) when the overpressure threshold of the primary system of the steam generator is reached, the horizontal fuel pipe is damaged, the calandria rupture disk is damaged, and outputting the damage progress status as the calandria core damage progresses; Methods for evaluating the impact of serious accidents. According to claim 1,
The step of confirming the pressure signal of step b) is outputting a state signal for breaking the water supply pipe due to steam overpressure due to the breakage of the water supply pipe located inside the reactor building when the nuclear reactor building damage occurrence signal is generated at the first set time; A method for evaluating the impact of serious accidents in heavy water power loss accidents. According to claim 1,
The step of confirming the pressure signal of step b) includes: outputting a state signal for breaking the main steam pipe with steam overpressure by breaking the main steam pipe located inside or outside the reactor building when the nuclear reactor building damage occurrence signal is generated at a second set time; A method for evaluating the impact of serious accidents in heavy water power loss accidents, including According to claim 1,
In step a), when the initial condition of the safety system is set, the water spray system among the safety system is set as non-operational and the reactor building damage pressure is set higher than the initial condition of the safety system to calculate the reactor building pressure behavior in step b) A method of evaluating the impact of a serious accident of a power loss accident in heavy water reactors, including the step of: According to claim 1,
When the secondary side cooling water reduction signal in step b) is generated, in the safety system, when rapid cooling using a steam generator is performed and the spray tank cooling water is used as the steam generator cooling water, the damage progress state in step e) is A method of evaluating the impact of a serious accident of a power loss accident in heavy water reactors, including a step of delaying at least for a certain period of time. According to claim 1,
When the initial condition of the safety system is set, when the terminal shielding cooling system is available, the step of outputting at least the impact on the reactor building damage or calandria damage in an undamaged state; . According to claim 1,
When setting the initial condition of the safety system, assuming that the bellows breaks when the pressure tube is damaged and high pressure safety injection and medium pressure safety injection are used, the step of delaying the damage progress state in step e) for at least 3 to 5 hours; A method for evaluating the impact of a serious accident in a power loss accident. According to claim 1,
The step a) is the step of setting the initial condition of the safety system as an initial condition in which the cooling device of the main system, including the steam generator water supply system and calandria, which are safety systems, does not operate; Impact Assessment Methods.

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