KR102255219B1 - Analysis Methodologies of Severe Accident Sensitivity Assesment for End Shield Failure - Google Patents

Abstract

The present invention relates to a severe accident influence evaluating method of a vertical shielding cooling accident which evaluates an influence due to a vertical shielding cooling system failure having a high core damage frequency. According to the present invention, provided is the severe accident influence evaluating method of the vertical shielding cooling accident which provides a safety system operation and non-operation influence evaluating method for the progress of a severe accident due to a vertical shielding cooling system failure having a high core damage frequency.

Description

Analysis Methodologies of Severe Accident Sensitivity Assesment for End Shield Failure

The present invention relates to a method for evaluating the impact of a critical accident of a terminal blocking cooling accident for evaluating the effect on the progress of a serious accident due to a failure of a terminal blocking cooling system having a large core damage frequency.

Conventionally, according to Korean Patent Laid-Open No. 2017-0016588, a radioactive information verification unit that checks whether radioactive materials are leaked, a major area in which one or more systems through which radioactive materials can leak out, and a nuclear power plant stop. It includes a zone classification part that categorizes as a dangerous zone with one or more systems that must be defended so that the initial event, which is all abnormal events resulting, can be resolved, and a major zone analysis part that analyzes through faulty trees.

The heavy water reactor critical accident evaluation code is performed by ISAAC (Integrated Severe Accident Analysis code for CANDU plants), and ISAAC developed KAERI/FAI in 1995, and modeled reflecting the characteristics of CANDU based on the severe accident analysis code (MAAP4). There is a problem in that there is no evaluation result of the safety system operation and non-operation effects for the progress of a serious accident due to a coolant leakage accident while the heavy water reactor is stopped at a heavy water reactor 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 Wolseong Unit 1, which was evaluated to supplement the Probabilistic Safety Assessment (PSA). There is a need to research and develop a method for evaluating the impact of severe accidents of terminal-shielded cooling accidents to evaluate the effects of the failure of the shielding cooling system.

The present invention is to solve the above-described problems, and the impact of the critical accident of the terminal shield cooling accident to provide a method for evaluating the impact of the safety system operation and non-operation against the progress of a serious accident caused by a failure of the terminal shield cooling system with a large core damage frequency. The purpose is to provide an evaluation method.

The present invention provides a method for evaluating the impact of a critical accident of a terminal blocking cooling accident, comprising the steps of: a) setting an initial condition of a safety system in the case of the terminal blocking cooling accident; b) a break signal confirmation step of generating a horizontal fuel pipe break signal when the pressure decrease signal is generated while maintaining the initial value of the primary pressure of the reactor coolant system for a first predetermined time; c) a rapid decompression operation confirmation step of generating a rapid decompression operation failure signal when the secondary system pressure of the reactor coolant system maintains the initial pressure and then the reactor is stopped and the pressure by the primary system low pressure signal decreases; And d) generating a calandria damage signal due to evaporation of shielded coolant in the reactor compartment after a second predetermined time according to the rapid decompression operation failure signal.

Preferably, in step a), the initial condition of the safety system is operated at 100% full power for the first predetermined time after the start of the accident, water is supplied to the steam generator, the moderator cooling system is also available, and only the terminal shielding cooling system has failed. It is assumed that, after the first predetermined time, some fuel pipes of the terminal shield are damaged, and thereby, the coolant and the moderator are set to leak to the calandria and the nuclear reactor building, respectively.

Preferably, in the safety system under the initial conditions of the safety system in step a), when rapid cooling of the steam generator is available, the integrity of the reactor building is maintained after step d).

Preferably, if the local air cooler fails in the safety system under the initial conditions of the safety system in step a), the depletion point of the moderator is shortened after step d), and the shielded coolant in the reactor compartment is heated and evaporated to damage the reactor building. It includes;

Preferably, in the initial condition of the safety system in step a), if auxiliary water supply to the steam generator is available in the safety system, the steam generator level is maintained after step d), but the heat transfer to the secondary side is rapidly reduced as the PHTS coolant is depleted. And the step of destroying the nuclear reactor building due to the failure of the local air cooler.

Preferably, if high and medium pressure ECC (Emergency Core Cooling) is available in the safety system under the initial conditions of the safety system in step a), the accident progression is delayed due to the injection of ECC in step b), but low pressure injection, which is a recirculation mode. It includes a; step in which the failure proceeds to core damage.

And preferably, in step a), in the initial conditions of the safety system, when water is supplied to the steam generator and the moderator cooling system is also available, or when the local air cooler is operated, damage to the reactor building is prevented.

According to the present invention, the effects of safety system operation, reactor building pressure on non-operation, hydrogen concentration, fuel pipe melting, callandria pipe damage, etc. at the time of progress due to a serious accident caused by a breakdown of the terminal shielding cooling system with a high core damage frequency are identified. By doing so, it can be reflected in the guidelines for mitigating serious accidents, and it has the effect of contributing to the license by submitting an evaluation for the continuous driving license of the Wolseong Unit 1.

1 is an exemplary view showing a heavy water reactor type nuclear power plant for evaluating the impact of a severe accident of a terminal shield cooling accident according to an embodiment of the present invention.
2 is an exemplary view showing a loop of a coolant system for evaluating the impact of a serious accident of a terminal shielding cooling accident according to an embodiment of the present invention.
3 is an exemplary view showing a schematic diagram of a heavy water reactor coolant system for evaluating the impact of a serious accident of a terminal shielding cooling accident according to an embodiment of the present invention.
4 is an exemplary view showing a system for evaluating the impact of a serious accident of a cooling accident with a terminal shield according to an embodiment of the present invention.
5 is a flow chart showing a method for evaluating the impact of a severe accident of a terminal shielding cooling accident according to an embodiment of the present invention.
6 is an exemplary view showing the event tree of the damaged group of a terminal-shielded cooling system loss accident (LOS) power plant.
7 and 8 are exemplary diagrams showing the pressure behavior and coolant mass of the primary system.
9 and 10 are exemplary diagrams showing the pressure behavior of the secondary system and the mass of cooling water.
11 is an exemplary view showing the amount of fuel material relocated to callandria in the horizontal fuel pipes of each closed circuit.
12 is an exemplary view showing the mass of the shielding coolant of the moderator of Calandria and the reactor compartment.
13 is an exemplary view showing a pressure change in a nuclear reactor building.
14 and 15 are exemplary views showing the release rate of an inert gas and CsI and CsOH outside the reactor building.
16 is an exemplary view showing the amount of hydrogen generated in the inner callandria and the core melt-concrete reaction (MCCI).
17 is an exemplary view showing the amount of hydrogen removed by a driven hydrogen recombiner (PAR).
18 is an exemplary diagram showing hydrogen concentration 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 known functions and configurations that may unnecessarily obscure the subject matter of the present invention will be omitted.

As shown in Figure 1, the heavy water reactor type nuclear power plant is a large calandria vessel surrounds fuel channels 20, 380 pressure tubes in the end shield 30 ( 10) is installed horizontally like the Callandria Tube, and the fuel supply is attached to individual fuel channels (40), and if the coolant cannot be filled in the core, it is critical. It can develop into thinking. As shown in FIG. 2, the coolant system of a heavy water reactor type nuclear power plant is composed of two loops, and it is possible to isolate an accident, thereby minimizing accidents.

3 is an exemplary view showing a schematic diagram of a heavy water reactor coolant system for evaluating the impact of a serious accident of a terminal shielding cooling accident according to an embodiment of the present invention. In FIG. 3, 1 is CALANDRIA, 2 is CALANDRIA END SHIELD, 3 is SHUT-OFF AND CONTROL RODS, 4 is POISON INJECTION, 5 is FUEL CHANNEL ASSEMBLIES, 6 for FEEDER PIPES, 7 for VAULT

4 is a block diagram showing a system for evaluating the impact of a serious accident of a cooling accident with a terminal shield according to an embodiment of the present invention. As shown in Figure 4, the critical accident impact evaluation system (1) of the terminal shielding cooling accident is a safety system setting unit 110, a break signal confirmation unit 120, a rapid decompression operation confirmation unit 130, the callandria damage A signal generator 140 and a control unit 150 are included.

The safety system setting unit 110 is a configuration for setting the initial conditions of the safety system in the event of a terminal shielding cooling accident. Here, the initial condition of the safety system is that after the start of the accident, it is operated at 100% full power for the first predetermined time, water is supplied to the steam generator, the moderator cooling system is also available, and only the terminal shielding cooling system has failed. After time, it is set that some of the fuel pipes in the end shields are damaged, resulting in a coolant and moderator leaking into the Callandria and the reactor building, respectively.

Also, since there is no model for a stationary cooling system pump in the ISAAC code, it is assumed that a break has occurred in the pump discharge line of each PHTS (Primary Heat Transport System) loop. At this time, the breaking area of each loop is 2.236 x 10-2 m ² . At this time, the initial total leakage is 10.83 kg/s. At the start of the accident, the PHTS (Primary Heat Transport System) pressure is 601 kPa(a), and the coolant temperature is 54℃ (327 K). The pressurizer is already isolated from the PHTS (Primary Heat Transport System) before the start of the accident, and the secondary side temperature of the steam generator is 54 °C (327 K). When the functions of each system such as rapid cooling of the steam generator, emergency core injection, local air cooler, sprinkling, operation of the shielded cooling system, and auxiliary water supply of the steam generator are maintained or not Evaluate.

In the initial conditions of the safety system, a failure of the end-shielding cooling system (LOS) is a failure of the end-shielding cooling system, resulting in loss of the cooling function of the reactor compartment and the end-shield, but the operator fails to recognize the accident and fails to stop the reactor due to low water level. It is an accident in which all of the reactor's continuous output deceleration (SETBACK) fails. In the basic accident, the steam generator rapid cooling, auxiliary water supply, and emergency core cooling systems failed, but the local air cooler and passive hydrogen recombiner, which are the safety systems of the nuclear reactor building, were available and the sprinkling system was not considered.

The break signal checking unit 120 is configured to maintain the initial value of the primary pressure of the reactor coolant system for a first predetermined time in the event of a terminal shielding cooling accident, and then generate a horizontal fuel pipe break signal when a pressure reduction signal is generated.

The rapid decompression operation check unit 130 is configured to generate a rapid decompression operation failure signal when the secondary system pressure of the reactor coolant system maintains the initial pressure and then the reactor is stopped and the pressure by the primary system low pressure signal decreases.

The calandria damage signal generation unit 140 generates a callandria damage signal due to evaporation of the shielding coolant in the reactor compartment after a second predetermined time in accordance with the rapid decompression operation failure signal of the rapid decompression operation confirmation unit 130.

The control unit 150 is a component that controls the above-described configurations, and a method of evaluating the impact of a critical accident of an end-shielding cooling accident with such a control unit will be described as follows.

5 is a flow chart showing a method for evaluating the impact of a severe accident of a terminal shielding cooling accident according to an embodiment of the present invention.

First, the control unit sets the initial conditions of the safety system in case of a terminal shielding cooling accident (a). Next, the initial value of the primary pressure of the reactor coolant system is maintained for a first predetermined period of time, and a break signal confirmation step (b) of generating a horizontal fuel pipe break signal when a pressure reduction signal occurs, and then the secondary system pressure of the reactor coolant system Rapid decompression operation confirmation step (c) to generate a rapid decompression operation failure signal when the pressure decreases due to the primary system low pressure signal while the reactor is stopped while maintaining this initial pressure, followed by a second predetermined time according to the rapid decompression operation failure signal. After that, a step (d) of generating a calandria damage signal due to evaporation of the shielding coolant in the reactor compartment is reached.

Here, in step a), the initial condition of the safety system is operated at 100% full power for the first predetermined time after the start of the accident, water supply is supplied to the steam generator, the moderator cooling system is also available, and the terminal shielding cooling system It is assumed that only failed, and after the first predetermined time, a part of the fuel pipe of the terminal shield is damaged, and thus the coolant and the moderator are leaked to the calandria and the reactor building, respectively.

In addition, the control unit maintains the integrity of the nuclear reactor building after step d) in step a), when rapid cooling of the steam generator is available in the safety system under the initial conditions of the safety system.

In addition, if the local air cooler fails in the safety system under the initial conditions of the safety system, the depletion time point of the moderator is shortened after step d), and the shielded coolant in the reactor compartment is heated and evaporated to damage the reactor building.

In addition, if the auxiliary water supply of the steam generator is available in the safety system under the initial conditions of the safety system, the water level of the steam generator is maintained after step d), but the heat transfer to the secondary side is rapidly reduced as the PHTS coolant is depleted, resulting in a failure of the local air cooler. The nuclear reactor building is destroyed.

If the safety system in the initial conditions of the safety system, high and medium pressure ECC (Emergency Core Cooling) is available, the accident progress is delayed due to ECC injection in step b), but the low pressure injection in the recirculation mode fails, leading to core damage. .

During the initial conditions of the safety system, when water is supplied to the steam generator and the moderator cooling system is also available, or when the local air cooler is operated, damage to the reactor building is prevented.

Hereinafter, with reference to the accompanying drawings, a detailed description of the method for evaluating the impact of a severe accident of a terminal shielding cooling accident will be described.

The severe accident analysis code used in the present invention will be described as follows. The ISAAC code was used to simulate the serious accident at Wolseong Unit 1. The ISAAC code uses MAAO4/PWR, developed for comprehensive serious accident simulation in light water reactors, as a reference code, and is an eight-shaped shape consisting of 380 horizontal fuel pipes and two independent closed circuits to simulate the characteristics of heavy accidents inherent in heavy water reactors. Of the primary system (PHTS), a pressurizer connecting each closed circuit, a steam header shared by four steam generators and the secondary side of the steam generator, a callandria with a horizontal fuel pipe locked, a reactor compartment, a terminal shield, a degassing condenser In addition, various safety systems were added for accident mitigation. Therefore, the ISAAC code is based on the temperature rise in the fuel rod and the horizontal fuel pipe according to the accident, the reaction with the deformation and moderator, the oxidation reaction between zircaloy and water vapor, the damage of the horizontal fuel pipe, the core relocation, the formation of a suspended debris bed and heat transfer. It is possible to predict the progress of the accident according to the characteristics of heavy water channels such as the behavior of fuel materials at the bottom of Calandria and damage to Calandria. In addition, when the bellows of the end joint pipe are simultaneously broken due to the breakage of the pressure tube in a high-pressure environment, in the case of a break in the supply pipe or a power outage in a high-pressure accident, the accident situation in which the bellows break, and the stationary cooling system in a stationary condition ISAAC was improved to simulate an accident in which the coolant was lost due to a leak in the cooling pump bearing of. In the present invention, accident analysis was performed using ISAAC version 4.03.

In one embodiment of the present invention, the initial case was reviewed by applying the analysis methodology at the PL (Point Lepreau, Canada) nuclear power plant to the latest result of the Wolseong Unit 1 Probabilistic Safety Assessment (PSA), and the Wolseong Unit 1 core damage frequency. Termination shielding cooling system failure (LOS), which accounts for about 15%, was added as the 6th initial accident. The six initial accidents are listed: supply pipe congestion rupture (SFB), power plant blackout (SBO), stationary accident (SSA), steam generator customs rupture (SGTR), small coolant loss (SLOCA), and termination. Loss of shielded cooling system (LOS). Here, when explaining the selection of an accident scenario for the end-shielding cooling system failure (LOS) as follows. The end-shielding cooling system failure (LOS) accident was an initial accident that was not included in the existing PL, and was classified as P-6 in this analysis. In the case of Wolseong Unit 1, the fraction of the end-shielding cooling system failure of the total core damage frequency is about 15%, and it is the fourth most contributing factor after class 4 power loss, cooling water system malfunction, and coolant distribution pipe failure. Therefore, P-6 was added to this analysis, and 6 accident scenarios analyzed according to the Loss of Terminal Shielding Cooling System Accident (LOS) and Plant Damage State Event Tree (PDSET, Plant Damage State Event Tree) in Fig. 6 (Fig. 7-13) (LOS-1, LOS-1B, LOS-2, LOS-2B, LOS-2C, LOS-2D) were selected. These scenarios will be described below.

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

-The reactor is in full power (2157.5 MW(th)) except for stationary accidents.

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

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

-Class 4 and 3 power sources are available except for SBO (Station Blackout) accidents.

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

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

-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 a closed circuit isolation valve when the pressure is lower than 5.516 MPa(a). Accordingly, the pressure is controlled at the set point of the liquid discharge valve until the pressure pipe is damaged.

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

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

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

-When the coolant inside the pressure pipe is depleted and the level of the moderator in Calandria is lowered and the fuel pipe is exposed to the inside and outside, the fuel pipe begins to deform, deviates from its original position, and relocates around a healthy horizontal fuel pipe in the lower space to solidify and solidify. (Suspended Debris Bed) is formed. If the level of the moderator is continuously lowered, the number of exposed fuel pipes increases, and when the healthy horizontal fuel pipe at the bottom cannot support any more due to the increase of fuel material transferred to the SDB (Suspended Debris Bed), the core collapses and the relocated fuel. It is assumed that all of the material is transferred to the bottom of Calandra.

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

-When the level of the shielding coolant decreases and the fuel material at the bottom of the callandria is exposed, cooling of the outer wall is no longer performed, and thus, damage to the callandria is expected.

-If the callandria is damaged, the fuel material inside the callandria is relocated to the reactor compartment and reacts with the concrete on the floor. Initially, the shielding coolant is evaporated, and then concrete erosion occurs and proceeds until the floor is punctured.

-The reactor building sprinkling system is a driven system and operates to stop when the reactor building pressure is higher than 14 kPa(g) and stops when it falls below 7 kPa(g), and the amount of spraying is about 1132 kg/s per each downcomer. The amount of cooling water that can be sprayed is 1,559 m3.

-While the sprinkling 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 total, but 16 with a large capacity are safety grades, and manual operation is possible in the main control room, and 12 out of 16 are used as success criteria for simulation.

-In Wolseong Unit 1, 27 passive material combiners (PARs) are installed in total (10 small, 15 medium, 2 large) for hydrogen control in the nuclear reactor building, and the installation location is 10 small in the upper dome. There are 5 medium-sized units and 5 medium-sized units, 5 medium-sized units in the boiler room, 3 medium-sized units in the moderator room, and 1 medium-sized and 1 large-sized unit in each fuel exchange room. The number and mode of operation can be adjusted by input, and in this analysis, when the hydrogen concentration reaches 2%, it delays for 30 minutes before operation, stops when the concentration falls below 0.2%, and operates without delay when restarting. In order for PAR to operate, adequate oxygen must be supplied.

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

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

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

① Bellows fracture model

If the pressure pipe is damaged in an accident where the PHTS (Primary Heat Transport System) is maintained at high pressure, the bellows of the end joint pipe may be damaged at the same time due to pressure propagation in the fuel pipe. Such bellows breakage may accompany supply pipe congestion breakage (SFB) and power plant power failure (SBO).

In ISAAC, the following input parameters were introduced to simulate the breakdown of the pressure tube and the leakage of the moderator from Callandria to the reactor building.

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

-ZFAILBELLO: The height of the break from the callandria floor

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

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

-ISTAGRUP (ICH, IL): Horizontal fuel pipe pressure pipe fracture area control variable after SFB

If the value of the ISTAGRUP variable is 1, the pressure pipe fracture area is limited to the maximum flow path area of the inlet and outlet supply pipe, and for other values, the pressure pipe fracture area is selected from the sum of twice the pressure pipe flow path and the sum of the inlet and outlet supply pipe flow channels. It is calculated by multiplying by the actual number of fuel pipes.

②Initial model of stop

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, the water level, and the PHTS, the temperature at the steam generator, the pressurizer, and the parameters indicating that the PHTS coolant is separated from the steam. . Based on these input variables, ISAAC defines the PHTS initial conditions. Under these conditions, the PHTS water level is located at the height of the capillary inlet and outlet, and the heat of decay is removed by the static cooling system. The initial conditions are similarly defined for steam generators and pressurizers. The variables used are:

-HALFLP: Control variable for half-loop operation

= 0 for normal operation

= 1 for half-loop operation

-TIHALF: Elapsed time from reactor shutdown (s)

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

-XWHLB(IL): The water level in the broken loop of the closed loop IL (based on the Calandria floor) (m)

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

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

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

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

-TWPBHLU(IL): Coolant temperature in the unbroken loop pump bowl of the closed circuit IL

-PPS0HL(IL): PHTS pressure (Pa)

-PPZ0HL: Pressurizer pressure

-TGPZ0HL: Initial gas pressure of pressurizer

-TWPZ0HL: Initial coolant temperature of pressurizer

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

-TSG0HL(2): Closed circuit-2 Coolant and gas temperature in broken SG

-TSG0HL(3): Closed circuit-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

Loss of Shield Cooling (LOS) according to this embodiment results in the loss of the cooling function of the reactor compartment and the end shield due to the failure of the end shielding cooling system, but the operator does not recognize the accident and fails to stop the reactor. It is an accident in which all reactor setbacks due to low water level fail. This accident was derived based on the PDSET #49 accident, which has the highest frequency in the first stage PSA (Probabilistic Safety Assessment). In this basic accident, the steam generator rapid cooling, auxiliary water supply, and ECC (Emergency Core Cooling) failed, but the local air cooler and passive hydrogen recombiner, which are the safety systems of the nuclear reactor building, were available and the trickling system was not considered. The reactor building breakdown pressure was assumed to be 426 kPa(a), and the accident analysis was performed up to 500,000 seconds (about 139 hours) after the start of the accident so that the concrete penetration of the reactor compartment floor could be seen.

The analysis of the basic accident situation (LOS-1) according to the present embodiment will be described as follows. Since there is no detailed thermal hydraulic analysis data for this accident, it is assumed that the fuel channel is damaged in the terminal shield after 1 hour after the start of the accident, using the results of the Probabilistic Safety Assessment (PSA) performed by AECL (Atomic Energy of Canada Limited). do. In this analysis, it is assumed that the operation is operated at 100% full power for 1 hour from the start of the accident, water supply is supplied to the steam generator, the moderator cooling system is also available, and only the end-shield cooling system has failed. After 1 hour, some of the fuel pipes of the terminal shield were damaged, and the coolant and moderator were simulated to leak into the Kallandria and the reactor building, respectively. For this, the breakage of the horizontal fuel pipe #16 was assumed, and the bellows breakage model was applied to reflect the moderator leakage.

Figure 7 (PHTS pressures for Case LOS-1 (Pa)) and Figure 8 (PHTS water masses for Case LOS-1 (kg)) show the pressure behavior and coolant mass of the primary system. For the first hour, even if the end-shielding cooling system fails, the pressure on the primary side maintains the initial value of 10 MPa(a), and after 1 hour, the pressure rapidly decreases by artificially stopping the reactor and breaking the horizontal fuel pipe. . Therefore, the mass of the moderator of the primary system also decreases over 3,600 seconds. The secondary system of the steam generator maintains the initial pressure and then the reactor is stopped, and the rapid decompression operation by the low pressure signal of the primary system fails, and the pressure gradually decreases, and the water supply to the steam generator is stopped. As this decreases, the secondary side coolant is not depleted, but it does not help to alleviate accidents. 9 (SG pressures for Case LOS-1 (Pa)) and FIG. 10 (SG water masses for Case LOS-1 (kg)) show the pressure behavior and cooling water mass of the secondary system.

Figure 11 (Corium masses in loops and calandria for Case LOS-1 (kg)) shows the amount of fuel material relocated to callandria in the horizontal fuel pipe of each closed circuit. In both Loops 1 and 2, the core collapses in about 6,340 seconds, relocating the fuel material to the bottom of the callandria, and relocating to the bottom of the reactor compartment in about 92,000 seconds when the callandria is damaged. Figure 12 (Water masses in calandria and reactor vault for Case LOS-1 (kg)) shows the masses of the shielding coolant in the moderator and reactor compartment of Kalandria, respectively. Leakage of the moderator was simulated after 1 hour, and the shielding coolant in the reactor compartment started to heat up after depletion of the moderator and evaporated from around 16,000 seconds. Eventually, the calandria was damaged in about 92,000 seconds as the calandria floor was exposed and shielded at the same time. The mass of cooling water also decreases sharply due to reaction with the melt.

The reactor building pressure change is shown in Fig. 13 (Boiler room pressure for Case LOS-1 (Pa)). The initial pressure increase is due to mass transfer due to the breakage of the horizontal fuel pipe over 1 hour, and the peak pressure at about 92,000 seconds and 320,000 seconds is caused by the reaction of the fuel material with the coolant due to the callandria breakage and the penetration of the concrete at the bottom of the reactor compartment. Because of the steam. In this accident, the local air cooler operates, so no damage to the reactor building occurs.

14 (Noble gas release fraction into environment for Case LOS-1) and 15 (CsI/CsOH release fraction into environment for Case LOS-1) show the release rates of inert gas and CsI and CsOH outside the reactor building. Although the nuclear reactor building is sound, the emission area was reflected in consideration of the design leakage rate, showing that about 0.82% of the inert gas and about 0.005% of the initial amount of CsI and CsOH were released. Figure 16 (Hydrogen mass generated from calandria and MCCI for Case LOS-1 (kg)) is the amount of hydrogen generated in the inner callandria and the core melt-concrete reaction (MCCI, Molten Core Concrete Interaction), respectively, in Loops 1 and 2 46 kg is produced, and about 92 kg is produced inside the callandria, and about 2940 kg of one hydrogen is produced by the reaction between the core melt relocated to the reactor compartment and the bottom concrete. No more hydrogen is produced after the reactor compartment floor is pierced. Figure 17 (Total mass of hydrogen removed by PARs for Case LOS-1 (kg)) shows the amount of hydrogen removed by a passive hydrogen recombiner (PAR), Figure 18 (Hydrogen and oxygen mole fractions in boiler room). for Case LOS-1) shows the hydrogen concentration in the boiler room of the reactor building. It removes up to 180 kg of the amount produced by the core in the beginning and the amount produced by the MCCI in the second half, but after 180,000 seconds, the oxygen depletion in the reactor building no longer removes hydrogen. At the beginning of the accident when the PAR (Passive Autocatalytic Recombiner) operates, the hydrogen concentration does not exceed 4%, but increases up to 33% as oxygen is depleted.

The details of the sensitivity accident are as follows. The sensitivity calculation result of analyzing the cases where the functions of each system such as bellows break, rapid decompression of the steam generator, emergency core injection, local air cooler, sprinkling, end-shielding cooling system, and auxiliary water supply of the steam generator are maintained are as follows It can be summarized together. Table 1 shows the summary of the main accident progression by failure scenario (LOS-1, LOS-1B, LOS-2, LOS-2B, LOS-2C, LOS-2D) of the end-shielding cooling system. (Calculation time: 500,000 seconds)

Broken pressure pipe Fuel material rearrangement
(Loop 1/2) Moderator
depletion Nuclear reactor building
damage Kalandria
damage Damage to the reactor compartment floor LOS-1 5,460 5,532
/5,534 7,107 No Fail 92,088 321,879 LOS-1B 5,452 5,527
/5,532 7,119 No Fail 92,124 322,539 LOS-2 6,007 6,089 7,606 46,462 93,174 325,220 LOS-2B 6,064 6,155 7,658 48,934 92,709 324,136 LOS-2C 6,007 6,089 7,606 36,836 92,980 325,199 LOS-2D 13,685 13,765
/13,956 20,873 65,459 116,522 361,538

-LOS-1B: Similar to the aforementioned basic accident scenario (LOS-1), but rapid cooling of the steam generator is possible. However, as auxiliary water supply and emergency core injection are not available, the effect of rapid cooling of the steam generator does not appear. The accident progression is almost similar to that of LOS-1, and the reactor building maintains soundness with the operation of the local air cooler.

-LOS-2: Similar to LOS-1, but this is an accident where the local air cooler has failed. The accident progression is almost similar to that of LOS-1, and the reactor building is damaged in about 12.9 hours due to the failure of the local air cooler. The reason why the reactor building is damaged quickly is that the depletion point of the moderator is fast (about 7,600 seconds), and this causes the shielding coolant in the reactor compartment to heat up and evaporate after 16,000 seconds.

-LOS-2B: Similar to LOS-2, but when auxiliary water supply to the steam generator is available. The steam generator level is maintained due to the availability of auxiliary water supply, but as the PHTS (Primary Heat Transport System) coolant is depleted after 4,000 seconds, heat transfer to the secondary side rapidly decreases, and the accident progression is similar to that of LOS-2. The reactor building is destroyed in about 13.6 hours due to the failure of the local air cooler.

-LOS-2C: With the same accident as LOS-2, the reactor building damaged in about 12.9 hours was damaged in about 10.2 hours by changing only the damage pressure of the reactor building from 426 kPa(a) to 358 kPa(a).

-LOS-2D: Similar to LOS-2, but with high and medium pressure ECC (Emergency Core Cooling) available, high pressure operates from about 3,610 seconds to 3,813 seconds, and medium pressure from 3,813 seconds to about 7,037 seconds. The accident progression is delayed due to ECC injection, but the low pressure injection, which is a recirculation mode, fails and eventually leads to core damage. Compared to LOS-2, moderator depletion is approximately 3.7 hours, callandria damage is approximately 6.5 hours, and reactor compartment floor penetration time is delayed by approximately 10 hours.

In the event of a failure of the terminal shield cooling system (LOS) according to an embodiment of the present invention, the cooling function of the reactor compartment and the terminal shield is lost due to the failure of the terminal shield cooling system, but the operator does not recognize the accident, so the reactor fails to stop and reaches a low water level It is an accident in which all of the reactor setback by failure. This incident was derived based on the PDSET #49 incident, which is the most frequent in the Phase 1 PSA. However, since there is no thermal-hydraulic analysis data for such an accident, from the existing PSA results, it was assumed that the fuel channel in the end shield was damaged after 1 hour from the start of the accident.

Due to the sudden progression of the accident after 1 hour, the moderator is depleted after about 2 hours, and if a local air cooler is available, the reactor building is healthy, but otherwise, the reactor building is damaged in about 10 to 18 hours. Steam generator rapid cooling operation hardly affects the accident progress, and if high and medium pressure ECC (Emergency Core Cooling System) is available, the test progress is delayed by 4 to 10 hours. In this basic accident, the auxiliary water supply also does not alleviate the accident, because both closed circuits are damaged from the beginning of the accident and heat transfer to the secondary side is insignificant.

If the reactor building is healthy, the inert gas leaked to the outside, CsI and CsOH are insignificant, at 0.8% and 0.005%, but when the reactor building is damaged, almost all of the inert gas is released, and CsI and CsOH are discharged up to 1.5% of the initial amount do. The hydrogen concentration of the nuclear reactor building does not exceed 4% at the beginning of the accident when the PAR (Passive Autocatalytic Recombiner) is operated, but increases up to 33% due to the generation of hydrogen in the second half from MCCI (Molten Core Concrete Interaction) as oxygen is depleted. . However, it is believed that hydrogen explosion inside the reactor building will not occur due to oxygen depletion.

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 within the scope of the technical spirit of the present invention. will be.

110: safety system setting unit
120: break signal confirmation unit
130: Rapid decompression operation confirmation unit
140: Callandria damage signal generation unit
150: control unit

Claims (7)

In the method of evaluating the impact of severe accidents of the terminal shielding cooling accident
a) setting an initial condition of the safety system in the event of the terminal shielding cooling accident;
b) a break signal confirmation step of generating a horizontal fuel pipe break signal when the pressure decrease signal is generated while maintaining the initial value of the primary pressure of the reactor coolant system for a first predetermined time;
c) a rapid decompression operation confirmation step of generating a rapid decompression operation failure signal when the secondary system pressure of the reactor coolant system maintains the initial pressure and then the reactor is stopped and the pressure by the primary system low pressure signal decreases; And
d) generating a callandria damage signal due to evaporation of the shielding coolant in the reactor compartment after a second predetermined time in accordance with the rapid decompression operation failure signal. The method of claim 1,
In step a), the initial condition of the safety system is that after the start of the accident, it is operated at 100% full power for the first predetermined time, water is supplied to the steam generator, the moderator cooling system is also available, and only the terminal shielding cooling system has failed. And, after the first predetermined time, a part of the fuel pipe of the terminal shield is damaged, and thereby, the coolant and the moderator are set to leak to the Callandria and the nuclear reactor building, respectively. The method of claim 1,
In the case where the rapid cooling of the steam generator is available in the safety system under the initial conditions of the safety system in step a), the method for evaluating the impact of the critical accident of the terminal shield cooling accident that maintains the integrity of the nuclear reactor building after the step d) . The method of claim 1,
If the local air cooler fails in the safety system under the initial conditions of the safety system in step a), the depletion time of the moderator is shortened after step d), and the shielding coolant in the reactor compartment is heated and evaporated to damage the reactor building. A method for evaluating the impact of a critical accident of a terminal shielding cooling accident comprising a step. The method of claim 1,
When the auxiliary water supply to the steam generator is available in the safety system under the initial conditions of the safety system in step a), the water level of the steam generator is maintained after step d), but as the PHTS coolant is depleted, heat transfer to the secondary side rapidly decreases. A method for evaluating the impact of a serious accident of a terminal shield cooling accident comprising; the step of destroying the nuclear reactor building due to a failure of the local air cooler. The method of claim 1,
In the case where high and medium pressure ECC (Emergency Core Cooling) is available in the safety system under the initial conditions of the safety system in step a), the accident progression is delayed due to ECC injection in step b), but low pressure injection, which is a recirculation mode, is A method for evaluating the impact of a critical accident of a cooling accident with a terminal shield including a failure and proceeding to a core damage. The method of claim 1,
In step a), during the initial conditions of the safety system, when water is supplied to the steam generator and the moderator cooling system is also available, or when the local air cooler is operated, the impact of a severe accident of a terminal shield cooling accident that prevents damage to the reactor building. Assessment Methods.

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