KR102351448B1 - Analysis Methodologies of Severe Accident Sensitivity Assesment for Steam Generator Tube Rupture in CANDU - Google Patents

Abstract

The present invention relates to a method for evaluating the impact of a severe accident of a heavy water reactor steam generator tube rupture accident for evaluating the effect of a steam generator tube rupture accident with a high frequency of core damage. The present invention provides a method for evaluating the impact of serious accidents in heavy water steam generator tube rupture accidents to provide a method for evaluating the effects of safety system operation and non-operation on the development of a serious accident due to a steam generator tube rupture accident with a high frequency of core damage.

Description

Analysis Methodologies of Severe Accident Sensitivity Assesment for Steam Generator Tube Rupture in CANDU}

The present invention relates to a method for evaluating the impact of a severe accident of a heavy water reactor steam generator tube rupture accident for evaluating the effect of a steam generator tube rupture accident with a high frequency of 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 dangerous zone where there is one or more systems that must be defended so that the initial events, which are all abnormal events that cause, can be resolved, and a major zone analysis part that analyzes through faulty trees.

The heavy water reactor serious accident evaluation code is implemented with ISAAC (Integrated Severe Accident Analysis code for CANDU plants), which was developed by KAERI/FAI in 1995 and modeled reflecting the CANDU characteristics based on MAAP4. There is a problem in the fact that there are no evaluation results of safety system operation and non-operation effects on the progress of serious accidents due to the rupture of the steam generator tube with a high frequency of damage.

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

The present invention is to solve the above problems, and to provide a method for evaluating the effect of safety system operation and non-operation on the development of a serious accident caused by a steam generator tube rupture accident with a high frequency of core damage. The purpose is to provide an accident impact assessment method.

The present invention provides a method for evaluating the impact of a serious accident of a customs rupture accident of a heavy water steam generator, comprising the steps of: a) setting initial conditions for a safety system in the case of a tubular rupture accident of the steam generator; b) generating a coolant loss accident signal when the primary pressure of the coolant system is reduced due to the occurrence of a reactor stop due to the tube breakage accident of the steam generator; c) isolating the two closed circuits of the coolant system so that the isolation valve is locked around the pressurizer when the coolant loss accident signal occurs; d) maintaining a constant pressure after the primary side pressure of the coolant system reaches the main steam safety valve opening/closing pressure after a predetermined time elapses; e) determining that a reverse flow of coolant has occurred from the secondary side of the coolant system to the primary side when the pressure tube temperature reaches a critical point under the pressure on the primary side of the coolant system; f) allowing the secondary-side liquid release valve to maintain a pressure higher than the open set value until the horizontal fuel pipe is damaged due to an increase in secondary-side pressure of the coolant system.

Preferably, when the secondary side water supply of the coolant system is not supplied due to the reactor shutdown in step b), the secondary side (sound closed circuit) pressure increases while heat transfer to the secondary side is reduced, and high pressure due to coolant exhaustion and generating a fuel pipe breakage occurrence signal in which a horizontal fuel pipe breakage on the secondary side occurs due to a rise in temperature of the bottom.

Preferably, after step a), among the safety systems, when a local air cooler (LAC) operates, a step of not generating a nuclear reactor building damage signal and a nuclear reactor stop occurrence signal.

Preferably, after step a), the primary side (Loop1) horizontal fuel compared to the failure of the rapid decompression (CC) operation until the primary side (Loop1) is broken by the rapid decompression (CC) operation of the safety system / until the cooling water of the sound steam generator is exhausted Including; the step of showing the low-pressure accident behavior of the pipe damage time is delayed.

Preferably, after step a), in the safety system, when auxiliary water supply is supplied to the steam generator, the stock of cooling water in the primary side (Loop1) is continuously discharged through the ruptured part of the customs until the primary side (Loop1) horizontal fuel pipe is damaged. In the secondary side (Loop2), the decay heat is removed by the secondary side heat removal method by supplying the auxiliary water supply, so that the calandria moderator depletion and damage time are delayed compared to the rapid decompression (CC) operation.

Preferably, after step a), when ECC (Emergency Core Cooling) is injected in the safety system, the ECC injection is started when the primary side pressure falls below a certain set value after the reactor is stopped at the secondary side (Loop2), and the primary side When the pressure increases above the preset value, the injection is stopped, and the primary side (Loop1) of the coolant system reaches the saturation temperature later than the secondary side (Loop2).

Preferably, after step a), when ECC (Emergency Core Cooling) is injected among the safety systems, the ECC is injected after the first channel damage that occurs on the primary side of the coolant system, and the main steam safety valve (MSSV) is fixed and opened ) through which the amount of loss of cooling water is reduced, and melting of the core of the primary side (Loop1) and the secondary side (Loop2) does not occur due to the injected cooling water.

According to the present invention, when the steam generator tube rupture accident progresses to a serious accident, the safety system operation and non-operation are in accordance with the serious accident mitigation guidelines by identifying 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 Wolsong Unit 1 is allowed to continue operation.

1 is an exemplary view showing a heavy water reactor type nuclear power plant for evaluation of the impact of a serious accident of a customs rupture accident of a heavy water steam generator according to an embodiment of the present invention.
2 is an exemplary view showing a loop of a coolant system for evaluation of the impact of a serious accident of a pipe rupture accident of a heavy water steam generator according to an embodiment of the present invention.
Figure 3 is an exemplary view showing a serious accident impact evaluation system of the heavy water steam generator customs rupture accident according to an embodiment of the present invention.
4 is a flowchart illustrating a method for evaluating the impact of a serious accident of a heavy water steam generator tube rupture accident according to an embodiment of the present invention.
5 is an exemplary view showing the event tree of the power plant damage group.
6 is an exemplary view showing the stock amount of coolant on the primary side through the rupture of the steam generator tube until the first pressure tube and calandria tube breakage.
7 is an exemplary view showing the pressure to maintain the pressure until a predetermined time due to FIG.
8 is a graph showing an example in which the pressure tube temperature reaches 900 k the fastest under the high pressure of the primary side (Loop1).
9 is a graph showing an example in which the water level of the fractured steam generator rapidly falls.
10 is an exemplary view illustrating an example in which a reverse flow of coolant occurs from the secondary side to the primary side (Loop 1).
11 is an exemplary view showing a change in the coolant level in the nuclear reactor compartment and the end shield.
12 is an exemplary view showing the pressure behavior of the boiler room, which is the largest compartment of the nuclear reactor building.
13 is an exemplary view showing the total amount of inert gases (Kr and Xe) discharged from the core to the reactor building (and later to the external environment).
14 is an exemplary view showing the total amount of CsI and CsOH released to the external environment.

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 Large calandria vessel surrounds fuel channels 20, 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 the 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.

Figure 3 is a block diagram showing a serious accident impact evaluation system of the heavy water steam generator customs rupture accident according to an embodiment of the present invention. As shown in FIG. 3 , the serious accident impact evaluation system 1 of the heavy water steam generator customs rupture accident includes the safety system setting unit 110 , the coolant loss accident signal generating unit 120 , the isolation processing unit 130 , the main steam It includes a safety valve 140 , a reverse flow occurrence determination unit 150 , a liquid discharge valve 160 , and a control unit 170 .

The safety system setting unit 110 is a configuration for setting the initial condition of the safety system in case of a rupture accident of the customs tube of the steam generator.

The coolant loss accident signal generating unit 120 generates a coolant loss accident signal when the primary pressure of the coolant system is reduced due to the occurrence of a reactor stop due to a tube breakage accident in the steam generator.

The isolation processing unit 130 is configured to isolate the two closed circuits of the coolant system so that the isolation valve is locked around the pressurizer when a coolant loss accident signal occurs.

The main steam safety valve 140 is configured to maintain a constant pressure after the primary side pressure of the coolant system reaches the opening/closing pressure of the main steam safety valve after a predetermined time elapses.

When the pressure tube temperature reaches a critical point under the pressure on the primary side of the coolant system, the reverse flow generation determining unit 150 determines that the reverse flow of the coolant has occurred from the secondary side of the coolant system to the primary side.

The liquid discharge valve 160 maintains a pressure equal to or greater than an open set value of the secondary side liquid discharge valve due to an increase in the secondary side pressure of the coolant system until the horizontal fuel pipe is damaged.

The control unit 170 is a component for controlling the above-mentioned components, and the method for evaluating the impact of a serious accident of a heavy water steam generator customs rupture accident with this control unit is described as follows.

4 is a flowchart illustrating a method for evaluating the impact of a serious accident of a heavy water steam generator tube rupture accident according to an embodiment of the present invention.

First, the control unit sets the initial conditions for the safety system in case of a rupture accident in the customs tube of the steam generator (a). Next, when the primary pressure of the coolant system is reduced due to a reactor shutdown due to a tube breakage accident in the steam generator, a coolant loss accident signal is generated (b). Next, when the coolant loss accident signal occurs, the two closed circuits of the coolant system are isolated so that the isolation valve is locked around the pressurizer (c). Next, after the primary pressure of the coolant system elapses for a certain period of time, the main steam safety valve reaches the opening/closing pressure, and then reaches the stage of maintaining the constant pressure (d). Next, when the pressure tube temperature reaches a critical point under the pressure on the primary side of the coolant system, it is determined that the reverse flow of coolant has occurred from the secondary side of the coolant system to the primary side (e). Next, due to the increase in the secondary pressure of the coolant system, the secondary side liquid release valve maintains the pressure above the open setting value until the horizontal fuel pipe is damaged (f).

Here, when the secondary side water supply of the coolant system is not supplied due to the reactor shutdown in step b), the control unit decreases the heat transfer to the secondary side and increases the pressure on the secondary side (sound closed circuit), Generates a fuel pipe breakage signal that causes horizontal fuel pipe breakage on the secondary side due to temperature rise under high pressure.

In addition, after step a), when the local air cooler (LAC) is operated among the safety systems, the nuclear reactor building damage signal and the nuclear reactor stop occurrence signal are not generated.

In addition, after step a), in the safety system, compared to the failure of the rapid decompression (CC) operation until the primary side (Loop1) is broken by the rapid decompression (CC) operation / until the cooling water of the sound steam generator is exhausted, the primary side (Loop1) horizontal fuel pipe It shows low-pressure accident behavior with delayed damage time. Such low pressure accident behavior will be described in detail below.

Also, after step a), when auxiliary water supply is supplied to the steam generator in the safety system, the stock of cooling water in the primary side (Loop1) is continuously leaked through the broken pipe until the primary side (Loop1) horizontal fuel pipe is damaged. In the secondary side (Loop2), the decay heat is removed by the secondary side heat removal method by supplying auxiliary water supply, so the calandria moderator depletion and damage time are delayed compared to the rapid decompression (CC) operation.

In addition, after step a), when ECC (Emergency Core Cooling) is injected in the safety system, the ECC injection is started when the primary side pressure drops below a certain set value after the reactor is stopped at the secondary side (Loop2), and the primary side pressure is the If the increase exceeds a certain set value, the injection is stopped, and the primary side (Loop1) of the coolant system reaches the saturation temperature later than the secondary side (Loop2), delaying the time to decrease the calandria moderator inventory.

In addition, after step a), when ECC (Emergency Core Cooling) is injected in the safety system, the ECC is injected after the first channel damage that occurs on the primary side of the coolant system and the main steam safety valve (MSSV) is fixed and opened The amount of loss of cooling water is reduced, and the core melt of the primary side (Loop1) and the secondary side (Loop2) does not occur due to the injected cooling water.

Hereinafter, a detailed description of a method for evaluating the impact of a heavy water steam generator customs pipe breakage 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 unique serious accident progression characteristics of 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 pipes, deformation and reaction with moderators, oxidation reaction between zircaloy and water vapor, damage to horizontal fuel pipes, 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 is 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 cooling system stopped in a stationary condition ISAAC was improved to simulate an accident in which coolant is lost due to leakage in the cooling pump bearing of In the present invention, accident analysis was performed using ISAAC version 4.03.

In an 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 result 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 6th initial accident. If the six initial accidents are listed, supply pipe stagnation rupture accident (SFB), power plant blackout accident (SBO), shutdown accident (SSA), steam generator customs rupture accident (SGTR), small coolant loss accident (SLOCA), 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 percentage of failure 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.

- Only the leakage due to the leakage rate is considered for the leakage of the reactor building.

- 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.

- The secondary side pressure of the steam generator is controlled from 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 is inflated and deformed, 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 starts to deform, moves away from its original position, and relocates around the sound 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. It is assumed that all material is 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 shielding 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 drilled.

- 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 which have a large capacity are safe and can be manually operated in the main control room, and 12 out of 16 are simulated as success criteria.

- A total of 27 (10 small, 15 medium, 2 large) passive hydrogen couplers (PARs) 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. In this analysis, when the hydrogen concentration reaches 2%, the operation is delayed for 30 minutes. 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 stationary cooling system under the stationary condition. The following new model has been added to ISAAC to simulate an accident in which coolant is lost due to leakage in the ISAAC.

① Bellows break model

If the pressure pipe is damaged in an accident in which the Primary Heat Transport System (PHTS) is maintained at high pressure, the bellows of the end joint 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 pressure tube breakage and the moderator leaking from Calandria to the reactor building.

- AFAILBELLOW: Bellows breaking 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: reactor building compartment number where moderator leaks through 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.

②Initial model in stationary state

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/outlet capillary level, 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 the unbroken loop pump bowl of the 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): Closed circuit-1 Coolant and gas temperature in 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

First, in the steam generator customs rupture (SGTR) accident scenario selection, four cases (SGTR-4~SGTR-7) with high frequency of steam generator customs rupture (SGTR) accidents are added based on the Wolseong Probabilistic Safety Assessment (PSA) results. did In Wolseong Unit 1, PDSET (Plant Damage State Event Tree) was developed by dividing multi-tube breakage and single-tube breakage into multi-tube breakage and single-tube breakage to simulate steam generator customs breakage accidents. An accident (multiple SGTR) was selected as the initial event. 5 shows the event tree of the power plant damage group. Ten accident scenarios (SGTR-1 to SGTR-10) to be used in the analysis will be described below.

(원자로입구 모관의 0.4% 값 해당)를 가정한다.The method for evaluating the impact of a serious accident of a customs rupture accident of a heavy water steam generator according to an embodiment of the present invention is based on a basic analysis of the safety system operation and non-operation impact evaluation methodology for the progress of a serious accident due to a customs rupture accident of a steam generator as follows. carry out The steam generator tube rupture (SGTR) accident is an accident that starts as the coolant on the primary side escapes to the secondary side in FIG. 2 due to the rupture of the tube. Assume that 10 tube ruptures occur at the tube sheet top of the low-temperature tube side of the Loop 1 (broken closed circuit) steam generator (SG), to obtain the maximum fracture release rate (even in reverse flow). it is for The total fracture area is 2.222 x corresponding to the average breaking flow rate of 80 kg/sec (maximum breaking flow rate of 88 kg/sec) until the time of reactor shutdown

(corresponding to the value of 0.4% of the reactor inlet capillary) is assumed.

Here, in order to determine the impact of a serious accident according to the progress of the accident, the cases in which the functions of each system such as rapid cooling of the steam generator, emergency core cooling injection, local air cooler, and auxiliary water supply to the steam generator are maintained or not are evaluated.

Steam Generator Tube Rupture (SGTR), SGTR-1 (default case)

In this case, in the event of an SGTR accident, simultaneous failure of active safety system and operator action (e.g., secondary side water supply failure of steam generator, secondary side rapid decompression (CC) operation failure, emergency core cooling failure, stop cooling failure, moderator/shield cooling failure, and Local air cooler (LAC) failure) is assumed. Instead, it relies on a passive heat sink (primary side stock, moderator, steam generator (SG) secondary, coolant stock in the reactor compartment) for core residual heat removal. It is also assumed that the reactor shutdown, the mutual isolation between the primary side closed loops, the reactor building sprinkling, and the passive hydrogen recombination machine are operational. It is assumed that the reactor building damage occurs at a pressure of 426 kPa(a), which corresponds to a 50% cumulative probability in the 50% confidence interval of the Wolsong Unit 1 reactor building damage curve. As for the results of major accidents, the basic case and the sensitivity case were partially compared in Table 1.

major events [seconds] PT/CT breakage
L1/L2 SG
Coolant exhaustion
BS1-US1 /BS2-US2 CV coolant exhaustion/
CV damage Shield/RV
Reaching the cooling water bath saturation temperature R/B damage in RV
Start/Stop MCCI SGTR-1 4374
/8767 5200-115000
/4900-4573 29005
/134738 30885
/44253 62570 164557
/395958 . . . . . . . . . . . . . . SGTR-10 10553
/No Fail NA NA NA No Fail NA

The stop time of the reactor was 430 seconds, which is the stop point of 10 customs rupture accidents described in the Wolsong Unit 1 FSAR (Final Safety Analysis report). After the reactor shutdown occurs at 430 s, the primary pressure decreases rapidly. When the primary pressure reaches 5.5 MPa(a), a Loss of Coolant Accident (LOCA) signal occurs at 471 seconds, and the isolation valve that isolates the two closed circuits around the pressurizer is closed at 491 seconds. 6 shows that the stock of coolant on the primary side steadily decreases through the rupture of the steam generator (SG) tube until the first pressure tube and calandria tube break. Due to this, the primary side pressure reaches the opening/closing pressure of the main steam safety valve (MSSV) in about 700 seconds, and maintains this pressure until about 4,000 seconds (see FIG. 7). At about 4,374 seconds, the pressure tube and calandria tube of Loop 1 are broken by the ballooning of the bundle, in which the pressure tube temperature reaches near 900 K (see FIG. 8) the fastest under the high pressure of Loop 1 at about 4,374 seconds. The water level of the broken steam generator (BS1) drops rapidly (refer to FIG. 9), and a reverse flow of cooling water occurs from the secondary side to Loop 1 (refer to FIG. 10).

Since the secondary side water supply is no longer supplied after the reactor is shut down, the secondary side coolant level of the sound steam generator (SG) approaches the bottom (see FIG. 9) and the heat transfer to the secondary side decreases, starting at about 4,000 seconds, Loop 2 (sound closed circuit) the pressure begins to increase. This pressure increase continues until Loop 2 liquid discharge valve opens at 6,623 seconds, and then maintains the liquid discharge valve opening set value (10.34 MPa(a)) or more until horizontal fuel line breakage occurs. This horizontal fuel pipe breakage of Loop 2 is caused by the temperature rise under high pressure due to the exhaustion of coolant in the highest nuclear fuel pipe (channel 1), which is already exposed from the moderator at 8,767 seconds.

When the melting temperature of the cladding material is reached due to the temperature rise, the fuel bundle sags downwards and the fuel rod or fuel pipe material is displaced from its original position and rearranged to the lower part, forming a suspended debris bed (SDB). this happens When these fuel channel disassembly conditions are satisfied, the fragments of the fuel pipe are relocated to holding bins and stay in the storage for a while as a suspended debris bed (SDB). Before the moderator reaches the saturation temperature (about 4,889 seconds) and immediately after the horizontal fuel line breakage occurs, the pressure difference between the calandria and the outside (ie, inside the reactor building) reaches the rupture disc set point (=20 psid). When the rupture disk ruptures, about 60% of the moderator inventory is released into the reactor building, exposing the fuel pipes located about one-third the height from the top. Most of the cooling water inside Calandria is exhausted by about 29,005 seconds before the relocation of the core is complete. The coolant in the reactor compartment then acts as a heat sink and cools the calandria outer wall, but eventually at 44,253 seconds the reactor compartment coolant reaches its saturation temperature and begins to boil.

After the moderator is depleted, the core fragments inside the calandria begin to heat up. When the reactor compartment water level drops below the level of the debris bed, the outer wall of the calandria floor heats up, causing creep failure at approximately 134,738 seconds. In the event of calandria failure, the debris is relocated to the bottom of the reactor compartment and cooled by the coolant still remaining inside the reactor compartment. A strong core melt-steam interaction (corium/steam interaction) inside the reactor compartment immediately after the core melt relocation was predicted by the code at about 135,976 seconds. Eventually, when the coolant remaining inside the reactor compartment is exhausted at 152,256 seconds (refer to FIG. 11), the core melt-concrete reaction (MCCI) starts at 164,557 seconds. The reactor compartment floor breaks at 395,958 seconds when the erosion depth of the concrete floor reaches 2 m. The thickness of the reactor compartment floor is 2.45 m, and in this analysis, it is considered that when the depth reaches 2 m, it can no longer withstand the load of eroded concrete and core melt, resulting in failure of the reactor compartment floor. The amount of hydrogen generated from MCCI was 2,938 kg. After the reactor compartment failure, the core melt reacts with the cooling water in the reactor building basement, and the level and quantity of the cooling water are gradually reduced.

12 shows a representative pressure behavior of a boiler room, which is the largest compartment (about 20,000 m3) of a nuclear reactor building. Initially, the reactor building pressure is atmospheric pressure. After the onset of the accident, the pressure gradually increases because steam is released into the reactor building from the Primary Heat Transport System (PHTS) through the liquid release valve and then from Calandria through the rupture disk. After the coolant in Calandria is exhausted at 29,005 seconds, there is no steaming source until 44,253 seconds, when the reactor compartment coolant reaches the saturation temperature and starts to boil (The reason why the cooling water boiling effect in the shield is not seen in FIG. 9 is that Because the coolant stock is very small compared to the reactor compartment). Outside of Calandria, steam from the reactor compartment enters the reactor building. The incoming steam further increases the reactor building pressure by 62,570 seconds, and when it reaches 426 kPa(a), the reactor building is damaged. Then, when calandria failure occurs at 134,738 seconds, the core melt is relocated to the reactor compartment and peak pressure occurs at this point (135,976 seconds). Thereafter, another peak pressure occurs at the point (401,650 seconds) when the core melt is relocated to the reactor building basement. At the time the core melt is relocated to the reactor building basement, the basement cooling water height and quantity are 2 m and 2,000 tons, respectively.

13 shows the total amount of inert gases (Kr and Xe) released from the core to the reactor building (and later to the external environment), including both stable and radioactive elements. These discharges start after Loop 1 horizontal fuel line breakage time (4,374 seconds), and the main discharge is between the 6,387 seconds (Loop 1) and 39,451 seconds (Loop 2) where the melting of the fuel tube/core starts and ends at 39,451 seconds (Loop 2). It is made from a Suspended Debris Bed (SDB). It shows a sudden release to the external environment at 62,570 seconds when the reactor building was damaged. It is noted that although the SGTR accident sequence is referred to as a bypass sequence, there is no discharge to the secondary side of the SG through the main steam safety valve (MSSV). This is because, in this case, the main steam safety valve (MSSV) does not open after the first breakage of the Loop 1 horizontal fuel pipe. 14 shows the total amount of CsI and CsOH released into the external environment, including RbI and RbOH nuclides. After the destruction of the reactor building, 1.97% and 2.03% of CsI and CsOH were cumulatively released, respectively.

SGTR-2: Since the reactor building failure pressure is low, failure occurs earlier (51,445 seconds) than SGTR-1, and the radiation source term to the external environment increases. At the end of the calculation, the accumulated CsI and CsOH emissions were 2.68% (0.035 kg) and 2.72% (0.92 kg), respectively.

SGTR-3: Because the local air cooler (LAC) operates, there is no damage to the reactor building, and the radiation source to the external environment is insignificant (the cumulative emission of CsI and CSOH until the end of the calculation is 4.80E-3%, respectively) and 4.83E-3%, which is mostly due to the design leakage (0.5%/day) of the nuclear reactor building. Assuming no design leakage, the cumulative emission of CsI and CsOH decreases to less than 1% of the calculated value)

SGTR-4: Because of rapid decompression (CC) operation, Loop 1 shows low-pressure accident behavior until the breakage of Loop 1/depletion of the cooling water of the sound steam generator (in the case of low-pressure accident, the nuclear fuel pipe is either a local melt penetration or a horizontal fuel pipe) In the case of melt penetration, if the melting temperature of zircaloy in a part of the nuclear fuel pipe exceeds the melting temperature (2,500 K), damage is assumed, and in case of deflection, it is assumed that the pressure tube temperature exceeds 2,200 K) . In this case, it does not reach 2,200 K until the damage of Loop 1 horizontal fuel pipe, and after loop 1 steam generator coolant depletion around 6,000 seconds, Loop 1 pressure increases and reaches 9 atmospheres (=90 kPa(a)) around 8,100 seconds. Therefore, under high pressure (9 atmospheres) around 8,105 seconds, the temperature of the pressure tube (the 7th bundle of the 10th channel, which has the highest temperature) reaches about 1,290 K, causing damage to the horizontal fuel line by swelling. Compared to SGTR-1 (CC operation failure), Loop 1 horizontal fuel line damage time is delayed due to the difference in the loop 1 pressure behavior (change from low pressure to high pressure) and the associated fuel line failure mode. On the other hand, unlike SGTR-1, the CV rupture disk does not open immediately after Loop 1 horizontal fuel line damage, but opens immediately after Loop 2 horizontal fuel line damage (9,629 seconds). In this case, more (80 to 250 tons) steam was produced compared to the SGTR-1 through the fixed open main steam safety valve (MSSV) (until the reactor building pressure drops to atmospheric pressure around 160,000 sec after the reactor building damage). Because it is discharged with a lot of decay heat, the time of damage to the reactor building is delayed by about 2.5 times compared to SGTR-1. Unlike SGTR-1, which emits little fission product (FP) through the SG secondary side, most CsI and CsOH are discharged to the environment through a fixed open main steam safety valve (MSSV) on the SG secondary side (of the total emissions). About 98% of CsI and CsOH were already released before the reactor building failure).

SGTR-5: Under high pressure (approximately 10.3 MPa(a)) (No. 8 bundle of channel 1), the pressure tube temperature reaches about 900 K and swell occurs around 9,808 seconds, when Loop 2 horizontal fuel tube damage occurs. Unlike SGTR-4, the loop 1 pressure reaches the maximum pressure of 4 atm around 10,000 sec. Instead, around 15,950 seconds, the horizontal fuel pipe temperature rises to over 2,100 K due to the exhaustion of coolant (the moderator level is already exposed) in Channel 1, the highest at near atmospheric pressure, and the damage to the Loop 1 horizontal fuel pipe is delayed. In this case, a similar amount of steam (about 250 tons) to SGTR-4 is discharged to the secondary side of SG, but the amount of CsI and CsOH deposited on the secondary side is large, so a smaller amount (about 3/4) than in case 4 is reduced to the secondary side (BS1). Except), it is discharged to the environment through the fixed open main steam safety valve (MSSV).

SGTR-6: The accident process is similar to that of SGTR-4, but the amount of CsI and CsOH generated from the two closed-loop cores is released due to the failure of the secondary closed-loop isolation. counted as a double

SGTR-7: Auxiliary water supply is supplied to all steam generators to maintain the coolant stock of the broken steam generator (BS1), but the stock on the primary side of Loop 1 continues to flow out through the ruptured part of the customs until the damage of the horizontal fuel pipe of Loop 1. Therefore, Loop 1 horizontal fuel pipe breakage occurs at 26,406 seconds when the horizontal fuel pipe temperature reaches 2,100 K or higher while the coolant in the highest channel (the moderator material level is already exposed) is depleted at low pressure near atmospheric pressure. Before the loop 1 horizontal fuel pipe breakage, the calandria moderator had already reached the saturation temperature (20,529 sec), and the rupture disk was already open (26,406 sec). On the other hand, Loop 2 is healthy throughout the accident because the decay heat is well removed by the secondary side heat removal method by supplying auxiliary water supply. After all, since the amount of core melt in calandria is about half that of SGTR-4, the depletion and damage time of calandria moderators are delayed more than twice. Continuous Loop 2 secondary side heat removal and fixation Due to the release of steam inside the reactor building through the open main steam safety valve (MSSV), the reactor building does not reach the damage pressure. Until the completion of the calculation (in fact, until about 100,000 seconds after Loop 1 core collapse is completed), about 1.42% of CsI is released through the secondary side of the SG, which corresponds to about 22% of the case of SGTR-4.

SGTR-8: In Loop 2, when the primary pressure drops below 4.14 MPa(a) after the reactor is shut down, high-pressure ECC (Emergency Core Cooling) injection starts (due to exhaustion of the Steam generator in Loop 2 before about 2,000 seconds) Injection is stopped when the pressure increases above 4.14 MPa(a). A little injection occurs immediately after the reactor is stopped, but the pressure of Loop 1 is kept lower than that of Loop 2, so the injection amount of Loop 2 is very small. Therefore, the first Loop 2 horizontal fuel pipe damage is delayed by about 45 minutes compared to SGTR-4 and occurs at about 12,333 seconds. At this time, the calandria rupture disc is also opened. After the channel damage, the ECCS (Emergency Core Cooling System) is injected again, and the injection is stopped in about 200,000 seconds when the recirculation is stopped due to the exhaustion of the coolant in the basement of the reactor building. In Loop 1, the primary side pressure is kept low and HPI/MPI/LPI are sequentially injected (the injected coolant is lost to the outside through the fixed and opened main steam safety valve (MSSV) before the Loop 1 channel is broken) Near 200,000 seconds After most of the coolant at the bottom of the reactor building is vaporized and depleted, the injection amount becomes insignificant. After that, the calandria moderator reaches the saturation temperature and the stock of Loop 2 and the calandria moderator starts to decrease. Since the coolant for Loop 1 and BS1 reaches the saturation temperature later than that of Loop 2, the timing of the stock decrease is delayed. Therefore, after ECCS injection is stopped and exposed to the moderator, the damage to the Loop 1 horizontal fuel pipe and the core starts very late at around 357,510 seconds at low pressure due to the exhaustion of the coolant inside the Loop 1 channel. When core collapse occurs at about 433,919 seconds, core damage is terminated. Since core damage occurs when the heat of collapse is reduced, the damage period is very long. The reactor building damage occurs at 423,873 seconds, when the damage to the core is not over. Before this point, fission products (FP) are released into the external environment through the secondary side of the SG, and after that, most of the emissions occur through the damaged reactor building. . By the end of the calculations, CsI emissions through the damaged reactor building accounted for about 87.5% of the total.

SGTR-9: Compared with SGTR-8, the total amount of loss of coolant through the fixed open main steam safety valve (MSSV) is smaller (water: 2,400? ?60 tons, steam: 100??350 tons) Since most of the injected coolant is inside the reactor building, the coolant at the bottom of the reactor building is not depleted until the calculation is complete. Therefore, core melting does not occur in both Loop 1 and Loop 2, and the radiation source to the external environment is insignificant.

SGTR-10: LP ECC could not be injected due to high pressure on the primary side of Loop 1, causing damage to the Loop 1 horizontal fuel pipe. However, compared to SGTR-1, it takes more time for the coolant exhaustion of the Loop 1 channel because the reverse flow occurs even before the damage of the Loop 1 horizontal fuel line (in this case, the pressure behaviors of Loop 1 and BS1 are almost similar. During the opening/closing period of the steam safety valve (MSSV), the SG secondary steam capillary pressure contracts little while the primary side pressure contracts (this instantaneous pressure differential allows for reverse flow). Eventually, the pressure tube temperature reached about 900 K under a high pressure of about 5.2 MPa(a) in 10,553 seconds, and swelling occurred, causing damage to the Loop 1 horizontal fuel tube. On the other hand, due to normal secondary heat removal due to the supply of auxiliary water supply, the Loop 2 coolant inventory does not change from the shutdown of the reactor until the end of the calculation, and there is no channel damage. Although the calandria moderator reaches the saturation temperature in 11,079 seconds, since LPI (Low Pressure Injection) is injected after the first damage to the horizontal fuel line in Loop 1, no further damage to the Loop 1 channel occurs and the radiation source to the external environment is also insignificant

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 without departing from the technical spirit of the present invention. will be.

110: safety system setting unit
120: coolant loss accident signal generating part
130: containment unit
140: main steam safety valve
150: reflux occurrence determination unit
160: liquid release valve
170: control unit

Claims (7)

In the method for evaluating the impact of heavy water vapor generator tube rupture in a serious accident,
a) setting the initial condition of the safety system in case of a rupture accident of the customs pipe of the steam generator;
b) generating a coolant loss accident signal when the primary pressure of the coolant system is reduced due to the occurrence of a reactor stop due to the tube breakage accident of the steam generator;
c) isolating the two closed circuits of the coolant system so that the isolation valve is locked around the pressurizer when the coolant loss accident signal occurs;
d) maintaining a constant pressure after the primary side pressure of the coolant system reaches the main steam safety valve opening/closing pressure after a predetermined time elapses;
e) determining that a reverse flow of coolant has occurred from the secondary side of the coolant system to the primary side when the pressure tube temperature reaches a critical point under the pressure on the primary side of the coolant system; and
f) allowing the secondary-side liquid discharge valve to maintain a pressure higher than the opening set value until the horizontal fuel pipe is damaged due to an increase in the secondary-side pressure of the coolant system; and
After step a), in the safety system, when auxiliary water supply is supplied to the steam generator, the stock of coolant in the primary side (Loop1) continuously flows out through the ruptured portion of the customs until the primary side (Loop1) horizontal fuel pipe is damaged. The secondary side (Loop2) is a step in which the decay heat is removed by the secondary side heat removal method by supplying the auxiliary water supply, so that the depletion of the calandria moderator and the damage time is delayed compared to the rapid decompression (CC) operation; heavy water vapor containing A method for evaluating the impact of serious accidents in the generator customs rupture accident. The method of claim 1,
When the secondary side water supply of the coolant system is not supplied due to the reactor shutdown in step b), the secondary side (sound closed circuit) pressure increases while heat transfer to the secondary side is reduced, and the temperature under high pressure due to coolant depletion Generating a fuel pipe breakage occurrence signal in which the horizontal fuel pipe on the secondary side is damaged by the rise; According to claim 1,
After step a), among the safety systems, when the local air cooler (LAC) operates, the reactor building damage signal and the nuclear reactor stop occurrence signal are not generated. The effect of a heavy water steam generator customs breakage accident including; Assessment Methods. According to claim 1,
After step a), the primary side (Loop1) horizontal fuel pipe compared to the failure of the rapid decompression (CC) operation until the primary side (Loop1) is broken by the rapid decompression (CC) operation in the safety system after step a) / until the cooling water of the sound steam generator is exhausted A method for evaluating the impact of a serious accident in a heavy water steam generator tube rupture accident, including a step of showing a low-pressure accident behavior in which the damage time is delayed. delete According to claim 1,
After step a), when ECC (Emergency Core Cooling) is injected in the safety system, the ECC injection is started when the primary side pressure falls below a certain set value after stopping the reactor at the secondary side (Loop2), and the primary side pressure When this increases above the predetermined value, the injection is stopped, and the primary side (Loop1) of the coolant system reaches the saturation temperature later than the secondary side (Loop2), so the calandria moderator inventory reduction time is delayed; A method for evaluating the impact of serious accidents in the tube rupture accident of a steam generator. According to claim 1,
After step a), when ECC (Emergency Core Cooling) is injected among the safety systems, the ECC is injected after the first channel damage that occurs on the primary side of the coolant system, and the main steam safety valve (MSSV) is fixed and opened A method for evaluating the impact of heavy water vapor generator tube rupture accidents, including a step in which the loss of cooling water through

Images