CN110991006B - Pressurized water reactor large LOCA accident reactor core damage evaluation method based on exposure time - Google Patents
Pressurized water reactor large LOCA accident reactor core damage evaluation method based on exposure time Download PDFInfo
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- CN110991006B CN110991006B CN201911076654.1A CN201911076654A CN110991006B CN 110991006 B CN110991006 B CN 110991006B CN 201911076654 A CN201911076654 A CN 201911076654A CN 110991006 B CN110991006 B CN 110991006B
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- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 title claims abstract description 42
- 238000011156 evaluation Methods 0.000 title claims abstract description 21
- 239000002826 coolant Substances 0.000 claims abstract description 23
- 238000000034 method Methods 0.000 claims abstract description 17
- 238000002347 injection Methods 0.000 claims description 23
- 239000007924 injection Substances 0.000 claims description 23
- 238000005253 cladding Methods 0.000 claims description 13
- 238000001816 cooling Methods 0.000 claims description 9
- 239000003381 stabilizer Substances 0.000 claims description 6
- ATYZRBBOXUWECY-UHFFFAOYSA-N zirconium;hydrate Chemical compound O.[Zr] ATYZRBBOXUWECY-UHFFFAOYSA-N 0.000 claims description 3
- 230000000694 effects Effects 0.000 description 3
- 238000004088 simulation Methods 0.000 description 2
- 238000004364 calculation method Methods 0.000 description 1
- 230000007547 defect Effects 0.000 description 1
- 230000007123 defense Effects 0.000 description 1
- 238000011161 development Methods 0.000 description 1
- 230000001747 exhibiting effect Effects 0.000 description 1
- 230000000977 initiatory effect Effects 0.000 description 1
- 238000002844 melting Methods 0.000 description 1
- 230000008018 melting Effects 0.000 description 1
- 238000012544 monitoring process Methods 0.000 description 1
- 238000002360 preparation method Methods 0.000 description 1
- 230000009257 reactivity Effects 0.000 description 1
- 238000011160 research Methods 0.000 description 1
Classifications
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
The invention provides a bare time-based pressurized water reactor large LOCA accident core damage evaluation method, which comprises the following steps: (1) The large LOCA accident emergency working condition process is predicted and analyzed by a model, and the accident scene comprises two types: serious accidents caused by SBO of double-end fracture superposition of a cold pipe section of a main pipeline of the loop coolant and serious accidents caused by SBO of double-end fracture superposition of a hot pipe section of the main pipeline of the loop coolant; (2) calculating a time node of the important event; (3) Establishing a relation between the exposure time and the core damage state; (4) And calculating the core bare time aiming at the accident to be evaluated, and evaluating the core damage. The method provided by the invention is based on the process of large LOCA accidents, and the accidents are specifically divided into two situations of serious accidents initiated by SBO of double-end fracture superposition of the cold pipe section of the main pipeline of the loop coolant and serious accidents initiated by SBO of double-end fracture superposition of the hot pipe section of the main pipeline of the loop coolant, and standards are respectively established for evaluation.
Description
Technical Field
The invention belongs to the technical field of emergency evaluation of reactors, and particularly relates to a pressurized water reactor large LOCA accident reactor core damage evaluation method based on exposure time.
Background
With the rapid development of nuclear power industry in China and the superposition of policies based on people, the nuclear safety is more and more important, and the unprecedented height is achieved. Nuclear emergency is used as the last defense line of nuclear safety, nuclear emergency preparation work must be constantly done, and nuclear emergency response capability is enhanced.
The core damage evaluation is important content of nuclear emergency, can be used as a basis for accident release source item estimation, and is also an important information basis for decision making of emergency directors. The large LOCA accident is a typical serious accident scenario of a pressurized water reactor nuclear power plant, and the core damage evaluation is very significant for the accident. As the nuclear power plant has a plurality of working condition data and the working environment of the monitoring instrument is bad when a nuclear accident occurs, the method which is conservative, convenient and quick is adopted to evaluate the damage of the large LOCA accident core, which is a necessary auxiliary decision support means. However, in the prior art, the core damage evaluation based on the bare time provided by foreign literature is not specific to the specific accident scenario.
The process of the large LOCA accident is specifically studied, and a core damage evaluation method based on the exposure time under different accident situations is formed.
Disclosure of Invention
Aiming at the defects existing in the prior art, the invention aims to provide a pressurized water reactor large LOCA accident core damage evaluation method based on the exposure time, which specifically divides accident situations into: (1) Serious accident of SBO initiation caused by double-end fracture superposition of cold pipe section of primary pipeline of primary circuit coolant; (2) And (3) evaluating the serious accidents caused by SBO in the superposition of the double-end fracture of the hot pipe section of the primary pipeline of the primary loop coolant respectively.
In order to achieve the above purpose, the invention adopts the technical scheme that:
a pressurized water reactor large LOCA accident core damage evaluation method based on bare time, the application comprising the steps of:
(1) The large LOCA accident emergency working condition process is predicted and analyzed by a model, and the accident scene comprises two types: serious accidents caused by SBO of double-end fracture superposition of a cold pipe section of a main pipeline of the loop coolant and serious accidents caused by SBO of double-end fracture superposition of a hot pipe section of the main pipeline of the loop coolant;
(2) Calculating a time node of the important event;
(3) Establishing a relation between the exposure time and the core damage state;
(4) And calculating the core bare time aiming at the accident to be evaluated, and evaluating the core damage.
Further, the step (1) specifically includes:
accident scenario assumes: the reactor is in a 100% power stable operation state before an accident; the large break and the power failure of the whole factory occur at the same time at the moment 0; reactor shutdown at 0s; the high-pressure safety injection, the low-pressure safety injection and the main and auxiliary water supply fail, and the passive medium-pressure safety injection can normally operate; the idle time of the main pump is 30s; containment leak rate is 0.1%/d; simulating and calculating to take the failure of the lower seal head as a termination event; the breach occurs in the loop where the voltage regulator is located.
Further, the step (1) specifically includes:
the 900MW typical pressurized water reactor full power operating parameters were used as modeling parameters for the severe accident program.
Further, the step (1) specifically includes:
the built model comprises three loops, each loop is modeled with 1 main pump, 1 steam generator and 1 safety injection box, and the three loops share a voltage stabilizer; the model includes 46 control bodies in total: the reactor core consists of 5 control bodies, each loop pipeline consists of 5 control bodies, each steam generator consists of 5 control bodies, and the containment and the environment are respectively represented by 1 control body; the reactor core is divided into 4 radial rings along the radial direction and 14 axial layers along the axial direction; the reactor core secondary support structure is located on the 1 st axial layer, the reactor core support plate is located on the 2 nd axial layer, the reactor core lower grid plate is located on the 3 rd axial layer, the reactor core active area is located on the 4 th-13 th axial layer, and the reactor core upper grid plate is located on the 14 th axial layer.
Further, the step (1) specifically includes: and judging the progress of the emergency working condition by taking the temperature of the cladding as a basis.
Further, the important events in the step (2) specifically include:
the reactor core is exposed, the voltage stabilizer is emptied, the medium-voltage safety injection is started to fill water, the safety injection box is emptied, the cladding is started to be damaged, the violent zirconium water reaction is started, the reactor core is melted, the reactor core is completely exposed for the first time, the reactor core is collapsed, and a molten pool and a lower seal head are formed to fail.
Further, the step (3) specifically includes: when the accident situation is a serious accident that the double ends of the cold pipe section of the main pipeline of the loop coolant are broken and superposed with SBO to start,
if the exposure time is 0 to 9.8min, the reactor core is not damaged;
if the exposure time is 9.8min to 14.3min, the cladding starts to be damaged;
if the exposure time is 14.3min to 31.8min, the reactor core starts to melt;
if the exposure time is more than 31.8min, the reactor core begins to collapse, and a molten pool is formed.
Further, the step (3) specifically further includes: when the accident scenario is a serious accident of the double-end fracture superposition SBO of the hot pipe section of the main pipeline of the loop coolant,
if the exposure time is 0 to 0.5min, the reactor core is not damaged;
if the exposure time is 0.5 to 8.1 minutes, the cladding starts to be damaged;
if the exposure time is 8.1min to 27.7min, the reactor core starts to melt;
if the bare time is more than 27.7min, the reactor core begins to collapse, and a molten pool is formed.
Further, the step (4) specifically includes:
4.1, acquiring the water level of the pressure vessel, the temperature of a core thermocouple, the temperature of a loop system, the pressure of the loop system, the negative cooling margin of the loop system and the water injection rate of the core of the accident to be evaluated;
4.2, estimating the moment of the bareness of the top of the reactor core, namely, assuming the moment of occurrence of a large LOCA accident as the moment of the bareness of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
4.4, calculating the bare time of the reactor core;
4.5 based on different accident scenarios, evaluating according to the core bare time and the relation between the bare time and the core damage state.
Further, the time at which the core is cooled in step 4.3 is the time at which (1), (2) and (3) are simultaneously satisfied, or (1), (2) and (4) are simultaneously satisfied:
(1) The water level is positioned at the top of the active section;
(2) CET is mostly less than 300 ℃;
(3) The primary circuit pressure and most core outlet thermocouples exhibit a positive cooling margin;
(4) The rate of water injected into the pressure vessel is greater than 3 times the rate of water injection required to the post-shutdown core due to decay heat.
The method provided by the invention has the effects that based on the process of the large LOCA accident, the accident is specifically divided into two situations of serious accident initiated by SBO of double-end fracture superposition of the cold pipe section of the main pipeline of the loop coolant and serious accident initiated by SBO of double-end fracture superposition of the hot pipe section of the main pipeline of the loop coolant, and standards are respectively established for evaluation.
Drawings
FIG. 1 is a schematic flow chart of the method of the invention.
Detailed Description
In order to make the technical problems solved, the technical scheme adopted and the technical effects achieved by the invention more clear, the technical scheme of the embodiment of the invention will be further described in detail with reference to the accompanying drawings. It will be apparent that the described embodiments are only some, but not all, embodiments of the invention. All other embodiments, which can be made by those skilled in the art based on the embodiments of the present invention without making any inventive effort, are intended to fall within the scope of the present invention.
Referring to fig. 1, fig. 1 is a schematic flow chart of the method of the present invention. The invention provides a bare time-based pressurized water reactor large LOCA accident core damage evaluation method which comprises the following steps:
(1) The large LOCA accident emergency working condition process is predicted and analyzed by a model, and the accident scene comprises two types: a severe accident that the double-end fracture of the cold pipe section of the main pipeline of the loop coolant is superposed with SBO and a severe accident that the double-end fracture of the hot pipe section of the main pipeline of the loop coolant is superposed with SBO are originated.
The patent adopts a severe accident analysis program to conduct prediction analysis on emergency working conditions of large LOCA accidents. The method specifically comprises the following steps:
accident scenario assumes: the reactor is in a 100% power stable operation state before an accident; the large break and the power failure of the whole factory occur at the same time at the moment 0; since the pressure of a loop in 0.1s after the occurrence of the large break can be reduced to the saturation pressure of the coolant, a large amount of steam is instantaneously generated, and the reactor is shut down due to the negative reactivity introduced by the cavitation effect, the reactor is assumed to be shut down in 0s; the high-pressure safety injection, the low-pressure safety injection and the main and auxiliary water supply fail, and the passive medium-pressure safety injection can normally operate; the idle time of the main pump is 30s; containment leak rate is 0.1%/d; the simulation calculation uses the failure of the lower seal head as a termination event, and the response of the containment is not in the research scope of the patent; the breach occurs in the loop where the voltage regulator is located.
The built model comprises three loops, each loop is modeled with 1 main pump, 1 steam generator and 1 safety injection box, and the three loops share a voltage stabilizer; the model includes 46 control bodies in total: the reactor core consists of 5 control bodies, each loop pipeline consists of 5 control bodies, each steam generator consists of 5 control bodies, and the containment and the environment are respectively represented by 1 control body; the reactor core is divided into 4 radial rings along the radial direction and 14 axial layers along the axial direction; the reactor core secondary support structure is located on the 1 st axial layer, the reactor core support plate is located on the 2 nd axial layer, the reactor core lower grid plate is located on the 3 rd axial layer, the reactor core active area is located on the 4 th to 13 th axial layers, and the reactor core upper grid plate is located on the 14 th axial layer.
The patent uses 900MW typical pressurized water reactor full power operation parameters as modeling parameters of a serious accident program, and main parameters are shown in table 1.
Table 1 main parameter table for modeling
And simulating two accident conditions according to the model, the parameters and the set conditions. And judging the emergency working condition progress based on the simulation result and the temperature of the cladding. Specific judgment data are shown in Table 2.
TABLE 2 judgment basis for Emergency working condition progress
(2) And calculating the time node of the important event.
The important events specifically include: the reactor core is exposed, the voltage stabilizer is emptied, the medium-voltage safety injection is started to fill water, the safety injection box is emptied, the cladding is started to be damaged, the violent zirconium water reaction is started, the reactor core is melted, the reactor core is completely exposed for the first time, the reactor core is collapsed, and a molten pool and a lower seal head are formed to fail. Specific important events and their time nodes are shown in Table 3.
Table 3 emergency condition contrast for severe accident with double-end broken cold and hot pipe section
(3) And establishing a relation between the bare time and the core damage state.
When the accident situation is a serious accident that the double ends of the cold pipe section of the main pipeline of the loop coolant are broken and SBO is initiated, if the exposure time is 0 to 9.8min, the reactor core is not damaged; if the exposure time is 9.8min to 14.3min, the cladding starts to be damaged; if the exposure time is 14.3min to 31.8min, the reactor core starts to melt; if the exposure time is more than 31.8min, the reactor core begins to collapse, and a molten pool is formed.
When the accident situation is a serious accident that the double ends of the hot pipe section of the primary loop coolant pipeline are broken and SBO is initiated, if the exposure time is 0 to 0.5min, the reactor core is not damaged; if the exposure time is 0.5 to 8.1 minutes, the cladding starts to be damaged; if the exposure time is 8.1min to 27.7min, the reactor core starts to melt; if the bare time is more than 27.7min, the reactor core begins to collapse, and a molten pool is formed.
The relationship between the exposure time of a large LOCA accident and the core damage status is shown in table 4.
TABLE 4 relationship between large LOCA bare time and core damage for pressurized water reactor
The core begins to collapse, forming a molten pool state at which time it is considered that 100% of the core is molten. For the accident of large LOCA of pressurized water reactor, the core damage evaluation based on the bare time is carried out, and the core damage state can be divided into core damage-free, cladding damage and core melting.
(4) And calculating the core bare time aiming at the accident to be evaluated, and evaluating the core damage.
4.1, acquiring the water level of the pressure vessel, the temperature of a core thermocouple, the temperature of a loop system, the pressure of the loop system, the negative cooling margin of the loop system and the water injection rate of the core of the accident to be evaluated;
4.2, estimating the moment of the bareness of the top of the reactor core, namely, assuming the moment of occurrence of a large LOCA accident as the moment of the bareness of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
in a particular embodiment, the time at which the core is cooled is at a time that satisfies both the water level at the top of the active section, the CET being mostly less than 300 ℃, and the loop pressure and most of the core outlet thermocouples exhibiting positive cooling margin.
In another specific embodiment, the core is cooled at a time that satisfies both the water level at the top of the active section, the CET being mostly less than 300℃, and the water rate injected into the pressure vessel being greater than 3 times the rate of the required make-up water due to decay heat for the post-shutdown core.
4.4, calculating the bare time of the reactor core.
The core bare time is estimated using equation (1).
T UC ＝t Cooling -t UC (1)
In which T is UC Core bare length of time; t is t Cooling Time when the core is cooled; t is t uc Time of core top bare.
4.5 based on different accident scenarios, evaluating according to the core bare time and the relation between the bare time and the core damage state.
Compared with the prior art, the method for evaluating the damage of the large LOCA accident core of the pressurized water reactor based on the exposure time is characterized in that based on the progress of the large LOCA accident, the accident is specifically divided into two situations of serious accident caused by double-end fracture superposition SBO of a cold pipe section of a main pipeline of a loop coolant and serious accident caused by double-end fracture superposition SBO of a hot pipe section of the main pipeline of the loop coolant, and standards are respectively established for evaluation.
It will be appreciated by persons skilled in the art that the methods of the present invention are not limited to the examples described in the detailed description, which are provided for the purpose of illustrating the invention and are not intended to limit the invention. Other embodiments will occur to those skilled in the art from a consideration of the specification and practice of the invention as claimed and as claimed in the claims and their equivalents.
Claims (7)
1. The method for evaluating the damage of the large LOCA accident core of the pressurized water reactor based on the exposure time is characterized by comprising the following steps of:
(1) The large LOCA accident emergency working condition process is predicted and analyzed by a model, and the accident scene comprises two types:
serious accidents caused by SBO of double-end fracture superposition of a cold pipe section of a main pipeline of the loop coolant and serious accidents caused by SBO of double-end fracture superposition of a hot pipe section of the main pipeline of the loop coolant;
the built model comprises three loops, each loop is modeled with 1 main pump, 1 steam generator and 1 safety injection box, and the three loops share a voltage stabilizer; the model includes 46 control bodies in total: the reactor core consists of 5 control bodies, each loop pipeline consists of 5 control bodies, each steam generator consists of 5 control bodies, and the containment and the environment are respectively represented by 1 control body; the reactor core is divided into 4 radial rings along the radial direction and 14 axial layers along the axial direction; the reactor core secondary support structure is positioned on the 1 st axial layer, the reactor core support plate is positioned on the 2 nd axial layer, the reactor core lower grid plate is positioned on the 3 rd axial layer, the reactor core active area is positioned on the 4 th-13 th axial layer, and the reactor core upper grid plate is positioned on the 14 th axial layer;
(2) Calculating a time node of the important event;
(3) Establishing a relation between the exposure time and the core damage state;
(4) Calculating the core bare time aiming at the accident to be evaluated, and evaluating the core damage;
establishing a relationship between bare time and core damage status includes:
when the accident situation is a serious accident that the double ends of the cold pipe section of the main pipeline of the loop coolant are broken and superposed with SBO to start,
if the exposure time is 0 to 9.8min, the reactor core is not damaged;
if the exposure time is 9.8min to 14.3min, the cladding starts to be damaged;
if the exposure time is 14.3min to 31.8min, the reactor core starts to melt;
if the exposure time is more than 31.8min, the reactor core begins to collapse, and a molten pool is formed;
when the accident scenario is a serious accident of the double-end fracture superposition SBO of the hot pipe section of the main pipeline of the loop coolant,
if the exposure time is 0 to 0.5min, the reactor core is not damaged;
if the exposure time is 0.5 to 8.1 minutes, the cladding starts to be damaged;
if the exposure time is 8.1min to 27.7min, the reactor core starts to melt;
if the bare time is more than 27.7min, the reactor core begins to collapse, and a molten pool is formed.
2. The bare time-based pressurized water reactor large LOCA accident core damage evaluation method as set forth in claim 1, wherein the step (1) specifically includes:
accident scenario assumes: the reactor is in a 100% power stable operation state before an accident; the large break and the power failure of the whole factory occur at the same time at the moment 0; reactor shutdown at 0s; the high-pressure safety injection, the low-pressure safety injection and the main and auxiliary water supply fail, and the passive medium-pressure safety injection can normally operate; the idle time of the main pump is 30s; containment leak rate is 0.1%/d; simulating and calculating to take the failure of the lower seal head as a termination event; the breach occurs in the loop where the voltage regulator is located.
3. The bare time-based pressurized water reactor large LOCA accident core damage evaluation method as set forth in claim 1, wherein the step (1) specifically includes:
the 900MW typical pressurized water reactor full power operating parameters were used as modeling parameters for the severe accident program.
4. The bare time-based pressurized water reactor large LOCA accident core damage evaluation method as set forth in claim 1, wherein the step (1) specifically includes: and judging the progress of the emergency working condition by taking the temperature of the cladding as a basis.
5. The bare time-based pressurized water reactor large LOCA accident core damage evaluation method as set forth in claim 1, wherein the important events in the step (2) specifically include:
the reactor core is exposed, the voltage stabilizer is emptied, the medium-voltage safety injection is started to fill water, the safety injection box is emptied, the cladding is started to be damaged, the violent zirconium water reaction is started, the reactor core is melted, the reactor core is completely exposed for the first time, the reactor core is collapsed, and a molten pool and a lower seal head are formed to fail.
6. The bare time-based pressurized water reactor large LOCA accident core damage evaluation method as set forth in claim 1, wherein the step (4) specifically includes:
4.1, acquiring the water level of the pressure vessel, the temperature of a core thermocouple, the temperature of a loop system, the pressure of the loop system, the negative cooling margin of the loop system and the water injection rate of the core of the accident to be evaluated;
4.2, estimating the moment of the bareness of the top of the reactor core, namely, assuming the moment of occurrence of a large LOCA accident as the moment of the bareness of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
4.4, calculating the bare time of the reactor core;
4.5 based on different accident scenarios, evaluating according to the core bare time and the relation between the bare time and the core damage state.
7. The bare time based large LOCA accident core damage assessment method as claimed in claim 6, wherein the core cooling time in step 4.3 is the time at which (1), (2) and (3) are simultaneously satisfied or the time at which (1), (2) and (4) are simultaneously satisfied:
(1) The water level is positioned at the top of the active section;
(2) CET is mostly less than 300 ℃;
(3) The primary circuit pressure and most core outlet thermocouples exhibit a positive cooling margin;
(4) The rate of water injected into the pressure vessel is greater than 3 times the rate of water injection required to the post-shutdown core due to decay heat.
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| 先进压水堆大破口始发严重事故下安全壳内氢气风险分析;温丽丽 等;《核科学与工程》;第37卷(第04期);第604-612页 * |
| 压力容器水位参数在堆芯损伤评价方法中的应用;魏严凇 等;《原子能科学技术》;第48卷;第385-388页 * |
| 大破口始发应急工况预测方法研究;王宁 等;《辐射防护》;第40卷(第6期);第671-676页 * |
| 应用CDAG方法进行秦山二期大破口LOCA严重事故堆芯损伤研究;魏玮 等;《核科学工程》;第28卷(第4期);第334-340页 * |
| 核事故应急堆芯损伤快速评价系统顶层设计方案研究;刘圆圆 等;《核动力工程》;第37卷(第03期);第142-145页 * |
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