CN110991006A - Core damage evaluation method for large LOCA accident of pressurized water reactor based on exposure time - Google Patents
Core damage evaluation method for large LOCA accident of pressurized water reactor based on exposure time Download PDFInfo
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- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 title claims abstract description 45
- 238000011156 evaluation Methods 0.000 title claims abstract description 24
- 239000002826 coolant Substances 0.000 claims abstract description 23
- 238000000034 method Methods 0.000 claims abstract description 23
- 230000008569 process Effects 0.000 claims abstract description 9
- 238000002347 injection Methods 0.000 claims description 27
- 239000007924 injection Substances 0.000 claims description 27
- 238000005253 cladding Methods 0.000 claims description 13
- 238000001816 cooling Methods 0.000 claims description 9
- 239000003381 stabilizer Substances 0.000 claims description 6
- 238000002844 melting Methods 0.000 claims description 4
- 230000008018 melting Effects 0.000 claims description 4
- 230000000153 supplemental effect Effects 0.000 claims description 3
- ATYZRBBOXUWECY-UHFFFAOYSA-N zirconium;hydrate Chemical compound O.[Zr] ATYZRBBOXUWECY-UHFFFAOYSA-N 0.000 claims description 3
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 2
- 239000001301 oxygen Substances 0.000 claims description 2
- 229910052760 oxygen Inorganic materials 0.000 claims description 2
- 239000003507 refrigerant Substances 0.000 claims description 2
- 230000000694 effects Effects 0.000 description 2
- 230000004044 response Effects 0.000 description 2
- 239000000243 solution Substances 0.000 description 2
- 238000005336 cracking 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
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 238000012544 monitoring process Methods 0.000 description 1
- 238000002360 preparation method Methods 0.000 description 1
- 230000002265 prevention Effects 0.000 description 1
- 230000009257 reactivity Effects 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 238000004088 simulation Methods 0.000 description 1
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
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- 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 core damage evaluation method for a large pressurized water reactor LOCA accident based on bare time, which comprises the following steps: (1) a model is established to predict and analyze the emergency working condition process of the large LOCA accident, and the accident scene comprises two types: a serious accident initiated by SBO superposition due to double-end fracture of a cold pipe section of a primary loop coolant main pipe and a serious accident initiated by SBO superposition due to double-end fracture of a heat pipe section of the primary loop coolant main pipe; (2) calculating time nodes of the important events; (3) establishing a relation between the exposure time and the damage state of the reactor core; (4) and calculating the core exposure 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 a large LOCA accident, specifically divides the accident into two situations of a serious accident initiated by the superposition of the fracture of the two ends of the cold pipe section of the main loop coolant pipeline and the SBO initiated serious accident initiated by the fracture of the two ends of the hot pipe section of the main loop coolant pipeline, and respectively establishes standards for evaluation.
Description
Technical Field
The invention belongs to the technical field of emergency evaluation of reactors, and particularly relates to a method for evaluating damage of a large LOCA accident reactor core of a pressurized water reactor based on bare time.
Background
With the rapid development of the nuclear energy cause of China and the superposition of people-oriented policies, the nuclear safety is more and more emphasized, and the unprecedented height is reached. The nuclear emergency is used as the last line of defense of nuclear safety, and nuclear emergency preparation work must be done constantly to strengthen the nuclear emergency response capability.
The core damage evaluation is an important content of nuclear emergency, can be used as a basis for estimating accident release source items, and is also an important information basis for making decisions by emergency commanders. The large LOCA accident is a typical serious accident situation of a pressurized water reactor nuclear power plant, and the core damage evaluation of the accident is very significant. Because the working condition data of the nuclear power plant are various and the working environment of the monitoring instrument equipment is severe during the emergency of the nuclear accident, the method for evaluating the damage of the reactor core of the large LOCA accident by using a conservative, convenient and quick method is a necessary auxiliary decision support means. However, in the prior art, the core damage evaluation based on the bare time provided by foreign documents does not aim at specific accident situations.
The process of a large LOCA accident is specifically researched, and a core damage evaluation method based on the exposure time under different accident situations is formed.
Disclosure of Invention
Aiming at the defects in the prior art, the invention aims to provide a method for evaluating the damage of the large LOCA accident reactor core of a pressurized water reactor based on the exposure time, which specifically comprises the following accident scenes: (1) overlapping the serious accident of SBO starting by the fracture of the two ends of the cold pipe section of the main loop coolant pipeline; (2) and (4) overlapping the serious accidents initiated by SBO by the fracture of the two ends of the heat pipe section of the main loop coolant pipeline, and respectively evaluating.
In order to achieve the above purposes, the invention adopts the technical scheme that:
a core damage evaluation method of a large pressurized water reactor LOCA accident based on bare time comprises the following steps:
(1) a model is established to predict and analyze the emergency working condition process of the large LOCA accident, and the accident scene comprises two types: a serious accident initiated by SBO superposition due to double-end fracture of a cold pipe section of a primary loop coolant main pipe and a serious accident initiated by SBO superposition due to double-end fracture of a heat pipe section of the primary loop coolant main pipe;
(2) calculating time nodes of the important events;
(3) establishing a relation between the exposure time and the damage state of the reactor core;
(4) and calculating the core exposure time aiming at the accident to be evaluated, and evaluating the core damage.
Further, the step (1) specifically comprises:
the accident scenario is assumed: the reactor is in a 100% power stable operation state before an accident; the large crevasses and the whole plant power failure occur at the same time at 0 moment; stopping the reactor at 0 s; high-pressure safety injection, low-pressure safety injection and failure of main and auxiliary water supply, and the passive medium-pressure safety injection can normally run; the idling time of the main pump is 30 s; the containment leak rate is 0.1%/d; simulating and calculating the failure of the lower end socket as a termination event; the breach occurs in the loop where the voltage regulator is located.
Further, the step (1) specifically comprises:
the full power operating parameters of a typical pressurized water reactor of 900MW are used as modeling parameters of a serious accident program.
Further, the step (1) specifically comprises:
the established 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 one voltage stabilizer; the model comprises 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 represented by 1 control body respectively; 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 region 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 comprises: and judging the emergency working condition process by taking the cladding temperature as a basis.
Further, the important events in step (2) specifically include:
the method comprises the following steps of exposing a reactor core, emptying a voltage stabilizer, filling water at the beginning of medium-pressure safety injection, emptying a safety injection tank, damaging a cladding, beginning a violent zirconium-water reaction, melting the reactor core, completely exposing the reactor core for the first time, collapsing the reactor core, and failing to form a molten pool and a lower end socket.
Further, the step (3) specifically comprises: when the accident situation is a serious accident initiated by superposition of SBO (cold pipe segment double-end fracture) of a main primary coolant pipeline,
if the exposure time is 0-9.8 min, the reactor core is not damaged;
if the exposure time is 9.8min to 14.3min, the cladding begins to be damaged;
if the exposure time is 14.3min to 31.8min, the reactor core begins to melt;
if the exposure time is more than 31.8min, the reactor core begins to collapse to form a molten pool.
Further, the step (3) specifically includes: when the accident situation is a serious accident originated by superposition of SBO (boundary layer of refrigerant and oxygen) of double-end fracture of a heat pipe section of a primary circuit coolant main pipeline,
if the exposure time is 0-0.5 min, the reactor core is not damaged;
if the exposure time is 0.5min to 8.1min, the cladding begins to be damaged;
if the exposure time is 8.1min to 27.7min, the reactor core begins to melt;
if the exposure time is more than 27.7min, the reactor core begins to collapse to form a molten pool.
Further, the step (4) specifically comprises:
4.1, acquiring the water level of the pressure vessel, the temperature of a reactor core thermocouple, the temperature of a primary circuit system, the pressure of the primary circuit system, the negative cooling margin of the primary circuit system and the water injection rate of the reactor core of the accident to be evaluated;
4.2 estimating the time of the exposure of the top of the reactor core, namely assuming the time of the occurrence of the large LOCA accident as the time of the exposure of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
4.4 calculating the exposure time of the reactor core;
and 4.5 evaluating according to the bare time of the reactor core and the relationship between the bare time and the damage state of the reactor core based on different accident scenes.
Further, in step 4.3, the time when the core is cooled is a time when (1), (2) and (3) are satisfied simultaneously, or when (1), (2) and (4) are satisfied simultaneously:
(1) the water level is positioned at the top of the active section;
(2) CET is mostly less than 300 ℃;
(3) primary loop pressure and most core outlet thermocouples showed positive cooling margins;
(4) the water injection rate into the pressure vessel is 3 times greater than the required supplemental water injection rate due to decay heat in the reactor core after shutdown.
The method has the advantages that based on the progress of the large LOCA accident, the accident is specifically divided into two situations, namely a serious accident initiated by the superposition of the double-end fracture of the cold pipe section of the main loop coolant pipeline and a serious accident initiated by the superposition of the double-end fracture of the heat pipe section of the main loop coolant pipeline, and the standards are respectively established for evaluation.
Drawings
FIG. 1 is a schematic flow chart of the method of the present invention.
Detailed Description
In order to make the technical problems solved, the technical solutions adopted, and the technical effects achieved by the present invention clearer, the technical solutions of the embodiments of the present invention will be further described in detail with reference to the accompanying drawings. It is to be understood that the described embodiments are merely exemplary of the invention, and not restrictive of the full scope of the invention. All other embodiments, which can be obtained by a person skilled in the art without any inventive step based on the embodiments of the present invention, are 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 core damage evaluation method of a large LOCA accident of a pressurized water reactor based on bare time, which comprises the following steps:
(1) a model is established to predict and analyze the emergency working condition process of the large LOCA accident, and the accident scene comprises two types: the severe accident originated by SBO superposition of the fracture of the two ends of the cold pipe section of the primary circuit coolant main pipe and the severe accident originated by SBO superposition of the fracture of the two ends of the heat pipe section of the primary circuit coolant main pipe.
The patent uses a serious accident analysis program to carry out prediction and analysis on emergency working conditions of a large LOCA accident. The method specifically comprises the following steps:
the accident scenario is assumed: the reactor is in a 100% power stable operation state before an accident; the large crevasses and the whole plant power failure occur at the same time at 0 moment; because the pressure of a loop can be reduced to the saturation pressure of the coolant within 0.1s after the large break occurs, a large amount of steam is generated instantly, and the reactor can be shut down due to the negative reactivity introduced by the cavitation effect, so that the reactor is shut down when the time of 0s is assumed; high-pressure safety injection, low-pressure safety injection and failure of main and auxiliary water supply, and the passive medium-pressure safety injection can normally run; the idling time of the main pump is 30 s; the containment leak rate is 0.1%/d; simulating and calculating the failure of the lower end socket as a termination event, wherein the response of the containment is out of the research range of the patent; the breach occurs in the loop where the voltage regulator is located.
The established 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 one voltage stabilizer; the model comprises 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 represented by 1 control body respectively; 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 region 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.
The full-power operation parameters of a 900MW typical pressurized water reactor are used as modeling parameters of a serious accident program, and main parameters are shown in a table 1.
TABLE 1 Primary 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 process by taking the cladding temperature as a basis based on the simulation result. The specific judgment data are shown in Table 2.
TABLE 2 judgment basis for Emergency operating Condition progress
(2) And calculating the time node of the important event.
The important events specifically include: the method comprises the following steps of exposing a reactor core, emptying a voltage stabilizer, filling water at the beginning of medium-pressure safety injection, emptying a safety injection tank, damaging a cladding, beginning a violent zirconium-water reaction, melting the reactor core, completely exposing the reactor core for the first time, collapsing the reactor core, and failing to form a molten pool and a lower end socket. Specific events of interest and their time nodes are shown in table 3.
TABLE 3 Severe accident emergency condition comparison of cold and hot pipe section double-end fracture
(3) And establishing a relation between the exposure time and the damage state of the reactor core.
When the accident situation is a serious accident initiated by superposition of SBO (cracking prevention of both ends of a cold pipe section of a primary circuit coolant main pipeline, if the exposure time is 0-9.8 min, the reactor core is not damaged; if the exposure time is 9.8min to 14.3min, the cladding begins to be damaged; if the exposure time is 14.3min to 31.8min, the reactor core begins to melt; if the exposure time is more than 31.8min, the reactor core begins to collapse to form a molten pool.
When the accident situation is a serious accident initiated by superposition of double-end fracture of a heat pipe section of a primary circuit coolant main pipeline and SBO, if the exposure time is 0-0.5 min, the reactor core is not damaged; if the exposure time is 0.5min to 8.1min, the cladding begins to be damaged; if the exposure time is 8.1min to 27.7min, the reactor core begins to melt; if the exposure time is more than 27.7min, the reactor core begins to collapse to form a molten pool.
The relationship between the exposure time to a large LOCA accident and the core damage status is shown in Table 4.
TABLE 4 relationship between large LOCA exposure time of pressurized water reactor and core damage
The core begins to collapse, creating a molten pool state at which the core is considered to be 100% molten. For a large LOCA accident of a pressurized water reactor, core damage evaluation based on bare time is carried out, and the core damage state can be divided into no core damage, cladding damage and core melting.
(4) And calculating the core exposure 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 reactor core thermocouple, the temperature of a primary circuit system, the pressure of the primary circuit system, the negative cooling margin of the primary circuit system and the water injection rate of the reactor core of the accident to be evaluated;
4.2 estimating the time of the exposure of the top of the reactor core, namely assuming the time of the occurrence of the large LOCA accident as the time of the exposure of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
in a specific embodiment, the core is cooled at a time that satisfies both a water level at the top of the active section, a CET of mostly less than 300 ℃, and a primary pressure and a majority of the core outlet thermocouples exhibiting a positive cooling margin.
In another specific embodiment, the core is cooled at a time that simultaneously satisfies the water level at the top of the active section, the CET is mostly less than 300 ℃, and the water injection rate into the pressure vessel is greater than 3 times the required supplemental water injection rate of the reactor core due to decay heat after shutdown.
4.4 calculate the core exposure time.
The core exposure time is estimated using equation (1).
TUC＝tCooling down-tUC(1)
In the formula, TUCThe length of the exposed time of the reactor core; t is tCooling downThe moment when the reactor core is cooled; t is tucThe time when the top of the reactor core is exposed.
And 4.5 evaluating according to the bare time of the reactor core and the relationship between the bare time and the damage state of the reactor core based on different accident scenes.
The method is characterized in that based on the progress of the large LOCA accident, the accident is specifically divided into two situations, namely a severe accident initiated by overlapping SBO due to double-end fracture of a cold pipe section of a primary coolant main pipe and a severe accident initiated by overlapping SBO due to double-end fracture of a heat pipe section of the primary coolant main pipe, and standards are respectively established for evaluation.
It will be appreciated by persons skilled in the art that the method of the present invention is not limited to the examples described in the specific embodiments, and that the above detailed description is for the purpose of illustrating the invention only and is not intended to limit the invention. Other embodiments will be apparent to those skilled in the art from the following detailed description, which is intended to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the appended claims.
Claims (10)
1. A core damage evaluation method for a large pressurized water reactor LOCA accident based on bare time is characterized by comprising the following steps:
(1) a model is established to predict and analyze the emergency working condition process of the large LOCA accident, and the accident scene comprises two types: a serious accident initiated by SBO superposition due to double-end fracture of a cold pipe section of a primary loop coolant main pipe and a serious accident initiated by SBO superposition due to double-end fracture of a heat pipe section of the primary loop coolant main pipe;
(2) calculating time nodes of the important events;
(3) establishing a relation between the exposure time and the damage state of the reactor core;
(4) and calculating the core exposure time aiming at the accident to be evaluated, and evaluating the core damage.
2. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (1) comprises:
the accident scenario is assumed: the reactor is in a 100% power stable operation state before an accident; the large crevasses and the whole plant power failure occur at the same time at 0 moment; stopping the reactor at 0 s; high-pressure safety injection, low-pressure safety injection and failure of main and auxiliary water supply, and the passive medium-pressure safety injection can normally run; the idling time of the main pump is 30 s; the containment leak rate is 0.1%/d; simulating and calculating the failure of the lower end socket as a termination event; the breach occurs in the loop where the voltage regulator is located.
3. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (1) comprises:
the full power operating parameters of a typical pressurized water reactor of 900MW are used as modeling parameters of a serious accident program.
4. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (1) comprises:
the established 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 one voltage stabilizer; the model comprises 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 represented by 1 control body respectively; 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 region 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.
5. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (1) comprises: and judging the emergency working condition process by taking the cladding temperature as a basis.
6. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the important events in the step (2) specifically include:
the method comprises the following steps of exposing a reactor core, emptying a voltage stabilizer, filling water at the beginning of medium-pressure safety injection, emptying a safety injection tank, damaging a cladding, beginning a violent zirconium-water reaction, melting the reactor core, completely exposing the reactor core for the first time, collapsing the reactor core, and failing to form a molten pool and a lower end socket.
7. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (3) comprises: when the accident situation is a serious accident initiated by superposition of SBO (cold pipe segment double-end fracture) of a main primary coolant pipeline,
if the exposure time is 0-9.8 min, the reactor core is not damaged;
if the exposure time is 9.8min to 14.3min, the cladding begins to be damaged;
if the exposure time is 14.3min to 31.8min, the reactor core begins to melt;
if the exposure time is more than 31.8min, the reactor core begins to collapse to form a molten pool.
8. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 1, wherein the step (3) further comprises: when the accident situation is a serious accident originated by superposition of SBO (boundary layer of refrigerant and oxygen) of double-end fracture of a heat pipe section of a primary circuit coolant main pipeline,
if the exposure time is 0-0.5 min, the reactor core is not damaged;
if the exposure time is 0.5min to 8.1min, the cladding begins to be damaged;
if the exposure time is 8.1min to 27.7min, the reactor core begins to melt;
if the exposure time is more than 27.7min, the reactor core begins to collapse to form a molten pool.
9. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the exposure time as claimed in claim 4, wherein the step (4) comprises:
4.1, acquiring the water level of the pressure vessel, the temperature of a reactor core thermocouple, the temperature of a primary circuit system, the pressure of the primary circuit system, the negative cooling margin of the primary circuit system and the water injection rate of the reactor core of the accident to be evaluated;
4.2 estimating the time of the exposure of the top of the reactor core, namely assuming the time of the occurrence of the large LOCA accident as the time of the exposure of the top of the reactor core;
4.3 estimating the time when the reactor core is cooled;
4.4 calculating the exposure time of the reactor core;
and 4.5 evaluating according to the bare time of the reactor core and the relationship between the bare time and the damage state of the reactor core based on different accident scenes.
10. The core damage evaluation method of the large pressurized water reactor LOCA accident based on the bare time as claimed in claim 9, wherein the core cooling in step 4.3 is performed at the time when (1), (2) and (3) are simultaneously satisfied, or when (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) primary loop pressure and most core outlet thermocouples showed positive cooling margins;
(4) the water injection rate into the pressure vessel is 3 times greater than the required supplemental water injection rate due to decay heat in the reactor core after shutdown.
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| CN110970142A (en) * | 2019-11-21 | 2020-04-07 | 中国辐射防护研究院 | Method for predicting emergency working condition of initiation of large-break water loss accident of pressurized water reactor |
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