CN105006259A - Core damage assessment method for nuclear power plant - Google Patents
Core damage assessment method for nuclear power plant Download PDFInfo
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- CN105006259A CN105006259A CN201510340796.XA CN201510340796A CN105006259A CN 105006259 A CN105006259 A CN 105006259A CN 201510340796 A CN201510340796 A CN 201510340796A CN 105006259 A CN105006259 A CN 105006259A
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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
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Abstract
The invention belongs to the technical field of nuclear power plant design, and particularly relates to a core damage assessment method for a nuclear power plant. The method includes the step of conducting core damage assessment through the parameters such as the core outlet thermocouple temperature (CET), the containment internal radiation level (CRM), the primary circuit heat pipe section temperature (RTD), the pressure vessel water level (RVL) and the source range detector reading (SRM), and further includes the step of conducting core damage assessment through containment internal hydrogen concentration (CH) and primary circuit coolant sampling concentration (SAM). According to the core damage assessment method for the nuclear power plant, the design characteristic of the latest nuclear power plant is fully considered, various monitoring parameters capable of reflecting the core state are adopted, hence, the core damage state and the degree are assessed through diversified means, and quick and visual support is provided for emergency aid decision making of the nuclear power plant.
Description
Technical field
The invention belongs to nuclear power plant design technology field, be specifically related to a kind of nuclear power plant core damage assessment method.
Background technology
Nuclear power plant's Core damage state is the important evidence of emergent accident process and alleviation and emergency protective action decision-making, and the exploitation of core damage assessment program is significant to nuclear power plant's aid decision making of meeting an urgent need.China's Nuclear safety guide rule " Emergency Preparedness of nuclear power plant operating organization and emergency response " (HAD002/01-2010) specifies, in crash programme, should have respective description to the method for core damage assessment and pattern." nuclear safety and radioactive contamination are prevented and treated " 12 " and planned and the year two thousand twenty long-range objectives " requires the Real-Time Evaluation expert system of Core damage situation under nuclear power generating sets accident conditions is built up at " 12 " end.Therefore, developing Core damage state evaluation system is also the important process that satisfied national relevant laws and regulations require.
NRC propose core damage assessment require after, Westinghouse Electric has carried out a large amount of core damage assessment analytical works, and after proposition accident in 1984 core damage assessment method (CDAM).The method evaluates state and the degree of Core damage mainly through post-accident sampling system (PASS).Due to the time delay of sampling, the method can not reflect the state of reactor core in real time, therefore can not effectively support emergency response decision-making.
1999, Westinghouse Electric developed new core damage assessment directive/guide (CDAG) on CDAM basis.First this directive/guide utilizes core exit electric thermo-couple temperature (CET) and containment radiation dose rate (CRM) two primary measured parameter to evaluate Core damage state and damage share, and recycling subsidiary parameter confirms the rationality of Core damage state.It is diagnosed Core damage state according to reactor stationary installation measurement parameter and evaluates Core damage degree, can provide core damage assessment comparatively promptly and accurately.
In the RTM-96 report of U.S.'s core pipe meeting, suggestion uses Core uncovering duration, radioactive level in containment, coolant activity and density of hydrogen to evaluate Core damage degree.Wherein, deficient cold degree allowance and pressure vessel water level may be used for the initial time determining Core uncovering.
France IRSN proposes the 3D/3P method for evaluating PWR nuclear power plant state when accident occurs.It, by nuclear power plant three road barrier circulation diagnosis and prediction, obtains power plant's faulted condition and release source item.In the method, core damage assessment is to the diagnosis of first barrier (i.e. fuel can) integrality, parameter selected by it is the radioactive level of core exit electric thermo-couple temperature, containment radiation dose rate and chimney effluent, wherein, the radioactive level of chimney effluent has certain time-delay.
In addition, give also a kind of core damage assessment method in the report TECDOC-955 of International Atomic Energy Agency (IAEA).This report thinks Core damage degree and Core uncovering time correlation, indicates, change that the remarkable increase of radiation level and PWR owe cold degree allowance can calculate the Core uncovering time according to the water level of reactor core active region.In addition, the reading of containment monitor and cooling medium sampling isotopes concentration also can be used as the foundation of core damage assessment.This evaluation method is quick, simple, but precision is not high.
At present, the 3D/3P method of Primary Reference France of domestic nuclear power plant and the CDAG method exploitation core damage assessment system of the U.S..In the article " nuclear power plant's core damage assessment research and software development " that inventor delivers early stage, give a kind of core damage assessment system, Core damage state and damage share can be evaluated according to part monitoring parameter.But this system does not consider up-to-date power plant design feature, other important parameters can monitoring reactor core state can not be made full use of; In addition, this system is developed for specific power plant only, does not have versatility.
Summary of the invention
The object of the present invention is to provide a kind of nuclear power plant core damage assessment method, it considers up-to-date design of nuclear power plant feature, and making full use of can the monitoring parameter of reactor core state, thus adopts diversified means to evaluate Core damage state and degree.
Technical scheme of the present invention is as follows: a kind of nuclear power plant core damage assessment method, at least comprises and utilizes concentration of hydrogen in containment vessel to crossing fire damage share K
cHevaluation, evaluation model is as follows:
If initial amount of oxygen is in containment
and early stage in accident, in containment, hydrogen is reacted by zirconium water and produces, and the oxygen in containment only reduces because of hydrogen and oxygen reaction;
The hydrogen efficiency m that disappears of t passive hydrogen recombiner
h (t)with amounts of hydrogen M in containment
h (t)with amount of oxygen M
o (t)relevant, i.e. m
h (t)=f (M
h (t), M
o (t)); Wherein M
h (t)measured by hydrogen monitor table, according to the hydrogen efficiency that disappears of hydrogen recombiner, calculate the hydrogen quality Δ M of t ~ t+ Δ t time internal consumption
h (t), calculate the oxygen quality Δ M of this Δ t time internal consumption according to oxyhydrogen reaction relation simultaneously
o (t),
ΔM
H(t)＝m
H(t)×Δt
ΔM
O(t)＝f(ΔM
H(t))
The then output M of t+ Δ t hydrogen reality
h(t+ Δ t) is:
M
H(t+Δt)＝M
H(t)+∑ΔM
H(t)
The residual mass M of t+ Δ t oxygen
o(t+ Δ t) is:
Above-mentioned formula forms closed circulation, adopts numerical analysis method to try to achieve actual hydrogen output, thinks and then calculate Core damage degree and actual hydrogen output positive correlation reactor core further and cross fire damage share K in model
cH:
Wherein, j=0,1, represent primary Ioops low pressure when j is 0, when j is 1, represent primary Ioops high pressure;
K=0,1, represent when k is 0 not to primary Ioops water filling, represent to primary Ioops water filling when k is 1.
Further, nuclear power plant as above core damage assessment method, also comprises and utilizes primary Ioops cooling medium to sample concentration SAM to Core damage share K
sAMievaluation, evaluation model is as follows:
When primary Ioops cooling medium sampling concentration SAM exceedes the corresponding setting valve of state i, just think that reactor core may be in the faulted condition corresponding to state i, described state i comprises: state 1--fuel can damages, and state 2--fuel superheater damages;
If Core damage degree and the positive correlation of primary Ioops cooling medium concentration, then Core damage share K
sAMifor:
Further, nuclear power plant as above core damage assessment method, also comprises and utilizes radiation level in core exit electric thermo-couple temperature, containment, primary Ioops heat pipe section temperature, pressure vessel water level, source range detector registration to the evaluation of Core damage.
Beneficial effect of the present invention is as follows:
1) the present invention adopts diversified monitoring parameter to diagnose rapidly Core damage state and degree, calculates accident source term, assessment emergency rating, for nuclear power plant's aid decision making of meeting an urgent need provides and supports fast, intuitively;
2) contemplated by the invention up-to-date power plant design feature, can be used in the nuclear power plant being designed with hydrogen monitoring system and hydrogen recombiner;
3) the present invention can, in the unavailable situation of Real-Time Monitoring parameter, utilize accident sample analysis result to evaluate Core damage state;
4) software systems designed according to evaluation method provided by the present invention can realize being connected with the facility of nuclear power plant other data platforms, and realize data information sharing in nuclear power plant's network platform; By modular design, different PWR nuclear power plants can be applied the present invention to.
Accompanying drawing explanation
Fig. 1 is Core damage method for evaluating state process flow diagram;
Fig. 2 is the frame diagram of core damage assessment software systems.
Embodiment
Below in conjunction with drawings and Examples, the present invention is described in detail.
For PWR nuclear power plant, can reflect that the parameter of reactor core faulted condition is mainly divided into two classes according to the priority of acquisition time, a Lei Shi nuclear power plant on-line monitoring parameter, can real time reaction reactor core state; Another kind of is sampling parameters after accident, can provide result after a period of time occurs accident.Although sampling parameters has certain time delay after accident, once there is the disabled situation of online monitoring parameter, after accident, sampling parameters can also as the means of Core damage state evaluation.
The reaction of zirconium water is the important phenomenon in major accident process, and zirconium water reaction hydrogen output becomes positive correlation with Core damage degree.For the nuclear power plant being provided with hydrogen monitor table and hydrogen recombiner, the density of hydrogen in containment also can be used as the important parameter of diagnosis reactor core state and estimation Core damage share.
In the core exit electric thermo-couple temperature (CET) that the present invention mentions in " nuclear power plant's core damage assessment research and software development " except publishing an article, containment except radiation level (CRM), primary Ioops heat pipe section temperature (RTD), pressure vessel water level (RVL), source range detector registration (SRM), the parameter finally for core damage assessment also comprises concentration of hydrogen in containment vessel (CH) and primary Ioops cooling medium sampling concentration (SAM).
Core damage state demarcation, mainly for the early stage emergent aid decision making of accident, is three classes by the present invention: damage (state 2) without Core damage (state 0), fuel can damage (state 1) and fuel superheater.Parameter CET and CRM is mainly used in judging reactor core state, and calculates damage share further when Core damage; Parameter RTD, RVL, SRM and CH are for confirming the rationality of Core damage condition diagnosing result.For SGTR accident conditions, parameter CRM and CH is not suitable for reactor core condition diagnosing, will give up this two kinds of parameters in evaluation procedure.
Can see " nuclear power plant's core damage assessment research and software development " published content for the concrete grammar utilizing parameter CET, CRM, RTD, RVL, SRM evaluate Core damage, the present invention establishes the model utilizing CH and SAM to carry out core damage assessment.
(1) utilize concentration of hydrogen in containment vessel CH to crossing fire damage share K
cHevaluation model as follows:
When reactor core started fire damage, it is temperature required that fuel can reached the reaction of zirconium water, therefore can utilize the further evaluated fire damage share K of concentration of hydrogen in containment vessel
cH.But if Installation in Plant passive hydrogen recombiner, recombiner then can eliminate hydrogen partial in containment, the therefore amounts of hydrogen that records of density of hydrogen monitor be not equal to the actual hydrogen output of zirconium water reaction.
Suppose that in containment, initial amount of oxygen is
and early stage in accident, in containment, hydrogen is reacted by zirconium water and produces, and the oxygen in containment only reduces because of hydrogen and oxygen reaction;
The hydrogen efficiency m that disappears of t passive hydrogen recombiner
h (t)with amounts of hydrogen M in containment
h (t)with amount of oxygen M
o (t)relevant, i.e. m
h (t)=f (M
h (t), M
o (t)), this is known technology; Wherein M
h (t)measured by hydrogen monitor table, according to the hydrogen efficiency that disappears of hydrogen recombiner, calculate the hydrogen quality Δ M of t ~ t+ Δ t time internal consumption
h (t), calculate the oxygen quality Δ M of this Δ t time internal consumption according to known oxyhydrogen reaction relation simultaneously
o (t),
ΔM
H(t)＝m
H(t)×Δt
ΔM
O(t)＝f(ΔM
H(t))
The then output M of t+ Δ t hydrogen reality
h(t+ Δ t) is:
M
H(t+Δt)＝M
H(t)+∑ΔM
H(t)
The residual mass M of t+ Δ t oxygen
o(t+ Δ t) is:
Above-mentioned formula forms closed circulation, and adopt numerical analysis method to try to achieve actual hydrogen output, this is general mathematical method, thinks and then calculate Core damage degree and actual hydrogen output positive correlation reactor core further and cross fire damage share K in model
cH:
Wherein, j=0,1, represent primary Ioops low pressure when j is 0, when j is 1, represent primary Ioops high pressure;
K=0,1, represent when k is 0 not to primary Ioops water filling, represent to primary Ioops water filling when k is 1.
CH
j,kbelong to power plant's setting valve category, need calculate respectively according to different power plant.Circular can be sketched and be: CH
j,kbe that 100% zirconium water reacts the hydrogen total amount be discharged in containment, consider the impact of primary Ioops water filling and primary Ioops delay effect simultaneously, belong to specific power plant setting valve, need calculate according to specific power plant.Concrete computing method are known for a person skilled in the art.
(2) primary Ioops cooling medium is utilized to sample concentration SAM to Core damage share K
sAMievaluation model as follows:
When being obtained primary Ioops cooling medium concentration by sample analysis, core damage assessment can be carried out as a reference further.When primary Ioops cooling medium sampling concentration SAM exceedes the corresponding setting valve of state i, just think that reactor core may be in the faulted condition corresponding to state i, described state i comprises: state 1--fuel can damages, and state 2--fuel superheater damages;
If Core damage degree and the positive correlation of primary Ioops cooling medium concentration, then Core damage share K
sAMifor:
subscript i in this formula is corresponding with Core damage state.
The present invention is based on the core damage assessment software systems that said method sets up is B/S framework, and system input and output are as the interface with managerial personnel, Database Systems and other emergent Decision Platforms.The user of native system is undertaken managing by network entry, data input, computational analysis, result are checked.This system can by monitoring parameter, reactor core state, result of calculation, prompting suggestion etc. by network implementation time, dynamically present to user.The framework of software systems as shown in Figure 2.
System adopts modular design concept, with core damage assessment module for core, is integrated with source item computing module and emergency rating evaluation module simultaneously.According to different input parameters, core damage assessment module is divided into again different core damage assessment submodules, to be applicable to different PWR nuclear power plants.That is, according to the design feature of specific power plant, by configuration input parameter kind and number, different core damage assessment submodules can being called, by configuring the initialization correlation parameter of specific power plant, being applicable to this power plant to make result of calculation.
Obviously, those skilled in the art can carry out various change and modification to the present invention and not depart from the spirit and scope of the present invention.Like this, if belong within the scope of the claims in the present invention and equivalent technology thereof to these amendments of the present invention and modification, then the present invention is also intended to comprise these change and modification.
Claims (3)
1. nuclear power plant's core damage assessment method, is characterized in that: at least comprise and utilize concentration of hydrogen in containment vessel to crossing fire damage share K
cHevaluation, evaluation model is as follows:
If initial amount of oxygen is in containment
and early stage in accident, in containment, hydrogen is reacted by zirconium water and produces, and the oxygen in containment only reduces because of hydrogen and oxygen reaction;
The hydrogen efficiency m that disappears of t passive hydrogen recombiner
h (t)with amounts of hydrogen M in containment
h (t)with amount of oxygen M
o (t)relevant, i.e. m
h (t)=f (M
h (t), M
o (t)); Wherein M
h (t)measured by hydrogen monitor table, according to the hydrogen efficiency that disappears of hydrogen recombiner, calculate the hydrogen quality Δ M of t ~ t+ Δ t time internal consumption
h (t), calculate the oxygen quality Δ M of this Δ t time internal consumption according to oxyhydrogen reaction relation simultaneously
o (t),
ΔM
H(t)＝m
H(t)×Δt
ΔM
O(t)＝f(ΔM
H(t))
The then output M of t+ Δ t hydrogen reality
h(t+ Δ t) is:
M
H(t+Δt)＝M
H(t)+∑ΔM
H(t)
The residual mass M of t+ Δ t oxygen
o(t+ Δ t) is:
Above-mentioned formula forms closed circulation, adopts numerical analysis method to try to achieve actual hydrogen output, thinks and then calculate Core damage degree and actual hydrogen output positive correlation reactor core further and cross fire damage share K in model
cH:
Wherein, j=0,1, represent primary Ioops low pressure when j is 0, when j is 1, represent primary Ioops high pressure;
K=0,1, represent when k is 0 not to primary Ioops water filling, represent to primary Ioops water filling when k is 1.
2. nuclear power plant as claimed in claim 1 core damage assessment method, is characterized in that: also comprise and utilize primary Ioops cooling medium to sample concentration SAM to Core damage share K
sAMievaluation, evaluation model is as follows:
When primary Ioops cooling medium sampling concentration SAM exceedes the corresponding setting valve of state i, just think that reactor core may be in the faulted condition corresponding to state i, described state i comprises: state 1--fuel can damages, and state 2--fuel superheater damages;
If Core damage degree and the positive correlation of primary Ioops cooling medium concentration, then Core damage share K
sAMifor:
3. nuclear power plant as claimed in claim 1 or 2 core damage assessment method, is characterized in that: also comprise and utilize radiation level in core exit electric thermo-couple temperature, containment, primary Ioops heat pipe section temperature, pressure vessel water level, source range detector registration to the evaluation of Core damage.
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| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN106680001A (en) * | 2016-10-25 | 2017-05-17 | 核动力运行研究所 | Pipeline three-way area on-line fatigue life monitoring device and method |
| CN107239606A (en) * | 2017-05-27 | 2017-10-10 | 国网福建省电力有限公司 | A kind of Sensitivity Analysis Method for presurized water reactor system dynamic model parameter evaluation |
| CN107808230A (en) * | 2017-09-28 | 2018-03-16 | 上海核工程研究设计院有限公司 | A kind of nuclear power plant's core damage assessment method |
| CN107844909A (en) * | 2017-11-17 | 2018-03-27 | 苏州热工研究院有限公司 | The rolling nargin risk control method and system of the daily production risk management of nuclear power plant |
| CN110991006A (en) * | 2019-11-06 | 2020-04-10 | 中国辐射防护研究院 | Core damage evaluation method for large LOCA accident of pressurized water reactor based on exposure time |
| CN111028965A (en) * | 2019-11-06 | 2020-04-17 | 中国辐射防护研究院 | Fuel element overheating damage evaluation method based on containment hydrogen concentration |
| CN112632873A (en) * | 2020-12-30 | 2021-04-09 | 西安交通大学 | Method for determining position arrangement mode of measuring points of temperature thermocouple of nuclear reactor hot leg |
| CN112735613A (en) * | 2019-10-28 | 2021-04-30 | 中核核电运行管理有限公司 | Tracking and monitoring method for integrity of domestic nuclear fuel assembly |
| CN112785064A (en) * | 2021-01-26 | 2021-05-11 | 中国人民解放军海军工程大学 | Nuclear accident handling strategy generation method based on accident state deduction optimization |
| CN113409971A (en) * | 2021-05-28 | 2021-09-17 | 中国原子能科学研究院 | Method, device, medium, and electronic apparatus for monitoring damage to core of nuclear reactor |
| CN114638464A (en) * | 2022-01-13 | 2022-06-17 | 中核武汉核电运行技术股份有限公司 | Intelligent evaluation method for reactor core burnout condition under nuclear accident |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104538070A (en) * | 2014-12-08 | 2015-04-22 | 中广核工程有限公司 | Method, device and system for measuring concentration of hydrogen in nuclear power station containment vessel |
-
2015
- 2015-06-18 CN CN201510340796.XA patent/CN105006259B/en active Active
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104538070A (en) * | 2014-12-08 | 2015-04-22 | 中广核工程有限公司 | Method, device and system for measuring concentration of hydrogen in nuclear power station containment vessel |
Non-Patent Citations (6)
| Title |
|---|
| 刘静: "核电厂事故后安全壳内氢气浓度测量方法分析", 《核安全》 * |
| 李文静 等: "核电厂堆芯损伤评价研究及软件开发", 《核科学与工程》 * |
| 王宏庆 等: "核电厂非能动氢复合器研制", 《核动力工程》 * |
| 邓坚 等: "氢气催化复合器对核电厂严重事故的缓解效果", 《原子能科学技术》 * |
| 魏玮 等: "中国百万千瓦级核电站严重事故下堆芯损伤评价", 《原子能科学技术》 * |
| 魏玮 等: "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", 《核科学与工程》 * |
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| CN106680001B (en) * | 2016-10-25 | 2019-09-17 | 核动力运行研究所 | A kind of online fatigue life monitor in branch pipe tee connection region and method |
| CN107239606B (en) * | 2017-05-27 | 2020-11-03 | 国网福建省电力有限公司 | Sensitivity analysis method for parameter evaluation of dynamic model of pressurized water reactor system |
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| CN112735613A (en) * | 2019-10-28 | 2021-04-30 | 中核核电运行管理有限公司 | Tracking and monitoring method for integrity of domestic nuclear fuel assembly |
| CN111028965A (en) * | 2019-11-06 | 2020-04-17 | 中国辐射防护研究院 | Fuel element overheating damage evaluation method based on containment hydrogen concentration |
| CN110991006A (en) * | 2019-11-06 | 2020-04-10 | 中国辐射防护研究院 | Core damage evaluation method for large LOCA accident of pressurized water reactor based on exposure time |
| CN111028965B (en) * | 2019-11-06 | 2022-05-20 | 中国辐射防护研究院 | Fuel element overheating damage evaluation method based on containment hydrogen concentration |
| CN110991006B (en) * | 2019-11-06 | 2024-01-23 | 中国辐射防护研究院 | Pressurized water reactor large LOCA accident reactor core damage evaluation method based on exposure time |
| CN112632873A (en) * | 2020-12-30 | 2021-04-09 | 西安交通大学 | Method for determining position arrangement mode of measuring points of temperature thermocouple of nuclear reactor hot leg |
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| CN112785064A (en) * | 2021-01-26 | 2021-05-11 | 中国人民解放军海军工程大学 | Nuclear accident handling strategy generation method based on accident state deduction optimization |
| CN113409971A (en) * | 2021-05-28 | 2021-09-17 | 中国原子能科学研究院 | Method, device, medium, and electronic apparatus for monitoring damage to core of nuclear reactor |
| CN113409971B (en) * | 2021-05-28 | 2024-03-22 | 中国原子能科学研究院 | Nuclear reactor core damage monitoring method, device, medium and electronic equipment |
| CN114638464A (en) * | 2022-01-13 | 2022-06-17 | 中核武汉核电运行技术股份有限公司 | Intelligent evaluation method for reactor core burnout condition under nuclear accident |
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