Design method and device for secondary loop heat extraction system of nuclear power unit

Abstract

The invention discloses a design method and device for a secondary loop heat removal system of a nuclear power unit, and the method comprises the steps: building a secondary loop heat removal system scheme of the nuclear power unit according to the measures of the secondary loop heat removal system; constructing a nuclear power heat extraction system model according to the secondary circuit heat extraction system scheme of the nuclear power unit; analyzing a secondary loop heat extraction system scheme of the nuclear power unit according to the nuclear power heat extraction system model to obtain risk frequency; if the risk frequency is greater than a preset risk value, updating a secondary loop heat removal system scheme of the nuclear power unit; and analyzing the updated secondary loop heat extraction system scheme of the nuclear power unit according to the nuclear power heat extraction system model to obtain the updated risk frequency.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear power plant safety and operation, and in particular to a design method and device for a secondary circuit heat removal system of a nuclear power unit. Background Art

[0002] The secondary circuit heat removal system of a large pressurized water reactor nuclear power unit is responsible for the important heat-carrying function of normal operation and the key safety function of removing residual heat from the core during accidents. Its design scheme directly determines the reliability of the system itself, thus affecting the overall operating performance and safety performance of the power plant.

[0003] The secondary circuit heat removal system usually includes: a large-flow water supply and heat removal system during power operation; a small-flow water supply and heat removal system during low power and hot shutdown; a system for exporting core residual heat under design basis accidents; and a system for exporting core residual heat under design extended operating conditions.

[0004] The four types of systems are independent or combined, active or passive solutions are selected, different power sources are selected for active solutions, etc., which form the first-level combination of alternative design solutions. Based on the first-level design solutions, the commissioning / exit logic of each system, the selection of support systems, injection channel solutions (main pipe or independent), redundancy, diversity, including the selection of key equipment, etc. need to be refined to form the second-level combination of solutions.

[0005] At present, in the field of nuclear power design, there is relatively sufficient design experience for single secondary circuit heat removal systems, and there are clear design principles for safety-level secondary circuit heat removal systems, but there is no standard approach to the overall solution of secondary circuit heat removal systems. The present invention is based on probability theory and system reliability evaluation technology, and provides a secondary circuit heat removal system design solution based on a forward system failure rate distribution method.

[0006] The existing patent CN111128414B discloses a safety system and method for a nuclear power plant that combines active and passive safety injection systems. The method includes a safety injection system composed of an active safety injection system and a passive safety injection system, a reactor cavity water injection system composed of an active reactor cavity water injection system and a passive reactor cavity water injection system, a containment heat removal system composed of an active containment spray system and a passive containment heat removal system, and a steam generator secondary side heat removal system composed of an auxiliary water supply system, a turbine bypass exhaust system, and a secondary side passive residual heat removal system. The method focuses on the passive safety system as a supplement to the active safety system, uses passive technology, and comprehensively considers the depth defense level and needs of the nuclear power plant. It does not solve the problem of combining the system failure rate formulated by the NRC with the latest technology development to design the overall solution of the secondary circuit heat removal system.

[0007] The existing patent CN214370112U discloses a nuclear power plant secondary circuit thermal system, which includes a heat exchange pipe connected between the condenser and the evaporator, a low-pressure heat recovery system for heating the condensate in the heat exchange pipe, a high-pressure heat recovery system, and a deaerator. The secondary circuit thermal system also includes a heating system, which includes a heat generating device and a heat collecting heat exchange device for heating the condensate in the heat exchange pipe with the heat generated by the heat generating device. This method focuses on heating the nuclear power plant feed water by using solar energy, and reducing the use of steam extraction by the nuclear power plant secondary circuit thermal system without the help of external energy. It does not solve the problem of the overall solution of the secondary circuit heat removal system designed by the current NRC in combination with the latest technology development based on the system failure rate.

[0008] In summary, neither of the above two existing patents solves the problem of combining the system failure rate currently formulated by the NRC with the latest technological developments to design an overall solution for the secondary circuit heat removal system. Summary of the invention

[0009] Based on the above technical problems, the present invention proposes a design method for a secondary circuit heat removal system of a nuclear power unit to solve the problem of combining the system failure rate currently formulated by NRC with the latest technological development to design an overall solution for the secondary circuit heat removal system.

[0010] To achieve the above object, the present invention proposes a design method for a secondary circuit heat removal system of a nuclear power unit.

[0011] A design method for a secondary circuit heat removal system of a nuclear power unit, comprising:

[0012] According to the secondary heat removal system measures, establish the secondary heat removal system plan for nuclear power units;

[0013] According to the secondary circuit heat removal system scheme of the nuclear power unit, a nuclear power heat removal system model is constructed;

[0014] Analyze the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the risk frequency;

[0015] If the risk frequency is greater than a preset risk value, updating the secondary circuit heat removal system solution of the nuclear power unit;

[0016] The updated secondary circuit heat removal system scheme of the nuclear power unit is analyzed according to the nuclear power heat removal system model to obtain the updated risk frequency.

[0017] Furthermore, according to the secondary circuit heat removal system measures, a secondary circuit heat removal system scheme for the nuclear power unit is established, including:

[0018] The secondary circuit heat removal system measures include first-layer measures and second-layer measures, wherein the second-layer measures include the type and / or capacity of equipment, valve model selection, battery model selection, and / or generator power supply plan formulation in the first-layer measures;

[0019] According to the first-layer measures and the second-layer measures, a secondary circuit heat removal system scheme for the nuclear power unit is formed.

[0020] Furthermore, the secondary circuit heat removal system measures include first-layer measures and second-layer measures, and also include:

[0021] The first-tier measures include a large-flow water supply and heat removal system, a small-flow water supply and heat removal system, a design basis accident heat removal system, and / or a design extended operating condition heat removal system.

[0022] Furthermore, according to the secondary circuit heat removal system scheme of the nuclear power unit, a nuclear power heat removal system model is constructed, including:

[0023] The nuclear power heat removal system model includes a CDF allocation model and / or a PSA model;

[0024] The reliability data and accident condition data of the equipment are collected, and the CDF allocation model and the PSA model are constructed according to the reliability data and the accident condition data of the equipment.

[0025] Furthermore, constructing the CDF allocation model and the PSA model further includes:

[0026] The CDF allocation model includes formula 1 and formula 2.

[0027] The formula 1 is

[0028] CDF IE(i) =F IE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……],

[0029] The formula 2 is

[0030] Among them, CDF IE(i) is the core damage frequency of the ith event; Q SYSA(i) , Q SYSB(i) , Q SYSC(i) F is the failure probability of the secondary circuit heat removal system measures described in the initiating event; IE(i) represents the occurrence frequency of the initiating event; CDF is the sum of the core damage frequencies of all events, i.e. the risk frequency;

[0031] The failure probabilities are respectively assigned to the secondary circuit heat removal system measures.

[0032] Furthermore, the secondary circuit heat removal system scheme of the nuclear power unit is calculated according to the nuclear power heat removal system model to obtain the risk frequency, including:

[0033] The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

[0034] Furthermore, if the risk frequency is greater than a preset risk value, the secondary circuit heat removal system scheme of the nuclear power unit is updated, including:

[0035] The secondary circuit heat removal system measures are adjusted to a redundant manner, and the secondary circuit heat removal system plan of the nuclear power unit is updated.

[0036] Further, the updated secondary circuit heat removal system scheme of the nuclear power unit is analyzed according to the nuclear power heat removal system model to obtain the updated risk frequency, including:

[0037] According to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model.

[0038] Furthermore, the PSA model in the nuclear power heat removal system model is analyzed to obtain a system failure rate, including:

[0039] The system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

[0040] To achieve the above-mentioned purpose, the present invention also proposes a design device for a secondary circuit heat removal system of a nuclear power unit.

[0041] A design device for a secondary circuit heat removal system of a nuclear power unit, comprising:

[0042] The first establishment module is used to establish a secondary circuit heat removal system plan for a nuclear power unit according to the secondary circuit heat removal system measures;

[0043] The second building module is used to build a nuclear power heat removal system model according to the secondary circuit heat removal system scheme of the nuclear power unit;

[0044] A first analysis module is used to analyze the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain a risk frequency;

[0045] A scheme updating module, configured to update the secondary circuit heat removal system scheme of the nuclear power unit if the risk frequency is greater than a preset risk value;

[0046] The second analysis module is used to analyze the updated secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the updated risk frequency.

[0047] Further, in the first establishment module:

[0048] The secondary circuit heat removal system measures include first-layer measures and second-layer measures, wherein the second-layer measures include the type and / or capacity of equipment, valve model selection, battery model selection, and / or generator power supply plan formulation in the first-layer measures;

[0049] Based on the first-layer measures and the second-layer measures, a secondary circuit heat removal system scheme for the nuclear power unit is established.

[0050] Further, in the first establishment module:

[0051] The first-tier measures include a large-flow water supply and heat removal system, a small-flow water supply and heat removal system, a design basis accident heat removal system, and / or a design extended operating condition heat removal system.

[0052] Further, in the second establishment module:

[0053] The nuclear power heat removal system model includes a CDF allocation model and / or a PSA model;

[0054] The reliability data and accident condition data of the equipment are collected, and the CDF allocation model and the PSA model are constructed according to the reliability data and the accident condition data of the equipment.

[0055] Further, in the second establishment module:

[0056] The CDF allocation model includes formula 1 and formula 2.

[0057] The formula 1 is

[0058] CDF IE(i) =F IE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……],

[0059] The formula 2 is

[0060] Wherein, CDFIE(i) is the core damage frequency of the i-th event; QSYSA(i), QSYSB(i), QSYSC(i) are the failure probabilities of the secondary heat removal system measures of the initiating event; FIE(i) represents the occurrence frequency of the initiating event; CDF is the sum of the core damage frequencies of all events, i.e., the risk frequency;

[0061] The failure probabilities are respectively assigned to the secondary circuit heat removal system measures.

[0062] Furthermore, in the first analysis module:

[0063] The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

[0064] Furthermore, in the scheme updating module:

[0065] The secondary circuit heat removal system measures are adjusted to a redundant manner, and the secondary circuit heat removal system plan of the nuclear power unit is updated.

[0066] Furthermore, in the second analysis module:

[0067] According to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model.

[0068] Furthermore, in the second analysis module:

[0069] The system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

[0070] Based on the above technical solution, the present invention has at least the following beneficial effects:

[0071] 1. The present invention is based on probability theory and system reliability evaluation technology, and on the principles of design simplification and safety benefit maximization. It provides a two-circuit heat removal system design method based on a positive system failure rate allocation method, which provides a more flexible response strategy for the two-circuit heat removal system solution and helps to avoid the limitations of conservative measures.

[0072] 2. The present invention combines the first-layer measures and the second-layer measures to obtain a preliminary plan for the secondary circuit heat removal system of the nuclear power unit, constructs a CDF allocation model, and updates the plan through risk analysis until the plan meets the risk frequency requirements of the nuclear power unit heat removal system. Based on probability theory and system reliability evaluation technology, the final nuclear power unit secondary circuit heat removal plan is more reliable.

[0073] 3. The present invention obtains the system failure rate by simulating the PSA models of different schemes, and makes adjustments in combination with the latest measures of the secondary circuit heat removal system of nuclear power units. It also provides a scheme design evaluation method for new technologies and new measures that may appear in the future, which is helpful for the application of new technologies in the heat removal system of nuclear power units. BRIEF DESCRIPTION OF THE DRAWINGS

[0074] The accompanying drawings constituting a part of the present invention are used to provide a further understanding of the present invention. The exemplary embodiments of the present invention and their descriptions are used to explain the present invention and do not constitute an improper limitation of the present invention. In the accompanying drawings:

[0075] Figure 1 A flow chart of a method for designing a secondary circuit heat removal system of a nuclear power unit according to an embodiment of the present invention;

[0076] Figure 2 A schematic diagram of a PSA model according to an embodiment of the present invention;

[0077] Figure 3 The present invention is a schematic diagram of a design device for a secondary circuit heat removal system of a nuclear power unit according to an embodiment of the present invention. DETAILED DESCRIPTION

[0078] It should be noted that, in the absence of conflict, the embodiments of the present invention and the features in the embodiments can be combined with each other. The present invention will be described in detail below with reference to the accompanying drawings and in combination with the embodiments.

[0079] The present invention is further described in detail below in conjunction with specific embodiments. These embodiments should not be construed as limiting the scope of protection claimed by the present invention.

[0080] Example

[0081] In order to solve the problem in the prior art of combining the system failure rate currently formulated by NRC with the latest technology development to design an overall solution for the secondary circuit heat removal system, the present invention proposes a design method and device for the secondary circuit heat removal system of a nuclear power unit.

[0082] To achieve the above object, the present invention proposes a design method for a secondary circuit heat removal system of a nuclear power unit.

[0083] like Figure 1FIG. 4 is a flow chart of a method for designing a secondary heat removal system of a nuclear power unit according to an embodiment of the present invention. Figure 1 As mentioned above, the process mainly includes the following steps:

[0084] S1. Establish a secondary circuit heat removal system plan for the nuclear power unit based on the secondary circuit heat removal system measures.

[0085] Furthermore, according to the secondary circuit heat removal system measures, a secondary circuit heat removal system scheme for the nuclear power unit is established, including:

[0086] The first-level measures of the secondary circuit heat removal system are shown in Table 1, including large-flow water supply and heat removal system, small-flow water supply and heat removal system, design basis accident heat removal system, and / or design extended operating condition heat removal system. The second-level measures include the selection of support systems, injection channel schemes (main pipe or independent), redundancy, diversity, including the selection of key equipment, etc.

[0087] In this implementation, taking the selection of power distribution to support the system as an example, the safety-level system for design basis accidents usually needs to set up two redundant power supplies to supply power to the redundant equipment, but some important valves may set up multiple power supplies at the same time to ensure the realization of important functions. Some valves perform different functions of opening or isolation under normal operating conditions and accident conditions. A variety of power distribution schemes are designed according to functional requirements. The secondary heat removal system scheme of the nuclear power unit can be combined through the first layer of measures, and the second layer is designed in a more detailed manner according to the combined first layer of measures to establish a preliminary secondary heat removal system scheme for the nuclear power unit.

[0088] In this implementation, according to the combination of measures of the secondary circuit heat removal system in Table 1, the active large-flow and small-flow water supply and heat removal systems are essential for operation, and accident mitigation should at least consider the design basis accident response measures, while the small-flow water supply and heat removal can just realize the decay heat removal function under shutdown. They can be combined according to the principles of minimal design and maximization of safety benefits, thus forming the first-level measure combination: (1) active large-flow water supply and heat removal systems are used for operation; (2) a passive secondary circuit heat removal system is used for design basis accidents; (3) an active small-flow water supply and heat removal system is used for operation and design expansion conditions.

[0089] Table 1 Combination of measures for the secondary circuit heat removal system

[0090]

[0091] The second level of measures is to establish a secondary heat removal system for the nuclear power unit by combining the first level of measures with a more detailed design: (1) The active large-flow feedwater and heat removal system is not related to accident mitigation and is based on the current standard design of a three-loop pressurized water reactor (three electric pumps, two in operation and one in standby, supplying water to three steam generators through the mother pipe); (2) The minimum capacity requirement for the passive secondary heat removal system for design basis accidents is 3×50%; (3) The minimum capacity requirement for the small-flow feedwater and heat removal system for design expansion conditions is 1×100%; (4) The main steam isolation valve, large-flow feedwater isolation valve and small-flow feedwater isolation valve that perform the design basis accident isolation function have a minimum power supply requirement of two safety-grade power supplies.

[0092] S2, constructing a nuclear power heat removal system model according to the secondary circuit heat removal system solution of the nuclear power unit.

[0093] The nuclear power heat removal system consists of a CDF allocation model and a PSA model. It collects equipment reliability data and accident condition data, establishes the CDF of each initiating event, and forms an overall CDF allocation model.

[0094] Furthermore, the nuclear power heat removal system model includes a CDF allocation model and / or a PSA model, including:

[0095] According to the secondary circuit heat removal system scheme of the nuclear power unit obtained in S1, the reliability data and accident condition data of the equipment are collected, and the CDF allocation model and PSA model are constructed.

[0096] The total CDF can be approximately expressed as Formula 2, the sum of the CDFs of all initiating events:

[0097]

[0098] The CDF of each initiating event is the combination of all system function failures that may lead to the failure of mitigation of the initiating event, which is approximately expressed as Formula 1:

[0099] CDF IE(i) =F IE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……],

[0100] Once all initiating events contributing to the total CDF and their frequencies have been identified, as well as the system functions required to mitigate these initiating events, a CDF allocation model can be established.

[0101] In the specific implementation, the accident conditions in Table 2 and the equipment reliability data table in Table 3 are used to establish Figure 2 The PSA model shown in Table 2 is substituted into Formula 1 and Formula 2 to obtain the CDF model.

[0102] S3, analyzing the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the risk frequency.

[0103] The CDF model is shown in Table 4 Analysis Results 1. Each row of data in the table represents an initiating event, and the combined frequency represents the CDF of a single initiating event. The sum of the CDFs of all initiating events is calculated as the risk frequency.

[0104] Table 2 Accident conditions considered in the model

[0105]

[0106] Table 3 Equipment reliability data table

[0107]

[0108]

[0109] Table 4 Analysis results 1

[0110]

[0111]

[0112] Further, according to the failure probability and the CDF allocation model in the nuclear power heat removal system model, the failure probability is respectively assigned to the secondary circuit heat removal system measures, and the risk frequency is obtained by calculation, including:

[0113] Based on the analysis results of Table 4 established by Formula 1 and Formula 2, the risk frequency is calculated.

[0114] According to empirical values, the first-level measures for secondary circuit heat removal of nuclear power units are assigned a system function failure probability: for passive systems (with redundancy), the failure probability is 1E-4; for passive systems (without redundancy), the failure probability is 3E-3; for active systems (with redundancy), the failure probability is 1E-3; and for active systems (without redundancy), the failure probability is 3E-2.

[0115] In this implementation, as shown in Table 4, taking the corresponding initiating event of complete loss of main water supply as an example, according to Formula 1, the CDF of the initiating event of complete loss of main water supply is (1) =2E-1*[1E-4*3E-2]=6.0E-7. According to Formula 1, the CDF of all initiating events is constructed, and the results are shown in Table 4. According to Formula 2, the risk frequency is calculated as CDF=[CDF (1) +CDF (2) ……]=3.17E-06.

[0116] S4: If the risk frequency is greater than a preset risk value, the secondary circuit heat removal system solution of the nuclear power unit is updated.

[0117] The setting of the operating risk threshold of nuclear power units is first decomposed according to the core damage frequency CDF target value. The design target of new nuclear power units is usually more stringent than the regulatory target. The current domestic regulatory target is CDF≤1E-05 / reactor-year. If it is reduced by one order of magnitude according to the design target, it will be CDF≤1E-06 / reactor-year. Based on the analysis experience of PSA of previous projects, internal events, internal fires and earthquakes are the three main contributions to CDF. Therefore, 1E-6 / reactor-year is divided into four parts (internal events, internal fires, earthquakes and necessary margins), and 2.5E-7 / reactor-year is used as the preset risk value.

[0118] The risk frequency obtained in S3 is compared with the preset risk value. If it is greater than the preset risk value, the measures in the secondary circuit heat removal system plan of the nuclear power unit are adjusted.

[0119] Furthermore, the secondary heat removal system measures are adjusted to a redundant manner, and the secondary heat removal system scheme of the nuclear power unit can be updated by increasing the minimum capacity requirements of the secondary heat removal system of the nuclear power unit, including:

[0120] The risk frequency after the initial value allocation in the secondary heat removal system of the nuclear power unit is 3.17E-6 / reactor-year, which exceeds the preset risk value of 2.5E-7 / reactor-year, and the design needs to be adjusted. The second-tier measures in the secondary heat removal system of the nuclear power unit are adjusted. Since the small-flow water supply and heat removal system is used in multiple initiating event conditions, increasing the redundancy of the system helps reduce the risk, and the capacity requirement of the small-flow water supply and heat removal system used for the design expansion condition is adjusted to 2×100%.

[0121] The updated plan is as follows: (1) The active large-flow feedwater and heat removal system is based on the current standard design of a three-loop pressurized water reactor (three electric pumps, two in operation and one in standby, supplying water to three steam generators through the main pipe); (2) The minimum capacity requirement for the passive secondary circuit heat removal system for design basis accidents is 3×50%; (3) The minimum capacity requirement for the small-flow feedwater and heat removal system for design expansion conditions is 2×100%; (4) The support system of important equipment (such as power supply, etc.) must also ensure redundancy. The main steam isolation valve, large-flow feedwater isolation valve and small-flow feedwater isolation valve that perform the design basis accident isolation function must have a minimum power supply requirement of two safety-grade power supplies.

[0122] S5, analyzing the updated secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the updated risk frequency.

[0123] Further, according to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model, including:

[0124] Redistribute the indicators of the updated secondary heat removal system of the nuclear power unit, update the CDF allocation model, recalculate the risk frequency according to the CDF model, and check whether the plan meets the requirements of the preset risk value. If not, continue to modify the secondary heat removal system of the nuclear power unit until the CDF meets the preset risk value. Figure 2 The PSA model shown performs probabilistic safety analysis on different schemes to obtain system failure rates of different schemes. The system failure rate may be a risk frequency and / or a CDF obtained through the PSA model.

[0125] In this implementation, since the second-layer measures of the secondary circuit heat removal system of the nuclear power unit have been updated and modified to be redundant, the failure probability of the small-flow water supply and heat removal system is changed to 1E-3, and the combination frequency is recalculated. The quantitative analysis results after redistribution are shown in Table 5 Analysis Result 2, which is 1.06E-7 / reactor-year, which is less than the preset risk value of 2.5E-7 / reactor-year, and has met the requirements.

[0126] Table 5 Analysis results 2

[0127]

[0128]

[0129] Furthermore, the system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

[0130] When designing the secondary heat removal system, a variety of secondary heat removal system schemes for different nuclear power units will be proposed. Risk probability estimation is required for the secondary heat removal system schemes for different nuclear power units to determine whether the system design can meet the requirements of the system's preset risk value.

[0131] Probabilistic safety analysis is used to analyze the secondary heat removal system scheme of the nuclear power unit, including the impact of the second-level measures of the secondary heat removal system scheme on the initiating event and accident mitigation, and the system failure rate of different schemes is obtained, and the configuration scheme with the optimal system failure rate is recommended.

[0132] In this implementation, taking the power distribution scheme of the main steam isolation valve (MSIV) in the updated scheme (4) as an example, the MSIV is provided with two columns of A / B fast-closing channels to ensure that when the main steam isolation valve is required to realize the safe fast-closing function under accident conditions, if one column of the fast-closing channels fails, the other column of the fast-closing channels can still ensure the realization of the fast-closing function of the main steam isolation valve. Two solenoid valves in series are provided on each column of the fast-closing channels. Only when the two solenoid valves are energized at the same time can the column of fast-closing channels be triggered to perform the fast-closing function. A total of four solenoid valves, SV1\SV2 and SV3\SV4, are provided on the two columns of MSIV fast-closing channels respectively. When the valve limit switch signal indicates that the valve has reached the fully closed position, the two solenoid valves on the corresponding fast-closing channels are de-energized. If a dual-channel fast closing is performed, the four solenoid valves are energized at the same time. The power distribution design schemes related to the main steam isolation are shown in Tables 6, 7, and 8:

[0133] 6Main steam isolation related power distribution scheme 1

[0134]

[0135] Table 7 Main steam isolation related power distribution scheme 2

[0136]

[0137] Table 8 Main steam isolation related power distribution scheme 3

[0138]

[0139] In the above three power distribution schemes, when one power supply is lost, the main steam isolation valve can be closed to achieve the isolation of the main steam. The main steam bypass isolation valve may fail. By equipping the main steam bypass regulating valve in series with different series of power supplies, the effective isolation function of the main steam can be achieved.

[0140] The risk results of the above three power distribution schemes are obtained by establishing the PSA model as shown in Table 9.

[0141] Table 9 Quantitative analysis of the impact of different power distribution schemes

[0142]

[0143] According to the quantitative analysis results in Table 9, it can be concluded that the scheme 1 will cause the relay protection system to fail if a safety-grade DC train is superimposed after the loss of a safety-grade DC train. Compared with other schemes, the redundancy is poor, and the CDF caused by the loss of EHC or EHD is significantly higher than the other two schemes. After calculation, the CDF of schemes 2 and 3 are not much different, and both can be used as design schemes.

[0144] To achieve the above-mentioned purpose, the present invention also proposes a design device for a secondary circuit heat removal system of a nuclear power unit.

[0145] like Figure 3 FIG. 1 is a schematic diagram of a design device for a secondary circuit heat removal system of a nuclear power unit according to an embodiment of the present invention. Figure 3 As shown, the device includes: a first establishment module, a second establishment module, a first analysis module, a solution update module, and a second analysis module. The functions of each module will be described in detail below.

[0146] The first establishment module is used to establish a secondary circuit heat removal system plan for a nuclear power unit based on the secondary circuit heat removal system measures.

[0147] Further, in the first establishment module:

[0148] The secondary circuit heat removal system measures include first-layer measures and second-layer measures, wherein the second-layer measures include the type and / or capacity of equipment, valve model selection, battery model selection, and / or generator power supply plan formulation in the first-layer measures;

[0149] Based on the first-layer measures and the second-layer measures, a secondary circuit heat removal system scheme for the nuclear power unit is established.

[0150] Further, in the first establishment module:

[0151] The first-tier measures include a large-flow water supply and heat removal system, a small-flow water supply and heat removal system, a design basis accident heat removal system, and / or a design extended operating condition heat removal system.

[0152] The second building module is used to construct a nuclear power heat removal system model according to the secondary circuit heat removal system scheme of the nuclear power unit.

[0153] Further, in the second establishment module:

[0154] The nuclear power heat removal system model includes a CDF allocation model and / or a PSA model;

[0155] The reliability data and accident condition data of the equipment are collected, and the CDF allocation model and the PSA model are constructed according to the reliability data and the accident condition data of the equipment.

[0156] Further, in the second establishment module:

[0157] The CDF allocation model includes formula 1 and formula 2.

[0158] The formula 1 is

[0159] CDF IE(i) =FIE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……],

[0160] The formula 2 is

[0161] Wherein, CDFIE(i) is the core damage frequency of the i-th event; QSYSA(i), QSYSB(i), QSYSC(i) are the failure probabilities of the secondary heat removal system measures of the initiating event; FIE(i) represents the occurrence frequency of the initiating event; CDF is the sum of the core damage frequencies of all events, i.e., the risk frequency;

[0162] The failure probabilities are respectively assigned to the secondary circuit heat removal system measures.

[0163] Furthermore, in the first analysis module:

[0164] The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

[0165] The first analysis module is used to analyze the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the risk frequency.

[0166] Furthermore, in the first analysis module:

[0167] The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

[0168] A scheme updating module is used to update the secondary circuit heat removal system scheme of the nuclear power unit if the risk frequency is greater than a preset risk value.

[0169] Furthermore, in the scheme updating module:

[0170] The secondary circuit heat removal system measures are adjusted to a redundant manner, and the secondary circuit heat removal system plan of the nuclear power unit is updated.

[0171] The second analysis module is used to analyze the updated secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the updated risk frequency.

[0172] Furthermore, in the second analysis module:

[0173] According to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model.

[0174] Furthermore, in the second analysis module:

[0175] The system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

[0176] It should be understood that a design device for a secondary circuit heat removal system of a nuclear power unit is consistent with the description of a corresponding embodiment of a design method for a secondary circuit heat removal system of a nuclear power unit, so it will not be repeated in this embodiment.

[0177] In summary, it can be seen from the above description that the above embodiments of the present invention achieve the following technical effects:

[0178] 1. The present invention is based on probability theory and system reliability evaluation technology, and on the principles of design simplification and safety benefit maximization. It provides a two-circuit heat removal system design method based on a positive system failure rate allocation method, which provides a more flexible response strategy for the two-circuit heat removal system solution and helps to avoid the limitations of conservative measures.

[0179] 2. The present invention combines the first-layer measures and the second-layer measures to obtain a preliminary plan for the secondary circuit heat removal system of the nuclear power unit, constructs a CDF allocation model, and updates the plan through risk analysis until the plan meets the risk frequency requirements of the nuclear power unit heat removal system. Based on probability theory and system reliability evaluation technology, the final nuclear power unit secondary circuit heat removal plan is more reliable.

[0180] 3. The present invention obtains the system failure rate by simulating the PSA models of different schemes, and makes adjustments in combination with the latest measures of the secondary circuit heat removal system of nuclear power units. It also provides a scheme design evaluation method for new technologies and new measures that may appear in the future, which is helpful for the application of new technologies in the heat removal system of nuclear power units.

[0181] The above description is only a preferred embodiment of the present invention and is not intended to limit the present invention. For those skilled in the art, the present invention may have various modifications and variations. Any modification, equivalent replacement, improvement, etc. made within the spirit and principle of the present invention shall be included in the protection scope of the present invention.

[0182] It should be noted that, in this article, relational terms such as first and second, etc. are only used to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Moreover, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, the elements defined by the sentence "comprise a ..." do not exclude the existence of other identical elements in the process, method, article or device including the elements.

[0183] The logic and / or steps represented in the flowchart or otherwise described herein, for example, can be considered as an ordered list of executable instructions for implementing logical functions, which can be embodied in any computer-readable medium for use by an instruction execution system, apparatus or device (such as a computer-based system, a system including a processor or other system that can fetch instructions from an instruction execution system, apparatus or device and execute instructions), or used in combination with these instruction execution systems, apparatuses or devices.

[0184] It should be understood that the various parts of the present invention can be implemented by hardware, software, firmware or a combination thereof. In the above-mentioned embodiments, a plurality of steps or methods can be implemented by software or firmware stored in a memory and executed by a suitable instruction execution system. For example, if implemented by hardware, as in another embodiment, it can be implemented by any one of the following technologies known in the art or their combination: a discrete logic circuit having a logic gate circuit for implementing a logic function for a data signal, a dedicated integrated circuit having a suitable combination of logic gate circuits, a programmable gate array (PGA), a field programmable gate array (FPGA), etc.

[0185] It should be noted that, in the description of this specification, the description with reference to the terms "one embodiment", "some embodiments", "example", "specific example", or "some examples" etc. means that the specific features, structures, materials or characteristics described in conjunction with the embodiment or example are included in at least one embodiment or example of the present invention. In this specification, the schematic representations of the above terms do not necessarily refer to the same embodiment or example. Moreover, the specific features, structures, materials or characteristics described may be combined in any one or more embodiments or examples in a suitable manner. In addition, those skilled in the art may combine and combine the different embodiments or examples described in this specification and the features of the different embodiments or examples, unless they are contradictory.

Claims

1. A design method for a secondary circuit heat removal system of a nuclear power unit, characterized in that: include: According to the secondary heat removal system measures, establish the secondary heat removal system plan for nuclear power units; According to the secondary circuit heat removal system scheme of the nuclear power unit, a nuclear power heat removal system model is constructed; Analyze the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the risk frequency; If the risk frequency is greater than a preset risk value, updating the secondary circuit heat removal system solution of the nuclear power unit; The updated secondary circuit heat removal system scheme of the nuclear power unit is analyzed according to the nuclear power heat removal system model to obtain the updated risk frequency.

2. The method according to claim 1, characterized in that: The secondary heat removal system scheme for the nuclear power unit is established according to the secondary heat removal system measures, including: The secondary circuit heat removal system measures include first-layer measures and second-layer measures, wherein the second-layer measures include the type and / or capacity of equipment, valve model selection, battery model selection, and / or generator power supply plan formulation in the first-layer measures; According to the first-layer measures and the second-layer measures, a secondary circuit heat removal system scheme for the nuclear power unit is formed.

3. The method according to claim 2, characterized in that: The secondary circuit heat removal system measures include first-layer measures and second-layer measures, and also include: The first-tier measures include a large-flow water supply and heat removal system, a small-flow water supply 53CA heat removal system, a design basis accident heat removal system, and / or a design extended operating condition heat removal system.

4. The method according to claim 1, characterized in that: According to the secondary circuit heat removal system scheme of the nuclear power unit, a nuclear power heat removal system model is constructed, including: The nuclear power heat removal system model includes a CDF allocation model and / or a PSA model; The reliability data and accident condition data of the equipment are collected, and the CDF allocation model and the PSA model are constructed according to the reliability data and the accident condition data of the equipment.

5. The method according to claim 4, characterized in that: The constructing of the CDF allocation model and the PSA model further includes: The CDF allocation model includes formula 1 and formula 2. The formula 1 is CDF IE(i) =F IE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……], The formula 2 is Among them, CDF IE(i) is the core damage frequency of the ith event; Q SYSA(i) , Q SYSB(i) , Q SYSC(i) F is the failure probability of the secondary circuit heat removal system measures described in the initiating event; IE(i) represents the occurrence frequency of the initiating event; CDF is the sum of the core damage frequencies of all events, i.e. the risk frequency; The failure probabilities are respectively assigned to the secondary circuit heat removal system measures.

6. The method according to claim 5, characterized in that: The step of calculating the secondary heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the risk frequency includes: The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

7. The method according to claim 1, characterized in that: If the risk frequency is greater than a preset risk value, the secondary circuit heat removal system scheme of the nuclear power unit is updated, including: The secondary circuit heat removal system measures are adjusted to a redundant manner, and the secondary circuit heat removal system plan of the nuclear power unit is updated.

8. The method according to claim 5, characterized in that: The step of analyzing the updated secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the updated risk frequency includes: According to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model.

9. The method according to claim 8, characterized in that: The system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model, including: The system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

10. A design device for a secondary circuit heat removal system of a nuclear power unit, characterized in that: include: The first establishment module is used to establish a secondary circuit heat removal system plan for a nuclear power unit according to the secondary circuit heat removal system measures; The second building module is used to build a nuclear power heat removal system model according to the secondary circuit heat removal system scheme of the nuclear power unit; A first analysis module is used to analyze the secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain a risk frequency; A scheme updating module, configured to update the secondary circuit heat removal system scheme of the nuclear power unit if the risk frequency is greater than a preset risk value; The second analysis module is used to analyze the updated secondary circuit heat removal system scheme of the nuclear power unit according to the nuclear power heat removal system model to obtain the updated risk frequency.

11. The device according to claim 10, characterized in that: In the first building block: The secondary circuit heat removal system measures include first-layer measures and second-layer measures, wherein the second-layer measures include the type and / or capacity of equipment, valve model selection, battery model selection, and / or generator power supply plan formulation in the first-layer measures; Based on the first-layer measures and the second-layer measures, a secondary circuit heat removal system scheme for the nuclear power unit is established.

12. The device according to claim 11, characterized in that: In the first building block: The first-tier measures include a large-flow water supply and heat removal system, a small-flow water supply and heat removal system, a design basis accident heat removal system, and / or a design extended operating condition heat removal system.

13. The device according to claim 10, characterized in that: In the second building block: The nuclear power heat removal system model includes a CDF allocation model and / or a PSA model; The reliability data and accident condition data of the equipment are collected, and the CDF allocation model and the PSA model are constructed according to the reliability data and the accident condition data of the equipment.

14. The device according to claim 13, characterized in that: In the second building block: The CDF allocation model includes formula 1 and formula 2. The formula 1 is CDF IE(i) =F IE(i) ×∑[Q SYSA(i) ×Q SYSB(i) ×Q SYSC(i) ……], The formula 2 is Among them, CDF IE(i) is the core damage frequency of the ith event; Q SYSA(i) , Q SYSB(i) , Q SYSC(i) F is the failure probability of the secondary circuit heat removal system measures described in the initiating event; IE(i) represents the occurrence frequency of the initiating event; CDF is the sum of the core damage frequencies of all events, i.e. the risk frequency; The failure probabilities are respectively assigned to the secondary circuit heat removal system measures.

15. The device according to claim 14, characterized in that: In the first analysis module: The risk frequency is obtained by calculation according to the failure probability and the CDF allocation model in the nuclear power heat removal system model.

16. The device according to claim 10, characterized in that: In the scheme update module: The secondary circuit heat removal system measures are adjusted to a redundant manner, and the secondary circuit heat removal system plan of the nuclear power unit is updated.

17. The device according to claim 14, characterized in that: In the second analysis module: According to the updated failure probability and the CDF allocation model of the nuclear power heat removal system model, the updated risk frequency is obtained by calculation. If the updated risk frequency is lower than the preset risk value, the system failure rate is obtained by analyzing the PSA model in the nuclear power heat removal system model.

18. The device according to claim 17, characterized in that: In the second analysis module: The system failure rate obtained by analyzing the secondary circuit heat removal system solutions of different nuclear power units composed of the secondary circuit heat removal system measures is selected as the best solution.

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