KR20230015030A - Method for analyzing interfacing system loss of coolant accident in nuclear power plant - Google Patents
Method for analyzing interfacing system loss of coolant accident in nuclear power plant Download PDFInfo
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- 239000002826 coolant Substances 0.000 title claims abstract description 45
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- 239000012857 radioactive material Substances 0.000 description 3
- 239000012530 fluid Substances 0.000 description 2
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Abstract
Description
본 발명은 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법에 관한 것이다. The present invention relates to a method for evaluating a loss-of-coolant accident at a low pressure boundary of a nuclear power plant.
원자력 발전소의 중대사고는 설계기준을 초과하여 노심의 현저한 손상이 발생하는 사고이다. A serious accident in a nuclear power plant is an accident in which significant damage to the core occurs beyond design standards.
사고관리 범위 및 사고관리능력 평가의 세부 기준에 관한 규정에 따르면, 노심의 현저한 손상 이후 발생하는 위협요인으로 인하여 방사성 물질의 대량방출을 방지하기 위한 원자로 건물의 방호벽 기능이 상실되지 않도록 해야 한다.According to the regulations on the detailed criteria for the evaluation of the accident management scope and accident management capability, it is necessary to prevent the loss of the barrier function of the reactor building to prevent the mass release of radioactive material due to the threat factors that occur after significant damage to the core.
또한, 증기 발생 전열관 크립파손 등 원자로 건물 격리 경계에 대한 평가가 이루어져야 하며, 저압경계부 냉각재 상실사고(interfacing system loss of coolant accident, ISLOCA)가 이에 해당한다.In addition, evaluation of the isolation boundary of the reactor building, such as steam generation heat pipe creep failure, should be performed, and the interfacing system loss of coolant accident (ISLOCA) corresponds to this.
저압경계부 냉각재 상실사고는 별도의 완화설비가 없어 평가 절차 및 결과가 매우 중요하나, 현재까지 이와 관련한 평가 방법(절차)이 정립되어 있지 않다. There is no separate mitigation facility for the low-pressure boundary coolant loss accident, so the evaluation procedure and results are very important, but no evaluation method (procedure) has been established so far.
본 발명의 목적은 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법을 제공하는 것이다. An object of the present invention is to provide a method for evaluating the loss of coolant accident at the low pressure boundary of a nuclear power plant.
본 발명은 원자력 발전소의 저압경계부 냉각재상실사고(ISLOCA)를 평가하는 방법에 있어서, 상기 원자력 발전소의 배관 중 조사 대상 배관을 파악하는 제1단계; 상기 조사 대상 배관이 상기 저압경계부 냉각재상실사고와 연관되어 있는 연관 배관인지를 판단하는 제2단계; 및 각각의 상기 연관 배관의 냉각재상실사고 평균 발생 빈도를 도출하는 제3단계;를 포함하는 저압경계부 냉각재상실사고를 평가하는 방법에 관한 것이다. The present invention is a method for evaluating a low pressure boundary loss of coolant accident (ISLOCA) of a nuclear power plant, comprising: a first step of identifying a pipe to be investigated among pipes of the nuclear power plant; a second step of determining whether the pipe to be investigated is a related pipe associated with the low-pressure boundary part coolant loss accident; and a third step of deriving an average frequency of loss of coolant accidents of each of the related pipes.
상기 제2단계에서는, 상기 배관이 다음을 만족하면 상기 연관 배관으로 판단하는 것을 포함할 수 있다. The second step may include determining that the pipe is the related pipe if the following conditions are satisfied.
㉠ 원자로냉각재계통과 연결㉠ connected to the reactor coolant system
㉡ 격납건물을 통과㉡ pass through the containment building
㉢ 저압경계부가 격납용기 외부로 확장㉢ the low pressure boundary extends outside the containment
㉣ 저압경계부의 설계압력이 일정 압력보다 낮음㉣ the design pressure of the low pressure boundary is lower than the certain pressure
㉤ 사고 발생에 따른 과압 발생에 의해 냉각재 방출 가능성이 있음 및㉤ There is a possibility of coolant release due to overpressure due to an accident, and
㉥ 내경이 일정 사이즈 이상.㉥ The inner diameter is larger than a certain size.
상기 일정 압력은, 상기 원자로냉각재계통의 압력을 포함할 수 있다.The predetermined pressure may include the pressure of the reactor coolant system.
상기 냉각재 방출은, 사고 발생에 따른 상기 연관 배관의 상기 저압경계부에서의 과압 발생에 의해 발생되는 것일 수 있다.The discharge of the coolant may be caused by an overpressure generated in the low pressure boundary portion of the associated pipe due to an accident.
상기 일정 사이즈는, 3/8인치일 수 있다.The predetermined size may be 3/8 inch.
상기 제3단계에서, 상기 평균 발생 빈도는, 상기 각각의 연관 배관을 차단할 수 있는 밸브 및 펌프 설비를 고려하여, 상기 각각의 연관 배관의 파단 빈도 값 결과를 토대로 도출될 수 있다.In the third step, the average frequency of occurrence may be derived based on a result of the break frequency value of each associated pipe in consideration of a valve and a pump facility capable of blocking each associated pipe.
상기 제3단계를 통해 도출된 평균 발생 빈도를 합산하는 제4단계를 더 포함하며, 상기 평균 발생 빈도의 합산 결과를 토대로, 상기 저압경계부 냉각재상실사고 빈도가 산출될 수 있다.A fourth step of summing the average occurrence frequencies derived through the third step is further included, and the frequency of the low pressure boundary loss coolant accident may be calculated based on a summation result of the average occurrence frequencies.
본 발명에 따르면, 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법이 제공된다. According to the present invention, a method for evaluating a loss of coolant accident at a low pressure boundary of a nuclear power plant is provided.
도 1은 본 발명의 일실시예에 따른 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법을 나타낸 순서도이고,
도 2는 본 발명의 일실시예에 따른 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법에서 제2단계를 상세히 나타낸 순서도이다. 1 is a flowchart showing a method for evaluating a loss of coolant accident at a low pressure boundary of a nuclear power plant according to an embodiment of the present invention;
2 is a flowchart showing in detail the second step in the method for evaluating the loss of coolant accident at the low pressure boundary part of a nuclear power plant according to an embodiment of the present invention.
이하 도면을 참조하여 본 발명을 더욱 상세히 설명한다. 첨부된 도면은 본 발명의 기술적 사상을 더욱 구체적으로 설명하기 위하여 도시한 일 예에 불과하므로 본 발명의 사상이 첨부된 도면에 한정되는 것은 아니다.The present invention will be described in more detail with reference to the following drawings. Since the accompanying drawings are only examples shown to explain the technical spirit of the present invention in more detail, the spirit of the present invention is not limited to the accompanying drawings.
도 1은 본 발명의 일실시예에 따른 원자력 발전소의 저압경계부 냉각재상실사고를 평가하는 방법을 나타낸 순서도이고, 도 2는 본 발명의 일실시예에 따른 제2단계를 나타낸 순서도이다.1 is a flowchart illustrating a method for evaluating a loss of coolant accident at a low pressure boundary of a nuclear power plant according to an embodiment of the present invention, and FIG. 2 is a flowchart showing a second step according to an embodiment of the present invention.
원자력 발전소의 배관 중에서 저압경계부 냉각재상실사고를 평가하기 위한 조사 대상 배관을 파악한다. (제1단계)(S100) Among the piping of a nuclear power plant, the piping to be investigated to evaluate the loss of coolant accident at the low pressure boundary is identified. (Step 1) (S100)
조사 대상 배관은 원자력발전소의 모든 배관을 대상으로 할 수 있다.Pipes to be surveyed may be all pipes of nuclear power plants.
제1단계는 원자력 발전소 내 설치된 배관과, 기존 원자력 발전소 설계 및 설계 변경된 도면과의 확인을 통해 이루어진다.The first stage is performed through confirmation of the piping installed in the nuclear power plant and the existing nuclear power plant design and design-changed drawings.
이후 조사 대상 배관이 저압경계부 냉각재상실사고와 연관되어 있는 연관 배관인지를 판단한다. (제2단계)(S200)Then, it is determined whether the piping to be investigated is a related piping related to the low pressure boundary part coolant loss accident. (Step 2) (S200)
도 2와 같이, 제2단계에서는 제1단계를 통해 조사 대상 배관으로 파악된 배관 중에서 다음과 같이 저압경계부 냉각재상실사고와 연관되어 있는 연관 배관으로 판단한다.As shown in FIG. 2, in the second step, among the pipes identified as the target pipes to be investigated through the first step, it is determined as a related pipe related to the low pressure boundary loss of coolant accident as follows.
첫 번째로 원자로냉각재계통과 연결되어 있는 배관인지를 확인한다. (S210)First, check whether the piping is connected to the reactor coolant system. (S210)
두 번째로 격납건물을 통과하는 배관인지를 확인한다. (S220)Second, check whether the pipe runs through the containment building. (S220)
세 번째로 저압경계부가 격납용기 외부로 확장되어 있는 배관인지를 확인한다. (S230)Thirdly, it is checked whether the low pressure boundary part is a pipe extending outside the containment container. (S230)
네 번째로 저압경계부의 설계압력이 일정 압력보다 낮은 배관인지를 확인한다. (S240)Fourth, check whether the design pressure of the low pressure boundary part is lower than the certain pressure. (S240)
일정 압력은 원자로냉각재계통의 압력을 의미할 수 있다. The constant pressure may mean the pressure of the reactor coolant system.
일정 압력은 원자로발전소 내 설치되어있는 배관은 각 배관별로 직경, 두께, 재질, 유체압력 및 온도 등 설비마스터에 등록되어 있는 속성정보를 통해 확인할 수 있다. The constant pressure can be checked through property information registered in the facility master, such as diameter, thickness, material, fluid pressure and temperature, for each pipe installed in the reactor power plant.
즉, 유체는 압력이 높은곳에서 낮은곳으로 이동하므로, 원자로냉각재계통보다 낮은 압력을 갖는 배관이 있을 경우, 해당 배관으로 1차측의 냉각재가 이동하게 되며 이로 인해 방사성 물질이 유출되며, 이때 방사성 물질이 유출되는 배관의 속성정보를 통해 일정 압력보다 낮은 배관인지를 확인할 수 있게 된다. That is, since the fluid moves from a place where the pressure is high to a place where the pressure is low, if there is a pipe having a lower pressure than the reactor coolant system, the coolant on the primary side moves to the pipe, and as a result, radioactive material leaks out. At this time, the radioactive material Through the attribute information of the outgoing pipe, it is possible to check whether the pipe is lower than a certain pressure.
다섯 번째로 사고 발생에 따른 저압경계부 내 밸브 및 펌프 설비의 차단 실패로 인해 발생되는 과압 발생에 따라, 냉각재 방출 가능성이 있는 배관인지를 확인한다. (S250) Fifth, it is checked whether the pipe has the possibility of releasing the coolant according to the occurrence of overpressure caused by the failure to shut off the valve and pump facilities in the low pressure boundary part due to the occurrence of the accident. (S250)
이때 냉각재 방출은 사고 발생에 따른 연관 배관의 저압경계부에서의 과압 발생에 의해 발생된다. At this time, the release of the coolant is caused by the occurrence of overpressure at the low pressure boundary of the associated pipe due to the occurrence of the accident.
여섯 번째로 배관 내경이 일정 사이즈 이상인지를 확인한다. (S260)Sixth, check whether the inner diameter of the pipe is larger than a certain size. (S260)
일정 사이즈는 1/8인치 내지 5/8인치일 수 있으며, 구체적으로는 3/8인치일 수 있다.A certain size may be 1/8 inch to 5/8 inch, and may be specifically 3/8 inch.
배관 내경 사이즈가 3/8인치 이하의 배관에서는, 이상이 발생하더라도 배관에서 누출되는 누출양이 많지 않기 때문에, 3/8인치 이상의 내경 사이즈를 가지고 있는 배관을 대상으로 확인한다. In pipes with an inner diameter of 3/8 inch or less, the amount of leakage from the pipe is not large even if an error occurs, so check the pipe with an inner diameter of 3/8 inch or more.
조사 대상 배관이 저압경계부 냉각재상실사고와 연관되어 있는 연관 배관인지를 판단한 이후, 도출된 각각의 연관 배관의 평균 발생 빈도를 도출한다. (제3단계)(S300)After determining whether the pipe to be investigated is a related pipe related to the low pressure boundary loss of coolant accident, the average occurrence frequency of each derived pipe is derived. (Step 3) (S300)
평균 발생 빈도는, 각각의 연관 배관을 차단할 수 있는 밸브 및 펌프 시설을 고려하여, 각각의 연관 배관의 파단 빈도 값 결과를 토대로 도출한다. 구체적으로, 파단 빈도 값 결과 도출은 격납건물 내부의 차단밸브나 펌프 등의 고장률 값을 고려하여 계산한다. The average occurrence frequency is derived based on the result of the break frequency value of each associated pipe, taking into account the valve and pump facilities capable of blocking each associated pipe. Specifically, the derivation of the fracture frequency value is calculated by considering the failure rate value of shut-off valves or pumps inside the containment building.
구체적으로, 파단 빈도 값은 기존 원자력 발전소 내 설치된 각각의 배관 파단 빈도 값 데이터베이스(DB)를 활용하여 도출할 수 있다. Specifically, the fracture frequency value may be derived using a database (DB) of each pipe fracture frequency value installed in an existing nuclear power plant.
이후 제3단계를 통해 도출된 평균 발생 빈도를 합산한다. (제4단계)(S400)Then, the average frequency of occurrence derived through the third step is summed. (Step 4) (S400)
평균 발생 빈도의 합산 결과를 토대로 저압경계부 냉각재상실사고 빈도가 산출되며, 이 빈도를 저압경계부 냉각재상실사고로 인한 노심손상 빈도로 사용하게 된다. Based on the result of summing up the average frequency of occurrence, the frequency of low pressure boundary loss of coolant accidents is calculated, and this frequency is used as the core damage frequency due to low pressure boundary loss of coolant accidents.
이상으로 본 발명 내용의 특정한 부분을 상세히 기술하였는바, 당업계의 통상의 지식을 가진 자에게 있어서, 이러한 구체적 기술은 단지 바람직한 실시 양태일 뿐이며, 이에 의해 본 발명의 범위가 제한되는 것이 아닌 점은 명백할 것이다. 따라서, 본 발명의 실질적인 범위는 첨부된 청구항들과 그것들의 등가물에 의하여 정의된다고 할 것이다. As above, specific parts of the present invention have been described in detail, and for those skilled in the art, it is clear that these specific descriptions are only preferred embodiments, and the scope of the present invention is not limited thereby. something to do. Accordingly, the substantial scope of the present invention will be defined by the appended claims and their equivalents.
Claims (7)
상기 원자력 발전소의 배관 중 조사 대상 배관을 파악하는 제1단계;
상기 조사 대상 배관이 상기 저압경계부 냉각재상실사고와 연관되어 있는 연관 배관인지를 판단하는 제2단계; 및
각각의 상기 연관 배관의 냉각재상실사고 평균 발생 빈도를 도출하는 제3단계;를 포함하는 저압경계부 냉각재상실사고를 평가하는 방법.In the method for evaluating the low pressure boundary loss of coolant accident (ISLOCA) of a nuclear power plant,
A first step of identifying a pipe to be investigated among pipes of the nuclear power plant;
a second step of determining whether the pipe to be investigated is a related pipe associated with the low-pressure boundary part coolant loss accident; and
A method for evaluating low-pressure boundary loss-of-coolant accidents, including a third step of deriving an average occurrence frequency of loss-of-coolant accidents of each of the related pipes.
상기 제2단계에서는,
상기 배관이 다음을 만족하면 상기 연관 배관으로 판단하는 것을 포함하는 저압경계부 냉각재상실사고를 평가하는 방법
㉠ 원자로냉각재계통과 연결
㉡ 격납건물을 통과
㉢ 저압경계부가 격납용기 외부로 확장
㉣ 저압경계부의 설계압력이 일정 압력보다 낮음
㉤ 사고 발생에 따른 과압 발생에 의해 냉각재 방출 가능성이 있음 및
㉥ 내경이 일정 사이즈 이상.In paragraph 1,
In the second step,
A method for evaluating a low-pressure boundary loss of coolant accident including determining that the pipe is the related pipe if the pipe satisfies the following
㉠ connected to the reactor coolant system
㉡ pass through the containment building
㉢ the low pressure boundary extends outside the containment
㉣ the design pressure of the low pressure boundary is lower than the certain pressure
㉤ There is a possibility of coolant release due to overpressure due to an accident, and
㉥ The inner diameter is larger than a certain size.
상기 일정 압력은,
상기 원자로냉각재계통의 압력을 포함하는 저압경계부 냉각재상실사고를 평가하는 방법.In paragraph 2,
The constant pressure,
A method for evaluating a low pressure boundary coolant loss accident including the pressure of the reactor coolant system.
상기 냉각재 방출은,
사고 발생에 따른 상기 연관 배관의 상기 저압경계부에서의 과압 발생에 의해 발생되는 것인 저압경계부 냉각재상실사고를 평가하는 방법. In paragraph 2,
The coolant discharge,
A method for evaluating a low-pressure boundary coolant loss accident caused by an overpressure occurring in the low-pressure boundary of the associated piping following the occurrence of an accident.
상기 일정 사이즈는,
3/8인치인 저압경계부 냉각재상실사고를 평가하는 방법.In paragraph 2,
The certain size,
A method for evaluating loss of coolant at the 3/8 inch low pressure boundary.
상기 제3단계에서,
상기 평균 발생 빈도는,
상기 각각의 연관 배관을 차단할 수 있는 밸브 및 펌프 설비를 고려하여, 상기 각각의 연관 배관의 파단 빈도 값 결과를 토대로 도출되는 저압경계부 냉각재상실사고를 평가하는 방법. In paragraph 1,
In the third step,
The average occurrence frequency is,
A method for evaluating the low pressure boundary loss of coolant accident derived based on the result of the fracture frequency value of each of the related pipes in consideration of the valve and pump equipment capable of blocking each of the related pipes.
상기 제3단계를 통해 도출된 평균 발생 빈도를 합산하는 제4단계를 더 포함하며,
상기 평균 발생 빈도의 합산 결과를 토대로, 상기 저압경계부 냉각재상실사고 빈도가 산출되는 저압경계부 냉각재상실사고를 평가하는 방법.In paragraph 1,
Further comprising a fourth step of summing the average frequency of occurrence derived through the third step,
A method for evaluating low pressure boundary coolant loss accidents in which the frequency of the low pressure boundary coolant loss accident is calculated based on the summation result of the average occurrence frequency.
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