STELLA-2 소듐 시험 시설 고온 배관 계통의 설계 및 건전성 평가

손석권(Seok-Kwon Son),이형연(Hyeong-Yeon Lee),주용선(Yong-Sun Ju),어재혁(JaeHyuk Eoh),김종범(Jong-Bum Kim),정지영(Ji-Young Jeong) 대한기계학회 2016 大韓機械學會論文集A Vol.40 No.9

본 연구에서는 한국원자력연구원이 개발 중인 소듐 열유동 종합효과 시험장치(STELLA-2)의 주요 배관 계통을 대상으로 고온 설계를 수행하고, 두 가지 설계기술기준에 따라 배관의 건전성 평가를 수행하였다. 배관 설계기술기준으로는 일반 압력배관에 관한 ASME B31.1과 프랑스의 원자력등급 배관 설계기술기준인 RCC-MRx RD-3600을 적용하였으며, 이들 기술기준의 보수성을 정량적으로 비교 및 분석하였다. STELLA-2 소듐시설에서는 모의 잔열제거계통(Model DHRS), 모의 중간열전달계통(Model IHTS) 및 펌프 모의계통(PSLS)에 배관이 설치되는데, 두 설계기술기준을 따라 이들 배관 계통에 대해 건전성 평가를 수행한 결과 설계 건전성이 확인되었으며, 설계 기술기준 간 비교분석 결과 유지하중에 대해서는 ASME B31.1이, 열하중에 대해서는 RCC-MRx RD-3600이 더 보수적인 것으로 평가되었다. In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.

Base-Isolated NPP의 Interface 배관계통에 대한 내진해석방법 연구

김대수(Daesoo Kim),이효승(Hyoseung Lee),류동원(Dongwon Lyu),김수겸(Sookyum Kim),장성철(Seongcheol Jang) 대한기계학회 2017 대한기계학회 춘추학술대회 Vol.2017 No.11

A study of the application of seismic isolation system have been widely researched in order to increase the safety of seismic analysis for nuclear power plant(NPP). In the design of base isolated NPPs, the seismic analysis of interface piping system multi-supported between base-isolated and fixed building is one of the key issues. For the seismic design of interface piping system, it is needed not only rigidity to withstand the force of the inertia response generated by excitations from base-isolated and fixed building but also flexibility to cope with the large relative displacements between two buildings. In order to meet seismic requirements of interface piping system, this paper utilized two methods to perform the seismic analysis of MS interface piping system. First, in order to satisfy the rigidity requirement of MS interface piping system, this paper adapts Multiple Input Response Spectrum (MIRS) method together with the SRSS combination between support groups, USNRC Regulatory Guide 1.61 damping as seismic analysis method to consider different vibration characteristics of base-isolated and fixed building. Second, in order to secure the flexibility of MS interface piping system, this paper use the method to adjust the stiffness of pipe supports installed on MS interface piping system, and performs the sensitivity study on the stiffness of pipe support to find the optimal stiffness that can distribute the excessive piping stress and support loads occurred in the local portion of MS interface piping system.

PIV와 신경망을 이용한 배관시스템 원격 미세변위 측정과 실시간 작동상태 진단

전민규,조경래,오정수,이창제,도덕희 한국수소및신에너지학회 2013 한국수소 및 신에너지학회논문집 Vol.24 No.3

Piping systems play an important role in gas and oil transferring system. In the piping system, there are many elements, such as valves and flow meters. In order to check their normal operating conditions, each signal from each element is displayed on the monitor in the pipe control room. By the way, there are several accidental cases in the piping system even if all signals from the local elements are judged to be normal on the monitor in the control room. Further, opposite cases often happen even the monitor shows abnormal while the local elements work normal. To overcome this abnormal functions, it is not so easy to construct the environment in which sensors detecting the working states of all elements installed in the piping system. In this paper, a new non-contact measurement technique which can calculate the elements' delicate displacements by using a PIV(particle image velocimetry) and diagnose their working states by using a neural network is proposed. The measurement system consists of a host computer, a micro system, a telescope and a high-resolution camera. As a preliminary test, the constructed measurement system was applied to measure delicate vibrations of mobile phones. For practical application, a pneumatic system was measured by the constructed system.

2차계통 배관의 특성을 고려한 균열평가 기법 분석

우승완(Seung-Wan Woo),장윤석(Yoon-Suk Chang),최재붕(Jae-Boong Choi),김영진(Young-Jin Kim),정명조(Myung-Jo Jhung),최영환(Young-Hwan Choi) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5

The well-known flaw evaluation criteria embodied in ASME Sec. XI are not applicable to secondary system piping of nuclear power plants, mainly, due to different R/t ratios comparing with those of primary system piping. Despite of previous activities to resolve this specific issue, there are still arguments to determine whether flaws detected in the secondary system are acceptable or not for continued service. In the present study, pros and cons of relevant researches are fully discussed and unique features of secondary system piping are assessed by retrieving a database enveloping eight nuclear power plants. Subsequently, for pipe geometries beyond the ASME applicable limits, finite element analyses are carried out by changing sizes of surface crack. The analysis results showed limitations of current flaw evaluation schemes and addressed necessity of establishing a new one which will be available from further detailed finite element analyses.

동흡진기를 사용한 원전 배관계 내진성능 상향에 대한 연구

곽신영,곽진성,이환호,오진호,구경회 한국압력기기공학회 2018 한국압력기기공학회 논문집 Vol.14 No.2

In this study, the dynamic absorber and the damper are applied to improve the seismic performance of the piping system, and their quantitative effects on the piping system performance are examined. For this purpose, the response performances of piping system applied with the dynamic absorber/damper are compared with those of the original piping system. Firstly, the frequency response analyses of the piping system with the presence or the absence of dynamic absorber/damper are performed and these results are compared. It has been shown that the maximum acceleration response per the frequency of the piping system is considerably reduced by installing the dynamic absorber and the damper. Secondly, the seismic responses of the piping systems with and without dynamic absorber/damper are compared. As a result of the numerical analyses, it is confirmed that key responses are reduced by 17%-63% due to the installation of the dynamic absorber and damper. Finally, as a result of the seismic performance evaluation, it is confirmed that the HCLPF (High Confidence of Low Probability of Failure) seismic performances are increased by 1.22 to 2.70 times with respect to the failure modes with an aid of the dynamic absorber and damper.

그루브 조인트가 설치된 수계소화설비 입상배관계통의 지진거동분석을 위한 실험적 연구

김성완 ( Sung-wan Kim ),윤다운 ( Da-woon Yun ),김재봉 ( Jae-bong Kim ),전법규 ( Bub-gyu Jeon ) 한국구조물진단유지관리공학회 2021 한국구조물진단유지관리공학회 논문집 Vol.25 No.2

본 연구에서는 NFPA 13을 참조하여 구조물 2층을 구현한 강재지그를 제작하였으며, 그루브 조인트의 배관연결재가 적용된 배관 계통을 구성하고 정적 반복가력에 의한 지진모사실험을 수행하였다. 지진모사실험은 엑츄에이터로 건축물 내진설계기준의 최대허용 층간변위에 대한 반복가력실험을 수행하였다. 지진하중 발생 시 입상배관의 구조물에 대한 변형 또는 구조부재간의 상대변위에 의한 변위지배적인 거동에 따른 배관계통과 주요 배관요소의 지진거동을 분석하였다. 배관계통의 변형각은 기존의 센서를 이용하여 측정하기가 어려우므로 이미 지측정시스템을 적용하였다. In this study, a steel frame that realized the second floor of a structure was fabricated in referring to NFPA 13. In addition, a riser pipe system with groove joints was installed, and a seismic simulation test was performed using static cyclic loading. Cyclic loading tests on the maximum allowable side sway of seismic design standards for buildings in Korea were conducted using actuators to analyze the seismic behavior of the riser pipe system and major piping elements due to the deformation of the steel frame structure or the displacement-dominant behavior caused by the relative displacement between the structural members in the event of a seismic load. Moreover, the deformation angle of the riser pipe system was measured using an image measurement system because it is difficult to measure using the conventional sensors.

Design Evaluation of Piping Systems in a Sodium Test Loop

Dong-Won Lee(이동원),Ji-Young Jeong(정지영),Hyeong-Yeon Lee(이형연) 대한기계학회 2013 대한기계학회 춘추학술대회 Vol.2013 No.12

A large scale sodium test loop of the STELLA(Sodium integral effect test loop for safety simulation and assessment)-1 has been installed at KAERI and the loop has main piping loop systems. In this study design and evaluation of the piping systems in the STELLA-1 loop have been conducted according to the design codes of the French code RCC-MRx RB-3600, ASME B31.1 and ASME NB-3200. The evaluation results have shown that the piping designs were well within all design allowable of the codes. It was also shown that the results from design by rules (RB-3600 and ASME B31.1) were more conservative than those from design by analysis (ASME NB-3200).

종단마개 주위의 유동해석을 통한 배관감육

곽승현(Seung-Hyun Kwag),정인동(In-Dong Jung) 한국해양환경·에너지학회 2012 한국해양환경공학회 학술대회논문집 Vol.2012 No.5

Flow analysis was carried out around the end cap in the nuclear piping system. The wall thinning was due to the liquid droplet impingement erosion, which gave the leakage in the piping system. They are composed of the control valve, orifice, expander, check valve, tie-type pipe and the md-cap. Numerical simulation has been carried out to make clear the effect of liquid droplet impingement erosion to the end cap, which is passed through the orifice in the two-phase of high pressure in a power plant For the analysis, three types of models (ID,2D,3D) are applied for the comparison with the base design. Computational result shows that 2D case is the best among them because it weakens the impingement erosion. FLUENT package code was used for the hybrid multi-phase with RNG k-e turbulence model.

장기간 크리프 영역에서 가동되는 압력 기기 및 배관계통의 ASME Code (VIII Div.2 & B31.1)를 따른 설계의 비보수성 리스크

이형연(Hyeong-Yeon Lee),이제환(Jewhan Lee),김태완(Tae-Wan Kim) 대한기계학회 2019 대한기계학회 춘추학술대회 Vol.2019 No.11

ASME Section VIII Division 2(‘ASME VIII(2)’) and ASME B31.1 are widely used for design of not only lowtemperature but also high-temperature pressure vessels and piping, respectively operating at high temperature in creep range. In high-temperature design of a pressure vessel according to ASME VIII(2) and of a piping according to ASME B31.1, creep effects are not considered explicitly and the design evaluation results do not change depending on hold times at high temperature. In case of French high temperature design code of RCC-MRx, it explicitly takes creep hold time into account like the case of ASME Section III Division 5. Sensitivity analyses with various hold times at high temperature in design evaluations were conducted as per RCC-MRx to see the effects of hold times on design evaluations for a heat exchanger of IHX (intermediate heat exchanger) and a piping system in a sodium test facility. The conservatism of ASME VIII(2) and ASME B31.1 in case of long-term operation at high temperature in creep range have been quantified based on the comparisons in design evaluations as per ASME Codes with those of RCC-MRx which explicitly takes hold time into account. It was shown from the comparison that the results from ASME VIII (2) and ASME B31.1 exceeded the design limits of RCC-MRx if hold times exceed certain limits, which means ASME VIII (2) and ASME B31.1 might give non-conservative results in case of long time operation in creep range.

가동원전 기기의 피로영향을 고려한 극한지진 재평가 부위 선정 방안 연구

이대영(Dae-Young Lee),박흥배(Heung-Bae Park),김윤재(Yun-Jae Kim),김진원(Jin-Weon Kim),박노철(No-Cheol Park) 대한기계학회 2021 대한기계학회 춘추학술대회 Vol.2021 No.4

동일본 대지진 이후, 원전 안전성에 대한 변화된 지침⁽¹⁾에 따라 원전 배관기기는 설계기준초과지진에 대한 최신 기준을 적용한 배관계 재평가를 고려하게 되었다. 이를 위해 대상 부위 선정부터 평가 방법 개발 등 관련 연구가 국내외에서 진행되고 있다. 동일본 대지진과 같은 설계기준초과지진은 최대반복하중이 짧은 시간 동안 작용하면서 원전 배관기기에 주로 저주기피로의 영향을 미친다. 원전 배관기기는 운전기간 동안 이러한 극한지진하중에 의한 저주기피로 영향과 함께 운전모드나 기하학적 형상 등에 따른 열피로 등의 영향을 누적하여 받게 된다. 주요 기기의 피로 영향을 관리하기 위하여 발전소 차원의 피로감시시스템이 설치되었으며, 환경피로의 영향을 받는 기기를 관리하기 위하여 피로 평가 곡선⁽²⁾ 등이 개발되었다. 환경피로에 대한 관리 방법은 미국과 유럽 등의 원전 사업자별로 다소 다른 접근 방법을 적용하고 있는 것으로 알려져 있다. 최근까지 발전소 환경피로를 관리하기 위한 관련 연구도 계속 진행되고 있다. 본 논문은 발전소 운전환경을 고려한 피로 관리 방법 등을 검토하고, 피로 조건에 취약한 부위들을 지진하중 재평가에 반영할 수 있는 선정 방법 등을 제시하고자 한다. After the Great East Japan Earthquake, Nuclear power plant piping should be re-evaluated according to the latest safety standards and requirements⁽¹⁾. In order to develop new evaluation methods, many studies are being conducted by domestic and international research groups. Beyond Design Basis Earthquakes such as Great East Japan Earthquake have load type of strong motions with maximum peak loads for a short interval. Seismic loads have an effect of low cycle fatigue failure on the piping system. Nuclear power plant piping systems are affected by low cycle fatigue due to beyond design seismic loads, as well as thermal fatigue according to an operation mode and geometry. A plant-level fatigue monitoring system was installed to manage the fatigue effects of major components, and fatigue evaluation curves⁽²⁾ were developed to manage equipment affected by environmentally-assisted fatigue. Nuclear power plant operators in the United States and Europe have different approaches to managing environmentally-assisted fatigue. The latest researches have been conducted to manage environmentally-assisted fatigue in power plants. This paper proposes the screening methods for re-evaluating seismic loads for piping component affected by fatigue.