• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• 대한기계학회논문집A
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• 제40권1호
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• pp.81-86
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• 2016

밸브는 배관의 유량을 차단 및 제어하기 위한 장치로써 게이트 밸브, 글로브 밸브, 체크 밸브 등 많은 종류가 사용되고 있다. 그 중 글로브 밸브는 고압력 조건에서의 유량조절이 용이하여 LNG 선박, 증기 배관 등에 사용된다. 본 논문에서는 글로브 밸브의 누설 문제를 구조적으로 해결하기 위해 시트의 형상을 변형하는 방법을 제시하였다. 또한 유한 요소 해석을 통해 각 모델의 응력분포와 변형량을 비교하고 이를 통하여 제안한 모델에 대한 검증을 진행하였다. 시뮬레이션 결과 제안된 모델에서 원주 방향의 변형이 줄어들고, 누설을 감소시킬 수 있는 Self-supporting 효과를 확인할 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.683-690
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• 1996

The MIDLOOP code has been developed for the evaluation of RES pressurization transients initiated from a loss-of-Residual Heat Removal System (RHRS) during mid-loop operation after reactor shutdown. It provides a fast running and realistic tool for studying parametrically the response of important plant parameters such as pressure, temperature, and level to various plant combinations of the primary side vent, makeup, and leakage procedures and the steam generator (SG) conditions. The code consists of ten nodes representing the primary and secondary sides of a nuclear power plant and can analyze the effect of air on the primary system pressurization and primary to secondary heat transfer. The analysis results of the MIDLOOP code are in good agreement with the ROSA-IV/LSTF experiment without opening in the RCS.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.475-479
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• 2011

Leak problem with large pressure drop occurrs non-periodic shock pulsation due to the deterioration of a isolation valve in feed-water recirculation piping system. This paper discusses on the shock vibration and noise occurred due to the effect of acoustical shock pulsations by degradation of the isolation valve in a power site.

• 방사선산업학회지
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• 제9권4호
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• pp.217-221
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• 2015

When reactor coolant leaks occur due to cracks of a steam generator's tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000~7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In supercritical carbon dioxide method, we found that it also had a high decontamination efficiency. According to the results of these experiments, almost all decontamination method had a high efficiency, but considering the amounts of the secondary waste productions and work environment of the nuclear power plant, we judged the ultrasound and supercritical carbon dioxide method are suitable for application to the plant and we established the plant applicable decontamination process system on the basis of these two methods.

• 비파괴검사학회지
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• 제36권5호
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• pp.384-390
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• 2016

플랜트 배관의 누설감시 시스템은 누설 유무 판단뿐만 아니라 누설의 위치를 신속히 파악하는 것 또한 매우 중요하다. 플랜트 배관의 누설 검출 방법에는 주로 AE(acoustic emission)센서, 마이크로폰어레이 그리고 카메라 영상을 이용한 방법들이 있다. 최근 광역감시 및 원거리감시의 이점이 있는 카메라 영상을 이용한 방법에 대한 연구가 진행되어 왔다. 하지만 기존 카메라 영상을 이용한 방법들은 누설 유무와 대략적인 누설의 위치를 판단하고 있으나 누설이 시작되는 정확한 위치 추정에 대한 연구는 아직 미흡한 상태이다. 따라서 본 논문에서는 카메라를 이용한 누설 검출 방법을 이용해 누설영역을 산출하고 누설 검출 결과를 분석하여 누설 위치를 추정하는 방법을 제안하였으며 실험을 통하여 성능을 평가하였다.

• 비파괴검사학회지
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• 제36권6호
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• pp.496-503
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• 2016

발전소나 석유화학 플랜트 구조물에 누설이 발생하면 인명 피해 및 경제적인 손실을 초래한다. 이러한 누설은 플랜트 배관의 진동으로 인한 피로파괴나 배관 감육으로 인해 발생한다. 플랜트 배관의 진동을 감시하기 위한 방법으로 주로 가속도센서나 레이저센서가 사용되지만 설치 및 운용의 어려움이 따르고 동시에 광범위 측정 시 비용 증가가 발생하게 된다. 이러한 문제점들을 해결하기 위해 최근 카메라를 이용한 누설 및 진동 변위 측정 방법에 대한 연구가 이루어지고 있다. 카메라를 이용한 누설 및 진동 변위 측정 방법은 설치가 간단하고 원거리 측정 및 넓은 범의의 측정이 가능한 장점을 가지고 있다. 따라서 본 연구에서는 카메라를 이용해 누설 및 진동 변위를 측정할 수 있는 시스템을 개발하였고 실험을 통해 성능을 검증하였다.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.61-66
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• 2011

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.368-375
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• 2016

A sodiume-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodiume-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called "self-wastage phenomenon." In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.