• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.33-38
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• 시스템엔지니어링학술지
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• 제12권1호
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• pp.89-103
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• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.820-830
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• 2016

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.13-19
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• 2015

Steam generator is one of the most important component of nuclear power plant, and it's integrity and reliability are to be assured to high level by pre-service inspection and in-service inspection. To improve the reliability of steam generator heat exchanger tubes and to secure the management of nuclear power plant safely, KHNP CRI recently has developed eddy current testing system for steam generator. KHNP CRI have performed a series of experimental verification and field application to confirm the performance of the developed ECT system in accordance with ASME Code requirements. The ECT system consists of a remote data acquisition unit, an ECT signal acquisition and analysis software, a water chamber robot controller and a probe push-puller. In this paper, we will details of the developed ECT system and the software and their experimental performance. And also we will report the field applying performance and the issues for further steps.

• 한국압력기기공학회 논문집
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• 제7권4호
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.49-58
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• 2018

The regulatory periodic inspection program (PSI) conducted at every overhaul period is the most important process for confirming the safety of nuclear power plants. The PSI for operating nuclear power plants in Korea mainly consist of component level performance check that had been developed based on deterministic approach putting the same degree of importance to all the inspection items. This inspection methodology is likely to be effective for preoperational inspection. However, once the plant is put into service, the PSI must be focused on whether to minimize the risk of accident using defense-in-depth concept and risk insight. The incorporation of defense-in-depth concept and risk insight into the deterministic based safety inspection has not been well studied so far. In this study, two track approaches are proposed to make sure that core damage be avoided: one is to secure success path and the other to block the failure path in a specific event tree of PSA. The investigation shows how to select safety important components and how to set up inspection group to ensure that core damage would not occur for a given initiating event, which results in strengthening defense-in-depth level 3.

• 한국산업융합학회 논문집
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• 제25권5호
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• pp.827-833
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• 2022

In this study, the prediction of the bending angle for the 350 A bellows-type expansion joints and the structural stability according to the load were determined. The stability of the 2km piping system was predicted by applying the allowable bending angle of the expansion pipe joint obtained from the analysis. The maximum bending angle was calculated through bending analysis of the bellows-type expansion joints, and the maximum bending angle by numerical calculation was about 1.8°, and the maximum bending angle of the bellows obtained by comparing the allowable strength of the material was about 0. 22°. This angle was very stable compared to the allowable bending angle (3°) of the expansion pipe joint regulation. By applying the maximum bending angle, the allowable maximum deflection of the 2 km pipe was about 3.8 m. When the seismic load was considered using regression analysis, the maximum deflection of the 2km pipe was about 142.3mm, and it was confirmed that the bellows-type expansion joints and the deflection were stable compared to the allowable maximum deflection of the pipe system. These research results are expected to present design and analysis guidelines for the construction of piping and the development of bellows systems, and to be used as basic data for systematic research.

• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.109-118
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• 2021

In the periodic surveillance material test for the spacer component of fuel channel assembly in CANDU, a microstructural characterization analysis is required in addition to the mechanical property evaluation test. In this study, detailed microstructure analysis and simple mechanical property evaluation of archive spacer parts were conducted to indirectly support the surveillance test and assist in the study of spacer material degradation. We investigated the microstructural characteristics of the spacer garter spring coil through comparative analysis with the plate material. The main microstructure characteristics of the garter spring coil X-750 are represented by the fine grain size distribution, the ordering phase distribution developed inside the matrix, the high dislocation density inside the grains, and the arrangement of coarse carbides. In addition, the yield strength of the garter spring coil X-750 was indirectly evaluated to be approximately 1 GPa. We also established an analytical method to elucidate the microstructural evolution of the radioactive spacer garter spring coil X-750 based on Canadian research experiences. Finally, we confirmed the measurement technique for helium bubble formation through TEM examination on the helium implanted X-750 material.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2009년 추계학술발표대회
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• pp.100-100
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• 2009

본 논문에서는 원자력발전소 1차 계통의 스테인리스강 저합금강 이종금속용접부 및 스테인리스강 동종용접부의 잔류응력을 평가하고 스테인리스강 용접부의 응력부식균열 민감성에 대해 고찰하였다. 노즐 안전단의 이종금속용접부 및 안전단 배관의 동종용접부 제작 및 소재가공에 의행 생성되는 잔류응력을 예측하기 위해 열 탄소성 유한요소법 수치해석을 수행하였으며, 용접공정과 함께 표면의 잔류응력에 기여하는 절삭 및 연삭가공과 소재의 담금질 공정을 열 탄소성적으로 모사하였다. 전산해석 결과, 스테인리스주강의 담금질 잔류응력은 무시할 수 없는 상당한 크기이므로 배관 용접잔류응력 평가 시 소재의 담금질 효과를 고려해야 할 것으로 판단된다. 이종금속 용접과 동종금속 용접공정이 보수용접 없이 정상적인 절차(내면에서 외면으로 적층)로 완성된다면, 냉각재 환경에 노출되는 용접부 내면의 잔류응력은 재료의 응력부식균열 민감성에 영향을 주지 않을 것으로 판단된다. 한편, 안전단 배관 동종용접부의 연삭가공에 의해 내면의 잔류응력이 크게 상승하는 것으로 예측되었으므로, 내면의 연삭가공 이후 표면잔류응력 완화처리(예, 버핑)가 필요하다.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.297-310
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• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.