격납건물 국부누설률시험 표준절차 개발

문용식,김창수,Moon, Yong-Sig,Kim, Chang-Soo 한국압력기기공학회 2012 한국압력기기공학회 논문집 Vol.8 No.2

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

회전코일 와전류신호를 이용한 증기발생기 곡관형 튜브의 축방향노치 신호의 특성

김창수,문용식,Kim, Chang-Soo,Moon, Yong-Sig 한국압력기기공학회 2012 한국압력기기공학회 논문집 Vol.8 No.3

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

원자력발전소 초음파검사자 기량검증시험 결과 검토

정남두,문용식,이승표,Jung, Nam-Du,Moon, Yong-Sig,Lee, Seung-Pyo 한국비파괴검사학회 2014 한국비파괴검사학회지 Vol.29 No.1

본 논문에서는 국내 원자력발전소 초음파검사자 기량검증시험 결과를 검토하였다. 원자력발전소 안전관련 설비에 대한 초음파검사 기량검증 요건은 ASME B&PV Code Section XI 1989 겨울 부록에서 최초로 언급되었다. 한국수력원자력(주)은 ASME Code Section XI, Appendix VIII의 요건 적용을 위해 국내 원자력발전소에 적합한 한국형 초음파검사 기량검증(KPD) 시스템을 개발, 규제기관의 승인을 받아 2004년 7월부터 운영하고 있다. 본 논문에서는 2004년부터 2013년까지 수행한 초음파검사자 기량검증시험 결과의 검토 내용으로 초음파검사 결함 검출 향상 및 기량훈련시스템 연구 개발에 도움을 줄 것으로 기대한다. In this paper, the result of an ultrasonic performance demonstration are analyzed. The requirements for an ultrasonic performance demonstration (PD) for a nuclear power plant were first described in ASME B&P Code Section XI, Appendix VIII (1989 winter addenda). In order to establish the performance demonstration scheme in Korean nuclear power plants, the Korea Hydro & Nuclear Power Co. Ltd (KHNP) has developed the Korean Performance Demonstration (KPD) system for the for the ultrasonic examination of nuclear power plants. An analysis of the ultrasonic performance demonstration results from 2004 through 2013 will improve the detection of flaws in an ultrasonic examination, as well as the further development of the KPD training system.

불규칙 표면 시편을 이용한 Flexible 위상배열초음파기술 적용 연구

이승표,문용식,정남두,Lee, Seung-Pyo,Moon, Yong-Sig,Jung, Nam-Du 한국비파괴검사학회 2015 한국비파괴검사학회지 Vol.15 No.1

원자력발전소에는 탄소강과 스테인리스강 용접을 위해 alloy600 용접재를 적용한 이종금속용접부가 다수 존재하며, 전 세계적으로 이종금속용접부에서 결함 발생 보고가 지속되고 있다. 주기적인 건전성 평가를 위해 이종금속용접부 초음파검사 일반절차서 (KPD-UT-10)를 적용하여 검사를 수행하고 있으며, 검사절차서에서는 탐촉자와 검사체 표면 사이의 간격을 최대 1/32"(0.8mm) 이내로 제한하고 있다. 국내의 일부 이종금속용접부는 테이퍼진 형상과 불규칙한 표면 형상을 가지고 있어, 가변형 위상배열초음파기술을 적용하여 검사 신뢰성을 높이고자 본 연구를 수행하였다. 연구 결과, 검사체 표면이 불규칙한 시편 내부의 인공결함에 대한 검출이 양호하였고, 이를 통해, 가변형 위상배열초음파기술의 현장 적용 가능성을 확인하였다. Nuclear power plant contain many dissimilar metal welds that connect carbon steel components with stainless steel pipes using alloy600 welding materials. Primary water stress corrosion cracks at dissimilar metal welds have been continuously reported around the world. In periodic integrity evaluations, dissimilar metal welds are examined using a generic ultrasonic testing procedure, KPD-UT-10. In this procedure, the gap between the probe and examination surface is limited to 1/32 inch (0.8mm). It is not easy to test some dissimilar metal welds in Korean plants applying ordinary technology because of their tapered shapes and irregular surface conditions. This paper introduces a method for applying a flexible phased array technology to improve the reliability of ultrasonic testing results for various shapes and surface conditions. The artificial flaws in specimens with irregular surfaces were completely detected using the flexible phased array ultrasonic technology. Therefore, it can be said that the technology is applicable to field examination.

격납건물 ILRT 본시험시간이 시험에 미치는 영향에 관한 연구

김창수,문용식,Kim, Chang-Soo,Moon, Yong-Sig 한국압력기기공학회 2012 한국압력기기공학회 논문집 Vol.8 No.3

The containment Integrated Leakage Rate Testing(ILRT) of nuclear power plants in Korea is performed in accordance with NSSC(Nuclear Safety and Security Commission) code 2012-16 and ANSI/ANS 56.8-1994. Nuclear power plants in Korea and the United States are to apply same test criteria, ANSI/ANS 56.8-1994, except type A testing time. NPPs in Korea apply 24 hours according to NSSC code 2012-16, but NPPs in United States apply 8 hours according to 10CFR50 App. J for type A test. So, there are many difficulties in order to perform ILRT in Korea. In this study, I review the impact on the ILRT results and the effect of ILRT due to type A testing time. The future, we will continue study to enhance the test reliability and improve these problems.

SG전열관 2차측 이물질 검출 및 특성분석을 위한 ETSS 개발

신기석,문용식,민경만,Shin, Ki Seok,Moon, Yong Sig,Min, Kyong Mahn 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.3

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

가압기 히터슬리브 용접부 PWSCC 검출을 위한 유도초음파 특성 평가

주경문,문용식,정우근,Joo, Kyung-Mun,Moon, Yong-Sig,Chung, Woo-Geun 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.2

Although many defects in PZR heater sleeve have been reported continually from operating experiences in oversea nuclear power plant, utilities get into difficulties in finding appropriate methods for diagnostics of the components due to the limited access or high radiation problems. Recently, as an alternative, diagnostics using Guided Wave Testing(GWT) are proposed and the attention of the methods has been growing gradually because of their long range inspection capability. This study is to investigate the effectiveness of GWT to detect PWSCC in welding points of PZR heater sleeve. Moreover, mode sensitivity analysis of GWT and optimal frequency for the diagnostics of PWSCC are presented by testing the mock-ups specimens that contain artificial flaws.

중·저준위 방사성폐기물 고화체의 압축강도 평가를 위한 초음파속도 측정

문균영,이태훈,문용식,Moon, Gyoon Young,Lee, Tae Hun,Moon, Yong Sig 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.4

In order to ship low & Intermediate level radioactive waste drums, which have been temporarily stored on site, to a disposal facility, their physical and chemical properties should be evaluated and proven to meet the acceptance guideline prior to their shipment. Ultrasonic velocity method, which has been used to estimate the strength of concrete, can be suggested to evaluate the compressive strength of solidified radioactive waste, which is one of the evaluated properties. The strength is estimated from acoustic velocity. However, a guided wave traveling along a drum is generated when applying ultrasonic method to the drum, and this makes it difficult to analyze the signal due to overlap between transmitted wave through the contents in drum and the guided wave. This paper reported feasibility of ultrasonic method to evaluate of the compressive strength of the solidified LLW. It is observed that the guide wave is greater than transmitted wave, and ultrasonic velocity could be estimated from transmitted wave signal arriving prior to the guided wave

리스크 정보를 활용한 배관 가동중검사 적용

진영복,진석홍,문용식,Jin, Young Bok,Jin, Seuk Hong,Moon, Yong Sig 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.4

Pressurized Water Reactor Owners Group(PWROG) proposed and applied a risk-informed inservice inspection(RI-ISI) program to alternate existing ASME Section XI periodic inspections. The RI-ISI programs enhance overall safety by focusing inspections of piping at high safety significant(HSS) and locations where failure mechanisms are likely to be present, and by improving the effectiveness on inspection of components because the examination methods are based on the postulated failure mode and the configuration of the piping structural element. The RI-ISI programs can reduce NDE, man-rem exposure, costs of engineering analysis, outage duration and chance of complicating plant operations etc. RI-ISI methods of piping inservice inspection were applied on 3 units(KSNP : Korea Standard Nuclear Power Plant) and are scheduled to apply on the other units. In this paper, we compared and showed the results of the 2 units and we concluded that the RI-ISI application could enhance and maintain plant safety and give unquantifiable benefits.

증기발생기 전열관 와전류검사용 국내 개발 보빈탐촉자 적용성 분석

김인철,주경문,문용식,Kim, In-Chul,Joo, Kyung-Mun,Moon, Yong-Sig 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.4

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.