• Journal of Radiation Protection and Research
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• 제28권2호
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.351-356
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• 1997

In the des19n of KNGR (Korea Next Generation Reactor), we believe that economy is one of the most important factors to be considered Thus, we reviewed and evaluated the consequences of designing the head area into an integrated package from an economical point of view. The refueling outage durations of the nuclear power plants currently in operation In Korea, some having and others not having integrated head package, are compared. This paper discusses the characteristics of head area design and the critical design issues of KNGR head area to evaluate the effect of the head area characteristics on the outage duration.

• Journal of Radiation Protection and Research
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• 제38권4호
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• pp.237-241
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• 2013

국내 가압경수로형 원전 계획예방정비기간에 수행되는 주요 방사선작업에 대한 감마선 에너지 분포를 측정하였다. 고방사선구역 작업시 종사자가 착용하는 대표적인 납 차폐복에 대하여 감마선 에너지 분포에 따른 차폐효과를 평가하기 위해 전산모사 방법을 이용하였다. 전산모사는 MIRD형 인체모형에 추가적으로 납 차폐복을 모델링하고 측정된 감마선 에너지 정보를 이용하여 수행하였다. 주요 방사선작업의 평균 감마선 에너지는 일반적으로 방사선방호 과정에서 적용되는 기준 방사선에너지 보다 낮은 것으로 평가되었다. 방사선 방호 목적을 달성하기 위한 효율적인 납 차폐복 착용을 위해 방사선작업 지역의 방사선에너지 분포평가의 필요성을 확인하였다.

• 시스템엔지니어링학술지
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• 제12권1호
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• pp.25-40
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• 2016

The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. In this study, the SE methodology is used to design a nondestructive inspection system for Bottom Mounted Instrumentation (BMI) nozzles. We developed a system that enables nondestructive inspection of BMI nozzles during regular refueling outage without removing the reactor internals. A special ultrasonic (UT) probe is introduced to scan and detect cracks within the weld region of the nozzle. A 3D model of the inspection structure system was developed along with the reactor pressure vessel (RPV) and internals which permits a virtual 3D simulation of the operation to check the design concept and effectiveness of the system and to provide a good visualization of the system. This approach allows for a virtual walk through to verify the proposed BMI nozzle inspection system.

• Nuclear Engineering and Technology
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• 제35권6호
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• pp.547-555
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• 2003

Fire hazard and risk analysis at Nuclear Power Plants is implemented on the basis of the normal operational configuration. This steady configuration, however, can be changed due to the temporary displacement of equipment, electric cable and irregular movement of workers through the fire compartments when the on-line maintenance is processed during the power operation mode or the scheduled outage mode for the refueling. With the consequence of this configuration change, the fire analysis condition and the evaluation result will be different from those that were analyzed based on the steady configuration. In this context, at this paper, the general items for the reassessment are categorized when the configuration has changed. The contemporary zone models for the detail fire analysis are also illustrated for their application for each classified condition.

• 한국산학기술학회논문지
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• 제14권8호
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• pp.3617-3624
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• 2013

원전 연료집합체의 연료봉 내에 잔존하여 연료봉의 손상을 발생시킬 수 있는 이물질의 잔존 여부 및 연료봉의 손상, 연료봉의 휨, 뒤틀림, 그리드 손상여부를 비젼기술과 레이저 스캔 기술을 응용한 원전 연료집합체 스크린 검사 방법을 개발하여 계획예방 정비 기간 중 검사가 가능하도록 연료집합체 검사의 신뢰성과 생산성을 확보하였다. 또한 검사 데이터를 집계, 분석하여, 연료집합체의 변형 상태를 지속적으로 감시함으로써, 국내 각 원자로별 노심 내 연료변형 패턴을 이해할 수 있다. 이는 연료 재장전 도중 발생 가능한 그리드 손상을 방지하는데 기술정보로 활용되어 국내외 원전 안전 운영의 중요한 데이터베이스를 제공하게 된다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.225-227
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• 2005

The in-core neutron Flux Mapping System in a pressurized water reactor yields information on the neutron flux distribution in the reactor core at selected core locations by means of movable detectors. The obtained data are used to verify the reactor core design parameters. The detector cables run through guide tubes(thimbles), and typically thirty-six to fifty-eight thimbles are allocated in the reactor depending on the number of fuel assemblies. These thimbles are inserted into nuclear fuel assemblies through conduits connected from the bottom of the reactor vessel to a seal table. During the plant refueling outage period, the thimbles are withdrawn up to 4m from the seal table, the height of a nuclear fuel. In spite of their importance, however, the thimble handling work has been performed by only human operators. In addition, its efficiency is very low due to narrow working environments on the seal table, thereby resulting in the excessive radiation exposure of maintenance personnel. To solve these problems, a new thimble handling equipment for in-core flux mapping system was developed, and we confirmed its effectiveness through experiments.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• Journal of Radiation Protection and Research
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• 제38권4호
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• pp.242-245
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• 2013

원전 방사선작업종사자의 내부선량평가를 위해 일반적으로 사용되고 있는 bed type 전신계측에 몬테카를로 방법을 적용하여 작업자에 대한 외부오염 계측특성을 평가하였다. 한국인 체형을 반영하는 voxel 모의 피폭체를 이용하여 내부오염시 측정 특성을 평가하였다. 외부오염 판별을 위해 BOMAB 모의 피폭체를 이용하여 신체 각 부위별 오염시 나타나는 측정 특성을 평가하였다. 가슴 부위 오염시 누운 자세와 엎드린 자세로 구분하여 외부오염시 계측 특성을 확인하였다. 정량적인 분석을 통해 bed type 전신계측기를 이용한 외부오염 판별이 가능함을 확인하였다.