• Journal of Radiation Protection and Research
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• v.28 no.2
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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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• v.36 no.6
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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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• v.38 no.4
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• pp.237-241
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• 2013

The gamma energy distributions at the major working places during refueling outage of Korean PWR nuclear power plants were measured. In order to estimate the dose reduction effect of a lead vest, Monte Carlo calculation method was used. For the simulations, the MIRD-V phantom with a lead vest was formed and exposed to the measured radiation field. The average measured gamma energy is lower than that of standard which is generally applied to radiation protection procedures. For the efficient use of a lead vest and achievement of radiation protection purpose, it is necessary to estimate the energy distribution of radiation field at working places.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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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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• v.35 no.6
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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.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.8
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• pp.3617-3624
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• 2013

Screen inspection system applied vision and laser scan technology which detect foreign materials caused fuel rod to be damaged, and which inspect fuel rod damage, bowing, distortion and grid damages, was developed to secure reliability and reproductivity of inspection method for nuclear fuel assembly during outage. In further, datum of inspection results will be continuously monitored and given understand the pattern of bowing and distorting for fuel assembly in reactor. Understanding of the pattern will be key technical information to avoid grid demage might be happened during refueling outage and provides important data base for safe operation of nuclear power plant in Korea and world wide.

• Proceedings of the KIEE Conference
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• 2005.10b
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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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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• v.38 no.4
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• pp.242-245
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• 2013

Monte Carlo method was applied to discriminate the external contamination on radiation workers in nuclear power plants for internal dose assessment generally used with a bed type scanning detector whole body counter. Korean voxel model with internal contamination was used to estimate the detection patterns of whole body scanning. Also, the BOMAB model with various external contamination was assumed to compare with detection of radionuclides inside the human body. From the comparison of detection efficiency between front and back side up, external contamination was easily distinguished.