• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.795-801
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• 1998

Metallic fatigue of Pressurized Water Reactor(PWR) materials is a generic safety issue for commercial nuclear power plants. It is very important to obtain the fatigue usage factor for component integrity and life extension. In this paper, fatigue usage was obtained at the inside surface of Kori unit 2, 3 and 4 RCP casing weld, based on the design transient. And it was intended to establish the procedure and the detailed method of fatigue evaluation in accordance with ASME Section III Code. According to this code rule, two methods to determine the stress cycle and the number of cycles could be applied. One method is the superposition of cycles of various design transients and the other is based on the assumption that a stress cycle correspond to only one design transient. Both method showed almost same fatigue usage in the RCP casing weld.

• Journal of the Korean Society for Nondestructive Testing
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• v.9 no.1
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• pp.106-110
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• 1989

Bolting degradation problems in primary coolant pressure boundary applications have become a major concern in the nuclear industry. In the bolts concerned, the failure mechanism was either corrosion wastage(loss of bolt diameter) or stress-corrosion cracking.(3) Here the manual ultrasonic testing of RPV(Reactor Pressure Vessel) and RCP(Reactor Coolant Pump) stud has been performed. But it is difficult to detect indications because examiner can not exactly control the rotation angle and can not distinguish the indication from signals of bolt. In many cases, the critical sizes of damage depth are very small(1-2 mm order). At critical size, the crack tends to propagatecompletly through the bolt under stress, Resulting in total fracture.(3) Automatic stud scanner for studs(bolts) was developed because the precise measurement of bolt diameter is required in this circumstance. By use of this scanner, the rotation angle of probe was exactly controlled and the exposure time of radiations was reduced.

• Plant Journal
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• v.12 no.4
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• pp.37-43
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• 2016

Power system which provides electricity to the accident mitigation load for nuclear power plant should be verified to maintain the proper voltage level under the various loading and source conditions. For this purpose, it was needed to collect the voltage data of safety related buses during operation of the Reactor Coolant Pump(RCP) motor and Component Cooling Water Pump(CCWP) motor, respectively, under the certain loading condition of the plant. The data (such as, voltage, current, power factor) collected from actual measurement were used to modify the existing ETAP model and then the reanalysis was conducted to simulate the testing conditions. Through these actual measurement and analysis, it ensures that the existing electrical system analysis including assumptions and methods was conducted properly. Finally, the voltage of safety related buses was not dropped below the acceptable level, and the discrepancy between two results was within the limit.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Korean Journal of Materials Research
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• v.17 no.7
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• pp.366-370
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• 2007

Silicon nitride ($Si_3N_4$) ceramics have been considered for various components of nuclear power plants such as the mechanical seal of a reactor coolant pump (RCP), the guide roller for a control rod drive mechanism (CRDM), and a seal support, etc. Corrosion behavior of $Si_3N_4$ ceramics in a high-temperature and high-pressure water must be elucidated before they can be considered as components for nuclear power plants. In this study, the corrosion behaviors of $Si_3N_4$ ceramics containing MgO and $Al_2O_3$ as sintering aids were investigated at a hydrothermal condition ($300^{\circ}C$, 9.0 MPa) in pure water and 35 ppm LiOH solution. The corrosion reactions were controlled by a diffusion of the reactive species and/or products through the corroded layer. The grain-boundary phase was preferentially corroded in pure water whereas the $Si_3N_4$ grain seemed to be corroded at a similar rate to the grain-boundary phase in LiOH solution. Flexural strengths of the $Si_3N_4$ ceramics were significantly degraded due to the corrosion reaction. Results of this study imply that a variation of the sintering aids and/or a control (e.g., crystallization) of the grain-boundary phase are necessary to increase the corrosion resistance of $Si_3N_4$ ceramics in a high-temperature water.