• Journal of the Earthquake Engineering Society of Korea
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• v.23 no.3
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• pp.149-157
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• 2019

The purpose of this study is to investigate the effects of the significant duration of ground motions on responses of base-isolated nuclear power plants (NPPs). Two sets of ground motion records with short duration (SD) and long duration (LD) motions, scaled to match the target response spectrum, are used to perform time-history analyses. The reactor containment building in the Advanced Power Reactor 1400 (APR1400) NPP is numerically modeled using lumped-mass stick elements in SAP2000. Seismic responses of the base-isolated NPP are monitored in forms of lateral displacements, shear forces, floor response spectra of the containment building, and hysteretic energy of the lead rubber bearing (LRB). Fragility curves for different limit states, which are defined based on the shear deformation of the base isolator, are developed. The numerical results reveal that the average seismic responses of base-isolated NPP under SD and LD motion sets were shown to be mostly identical. For PGA larger than 0.4g, the mean deformation of LRB for LD motions was bigger than that for SD ones due to a higher hysteretic energy of LRB produced in LD shakings. Under LD motions, median parameters of fragility functions for three limit states were reduced by 12% to 15% compared to that due to SD motions. This clearly indicates that it is important to select ground motions with both SD and LD proportionally in the seismic evaluation of NPP structures.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2015.05a
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• pp.264-266
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• 2015

Currently as cyber threats grow up targeting nuclear power plants(NPP), licensees must guarantee that computer and information systems of nuclear facilities can be adequately protected against cyber attack. Especially critical system that cause illegal transfer of nuclear material and adverse impact to public safety need protecting. In this paper, we surveying the cyber threat examples targeted at NPP, and taxonomy the method of cyber security for NPPs in korea through analyzing the methodology to identify critical system and address cyber security controls for nuclear facilities.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.125-134
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• 2009

This paper presents details on a tube inspection robotic system and a positioning method of the robot for a steam generator (SG) in nuclear power plants (NPPs). The robotic system is separated into three parts for easy handling, which reduces the radiation exposure during installation. The system has a supporting leg to increase the rigidity of the robot base. Since there are several thousands of tubes to be inspected inside a SG, it is very important to position the tool of the robot at the right tubes even if the robot base is positioned inaccurately during the installation. In order to obtain absolute accuracy of a position, the robot kinematics was mathematically modeled with the modified DH(Denavit-Hartenberg) model and calibrated on site using tube holes as calibration points. To tune the PID gains of a commercial motor driver systematically, the time delay control (TDC) based gain tuning method was adopted. To verify the performance of the robotic system, experiments on a Framatomes 51B Model type SG mockup were undertaken.

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.117-121
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• 2016

Background: Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Materials and Methods: Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. Results and Discussion: The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. Conclusion: The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1686-1697
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• 2022

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. Accordingly, HRA community has emphasized the accumulation of HRA data to support HRA practitioners for many decades. To this end, it is critical to resolve practical problems including (but not limited to): (1) how to collect HRA data from available information sources, and (2) how to inform HRA practitioners with the collected HRA data. In this regard, the U.S. Nuclear Regulatory Commission (NRC) and Korea Atomic Energy Research Institute (KAERI) independently initiated two large projects to accumulate HRA data by using full-scale simulators (i.e., simulator data). In terms of resolving the first practical problem, the NRC and KAERI developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. In addition, to inform HRA practitioners, the NRC and KAERI proposed several ideas to extract useful information from simulator data. This paper is the first of two papers to discuss the technical underpinnings of the development of the SACADA and HuREX systems.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• Smart Structures and Systems
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• v.9 no.5
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• pp.461-472
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• 2012

This study proposes an optics-based active sensing system for continuous monitoring of underground pipelines in nuclear power plants (NPPs). The proposed system generates and measures guided waves using a single laser source and optical cables. First, a tunable laser is used as a common power source for guided wave generation and sensing. This source laser beam is transmitted through an optical fiber, and the fiber is split into two. One of them is used to actuate macro fiber composite (MFC) transducers for guided wave generation, and the other optical fiber is used with fiber Bragg grating (FBG) sensors to measure guided wave responses. The MFC transducers placed along a circumferential direction of a pipe at one end generate longitudinal and flexural modes, and the corresponding responses are measured using FBG sensors instrumented in the same configuration at the other end. The generated guided waves interact with a defect, and this interaction causes changes in response signals. Then, a damage-sensitive feature is extracted from the response signals using the axi-symmetry nature of the measured pitch-catch signals. The feasibility of the proposed system has been examined through a laboratory experiment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.419-427
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• 2019

The application of ion crystallization technology was considered as a way to increase the operating efficiency and improve the operating performance of a liquid waste management system (LWMS) in the Advanced Power Reactor 1400 (APR 1400). Although ion crystallization technology has not been practically applied to Nuclear Power Plants (NPPs) until now, a previous experimental study demonstrated that it is possible to selectively remove at least 95% of various nuclide ions present in the liquid radioactive waste of NPPs. We reviewed the possibility of applying ion crystallization technology to the existing LWMS by applying the nuclide removal rate of ion crystallization technology and prepared a way to improve the existing LWMS in the APR 1400. Furthermore, we determined the optimized application location of ion crystallization technology in the existing LWMS by considering decontamination characteristics of the ion crystallization technology and the existing LWMS design features and operating experiences. The application of ion crystallization technology to the liquid waste collection tank, where liquid radioactive materials are collected, will have the least impact on the existing design while providing the greatest improvement. It is expected that the application of ion crystallization technology to the current APR 1400 or new NPPs would increase the operating efficiency of the LWMS and result in an improvement of system performance.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.35 no.2
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• pp.21-29
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• 2012

Person characteristics and the perception of organization climate affect on the performances associated with causes of job stress, and may induce various types of human errors. For incidence, a person's adaptation to organizational traits can influence on a person's performance with tasks assigned as well as job satisfaction, a change of occupation, etc. There are several evaluation methods such as an aptitude test to evaluate the suitability of department allocation. However, only person's traits such as an aptitude has been associated with department allocation and job assignment for the personnel management. This paper shows not only an evaluation result on the job-stress of the workers in nuclear power plants (NPPs), but also the relationships between the job-stress and the coincidences of group-personal, team-personal traits. Then, we systematically deduct a basic information on the factors to be considered to manage the organizational traits and job stress. We expect this result can contribute to enhance the organizational management against to the human errors as well as for the promotion of safety and efficiency of NPPs.