• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1307-1313
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• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.866-872
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• 2024

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.842-850
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• 2021

Real Time Simulation (RTS) has long been used in the nuclear power industry for operator training and engineering purposes. And, Online Simulation (OLS) is based on RTS and with connection to the plant information system to acquire the measurement data in real time for calibrating the simulation models and following plant operation, for the purposes of analyzing plant events and providing indicative signs of malfunctioning. An OLS system has been developed to support PWR operations for CPR1000 plants. The OLS system provides graphical user interface (GUI) for operators to monitor critical plant operations for preventing faulty operation or analyzing plant events. Functionalities of the OLS system are depicted through the maneuvering of the GUI for various OLS functional modules in the system.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.254-259
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• 1995

Large scale display panel(LSDP) is a main component in the next generation main control rooms. LSDP is located at the front of VDU-based operator's workstation and plays an important role in providing operators with overall information of plant status through mimic diagram, text/digit, graph, and so on. A critical matter determined at the first stage of LSDP design is how much information is displayed, because the information density of LSDP affects operator's performance. Many human factors guidelines recommend low information density of displays to avoid degrade of operator's performance, but doesn't provide a useful limit of information density. In this paper, we considered information density as the number of plant parameters and investigated the proper number of plant parameters through a human factors experiment. The experiment with 4 subjects was carried out and response time, error, and heart rate variation as criterion measures were recorded and analyzed. As the results, it is identified that the proper number of parameters in a LSDP is about thirty.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.127-138
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• 2005

Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.

• Proceedings of the ESK Conference
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• 1995.04a
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• pp.30-34
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• 1995

The operator's performance of nuclear power plants is affected by many performance shaping factors(PSF). The objective of this study is to find out the PSFs and their effect on the nuclear power plant operations. We extracted PSFs in five category, and identified the relationships between PSFs and performance using the four survey methods; literture survey, case study, video task analysis and questionnaire survey. Finally the knowledge on PSFs and their effect was represented as rule form for cognitive simulation.

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.368-374
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• 1997

This paper presents a new approach to the evaluation of an accident management strategy when an operator action is involved. This approach classifies the failure in implementing a given strategy into 4 possible mechanisms, and provides their corresponding quantification methods : 1) the failure to formulate correct intention by operators, 2) the failure to take an adequate action following a correct diagnosis, 3) the failure of a system operation following an adequate action, and 4) the failure due to a delayed action. The proposed method was applied to assess a cavity flooding strategy that uses containment spray system (CSS), and the result shows that the method is more appropriate in evaluating accident management strategies when human action is involved.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.141-150
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• 2013

There is mounting evidence that complex computer system displays in control rooms contribute to cognitive complexity and, thus, to the probability of human error. Research shows that reaction time increases and response accuracy decreases as the number of elements in the display screen increase. However, in terms of supporting the control room operator, approaches focusing on addressing display complexity solely in terms of information density and its location and patterning, will fall short of delivering a properly designed interface. This paper argues that information complexity and semantic complexity are mandatory components when considering display complexity and that the addition of these concepts assists in understanding and resolving differences between designers and the preferences and performance of operators. This paper concludes that a number of simplified methods, when combined, can be used to estimate the impact that a particular display may have on the operator's ability to perform a function accurately and effectively. We present a mixed qualitative and quantitative approach and a method for complexity estimation.

• Nuclear Engineering and Technology
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• v.50 no.4
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• pp.613-618
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• 2018

We are developing a shared remote control mobile robot for aerial work in nuclear power plants (NPPs); a robot consists of a mobile platform, a telescopic mast, and a dual-arm slave with a working tool. It is used at a high location operating the manual operation mechanism of a fuel changer of a heavy water NPP. The robot system can cut/weld a pipe remotely in the case of an emergency or during the dismantling of the NPP. Owing to the challenging control mission considering limited human operator cognitive capability, some remote tasks require a shared control scheme, which demands systematic software design and integration. Therefore, we designed the architecture of the software systematically.