• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.11-21
• /
• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.3990-4002
• /
• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Journal of the Ergonomics Society of Korea
• /
• v.32 no.1
• /
• pp.37-45
• /
• 2013

Objective: This study describes the change of operator performance in drastic change of illumination level, and proposes an alternative method to cope with it. Background: The control standard of illumination for nuclear power plants(NPPs) is based on the set of limit criteria for maintaining a specific illumination level. However, there is a possibility to cause human errors according to the psychological and physiological influences to operators in the situation of drastic change of illumination such as SBO(Station Black Out), so a basic study is necessary to review the current approach. Method: We assessed the visual fatigue, subjective work load and task performance according to the three illumination situations(Normal Illumination, Emergency Illumination, and Drastic Change of Illumination). Result: Research finding shows that there are not significant differences in task performance between normal illumination (1,000lx level) and emergency illumination (100lx level), only if beyond the dark adaptation limit. However, subjective work load on mental demand and visual fatigue show a potential challenge to visual performance in drastic change of illumination. Conclusion/Application: Several trials can complement this challenge in NPPs by applying 3-way communication, enhancing readability of procedures, and managing the visual factors affecting the operators' performance through a Visual Environment Management Program including visual health aspects, etc.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.13 no.3
• /
• pp.42-48
• /
• 2014

The auxiliary-feed-water pump (AFWP) used to supply water during a station black out situation at nuclear power plants should meet the seismic qualification regulations stipulated in IEEE Std 323 and 344, so as to withstand earthquakes or dangerous situations. Here, we establish a model for the estimation of the structural integrity of this type of pump. If the natural frequency that results from a modal analysis is less than 33 Hz, we adopt a dynamic analysis, instead of a static analysis. A dynamic analysis was carried out taking into consideration seismic conditions such as the floor response spectra (FRS), an operation-base earthquake (OBE), and a safe-shutdown earthquake (SSE). Finally, an analytical estimation of the structural integrity of an AFWP is made through a comparison of calculated values and allowable values. If the result is less than the allowable stress, the pump is deemed to have good structural integrity. In addition, future studies will involve a stability check for rotor accidents that may occur during the operation of the pump.

• 제어로봇시스템학회:학술대회논문집
• /
• 1997.10a
• /
• pp.179-182
• /
• 1997

In a training simulator for power plant, operator's action in the MCR(Main Control Room) are given to plant process and computer system model as an inputs, and the same response as in real power plant is provided in real time. Inter-process communication and synchronization are especially important among various inputs. In the plant simulator, to simulate the digital control system such as FOXBORO SPEC-200 Micro control system, modification and adaptation of control card(CCC) and its continuous display station(CDS) is necessary. This paper describes the modeling and simulation of FOXBORO SPEC-200 Micro control system applied to Younggwang nuclear power plant unit #3 & 4, and its integration process to the full-scope replica type training simulator. In a simulator, display station like CDS of FOXBORO SPEC-200 Micro control system is classified as ITI(Intelligent Type Instrument), which has a micro processor inside to process information and the corresponding alphanumeric display, and the stimulation of ITI limits the important functions in a training simulator such as backtrack, replay, freeze and IC reset. Therefore, to achieve the better performance of the simulator, modification of CDS and special firmware is developed to simulate the FOXBORO SPEC-200 Micro control system. Each control function inside control card is modeled and simulated in generic approach to accept the plant data and control parameter conveniently, and debugging algorithms are applied for massive coding developed in short period.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
• /
• 2002.05a
• /
• pp.314-317
• /
• 2002

This paper present a methods for performance evaluation of digital governor and we treat the way to reduce the trial and error when the developed digital governing system is applied to station. The parameters of the model are estimated using input-output data, the model adjustment technique and a genetic algorithm(GA). To verify validity of the completed model, the comparision model and real output signals have been carried out. The developed digital governing system is applied to Sumjingang hydro-power plant, Korea Hydro Nuclear Power Corporation, it has replaced mechanical governing system. Test was performed in order to demonstrate the performance evaluation of the digital governor parameters.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.191-196
• /
• 1995

The intelligent human-machine interface (HMI) has been developed to enhance the safety and availability of a nuclear power plant by improving operational reliability The key elements of the HMI are the large display panels which present synopsis of the plant status and the compact, digital work stations for the primary operator control and monitoring functions. The work station consists of four consoles such as a dynamic alarm console (DAC), a system information console (SIC), a computerized operating-procedure console (COC), and a safety related information console (SRIC). The DAC provides clean alarm pictures, in which information overlapping is excluded and alarm impacts are discriminated, for quick situation awareness. The SIC covers a normal operation by offering all necessary plant information and control functions. In addition, it is closely linked with the DAC and the COC to automatically display related system information under the request of these consoles. The COC aids the operator with proper emergency operation guidelines so as to shutdown the plant safely, and it also reduces his physical/mental burden by automating the operating procedures. The SRIC continuously displays safety related information to allow the operator to assess the plant status focusing on plant safety. The proposed HMI has been validated and demonstrated with on-line data obtained from the full-scope simulator for Yonggwang Units 1,2.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.579-585
• /
• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Proceedings of the KIEE Conference
• /
• 2000.07d
• /
• pp.2831-2833
• /
• 2000

Since the access to Station Control Computers(DCCs) is restricted to the main control room(MCR). the operating data of power plants can't be easily analyzed and effectively managed. It is possible to reduce waste of time and human energy by means of building the Remote Monitoring Network of DCCs connected to Local Area Network. automatizing collection and analysis of DCC data. gathering the operating state of power plants. and managing systematically. Furthermore. this system help preventing trip by means of analyzing the data promptly and watching main system continuously.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4335-4349
• /
• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.