Go to the main menu
Skip to content
Go to bottom
REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
Editor in Chief :
Volume & Issues
Volume 44, Issue 8 - Dec 2012
Volume 44, Issue 7 - Oct 2012
Volume 44, Issue 6 - Aug 2012
Volume 44, Issue 5 - Jun 2012
Volume 44, Issue 4 - May 2012
Volume 44, Issue 3 - Apr 2012
Volume 44, Issue 2 - Mar 2012
Volume 44, Issue 1 - Feb 2012
Selecting the target year
PREDICTION OF DIAMETRAL CREEP FOR PRESSURE TUBES OF A PRESSURIZED HEAVY WATER REACTOR USING DATA BASED MODELING
Lee, Jae-Yong ; Na, Man-Gyun ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 355~362
DOI : 10.5516/NET.04.2012.507
The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.
DESIGN OF A FPGA BASED ABWR FEEDWATER CONTROLLER
Huang, Hsuanhan ; Chou, Hwaipwu ; Lin, Chaung ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 363~368
DOI : 10.5516/NET.04.2012.508
A feedwater controller targeted for an ABWR has been implemented using a modern field programmable gate array (FPGA), and verified using the full scope simulator at Taipower's Lungmen nuclear power station. The adopted control algorithm is a rule-based fuzzy logic. Point to point validation of the FPGA circuit board has been executed using a digital pattern generator. The simulation model of the simulator was employed for verification and validation of the controller design under various plant initial conditions. The transient response and the steady state tracking ability were evaluated and showed satisfactory results. The present work has demonstrated that the FPGA based approach incorporated with a rule-based fuzzy logic control algorithm is a flexible yet feasible approach for feedwater controller design in nuclear power plant applications.
DESIGN OF A LOAD FOLLOWING CONTROLLER FOR APR+ NUCLEAR PLANTS
Lee, Sim-Won ; Kim, Jae-Hwan ; Na, Man-Gyun ; Kim, Dong-Su ; Yu, Keuk-Jong ; Kim, Han-Gon ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 369~378
DOI : 10.5516/NET.04.2012.509
A load-following operation in APR+ nuclear plants is necessary to reduce the need to adjust the boric acid concentration and to efficiently control the control rods for flexible operation. In particular, a disproportion in the axial flux distribution, which is normally caused by a load-following operation in a reactor core, causes xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. A model predictive control (MPC) method was used to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for APR+ nuclear plants. Some tracking controllers employ the current tracking command only. On the other hand, the MPC can achieve better tracking performance because it considers future commands in addition to the current tracking command. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at the current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The support vector regression (SVR) model that is used widely for function approximation problems is used to predict the future outputs based on previous inputs and outputs. In addition, a genetic algorithm is employed to minimize the objective function of a MPC control algorithm with multiple constraints. The power level and ASI are controlled by regulating the control banks and part-strength control banks together with an automatic adjustment of the boric acid concentration. The 3-dimensional MASTER code, which models APR+ nuclear plants, is interfaced to the proposed controller to confirm the performance of the controlling reactor power level and ASI. Numerical simulations showed that the proposed controller exhibits very fast tracking responses.
OPERATOR BEHAVIORS OBSERVED IN FOLLOWING EMERGENCY OPERATING PROCEDURE UNDER A SIMULATED EMERGENCY
Choi, Sun-Yeong ; Park, Jin-Kyun ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 379~386
DOI : 10.5516/NET.04.2012.510
A symptom-based procedure with a critical safety function monitoring system has been established to reduce the operator's diagnosis and cognitive burden since the Three-Mile Island (TMI) accident. However, it has been reported that a symptom-based procedure also requires an operator's cognitive efforts to cope with off-normal events. This can be caused by mismatches between a static model, an emergency operating procedure (EOP), and a dynamic process, the nature of an ongoing situation. The purpose of this study is to share the evidence of mismatches that may result in an excessive cognitive burden in conducting EOPs. For this purpose, we analyzed simulated emergency operation records and observed some operator behaviors during the EOP operation: continuous steps, improper description, parameter check at a fixed time, decision by information previously obtained, execution complexity, operation by the operator's knowledge, notes and cautions, and a foldout page. Since observations in this study are comparable to the results of an existing study, it is expected that the operational behaviors observed in this study are generic features of operators who have to cope with a dynamic situation using a static procedure.
PREJOB BRIEFING USING PROCESS DATA AND TAG-OUT / LINE-UP DATA ON 2D DRAWINGS
Dionis, Francois ; Alain, Ribiere ; Renaud, Aubin ; Romain, Catteau ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 387~392
DOI : 10.5516/NET.04.2012.511
This paper discusses the concept of merging multiple sources of data on the same 2D CAD drawing(s) for power plant operation. It also presents the concepts and tools used in this project.
MONITORING SEVERE ACCIDENTS USING AI TECHNIQUES
No, Young-Gyu ; Kim, Ju-Hyun ; Na, Man-Gyun ; Lim, Dong-Hyuk ; Ahn, Kwang-Il ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 393~404
DOI : 10.5516/NET.04.2012.512
After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.
RELIABILITY ESTIMATION FOR A DIGITAL INSTRUMENT AND CONTROL SYSTEM
Yaguang, Yang ; Russell, Sydnor ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 405~414
DOI : 10.5516/NET.04.2012.513
In this paper, we propose a reliability estimation method for DI&C systems. At the system level, a fault tree model is suggested and Boolean algebra is used to obtain the minimal cut sets. At the component level, an exponential distribution is used to model hardware failures, and Bayesian estimation is suggested to estimate the failure rate. Additionally, a binomial distribution is used to model software failures, and a recently developed software reliability estimation method is suggested to estimate the software failure rate. The overall system reliability is then estimated based on minimal cut sets, hardware failure rates and software failure rates.
ESTIMATING THE OPERATOR'S PERFORMANCE TIME OF EMERGENCY PROCEDURAL TASKS BASED ON A TASK COMPLEXITY MEASURE
Jung, Won-Dea ; Park, Jin-Kyun ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 415~420
DOI : 10.5516/NET.04.2012.514
It is important to understand the amount of time required to execute an emergency procedural task in a high-stress situation for managing human performance under emergencies in a nuclear power plant. However, the time to execute an emergency procedural task is highly dependent upon expert judgment due to the lack of actual data. This paper proposes an analytical method to estimate the operator's performance time (OPT) of a procedural task, which is based on a measure of the task complexity (TACOM). The proposed method for estimating an OPT is an equation that uses the TACOM as a variable, and the OPT of a procedural task can be calculated if its relevant TACOM score is available. The validity of the proposed equation is demonstrated by comparing the estimated OPTs with the observed OPTs for emergency procedural tasks in a steam generator tube rupture scenario.
FAULT DETECTION COVERAGE QUANTIFICATION OF AUTOMATIC TEST FUNCTIONS OF DIGITAL I＆C SYSTEM IN NPPS
Choi, Jong-Gyun ; Lee, Seung-Jun ; Kang, Hyun-Gook ; Hur, Seop ; Lee, Young-Jun ; Jang, Seung-Cheol ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 421~428
DOI : 10.5516/NET.04.2012.515
Analog instrument and control systems in nuclear power plants have recently been replaced with digital systems for safer and more efficient operation. Digital instrument and control systems have adopted various fault-tolerant techniques that help the system correctly and safely perform the specific required functions regardless of the presence of faults. Each fault-tolerant technique has a different inspection period, from real-time monitoring to monthly testing. The range covered by each faulttolerant technique is also different. The digital instrument and control system, therefore, adopts multiple barriers consisting of various fault-tolerant techniques to increase the total fault detection coverage. Even though these fault-tolerant techniques are adopted to ensure and improve the safety of a system, their effects on the system safety have not yet been properly considered in most probabilistic safety analysis models. Therefore, it is necessary to develop an evaluation method that can describe these features of digital instrument and control systems. Several issues must be considered in the fault coverage estimation of a digital instrument and control system, and two of these are addressed in this work. The first is to quantify the fault coverage of each fault-tolerant technique implemented in the system, and the second is to exclude the duplicated effect of fault-tolerant techniques implemented simultaneously at each level of the system's hierarchy, as a fault occurring in a system might be detected by one or more fault-tolerant techniques. For this work, a fault injection experiment was used to obtain the exact relations between faults and multiple barriers of faulttolerant techniques. This experiment was applied to a bistable processor of a reactor protection system.
CRITICAL HEAT FLUX ENHANCEMENT IN FLOW BOILING OF Al
AND SiC NANOFLUIDS UNDER LOW PRESSURE AND LOW FLOW CONDITIONS
Lee, Seung-Won ; Park, Seong-Dae ; Kang, Sa-Rah ; Kim, Seong-Man ; Seo, Han ; Lee, Dong-Won ; Bang, In-Cheol ;
Nuclear Engineering and Technology, volume 44, issue 4, 2012, Pages 429~436
DOI : 10.5516/NET.04.2012.516
Critical heat flux (CHF) is the thermal limit of a phenomenon in which a phase change occurs during heating (such as bubbles forming on a metal surface used to heat water), which suddenly decreases the heat transfer efficiency, thus causing localized overheating of the heating surface. The enhancement of CHF can increase the safety margins and allow operation at higher heat fluxes; thus, it can increase the economy. A very interesting characteristic of nanofluids is their ability to significantly enhance the CHF. Nanofluids are nanotechnology-based colloidal dispersions engineered through the stable suspension of nanoparticles. All experiments were performed in round tubes with an inner diameter of 0.01041 m and a length of 0.5 m under low pressure and low flow (LPLF) conditions at a fixed inlet temperature using water, 0.01 vol.%
/water nanofluid, and SiC/water nanofluid. It was found that the CHF of the nanofluids was enhanced and the CHF of the SiC/water nanofluid was more enhanced than that of the