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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 45, Issue 7 - Dec 2013
Volume 45, Issue 6 - Nov 2013
Volume 45, Issue 5 - Oct 2013
Volume 45, Issue 4 - Aug 2013
Volume 45, Issue 3 - Jun 2013
Volume 45, Issue 2 - Apr 2013
Volume 45, Issue 1 - Feb 2013
Selecting the target year
ADVANCED MMIS TOWARD SUBSTANTIAL REDUCTION IN HUMAN ERRORS IN NPPS
Seong, Poong Hyun ; Kang, Hyun Gook ; Na, Man Gyun ; Kim, Jong Hyun ; Heo, Gyunyoung ; Jung, Yoensub ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 125~140
DOI : 10.5516/NET.04.2013.700
This paper aims to give an overview of the methods to inherently prevent human errors and to effectively mitigate the consequences of such errors by securing defense-in-depth during plant management through the advanced man-machine interface system (MMIS). It is needless to stress the significance of human error reduction during an accident in nuclear power plants (NPPs). Unexpected shutdowns caused by human errors not only threaten nuclear safety but also make public acceptance of nuclear power extremely lower. We have to recognize there must be the possibility of human errors occurring since humans are not essentially perfect particularly under stressful conditions. However, we have the opportunity to improve such a situation through advanced information and communication technologies on the basis of lessons learned from our experiences. As important lessons, authors explained key issues associated with automation, man-machine interface, operator support systems, and procedures. Upon this investigation, we outlined the concept and technical factors to develop advanced automation, operation and maintenance support systems, and computer-based procedures using wired/wireless technology. It should be noted that the ultimate responsibility of nuclear safety obviously belongs to humans not to machines. Therefore, safety culture including education and training, which is a kind of organizational factor, should be emphasized as well. In regard to safety culture for human error reduction, several issues that we are facing these days were described. We expect the ideas of the advanced MMIS proposed in this paper to lead in the future direction of related researches and finally supplement the safety of NPPs.
A QUALITATIVE METHOD TO ESTIMATE HSI DISPLAY COMPLEXITY
Hugo, Jacques ; Gertman, David ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 141~150
DOI : 10.5516/NET.04.2013.701
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.
MEASURING THE INFLUENCE OF TASK COMPLEXITY ON HUMAN ERROR PROBABILITY: AN EMPIRICAL EVALUATION
Podofillini, Luca ; Park, Jinkyun ; Dang, Vinh N. ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 151~164
DOI : 10.5516/NET.04.2013.702
A key input for the assessment of Human Error Probabilities (HEPs) with Human Reliability Analysis (HRA) methods is the evaluation of the factors influencing the human performance (often referred to as Performance Shaping Factors, PSFs). In general, the definition of these factors and the supporting guidance are such that their evaluation involves significant subjectivity. This affects the repeatability of HRA results as well as the collection of HRA data for model construction and verification. In this context, the present paper considers the TAsk COMplexity (TACOM) measure, developed by one of the authors to quantify the complexity of procedure-guided tasks (by the operating crew of nuclear power plants in emergency situations), and evaluates its use to represent (objectively and quantitatively) task complexity issues relevant to HRA methods. In particular, TACOM scores are calculated for five Human Failure Events (HFEs) for which empirical evidence on the HEPs (albeit with large uncertainty) and influencing factors are available - from the International HRA Empirical Study. The empirical evaluation has shown promising results. The TACOM score increases as the empirical HEP of the selected HFEs increases. Except for one case, TACOM scores are well distinguished if related to different difficulty categories (e.g., "easy" vs. "somewhat difficult"), while values corresponding to tasks within the same category are very close. Despite some important limitations related to the small number of HFEs investigated and the large uncertainty in their HEPs, this paper presents one of few attempts to empirically study the effect of a performance shaping factor on the human error probability. This type of study is important to enhance the empirical basis of HRA methods, to make sure that 1) the definitions of the PSFs cover the influences important for HRA (i.e., influencing the error probability), and 2) the quantitative relationships among PSFs and error probability are adequately represented.
A REVIEW ON DEVELOPING INDUSTRIAL STANDARDS TO INTRODUCE DIGITAL COMPUTER APPLICATION FOR NUCLEAR I&C AND HMIT IN JAPAN
Yoshikawa, Hidekazu ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 165~178
DOI : 10.5516/NET.04.2013.703
A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.
ANALYSIS OF A STATION BLACKOUT SCENARIO WITH AN ATLAS TEST
Kim, Yeon-Sik ; Yu, Xin-Guo ; Kang, Kyoung-Ho ; Park, Hyun-Sik ; Cho, Seok ; Choi, Ki-Yong ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 179~190
DOI : 10.5516/NET.02.2012.007
A station blackout experiment called SBO-01 was performed at the ATLAS facility. From the SBO-01 test, the station blackout scenario can be characterized into two typical phases: A first phase characterized by decay heat removal through secondary safety valves until the SG dryouts, and a second phase characterized by an energy release through a blowdown of the primary system after the SG dryouts. During the second phase, some physical phenomena of the change over a pressurizer function, i.e., the pressurizer being full before the POSRV
opening and then its function being taken by the RV, and the termination of normal natural circulation flow were identified. Finally, a core heatup occurred at a low core water level, although under a significant amount of PZR inventory, whose drainage seemed to be hindered owing to the pressurizer function by the RV. The transient of SBO-01 is well reproduced in the calculation using the MARS code.
NUMERICAL ANALYSIS OF THERMAL STRATIFICATION IN THE UPPER PLENUM OF THE MONJU FAST REACTOR
Choi, Seok-Ki ; Lee, Tae-Ho ; Kim, Yeong-Il ; Hahn, Dohee ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 191~202
DOI : 10.5516/NET.02.2012.050
A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.
IODINE REMOVAL EFFICIENCY IN NON-SUBMERGED AND SUBMERGED SELF-PRIMING VENTURI SCRUBBER
Ali, Majid ; Yan, Changqi ; Sun, Zhongning ; Gu, Haifeng ; Wang, Junlong ; Khurram, Mehboob ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 203~210
DOI : 10.5516/NET.03.2012.047
The objective of this conducted research is to study the iodine removal efficiency in a self-priming venturi scrubber for submerged and non-submerged operating conditions experimentally and theoretically. The alkaline solution is used as an absorbent, which is prepared by dissolving sodium hydroxide (NaOH) and sodium thiosulphate (
) in water to remove the gaseous iodine (
) from the gas. Iodine removal efficiency is examined at various gas flow rates and inlet concentrations of iodine for submerged and non-submerged operating conditions. In the non-submerged venturi scrubber, only the droplets take part in iodine removal efficiency. However, in a submerged venturi scrubber condition, the iodine gas is absorbed from gas to droplets inside the venturi scrubber and from bubbles to surrounding liquid at the outlet of a venturi scrubber. Experimentally, it is observed that the iodine removal efficiency is greater in the submerged venturi scrubber as compare to a non-submerged venturi scrubber condition. The highest iodine removal efficiency of
has been achieved in a submerged self-priming venturi scrubber condition. A mathematical correlation is used to predict the theoretical iodine removal efficiency in submerged and non-submerged conditions, and it is compared against the experimental results. The Wilkinson et al. correlation is used to predict the bubble diameter theoretically whereas the Nukiyama and Tanasawa correlation is used for droplet diameter. The mass transfer coefficient for the gas phase is calculated from the Steinberger and Treybal correlation. The calculated results for a submerged venturi scrubber agree well with experimental results but underpredicts in the case of the non-submerged venturi scrubber.
LIMITED OXIDATION OF IRRADIATED GRAPHITE WASTE TO REMOVE SURFACE CARBON-14
Smith, Tara E. ; Mccrory, Shilo ; Dunzik-Gougar, Mary Lou ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 211~218
DOI : 10.5516/NET.06.2012.025
Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (
), with a half-life of 5730 years. Fachinger et al.  have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the
, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create
gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice  and later incorporated into an oxidation kinetics model by El-Genk and Tournier . Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide (
). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO .
EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE
Ryu, Woo Seog ; Park, Dae Gyu ; Song, Ung Sup ; Park, Jin Seok ; Ahn, Sang Bok ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 219~222
DOI : 10.5516/NET.07.2012.048
Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28)
, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at
did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28)
STRUCTURAL TEST AND ANALYSIS OF RC SLAB AFTER FIRE LOADING
Chung, Chul-Hun ; Im, Cho Rong ; Park, Jaegyun ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 223~236
DOI : 10.5516/NET.09.2011.072
In the present study the behavior of fire and the residual strength of fire-ignited RC slabs are investigated by experimental tests and numerical simulations. The fire tests of RC slabs were carried out in a furnace using the ISO 834 standard fire. The load capacity of the cooled RC slabs that were not loaded during the fire tests was evaluated by additional 3 point bending tests. The influence of the proportion of PP (polypropylene) fibers in the RC slabs on the structural behavior of the RC slabs after the fire loading was investigated. The results of the fire tests showed that the maximum temperature of concrete with PP fiber was lower than that of concrete without PP fiber. As the concrete was heated, the ultimate compressive strength decreased and the ultimate strain increased. The load-deflection relations of RC slabs after fire loading were compared by using existing stress-strain-temperature models. The comparison between the numerical analysis and the experimental tests showed that some numerical analyses were reliable and therefore, can be applied to evaluate the ultimate load of RC slabs after fire loading. The ultimate load capacity after cooling down the RC slabs without PP fiber showed a considerable reduction from that of the RC slabs with PP fiber.
A COUPLED CFD-FEM ANALYSIS ON THE SAFETY INJECTION PIPING SUBJECTED TO THERMAL STRATIFICATION
Kim, Sun-Hye ; Choi, Jae-Boong ; Park, Jung-Soon ; Choi, Young-Hwan ; Lee, Jin-Ho ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 237~248
DOI : 10.5516/NET.09.2012.038
Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.
DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM
Ko, Do-Young ; Kim, Kyu-Hyung ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 249~256
DOI : 10.5516/NET.09.2012.049
In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.
DEVELOPMENT OF AN IMPROVED FARE TOOL WITH APPLICATION TO WOLSONG NUCLEAR POWER PLANT
Lee, Sun Ki ; Hong, Sung Yull ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 257~264
DOI : 10.5516/NET.09.2012.071
In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.
DETERMINISTIC EVALUATION OF DELAYED HYDRIDE CRACKING BEHAVIORS IN PHWR PRESSURE TUBES
Oh, Young-Jin ; Chang, Yoon-Suk ;
Nuclear Engineering and Technology, volume 45, issue 2, 2013, Pages 265~276
DOI : 10.5516/NET.09.2012.027
Pressure tubes made of Zr-2.5 wt% Nb alloy are important components consisting reactor coolant pressure boundary of a pressurized heavy water reactor, in which unanticipated through-wall cracks and rupture may occur due to a delayed hydride cracking (DHC). The Canadian Standards Association has provided deterministic and probabilistic structural integrity evaluation procedures to protect pressure tubes against DHC. However, intuitive understanding and subsequent assessment of flaw behaviors are still insufficient due to complex degradation mechanisms and diverse influential parameters of DHC compared with those of stress corrosion cracking and fatigue crack growth phenomena. In the present study, a deterministic flaw assessment program was developed and applied for systematic integrity assessment of the pressure tubes. Based on the examination results dealing with effects of flaw shapes, pressure tube dimensional changes, hydrogen concentrations of pressure tubes and plant operation scenarios, a simple and rough method for effective cooldown operation was proposed to minimize DHC risks. The developed deterministic assessment program for pressure tubes can be used to derive further technical bases for probabilistic damage frequency assessment.