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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 41, Issue 10 - Dec 2009
Volume 41, Issue 9 - Nov 2009
Volume 41, Issue 8 - Oct 2009
Volume 41, Issue 7 - Sep 2009
Volume 41, Issue 6 - Aug 2009
Volume 41, Issue 5 - Jun 2009
Volume 41, Issue 4 - May 2009
Volume 41, Issue 3 - Apr 2009
Volume 41, Issue 2 - Mar 2009
Volume 41, Issue 1 - Feb 2009
Selecting the target year
LESSONS LEARNED FROM HALDEN PROJECT RESEARCH ON HUMAN SYSTEM INTERFACES
Braseth, Alf Ove ; Nihlwing, Christer ; Svengren, Hakan ; Veland, Oystein ; Hurlen, Lars ; Kvalem, Jon ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 215~224
DOI : 10.5516/NET.2009.41.3.215
Innovative Human System Interfaces (HSIs) has been a major topic of research of the international Halden Reactor Project (HRP) for many years. Different design concepts have been addressed and prototypes have been implemented and evaluated in the experimental control room facility of HRP. Many of the concepts go far beyond traditional P&ID type displays, and utilize advanced computer graphics and animations. The paper briefly describes some of the concepts, their advantages and disadvantages experienced through evaluations and feedback from users.
IDENTIFICATION AND EVALUATION OF HUMAN FACTORS ISSUES ASSOCIATED WITH EMERGING NUCLEAR PLANT TECHNOLOGY
O'Hara, John M. ; Higgins, James C. ; Brown, William S. ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 225~236
DOI : 10.5516/NET.2009.41.3.225
This study has identified human performance research issues associated with the implementation of new technology in nuclear power plants (NPPs). To identify the research issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were prioritized into four categories based on evaluations provided by 14 independent subject matter experts representing vendors, utilities, research organizations and regulators. Twenty issues were categorized into the top priority category. The study also identifies the priority of each issue and the rationale for those in the top priority category. The top priority issues were then organized into research program areas of: New Concepts of Operation using Multi-agent Teams, Human-system Interface Design, Complexity Issues in Advanced Systems, Operating Experience of New and Modernized Plants, and HFE Methods and Tools. The results can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas to support the safe operation of new NPPs.
DISTRIBUTED HMI SYSTEM FOR MANAGING ALL SPAN OF PLANT CONTROL AND MAINTENANCE
Yoshikawa, Hidekazu ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 237~246
DOI : 10.5516/NET.2009.41.3.237
Digitalization of not only non-safety but also safety-grade I &C systems with full computerized Main Control Room (MCR) is the recent trend of I&C systems of nuclear power plants (NPP) around the world, while plant maintenance has been shifting from traditional time based maintenance to condition based maintenance. In order to cope with the new trend of operation and maintenance in NPP, a concept of online distributed diagnostic system for both plant operation and maintenance has been proposed in order to further improve both the plant efficiency and the work environment of plant operation staff members by organizational learning. In this respect, the research subjects of human machine interface (HMI) for the online distributed diagnostic system are also discussed for supporting the plant personnel at both MCR and local working places in the plant by the application of advanced ICT (Information and Communication Technologies).
A REVIEW OF STUDIES ON OPERATOR'S INFORMATION SEARCHING BEHAVIOR FOR HUMAN FACTORS STUDIES IN NPP MCRS
Ha, Jun-Su ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 247~270
DOI : 10.5516/NET.2009.41.3.247
This paper reviews studies on information searching behavior in process control systems and discusses some implications learned from previous studies for use in human factors studies on nuclear power plants (NPPs) main control rooms (MCRs). Information searching behavior in NPPs depends on expectancy, value, salience, and effort. The first quantitative scanning model developed by Senders for instrument panel monitoring considered bandwidth (change rate) of instruments as a determining factor in scanning behavior. Senders' model was subsequently elaborated by other researchers to account for value in addition to bandwidth. There is also another type of model based on the operator's situation awareness (SA) which has been developed for NPP application. In these SA-based models, situation-event relations or rules on system dynamics are considered the most significant factor forming expectancy. From the review of previous studies it is recommended that, for NPP application, (1) a set of symptomatic information sources including both changed and unchanged symptoms should be considered along with bandwidth as determining factors governing information searching (or visual sampling) behavior; (2) both data-driven monitoring and knowledge-driven monitoring should be considered and balanced in a systematic way; (3) sound models describing mechanisms of cognitive activities during information searching tasks should be developed so as to bridge studies on information searching behavior and design improvement in HMI; (4) the attention-situation awareness (A-SA) modeling approach should be recognized as a promising approach to be examined further; and (5) information displays should be expected to have totally different characteristics in advanced control rooms. Hence much attention should be devoted to information searching behavior including human-machine interface (HMI) design and human cognitive processes.
EFFECTS OF GEOMETRIC PARAMETERS ON NUCLEATE POOL BOILING OF SATURATED WATER IN VERTICAL ANNULI
Kang, Myeong-Gie ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 271~278
DOI : 10.5516/NET.2009.41.3.271
Nucleate pool boiling of water in vertical annuli at atmospheric pressure has been studied experimentally and two empirical correlations have been suggested to obtain effects of geometric parameters on heat transfer. Data of the present and the previous tests range over a tube length of 0.50-0.57 m, a diameter of 16.5-34.0 mm, and an annular gap size of 3.7-44.3 mm. Through the analysis, tube bottom confinement (open or closed) has been investigated, as well. The developed correlations predict experimental data within a
error bound. It has been identified that effects of the diameter and the length of heated tubes as well as the annular gap size should be counted into the analyses to estimate heat transfer coefficients accurately.
AN IMPROVED MONTE CARLO METHOD APPLIED TO THE HEAT CONDUCTION ANALYSIS OF A PEBBLE WITH DISPERSED FUEL PARTICLES
Song, Jae-Hoon ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 279~286
DOI : 10.5516/NET.2009.41.3.279
Improving over a previous study , this paper provides a Monte Carlo method for the heat conduction analysis of problems with complicated geometry (such as a pebble with dispersed fuel particles). The method is based on the theoretical results of asymptotic analysis of neutron transport equation. The improved method uses an appropriate boundary layer correction (with extrapolation thickness) and a scaling factor, rendering the problem more diffusive and thus obtaining a heat conduction solution. Monte Carlo results are obtained for the randomly distributed fuel particles of a pebble, providing realistic temperature distributions (showing the kernel and graphite-matrix temperatures distinctly). The volumetric analytic solution commonly used in the literature is shown to predict lower temperatures than those of the Monte Carlo results provided in this paper.
SIMULATION OF THERMAL STRATIFICATION IN INLET NOZZLE OF STEAM GENERATOR
Ji, Joon-Suk ; Youn, Bum-Su ; Jeong, Hyun-Chul ; Kim, Sang-Nyung ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 287~294
DOI : 10.5516/NET.2009.41.3.287
Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.
NUPEC BFBT SUBCHANNEL VOID DISTRIBUTION ANALYSIS USING THE MATRA AND MARS CODES
Hwang, Dae-Hyun ; Jeong, Jae-Jun ; Chung, Bub-Dong ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 295~306
DOI : 10.5516/NET.2009.41.3.295
The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.
THE BENCHMARK CALCULATIONS OF THE GAMMA+ CODE WITH THE HTR-10 SAFETY DEMONSTRATION EXPERIMENTS
Jun, Ji-Su ; Lim, Hong-Sik ; Lee, Won-Jae ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 307~318
DOI : 10.5516/NET.2009.41.3.307
KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.
CO-SEPARATION OF Am AND RARE EARTH ELEMENTS FROM A HIGHLY ACIDIC RADWASTE SOLUTION BY A SOLVENT EXTRACTION WITH (DIMETHYLDIBUTYL TETRADECYLMALONAMIDE-DIHEXYLOCTANAMIDE)/N-DODECANE
Lee, Eil-Hee ; Lim, Jae-Gwan ; Chung, Dong-Yong ; Yoo, Jae-Hyung ; Kim, kwang-Wook ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 319~326
DOI : 10.5516/NET.2009.41.3.319
This study was carried out to investigate the high-acidity co-separation of Am and RE from a simulated radwaste solution by a solvent extraction using a mixture of Dimethyldibutyltetradecylmalonamide (DMDBTDMA, as an extractant) and dihexyl octanamide (DHOA, as a phase modifier) diluted with n-dodecane (NDD). All the experiments were conducted as a batch type. First, the environmentally friendly DMDBTDMA and DHOA composed of only CHON atoms were self-synthesized. Then, the conditions for the prevention of a third phase, generated in the organic phase were examined. In addition, the effects of the concentration of nitric acid, DHOA, oxalic acid and
on the co-extraction of Am and RE were elucidated. Consequently, the optimum condition of (0.5M DMDBTDMA+0.5M DHOA)/NDD-0.3M
and O/A=2 was obtained through experimental work. Under this condition, the extraction yields were found to be about 80% for Am, more than 70% for RE such as La, Eu, Nd, Ce, etc., 3% for Cs and Sr, 69% for Fe and less than 11% for Mo and Ru. For the co-extraction of Am and RE, Fe should be removed in advance or prevented from a co-extraction with Am by controlling the different extraction rates of Am and Fe. About 95% of the Am and RE in the organic phase were stripped using a 0.5M
AN IN-SITU YOUNG'S MODULUS MEASUREMENT TECHNIQUE FOR NUCLEAR POWER PLANTS USING TIME-FREQUENCY ANALYSIS
Choi, Young-Chul ; Yoon, Doo-Byung ; Park, Jin-Ho ; Kwon, Hyun-Sang ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 327~334
DOI : 10.5516/NET.2009.41.3.327
Elastic wave is one of the most useful tools for non-destructive tests in nuclear power plants. Since the elastic properties are indispensable for analyzing the behaviors of elastic waves, they should be predetermined within an acceptable accuracy. Nuclear power plants are exposed to harsh environmental conditions and hence the structures are degraded. It means that the Young's modulus becomes unreliable and in-situ measurement of Young's modulus is required from an engineering point of view. Young's modulus is estimated from the group velocity of propagating waves. Because the flexural wave of a plate is inherently dispersive, the group velocity is not clearly evaluated in temporal signal analysis. In order to overcome such ambiguity in estimation of group velocity, Wigner-Ville distribution as the time-frequency analysis technique was proposed and utilized. To verify the proposed method, experiments for steel and acryl plates were performed with accelerometers. The results show good estimation of the Young's modulus of two plates.
HYDROELASTIC VIBRATION ANALYSIS OF TWO FLEXIBLE RECTANGULAR PLATES PARTIALLY COUPLED WITH A LIQUID
Jeong, Kyeong-Hoon ; Kim, Jong-Wook ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 335~346
DOI : 10.5516/NET.2009.41.3.335
This paper deals with a hydroelastic vibration analysis of two rectangular plates partially coupled with a liquid, which is bounded by two plates and two rigid side walls. The wet displacement of each plate is assumed to be a combination of the modal functions of a dry uniform beam with a clamped boundary condition. As the liquid is assumed to be an ideal liquid, the displacement potential satisfying the Laplace equation is determined so that the liquid boundary conditions can meet the requirements at the rigid surfaces and the free liquid surface. The wet dynamic modal functions of each plate are expanded by using the finite Fourier transform to obtain an appropriate form of the compatibility requirement along the contacting surfaces between the plates and the liquid. The liquid-coupled natural frequencies of the plates are derived by using the Rayleigh-Ritz method. Finite element analyses using commercial software are carried out to verify the proposed theory. It is observed that the theoretical method agrees excellently with the three-dimensional finite element analyses results. The effects of the liquid depth and the liquid thickness on the normalized natural frequencies are investigated to identify the dynamic characteristics of the liquid coupled system.
STRUCTURAL INTEGRITY EVALUATION OF NUCLEAR FUEL WITH REDUCED WELDING CONDITIONS
Park, Nam-Gyu ; Park, Joon-Kyoo ; Suh, Jung-Min ; Kim, Kyu-Tae ; Jeon, Kyeong-Lak ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 347~354
DOI : 10.5516/NET.2009.41.3.347
Welding is required for a connection between two different components in the nuclear fuel of a pressurized water reactor. This work relies on a mechanical experiment and analytic results to investigate the structural integrity of nuclear fuel in a situation where some components are not welded to each other. A series of lateral vibration tests are performed in a test facility, and the test structures are examined in terms of dynamic behavior. In the tests, the displacement signal at every grid structure that sustains fuel rods is measured and processed to identify the dynamic properties. The fluid-elastic stability of the structure is also analyzed to evaluate susceptibility to a cross flow with an assumed conservative cross flow distribution. The test and analysis results confirm that the structural integrity can be maintained even in the absence of some welding connections.
FREE VIBRATION ANALYSIS OF CIRCULAR PLATE WITH ECCENTRIC HOLE SUBMERGED IN FLUID
Jhung, Myung-Jo ; Choi, Young-Hwan ; Ryu, Yong-Ho ;
Nuclear Engineering and Technology, volume 41, issue 3, 2009, Pages 355~364
DOI : 10.5516/NET.2009.41.3.355
Circular plates with holes are extensively used in mechanical components. The existence of a hole in a circular plate results in a significant change in the natural frequencies and mode shapes of the structure. Especially if the hole is located eccentrically, the vibration behavior of these structures is expected to deviate significantly from that of a plate with a concentric hole. In addition, if the plate is in contact with or submerged in fluid, the situation is more complex. Therefore, in this study, an analytical method to determine the modal characteristics of a plate submerged in fluid is developed based on the finite Fourier-Bessel series expansion and Rayleigh-Ritz method and is verified by the finite element analysis using a commercial program. Also, the relationship between parameter variations and vibration modes is investigated. These results can be used as guidance for the modal analysis and damage detection of a circular plate with a hole.