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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 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
MULTI-SCALE THERMAL-HYDRAULIC ANALYSIS OF PWRS USING THE CUPID CODE
Yoon, Han Young ; Cho, Hyoung Kyu ; Lee, Jae Ryong ; Park, Ik Kyu ; Jeong, Jae Jun ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 831~846
DOI : 10.5516/NET.02.2012.716
KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.
BRIEF REVIEW OF LATEST DIRECT NUMERICAL SIMULATION ON POOL AND FILM BOILING
Kunugi, Tomoaki ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 847~854
DOI : 10.5516/NET.02.2012.717
Despite extensive research efforts, the mechanism of the nucleate boiling phenomena is still not clear. A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify its heat transfer characteristics and discuss their mechanism. Therefore, many DNS procedures have been developed based on recent highly advancing computer technologies. This brief review focuses on the state of the art in direct numerical simulation of the pool boiling phenomena over the past two decades. In this review, the fundamentals of the boiling phenomena and the bubble departure and micro-layer models are briefly introduced, and then the numerical procedures for tracking or capturing interface/surface shape such as the front tracking method, level set method, volume of fluid treatments, and other methods (Lattice Boltzmann method, phase-field method and so on) are briefly reviewed.
PERSPECTIVES IN SYSTEM THERMAL-HYDRAULICS
D'auria, F. ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 855~870
DOI : 10.5516/NET.02.2012.718
The paper deals with three main topics: a) the definition of System Thermal-Hydraulics (SYS TH), b) a historical outline for SYS TH and, c) the description of elements for reflection when planning research projects or improvement activities, this last topic being the main reason for the paper. Distinctions between basic thermal-hydraulics and computational Fluid-Dynamics (CFD) on the one side and SYS TH on the other side are considered under the first topic; stakeholders in the technology are identified. The proposal of Interim Acceptance Criteria for Emergency Core Cooling Systems in 1971 by US NRC (AEC at the time) is recognized as the starting date or the triggering event for SYS TH (second topic). The complex codes and the main experimental programs (list provided in the paper) constitute the pillars for SYS TH. Caution or warning statements are introduced in advance when discussing the third topic: a single person (or a researcher) has little to no possibility, or capability, of streamlining the forthcoming investments or to propose a roadmap for future activities. Nevertheless, the ambitious attempt to foresee developments in this area has been pursued without constraints connected with the availability of funds and with industrial benefits or interests. Demonstrating the acceptability of current SYS TH limitations and training in the application of those codes are mentioned as the main challenges for forthcoming research activities.
ON THE MODELLING OF TWO-PHASE FLOW IN HORIZONTAL LEGS OF A PWR
Bestion, D. ; Serre, G. ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 871~888
DOI : 10.5516/NET.02.2012.719
This paper aims at presenting the state of the art, the recent progress, and the perspective for the future, in the modelling of two-phase flow in the horizontal legs of a PWR. All phenomena relevant for safety analysis are listed first. The selection of the modelling approach for system codes is then discussed, including the number of fluids or fields, the space and time resolution, and the use of flow regime maps. The classical two-fluid six-equation one-pressure model as it is implemented in the CATHARE code is then presented and its properties are described. It is shown that the axial effects of gravity forces may be correctly taken into account even in the case of change of the cross section area or of the pipe orientation. It is also shown that it can predict both fluvial and torrential flow with a possible hydraulic jump. Since phase stratification plays a dominant role, the Kelvin-Helmholtz instability and the stability of bubbly flow regime are discussed. A transition criterion based on a stability analysis of shallow water waves may be used to predict the Kelvin-Helmholtz instability. Recent experimental data obtained in the METERO test facility are analysed to model the transition from a bubbly to stratified flow regime. Finally, perspectives for further improvement of the modelling are drawn including dynamic modelling of turbulence and interfacial area and multi-field models.
TOWARD MECHANISTIC MODELING OF BOILING HEAT TRANSFER
Podowski, Michael Z. ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 889~896
DOI : 10.5516/NET.02.2012.720
Recent progress in the computational fluid dynamics methods of two- and multiphase phase flows has already started opening up new exciting possibilities for using complete multidimensional models to simulate boiling systems. Combining this new theoretical and computational approach with novel experimental methods should dramatically improve both our understanding of the physics of boiling and the predictive capabilities of models at various scale levels. However, for the multidimensional modeling framework to become an effective predictive tool, it must be complemented with accurate mechanistic closure laws of local boiling mechanisms. Boiling heat transfer has been studied quite extensively before. However, it turns out that the prevailing approach to the analysis of experimental data for both pool boiling and forced-convection boiling has been associated with formulating correlations which normally included several adjustable coefficients rather than based on first principle models of the underlying physical phenomena. One reason for this has been the tendency (driven by practical applications and industrial needs) to formulate single expressions which encompass a broad range of conditions and fluids. This, in turn, makes it difficult to identify various specific factors which can be independently modeled for different situations. The objective of this paper is to present a mechanistic modeling concept for both pool boiling and forced-convection boiling. The proposed approach is based on theoretical first-principle concepts, and uses a minimal number of coefficients which require calibration against experimental data. The proposed models have been validated against experimental data for water and parametrically tested. Model predictions are shown for a broad range of conditions.
DEVELOPMENT OF NEW TAXONOMY OF INAPPROPRIATE COMMUNICATION AND ITS APPLICATION TO OPERATING TEAMS IN NUCLEAR POWER PLANTS
Kim, Ar Ryum ; Park, Jinkyun ; Lee, Seung Woo ; Jang, Inseok ; Kang, Hyun Gook ; Seong, Poong Hyun ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 897~910
DOI : 10.5516/NET.04.2011.068
Inappropriate communications can cause a lack of necessary information exchange between operators and lead to serious consequences in large process systems such as nuclear power plants (NPPs). In this regard, various kinds of taxonomies of inappropriate communications have been developed to prevent inappropriate communications. However, there seems to be difficult to identify inappropriate communications from verbal protocol data between operators. Because the existing taxonomies were developed for use in report analysis, there is a problem of 'uncertainty'. In consequence, this paper proposes a new taxonomy of inappropriate communications and provides some insights to prevent inappropriate communications. In order to develop the taxonomy, existing taxonomies for four industries from 1980 to 2010 were collected and a new taxonomy is developed based on the simplified one-way communication model. In addition, the ratio of inappropriate communications from 8 samples of audio-visual format verbal protocol data recorded during emergency training sessions by operating teams is compared with performance scores calculated based on the task analysis. As a result, inappropriate communications can be easily identified from the verbal protocol data using the suggested taxonomy, and teams with a higher ratio of inappropriate communications tend to have a lower performance score.
THE MODEL PREDICTIVE CONTROLLER FOR THE FEEDWATER AND LEVEL CONTROL OF A NUCLEAR STEAM GENERATOR
Lee, Yoon Joon ; Oh, Seung Jin ; Chun, Wongee ; Kim, Nam Jin ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 911~918
DOI : 10.5516/NET.04.2012.002
Steam generator level control at low power is difficult due to its adverse thermal hydraulic properties, and is usually conducted by an operator. The basic model predictive control (MPC) is similar to the action of an operator in that the operator knows the desired reference trajectory for a finite period of time and takes the necessary control actions needed to ensure the desired trajectory. An MPC is based on a model; the performance as well as the efficiency of the MPC depends heavily on the exactness of the model. In this study, steam generator models that can describe in detail its thermal hydraulic behaviors, particularly at low power, are used in the MPC design. The design scope is divided into two parts. First, the MPC feedwater controller of the feedwater station is determined, and then the MPC level controller for the overall system is designed. Because the dynamic properties of a steam generator change with the power levels, a realistic situation is simulated by changing the transfer functions of the steam generator at every time step. The resulting MPC controller shows good performance.
A CYBER SECURITY RISK ASSESSMENT FOR THE DESIGN OF I&C SYSTEMS IN NUCLEAR POWER PLANTS
Song, Jae-Gu ; Lee, Jung-Woon ; Lee, Cheol-Kwon ; Kwon, Kee-Choon ; Lee, Dong-Young ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 919~928
DOI : 10.5516/NET.04.2011.065
The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.
STATE TOKEN PETRI NET MODELING METHOD FOR FORMAL VERIFICATION OF COMPUTERIZED PROCEDURE INCLUDING OPERATOR'S INTERRUPTIONS OF PROCEDURE EXECUTION FLOW
Kim, Yun Goo ; Seong, Poong Hyun ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 929~938
DOI : 10.5516/NET.04.2012.019
The Computerized Procedure System (CPS) is one of the primary operating support systems in the digital Main Control Room. The CPS displays procedure on the computer screen in the form of a flow chart, and displays plant operating information along with procedure instructions. It also supports operator decision making by providing a system decision. A procedure flow should be correct and reliable, as an error would lead to operator misjudgment and inadequate control. In this paper we present a modeling for the CPS that enables formal verification based on Petri nets. The proposed State Token Petri Nets (STPN) also support modeling of a procedure flow that has various interruptions by the operator, according to the plant condition. STPN modeling is compared with Coloured Petri net when they are applied to Emergency Operating Computerized Procedure. A converting program for Computerized Procedure (CP) to STPN has been also developed. The formal verification and validation methods of CP with STPN increase the safety of a nuclear power plant and provide digital quality assurance means that are needed when the role and function of the CPS is increasing.
DEVELOPMENT OF ANODIC STRIPPING VOLTAMMETRY FOR THE DETERMINATION OF PALLADIUM IN HIGH LEVEL NUCLEAR WASTE
Bhardwaj, T.K. ; Sharma, H.S. ; Jain, P.C. ; Aggarwal, S.K. ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 939~944
DOI : 10.5516/NET.06.2011.060
Deposition potential, deposition time, square wave frequency, rotation speed of the rotating disc electrode, and palladium concentration were studied on a Glassy Carbon Electrode (GCE) in 0.01M HCl for the determination of palladium in High Level Nuclear Waste (HLNW) by anodic stripping voltammetry. Experimental conditions were optimized for the determination of palladium at two different,
, levels. Error and standard deviation of this method were under 1% for all palladium standard solutions. The developed technique was successfully applied as a subsidiary method for the determination of palladium in simulated high level nuclear waste with very good precision and high accuracy (under 1 % error and standard deviation).
ESTIMATION OF THE BEHAVIORS OF SELENIUM IN THE NEAR FIELD OF REPOSITORY
Kim, Seung-Soo ; Min, Jae-Ho ; Baik, Min-Hoon ; Kim, Gye-Nam ; Choi, Jong-Won ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 945~952
DOI : 10.5516/NET.06.2012.010
The sorption of selenium ions onto iron and iron compounds as a disposal container material and its corrosion products, and onto bentonite as a buffer material, was studied to understand the behaviors of selenium in a waste repository. Selenite was sorbed onto commercial magnetite very well in solutions at around pH 9, but silicate hindered their sorption onto both magnetite and ferrite. Unlike commercial magnetite and ferrite, flesh synthesized magnetite, green rust and iron greatly decreased selenium concentration even in a silicate solution. These results might be due to the formation of precipitates, or the sorption of selenide or selenite onto an iron surface at below Eh= -0.2 V. Red-colored Se(cr) was observed on the surface of a reaction bottle containing iron powder added into a selenite solution. Silicate influences on the sorption onto magnetite and iron for selenide are the same as those for selenite. Even though bentonite adsorbed a slight amount of selenite, the sorption cannot be ignored in the waste repository since a very large quantity of bentonite is used.
ANALYSIS OF NECKING DEFORMATION AND FRACTURE CHARACTERISTICS OF IRRADIATED A533B RPV STEEL
Kim, Jin Weon ; Byun, Thak Sang ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 953~960
DOI : 10.5516/NET.07.2012.012
This paper reports the irradiation effect on the deformation behavior and tensile fracture properties of A533B RPV steel. An inverse identification technique using iterative finite element (FE) simulation was used to determine those properties from tensile data for the A533B RPV steel irradiated at 65 to
and deformed at room temperature. FE simulation revealed that the plastic instability at yield followed by softening for higher doses was related to the occurrence of localized necking immediately after yielding. The strain-hardening rate in the equivalent true stress-true strain relationship was still positive during the necking deformation. The tensile fracture stress was less dependent on the irradiation dose, whereas the tensile fracture strain and fracture energy decreased with increasing dose level up to 0.1 dpa and then became saturated. However, the tensile fracture strain and fracture energy still remained high after high-dose irradiation, which is associated with a large amount of ductility during the necking deformation for irradiated A533B RPV steel.
PFM APPLICATION FOR THE PWSCC INTEGRITY OF Ni-BASE ALLOY WELDS-DEVELOPMENT AND APPLICATION OF PINEP-PWSCC
Hong, Jong-Dae ; Jang, Changheui ; Kim, Tae Soon ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 961~970
DOI : 10.5516/NET.07.2012.017
Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.
SYNTHESIS OF SILICA-COATED Au WITH Ag, Co, Cu, AND Ir BIMETALLIC RADIOISOTOPE NANOPARTICLE RADIOTRACERS
Jung, Jin-Hyuck ; Jung, Sung-Hee ; Kim, Sang-Ho ; Choi, Seong-Ho ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 971~976
DOI : 10.5516/NET.08.2012.001
Silica-coated Au with Ag, Co, Cu, and Ir bimetallic radioisotope nanoparticles were synthesized by neutron irradiation, after coating
onto the bimetallic particles by the sol-gel St
ber process. Bimetallic nanoparticles were synthesized by irradiating aqueous bimetallic ions at room temperature. Their shell and core diameters were recorded by TEM to be 100 - 112 nm and 20 - 50 nm, respectively. The bimetallic radioisotope nanoparticles' gamma spectra showed that they each contained two gamma-emitting nuclides. The nanoparticles could be used as radiotracers in petrochemical and refinery processes that involve temperatures that would decompose conventional organic radioactive labels.
IMPROVEMENT OF CROSS-CORRELATION TECHNIQUE FOR LEAK DETECTION OF A BURIED PIPE IN A TONAL NOISY ENVIRONMENT
Yoon, Doo-Byung ; Park, Jin-Ho ; Shin, Sung-Hwan ;
Nuclear Engineering and Technology, volume 44, issue 8, 2012, Pages 977~984
DOI : 10.5516/NET.09.2011.067
The cross-correlation technique has been widely used for leakage detection of buried pipes, and this technique can be successfully applied when the leakage signal has a high signal-to-noise ratio. In the case of a power plant, the measured leakage signals obtained from the sensors may contain background noise and mechanical noise generated by adjacent machinery. In such a case, the conventional method using the cross-correlation function may fail to estimate the leakage point. In order to enhance the leakage estimation capability of a buried pipe in a noisy environment, an improved cross-correlation technique is proposed. It uses a noise rejection technique in the frequency domain to effectively eliminate the tonal noise due to rotating machinery. Experiments were carried out to verify the validity of the proposed method. The results show that even in a tonal noisy environment, the proposed method can provide more reliable means for estimating the time delay of the leakage signals.