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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
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
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NUCLEAR ENERGY MATERIALS PREDICTION: APPLICATION OF THE MULTI-SCALE MODELLING PARADIGM
Samaras, Maria ; Victoria, Maximo ; Hoffelner, Wolfgang ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 1~10
DOI : 10.5516/NET.2009.41.1.001
The safe and reliable performance of fusion and fission plants depends on the choice of suitable materials and an assessment of long-term materials degradation. These materials are degraded by their exposure to extreme conditions; it is necessary, therefore, to address the issue of long-term damage evolution of materials under service exposure in advanced plants. The empirical approach to the study of structural materials and fuels is reaching its limit when used to define and extrapolate new materials, new environments, or new operating conditions due to a lack of knowledge of the basic principles and mechanisms present. Materials designed for future Gen IV systems require significant innovation for the new environments that the materials will be exposed to. Thus, it is a challenge to understand the materials more precisely and to go far beyond the current empirical design methodology. Breakthrough technology is being achieved with the incorporation in design codes of a fundamental understanding of the properties of materials. This paper discusses the multi-scale, multi-code computations and multi-dimensional modelling undertaken to understand the mechanical properties of these materials. Such an approach is envisaged to probe beyond currently possible approaches to become a predictive tool in estimating the mechanical properties and lifetimes of materials.
MULTISCALE MODELING OF RADIATION EFFECTS ON MATERIALS: PRESSURE VESSEL EMBRITTLEMENT
Kwon, Jun-Hyun ; Lee, Gyeong-Geun ; Shin, Chan-Sun ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 11~20
DOI : 10.5516/NET.2009.41.1.011
Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multi scale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.
METALLIC INTERFACES IN HARSH CHEMO-MECHANICAL ENVIRONMENTS
Yildiz, Bilge ; Nikiforova, Anna ; Yip, Sidney ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 21~38
DOI : 10.5516/NET.2009.41.1.021
The use of multi scale modeling concepts and simulation techniques to study the destabilization of an ultrathin layer of oxide interface between a metal substrate and the surrounding environment is considered. Of particular interest are chemo-mechanical behavior of this interface in the context of a molecular-level description of stress corrosion cracking. Motivated by our previous molecular dynamics simulations of unit processes in materials strength and toughness, we examine the challenges of dealing with chemical reactivity on an equal footing with mechanical deformation, (a) understanding electron transfer processes using first-principles methods, (b) modeling cation transport and associated charged defect migration kinetics, and (c) simulation of pit nucleation and intergranular deformation to initiate the breakdown of the oxide interlayer. These problems illustrate a level of multi-scale complexity that would be practically impossible to attack by other means; they also point to a perspective framework that could guide future research in the broad computational science community.
MULTI-SCALE MODELS AND SIMULATIONS OF NUCLEAR FUELS
Stan, Marius ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 39~52
DOI : 10.5516/NET.2009.41.1.039
Theory-based models and high performance simulations are briefly reviewed starting with atomistic methods, such as Electronic Structure calculations, Molecular Dynamics, and Monte Carlo, continuing with meso-scale methods, such as Dislocation Dynamics and Phase Field, and ending with continuum methods that include Finite Element and Finite Volume. Special attention is paid to relating thermo-mechanical and chemical properties of the fuel to reactor parameters. By inserting atomistic models of point defects into continuum thermo-chemical calculations, a model of oxygen diffusivity in
is developed and used to predict point defect concentrations, oxygen diffusivity, and fuel stoichiometry at various temperatures and oxygen pressures. The simulations of coupled heat transfer and species diffusion demonstrate that including the dependence of thermal conductivity and density on composition can lead to changes in the calculated centerline temperature and thermal expansion displacements that exceed 5%. A review of advanced nuclear fuel performance codes reveals that the many codes are too dedicated to specific fuel forms and make excessive use of empirical correlations in describing properties of materials. The paper ends with a review of international collaborations and a list of lessons learned that includes the importance of education in creating a large pool of experts to cover all necessary theoretical, experimental, and computational tasks.
EXPERIMENTS ON THE PERFORMANCE SENSITIVITY OF THE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM OF AN ADVANCED INTEGRAL TYPE REACTOR
Park, Hyun-Sik ; Choi, Ki-Yong ; Choi, Seok ; Yi, Sung-Jae ; Park, Choon-Kyung ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 53~62
DOI : 10.5516/NET.2009.41.1.053
A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.
APPLICATION OF SEVERE ACCIDENT MANAGEMENT GUIDANCE IN THE MANAGEMENT OF AN SGTR ACCIDENT AT THE WOLSONG PLANTS
Jin, Young-Ho ; Park, Soo-Yong ; Song, Yong-Mann ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 63~70
DOI : 10.5516/NET.2009.41.1.063
A steam generator tube rupture (SGTR) accident, which is a partial reactor building bypass scenario, has a low probability and high consequences. SAMG has been used to manage the progression of severe accidents and the release of fission products induced by an SGTR at the Wolsong plants. Four of the six SAGs in the SAMG are used to manage the progression of a severe accident induced by an SGTR at the Wolsong plants. The results of the ISAAC code calculation have shown that the proper use the SAMG can stop a severe accident from progressing and keep the reactor building intact during a severe accident. These results confirm that the SAMG is an effective means of managing the progression of severe accidents initiated by an SGTR at the Wolsong plants.
DESIGN OF DELAY-TOLERANT CONTROLLER FOR REMOTE CONTROL OF NUCLEAR REACTOR POWER
Lee, Yoon-Joon ; Na, Man-Gyun ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 71~78
DOI : 10.5516/NET.2009.41.1.071
One of main concepts involved in regional small nuclear reactors is unmanned remote control. Internet-based virtual private networks provide environments for the remote monitoring and control of geographically-dispersed systems, and with the advances in communication technologies, the potential of networks for real time control and automation becomes enormous. However, networked control has some problems. The most critical is delay in signal transmission, which degrades system stability and performance. Therefore, a networked control system should be designed to account for delay. This paper proposes some design approaches for a delay-tolerant system that can guarantee predetermined stability margins and performance. To accomplish this, the reactor plant is modeled with consideration of uncertainties. With this model, three kinds of controllers are developed using different methods. The designed systems are compared with respect to stability and performance, and a second-order controller designed using the table lookup method was found to give the most satisfactory results.
VERIFICATION OF PLC PROGRAMS WRITTEN IN FBD WITH VIS
Yoo, Jun-Beom ; Cha, Sung-Deok ; Jee, Eun-Kyung ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 79~90
DOI : 10.5516/NET.2009.41.1.079
Verification of programmable logic controller (PLC) programs written in IEC 61131-3 function block diagram (FBD) is essential in the transition from the use of traditional relay-based analog systems to PLC-based digital systems. This paper describes effective use of the well-known verification tool VIS for automatic verification of behavioral equivalences between successive FBD revisions. We formally defined FBD semantics as a state-transition system, developed semantic-preserving translation rules from FBD to Verilog programs, implemented a software tool to support the process, and conducted a case study on a subset of FBDs for APR-1400 reactor protection system design.
"3+3 PROCESS" FOR SAFETY CRITICAL SOFTWARE FOR I&C SYSTEM IN NUCLEAR POWER PLANTS
Jung, Jae-Cheon ; Chang, Hoon-Sun ; Kim, Hang-Bae ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 91~98
DOI : 10.5516/NET.2009.41.1.091
The "3+3 Process" for safety critical software for nuclear power plants' I&C (Instrumentation and Control system) has been developed in this work. The main idea of the "3+3 Process" is both to simplify the software development and safety analysis in three steps to fulfill the requirements of a software safety plan . The "3-Step" software development process consists of formal modeling and simulation, automated code generation and coverage analysis between the model and the generated source codes. The "3-Step" safety analysis consists of HAZOP (hazard and operability analysis), FTA (fault tree analysis), and DV (design validation). Put together, these steps are called the "3+3 Process". This scheme of development and safety analysis minimizes the V&V work while increasing the safety and reliability of the software product. For assessment of this process, validation has been done through prototyping of the SDS (safety shut-down system) #1 for PHWR (Pressurized Heavy Water Reactor).
DETERMINATION OF THE TRANSURANIC ELEMENTS INVENTORY IN HIGH BURNUP PWR SPENT FUEL SAMPLES BY ALPHA SPECTROMETRY-II
Joe, Kih-Soo ; Song, Byung-Chul ; Kim, Young-Bok ; Jeon, Young-Shin ; Han, Sun-Ho ; Jung, Euo-Chang ; Song, Kyu-Seok ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 99~106
DOI : 10.5516/NET.2009.41.1.099
The contents of transuranic elements (
) in high-burnup spent fuel samples (
) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about
on average according to each isotope, and those for americium and curium were also higher by about
. However, for
, the measurements were lower by about 52% on average for the samples.
CORROSION BEHAVIOR OF NI-BASE ALLOYS IN SUPERCRITICAL WATER
Zhang, Qiang ; Tang, Rui ; Li, Cong ; Luo, Xin ; Long, Chongsheng ; Yin, Kaiju ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 107~112
DOI : 10.5516/NET.2009.41.1.107
Corrosion of nickel-base alloys (Hastelloy C-276, Inconel 625, and Inconel X-750) in
, 25MPa supercritical water (with 10 wppb oxygen) was investigated to evaluate the suitability of these alloys for use in supercritical water reactors. Oxide scales formed on the samples were characterized by gravimetry, scanning electron microscopy/energy dispersive spectroscopy, X-ray diffraction, and X-ray photoelectron spectroscopy. The results indicate that, during the 1000h exposure, a dense spinel oxide layer, mainly consisting of a fine Cr-rich inner layer (
) underneath a coarse Fe-rich outer layer (
), developed on each alloy. Besides general corrosion, nodular corrosion occurred on alloy 625 possibly resulting from local attack of
" clusters in the matrix. The mass gains for all alloys were small, while alloy X -750 exhibited the highest oxidation rate, probably due to the absence of Mo.
ANALYSIS OF DOPPLER-BROADENED PEAK IN THERMAL NEUTRON INDUCED
Li REACTION USING HYPERGAM
Choi, Hee-Dong ; Jung, Nam-Suk ; Park, Byung-Gun ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 113~124
DOI : 10.5516/NET.2009.41.1.113
The line shape functions for the Doppler-broadened gamma ray spectrum are considered in the
reaction occurring in a surrounding medium where the excited
nucleus is slowed down and stopped before decay. The phenomenological form of the stopping power was used for the broadening effect. Convolution with the detailed response of a germanium detector is taken into consideration for the simplest case of solely electronic stopping. A numerical study for the analysis of
by thermal neutron capture is conducted by performing a parametric search and fitting the measured spectrum in a least-squares approach. In comparison with the previous numerical approach using the same analysis, the computational speed is increased and reliable information concerning the stopping power of the medium is obtained while estimating the uncertainty. Implementation of the routine analysis of
is facilitated on a recent version of the gamma ray spectrum analysis package HyperGam.
DEVELOPMENT OF A STEAM GENERATOR TUBE INSPECTION ROBOT WITH A SUPPORTING LEG
Shin, Ho-Cheol ; Jeong, Kyung-Min ; Jung, Seung-Ho ; Kim, Seung-Ho ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 125~134
DOI : 10.5516/NET.2009.41.1.125
This paper presents details on a tube inspection robotic system and a positioning method of the robot for a steam generator (SG) in nuclear power plants (NPPs). The robotic system is separated into three parts for easy handling, which reduces the radiation exposure during installation. The system has a supporting leg to increase the rigidity of the robot base. Since there are several thousands of tubes to be inspected inside a SG, it is very important to position the tool of the robot at the right tubes even if the robot base is positioned inaccurately during the installation. In order to obtain absolute accuracy of a position, the robot kinematics was mathematically modeled with the modified DH(Denavit-Hartenberg) model and calibrated on site using tube holes as calibration points. To tune the PID gains of a commercial motor driver systematically, the time delay control (TDC) based gain tuning method was adopted. To verify the performance of the robotic system, experiments on a Framatomes 51B Model type SG mockup were undertaken.
EXPERIMENTAL APPROACHES FOR WATER DISCHARGE CHARACTERISTICS IN PEMFC USING NEUTRON IMAGING TECHNIQUE AT CONRAD, HMI
Kim, Tae-Joo ; Kim, Jong-Rok ; Sim, Cheul-Muu ; Lee, Sung-Ho ; Son, Young-Jin ; Kim, Moo-Hwan ;
Nuclear Engineering and Technology, volume 41, issue 1, 2009, Pages 135~142
DOI : 10.5516/NET.2009.41.1.135
In this investigation, we prepared a 1 and 3-parallel serpentine single PEMFC, which has an active area of
and a flow channel cross section of
. Distribution and transport of water in a non-operating PEMFC were observed by varying flow types and the flow rates (250, 400, and 850 cc/min). This investigation was performed at the neutron imaging facility at the CO1d Neutron RAdiography facility (CONRAD), HMI, Germany of which the collimation ratio and neutron fluence rate are 250,
, respectively. The neutron image was continuously recorded by a scintillator and lens-CCD coupled detector system every 10 seconds. It has been observed that although the distilled water was supplied into the cathode channel only, the neutron image showed a water movement from the cathode to the anode channel. The water at the cathode channel was completely discharged as soon as the pressurized air was supplied. But the water at the anode channel was not easily removed by the pressurized air except for the 3-parallel serpentine type with 850cc/min of air flow rate. Moreover, the water at the MEA wasn't removed for any of the cases.