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 31, Issue 6 - Dec 1999
Volume 31, Issue 5 - Oct 1999
Volume 31, Issue 4 - Aug 1999
Volume 31, Issue 3 - Jun 1999
Volume 31, Issue 2 - Apr 1999
Volume 31, Issue 1 - Feb 1999
Selecting the target year
Development of Self-Actuated Shutdown System Using Curie Point Electromagnet
Kim, Tae-Ryong ; Park, Jin-Ho ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 1~7
An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system(SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet(CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid MEtal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design.
Comparative Evaluation of Three Cognitive Error Analysis Methods Through an Application to Accident Management Tasks in NPPs
Wondea Jung ; Kim, Jaewhan ; Jaejoo Ha ; Wan C. Yoon ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 8~22
This study was performed to comparatively evaluate selected Human Reliability Analysis (HRA) methods which mainly focus on cognitive error analysis, and to derive the requirement of a new human error analysis (HEA) framework for Accident Management (AM) in Nuclear Power Plants (NPPs). In order to achieve this goal, we carried out a case study of human error analysis on an AM task in NPPs. In the study we evaluated three cognitive HEA methods, HRMS, CREAM and PHECA, which were selected through the review of the currently available seven cognitive HEA methods. The task of reactor cavity flooding was chosen for the application study as one of typical tasks of AM in NPPs. From the study, we derived seven requirement items for a new HEA method of AM in NPPs. We could also evaluate the applicability of three cognitive HEA methods to AM tasks. CREAM is considered to be more appropriate than others for the analysis of AM tasks, HRMS is also applicable to the error analysis of AM tasks. But, PHECA is regarded less appropriate for the predictive HEA technique as well as for the analysis of AM tasks. In addition to these, the advantages and disadvantagesofeachmethodaredescribed.
Reference Spent Fuel and Its Characteristics for a Deep Geological Repository Concept Development
Choi, Jong-Won ; Ko, Won-Il ; Kang, Chul-Hyung ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 23~38
This study addresses the reference spent fuel and its characteristics for developing a geological repository concept. As a disposal capacity of the reference repository system to be developed, spent fuel inventories were projected based on the basis of the Nuclear Energy Plan of the Long-term National Power Program. The reference spent fuel encompassing a variability in characteristics of all existing and future spent fuels of interest was defined. Key parameters in the reference fuel screening processes were the nuclear and mechanical design parameters and the burnup histories for existing spent fuels as of 1996 and for future spent fuels with the more extended burnup the initial enrichment and its expected turnup. The selected reference fuel was characterized in terms of initial enrichment, bumup, dimension, gross weight and age. Also the isotopic composition and the radiological properties are quantitatively identified. This information provided in this study could be used as input for repository system development and performance assessment and applied in fuel material balance evaluation for the various types of back-end fuel cycle studies.
Basic Physicochemical and Mechanical Properties of Domestic Bentonite for Use as a Buffer Material in a High-level Radioactive Waste Repository
W. J. Cho ; Lee, J. O. ; K. S. Chun ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 39~50
The physicochemical, mineralogical, hydraulic, swelling and mechanical properties of a domestic bentonite for use as the buffer material in a high-level waste repository have been measured. The bentonite is identified to be a Ca-bentonite, and the hydraulic conductivity of the compacted bentonite with the dry density higher than 1.4 Mg/㎥ is lower than 10
m/s When the dry densities are 1.4 to 1.8 Mg/㎥, the swelling pressures are in the range of 6.6 to 143.5 kg/
. The unconfined compressive strength is about 94 kg/
, and the coefficient of volume change and the coefficient of consolidation are in the range of 0.O0249 to 0.02142
/MN and 0.018 to 0.115
Review on Gas-Voiding Models for HCDA(Hypothetical Core Disruptive Accident) Initiating Phase in LMR Analysis (I)
Chang, W.P. ; Kwon, Y.M. ; Hahn, D.H. ; Suk, S.D. ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 51~65
The present review report introduces the existing analysis codes and physical modeling of two-phase flow associated with initiating event of HCDA in Liquid Metal Reactors for the effective study in the future, because the related research has not been systematically carried out in Korea compared with other areas. The description in this report is specifically addressed to the results yielded from careful review of the technical concepts on the two-phase flow modeling in the SAS2A code which was developed in ANL. The report is prepared in 2 parts based on the definite physical phenomena. The liquid slug and gas behavior models are main representations in the part (I) and (II), respectively. In this regard, it is expected that this report provide a fundamental knowledge on the two-phase flow model in LMR and, thus, contribute to establishment of the necessary HCDA analysis technology concerned with the LMR development in Korea.
Conceptual Safety Design Analyses of Korea Advanced Liquid Metal Reactor
Suk, S.D. ; Park, C.K. ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 66~82
The national long-term R&D program, updated in 1997, requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor(KALIMER), along with supporting R&D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R&D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of HAMMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation.
Technology Assessment of the Repository Alternatives to Establish a Reference HLW Disposal Concept
Choi, Jong-Won ; Choi, Young-Sung ; Kwon, Sang-Ki ; Kuh, Jung-Eui ; Kang, Chul-Hyung ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 83~100
As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than
in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.
Evaluation of Photonuclear Data of Mo, Zn, S and Cl for Applications
Lee, Young-Ouk ; Han, Yin-Lu ; Lee, Jeong-Yeon ; Chang, Jogn-Hwa ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 529~540
As part of IAEA CRP on "Compilation and evaluation of photonuclear data for applications", we evaluated photoproduction data of Mo, Zn, S and Cl isotopes for medical use and biological applications. Available experimental data were collected and their discrepancies were analyzed to select or reconstruct the representative data set. The photoabsorption cross sections were then evaluated tv applying the Giant Dipole Resonance (GDR) model for the energies below about 30 MeV and the quasi-deuteron model for energies below 140 MeV. The resulting representative photoabsorption data were given as input for the theoretical calculations for the emission process of light nuclei including neutron, proton, deuteron, triton,
, alpha particles and gamma rays by use of the Hauser-Feshbach and the preequilibrium model.
Removal of Dissolved Oxygen from the Make-up Water of NPP Using Membrane-based Oxygen Removal System
Chung, Kun-Ho ; Kang, Duck-Won ; Hong, Sung-Yull ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 541~547
Corrosion control, in the end-shield cooling system of Wolsung Nuclear Power Plant, is directly related to the control of dissolved oxygen (DO). The current method, being used to deoxygenate the end-shield cooling water, is a chemical treatment by addition of reducing agent, hydrazine, to react with DO. This method has several limitations including high reaction temperature of hydrazine , unwanted explosive hydrogen gas production, and its intrinsic harmful property. A new approach to remove DO using a membrane-based oxygen removal system (MORS) was tried to overcome limitations of the hydrazine treatment. The DO removal efficiency of the MORS was found to be in the range 87% to 98%: The higher vacuum, the lower water flow rate and the higher water temperature tend to increase the DO removal efficiency.
Air-Water Countercurrent Flow Limitation in a Horizontal Pipe Connected to an Inclined Riser
Kang, Seong-Kwon ; Chu, In-Cheol ; No, Hee-Cheon ; Chun, Moon-Hyun ; Sung, Chang-Kyung ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 548~560
An experimental investigation has been peformed to examine the effects of various geometrical parameters and an initial operating condition on the air-water countercurrent How limitation (CCFL) in a simulated PWR hot leg. A total of 118 experimental data for the onset of CCFL and zero liquid penetration were obtained for various combinations of test parameters. It was observe that the CCFL can be classified into three different categories: (the onset of CCFL, (the partial liquid delivery, and (r) the zero liquid penetration. The observed mechanisms of the onset of CCFL were different depending on the inlet water flow rate. The parametric effects of pipe diameter, horizontal pipe length, horizontal pipe length-to-diameter (L/D) ratio, and initial water level in the horizontal pipe of the test section on the onset of air-water CCFL were also examined. An empirical correlation for the onset of CCFL in a horizontal pipe connected to an inclined riser was developed in terms of Wallis flooding parameters for the low inlet water flow rate region. Comparisons of the present empirical correlation with the air-water CCFL data of large pipe diameters show that the present correlation agrees more closely with the experimental data than the existing CCFL correlations.
Microstructure and Mechanical Properties of Cr-Mo Steels for Nuclear Industry Applications
Kim, Sung-Ho ; Ryu, Woo-Seong ; Kuk, Il-Hiun ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 561~571
Microstructure and mechanical properties of five Cr-Mo steels for nuclear industry applications have been investigated. Transmission electron microscopy, energy dispersive spectrometer, differential scanning calorimeter, hardness, tensile, and impact test were used to evaluate the Cr and W effect on the microstructure and mechanical properties. Microstructures of Cr-Mo steels after tempering are classified into three types : bainitic 2.25Cr-lMo steel, martensitic Mod.9Cr-lMo, HT9M, and HT9W steels, and dual phase HT9 steel. The majority of the precipitates were found to be M
carbides. As minor phases, fine needle-like V(C,N), spherical NbC, fine needle-like Cr-rich Cr
N, and Cr-rich M
were also found. Addition of 2wt.% W in Cr-Mo steels retarded the formation of subgrain and dissolution of Cr
N precipitates. Hardness and ultimate tensile strength increased with increasing Cr content. Though Cr content of HT9W steel was lower than that of HT9 steel, the hardness of HT9W was higher due to the higher W content. W added HT9W steel had the highest ultimate tensile strength above
. But impact toughness of W added steel (HT9W) and high Cr steel (HT9) was low.w.w.
Development of a Linear Stability Analysis Model for Vertical Boiling Channels Connecting with Unheated Risers
Hwang, Dae-Hyun ; Yoo, Yeon-Jong ; Zee, Seong-Quun ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 572~585
The characteristics of two-phase flow instability in a vertical boiling channel connecting with an unheated riser are investigated through the linear stability analysis model. Various two-phase flow models, including thermal non-equilibrium effects, are taken into account for establishing a physical model in the time domain. A classical approach to the frequency response method is adopted for the stability analysis by employing the D-partition method. The adequacy of the linear model is verified by evaluating experimental data at high quality conditions. It reveals that the flow-pattern-dependent drift velocity model enhances the prediction accuracy while the homogeneous equilibrium model shows the most conservative predictions. The characteristics of density wave oscillations under low-power and low-quality conditions are investigated by devising a simple model which accounts for the gravitational and frictional pressure losses along the channel. The necessary conditions for the occurrences of type-I instability and flow excursion are deduced from the one-dimensional D-partition analysis. The parametric effects of some design variables on low quality oscillations are also investigated.
Study on the Effective Stiffness of Base Isolation System for Reducing Acceleration and Displacement Responses
Kim, Young-Sang ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 586~594
To limit both the large displacement and acceleration response of the structure efficiently, the relationships between acceleration and displacement responses of the structure under several earthquakes are investigated for various horizontal stiffness of the base isolation system to determine the effective stiffness of the base isolation system in this paper. An example structure is a five-storey steel frame building as the primary structure and the secondary structures are assumed to be located on the fifth floor of the primary structure. Input motions used in the structural analysis are El Centre 1940, Taft 1952, Mexico 1985, San Fernando 1971 Pacoima Dam, and artificially generated earthquakes. The relationships of the absolute peak acceleration and the displacement at the top of the structure are calculated for various natural periods of base isolators under various earthquakes. The peak acceleration response of the fifth floor in the base isolated structure is significantly reduced by a factor of 2.1 through 6.25. Also, the relative displacement response of the floor to the base of the superstructure is very small. The results of this study can be utilized to determine the effective stiffness of the base isolation system.
Calculation of Proton-Induced Reactions on Ti, Fe, Cu and Mo up to 60 MeV for TLA Application
Kim, Doohwan ; Lee, Young-Ouk ; Jonghwa Chang ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 595~607
The reaction cross-sections of
Tc for TLA application are calculated in the frame of the ECIS-GNASH code system up to 60 MeV. The calculated results are compared with the experimental data taken from the EXFOR at the NEA Data Bank. A preliminary calculation with the global optical parameters of Varner et al. shows considerable differences from the experimental data at low energy range. The global optical parameters for the imaginary volume potential and the diffuseness of the imaginary potential are adjusted to achieve a better description of the experimental data in the vicinities of peak position below 16 MeV. 16 MeV.
Properties of Compacts and Pellets Made Using Bimodal- Sized
Kim, Keon-Sik ; Song, Kun-Woo ; Kang, Ki-Won ; Kim, Jong-Hun ; Kim, Young-Min ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 608~617
The powder mixture which has a bimodal size distribution, with a large mode corresponding to AUC-UO
powder and a small one corresponding to ADU-UO
powder, was prepared, pressed into compacts, and sintered at 1680t for 4 hours in hydrogen gas. The compact density of the powder mixture increases with increasing ADU-UO
content within a content of 20 wt %, since small ADU-UO
particles can fill interstices between large AUC-UO
particles. The UO
pellet made using the powder mixture has a lower open porosity than that made using AUC-UO
powder alone. The mechanism for the formation of a flake-like pore is proposed, and the decrease in open porosity may be ascribed to the decrease in the number of flake-like pores.
A Theoretical and Experimental Study of the Steam Condensation Effect on the CCFL in Nearly Horizontal Two- phase Flow
Chun, Moon-Hyun ; Yu, Seon-Oh ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 618~630
An analytical model that includes the steam condensation effect has been derived and a parametric study has been performed. In addition, a series of experiments were performed and a total of 34 experimental data for the onset of CCFL in nearly horizontal countercurrent two-phase How have been obtained for various flow rates of water. Comparisons of the present CCFL data with slug formation models show that the agreement between the present as well as the existing model and the data is about the same. However, the deviation between the Taitel and Dukler's model predictions and the data is the largest when if j
<0.04 m/s. A parametric study of the effect of the steam condensation using the present model shows that, when all local conditions are similar, the model predicted local gas velocities that cause the onset of flooding are slightly lower when condensation occurred. Based on the visual observation and the evaluation of the present work, it has been concluded that the criterion derived for the onset of slug flow can be directly used to predict the onset of inner flooding in nearly horizontal two-phase flow within the experimental ranges of the present work.
The Calculation of Neutron Scattering Cross Sections for Silicon Crystal at the Thermal Energies
Cho, Young-Sik ; Gil, Choong-Sup ; Jonghwa Chang ;
Nuclear Engineering and Technology, volume 31, issue 6, 1999, Pages 631~637
The module LEAPR of NJOY data processing system has been improved to have the capability of computing the thermal elastic scattering cross sections for silicon, which has a diamond-like structure. Silicon lattice was assumed as an fcc lattice with two atoms at each lattice point. The calculation formulas for thermal neutron elastic scattering by silicon were introduced and incorporated into LEAPR, and then the scattering cross sections for silicon were computed. The results were compared with experimental data, and they were found to give a good agreement with experimental data.