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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 20, Issue 4 - Dec 1988
Volume 20, Issue 3 - Sep 1988
Volume 20, Issue 2 - Jun 1988
Volume 20, Issue 1 - Mar 1988
Selecting the target year
Energy and Angular Response of CR-39 Neutron Track Detector
Kim, Jang-Lyul ; Ha, Chung-Woo ; Yoon, Yea-Chang ; W.G. Cross ; A. Arneja ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 71~79
Published data of the efficiency of CR-39 detectors as a function of neutron energy from different laboratories show wide variations in the response obtained. These variations result from differences in etching conditions, in the materials and thickness of the radiator, in the sensitivities of CR-39 from different manufacturers and perhaps in criteria used for the size of spots that are counted. This paper describes some effects of these factors on the energy and angular response with calculational results. Calculated and measured results of the variations of response with neutron incident angie are more consistent than those of energy response. The data calculated show that the angular response is not a strong function of neutron energy except below about 0.3 MeV.
Analysis of Inter-channel Cross Flow Effect on PWR LOCA
Park, Jong-Ho ; Lee, Sang-Yong ; Han, Ki-In ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 80~87
Predicted in this paper are flow distributions in average and hot channels of the reactor core during small and large break LOCAs. Also estimated based on REALP5/MOD2 calculations are the effects of cross flow between channels on LOCA analysis results. It has been so far generally accepted that a single average channel is sufficient for small break LOCA core hydraulic modelling. However, based on these calculation results, hot channel modeling (two channel modeling) is found necessary in order to guarantee more reliable and conservative results. In large break LOCA blowdown phase, the hot channel thermal hydraulics is worse than that of average channel in both cases with the without consideration of cross flow.
Development of FURA Code and Application for Load Follow Operation
Park, Young-Seob ; Lee, Byong-Whi ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 88~104
The FUel Rod Analysis(FURA) code is developed using two-dimensional finite element methods for axisymmetric and plane stress analysis of fuel rod. It predicts the thermal and mechanical behavior of fuel rod during normal and load follow operations. To evaluate the exact temperature distribution and the inner gas pressure, the radial deformation of pellet and clad, the fission gas release are considered over the full-length of fuel rod. The thermal element equation is derived using Galerkin's techniques. The displacement element equation is derived using the principle of virtual works. The mechanical analysis can accommodate various components of strain: elastic, plastic, creep and thermal strain as well as strain due to swelling, relocation and densification. The 4-node quadratic isoparametric elements are adopted, and the geometric model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The pellet cracking and crack healing, pellet-cladding interaction are modelled. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behavior accurately and stably. The pellet and cladding model has been compared with both analytical solutions and experimental results. The observed and predicted results are in good agreement. The general behavior of fuel rod is calculated by axisymmetric system and the cladding behavior against radial crack is used by plane stress system. The sensitivity of strain aging of PWR fuel cladding tube due to load following is evaluated in terms of linear power, load cycle frequency and amplitude.
Recoil Effects of Neutron-Irradiated Metal Permanganates
Lee, Byung-Hun ; Kim, Jung-Gwan ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 105~111
The chemical effects resulting from the capture of the thermal neutron by manganese in various crystalline permanganates, that is, potassium permanganate ammonium permangante and barium permanganate, have been investigated. The effect of pH of solvent on the distribution of radioactive manganese chemical species, that is, cationic
produced in the permanganates by
Mn reaction was studied by using various adsorbents and ion-exchanger, that is, zeolite A-3, kaolinite, alumina, manganese dioxide and Dowex-50 The distribution of radioactive MnO
in kaolinite and alumina has higher than that in other adsorbents and ion-exchanger at a representative pH value of 4, 7 and 9, respectively. The yield of radioactive MnO
is higher at pH 4 End pH 9 than at pH 7. The thermal annealing behavior of recoil manganese atoms produced in the permanganates by
Mn reaction was also studied. The retention of MnO
in the thermal annealing is increased as annealing temperature increases when it was treated at 10
. The recoil effect of permanganates was explained by the hot zone model.
Forced Flow Dryout Heat Flux in a Stratified Debris Bed
Cha, Jong-Hee ; Chung, Moon-Ki ; Jin, Yong-Suk ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 112~119
The purpose of this work is to obtain the experimental data for the forced How dryout heat flux in a heat generating stratified debris bed which simulates the degraded nuclear reactor core after severe accident. The present observations were mainly focused on the effect of coolant mass flux on the dryout heat flux in the stratified debris bed which consists of several layers with selected particle sizes under constant bed depth and temperature of inlet coolant flow conditions. The following results were obtained from this experimental work: (1) The dryout heat flux in the stratified debris bed increases with increase of upward forcing mass flux of coolant. The similar trend of increase rate of dryout heat flux in the stratified bed was observed in the uniform particle size bed. (2) For the comparison of theoretical values and experimental data, the use of surface area mean diameter as a particle diameter was suitable for the calculation of dryout heat flux.
Mathematical Modeling for Leaching and dissolution of Solidified Radioactive Waste in a Geologic Reposiory
Kim, Chang-Lak ; Park, Kwang-Sub ; Cho, Chan-Hee ; Kim, Jhinwung ; Suh, In-Suk ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 120~131
A souce term model describes mathematically the source of radionuclides as they begin slow migration and decay in deep groundwater. Various source term models based on mass-transfer analysis and measurement-based source term models are reviewed. Ganerally, two processes are involved in leaching or dissolution: (1) chemical reactions and (2) mass transfer by diffusion. The chemical reaction controls the dissolution rates only during the early stage of exposure to groundwater. The exterior-field mass transfer may control the long term dissolution rates from the waste solid in a geologic repository. Mass-transfer analyses re3y on detailed and careful application of the governing equations that describe the mechanistic processes of transport of material between and within phases. If used correctly, source term models based on mass-transfer theory are valuable and necessary tools for developing reliable predictions.
Approximations for Array of Point Sources in Groundwater Contaminant Transport Modeling
Kim, Chang-Lak ;
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 132~136
A strategic question in groundwater contaminant transport modeling is whether we need to treat waste packages or drums as individual, discrete sources or as approximately lumped sources. In this paper we present analyses of array sources in porous media. We analyze a planar array of sources in porous media with groundwater flow. We compare the concentration field predicted by a detailed model of individual point sources to concentration fields predicted by an infinite plane source and a single point source, all of the same equivalent strength. From this study we identified three regions: (1) a region close to the sources where the effects of adjacent sources are significant and individual source models should be used, (2) a region extending from a few meters to hundreds to thousands of meters downstream, where an equivalent source of infinite extent gives accurate results, and (3) a far-field region, where in an equivalent source of finite extent gives accurate results.
우리나라 원전의 현황과 과제
Nuclear Engineering and Technology, volume 20, issue 2, 1988, Pages 137~143