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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 36, Issue 6 - Dec 2004
Volume 36, Issue 5 - Oct 2004
Volume 36, Issue 4 - Aug 2004
Volume 36, Issue 3 - Jun 2004
Volume 36, Issue 2 - Apr 2004
Volume 36, Issue 1 - Feb 2004
Selecting the target year
Application of Geometry-Efficiency Variation Technique to Activity Measurement of
for 3-PM Liquid Scintillation Counting
Lee Hwa Yong ; Seo Ji Suk ; Kwak Ji Yeon ; Hwang Han-Yull ; Lee K. B. ; Lee Jong Man ; Park Tae Soon ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 121~126
3-PM liquid scintillation counting using the geometry-efficiency variation technique has been applied to the activity measurement of
, which decays to
and E.C., respectively. The TDCR values K have been derived over a wide range, 0.78 < K < 0.97, by displacing the detectors up to 50 mm away from an unquenched liquid scintillation sample
. The derived plots of the logic sums of double coincidences
very K vary linearly in the observed regions. The fractions of losses due to electron capture decay have been taken into account by employing a PENELOPE Monte Carlo simulation. The calibrated activity is 102.3 kBq at a reference date of July 1st, 2002 (UT) with a combined uncertainty of
. This is consistent with the value determined by means of the CIEMAT/NIST method at KRISS.
Characteristics of the Integrated Steam Generators for a Liquid Metal Reactor
Sim Yoon Sub ; Kim Eui Kwang ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 127~141
Various types of integrated steam generators, which integrate IHTS and a steam generator into a single unit of equipment for an LMR, were analyzed using an analytic solution with some simplification. The analysis showed that the undesirable reversed heat transfer, of which occurrence was previously observed only in an integrated single-region bundle type, can also occur in an integrated double-region bundle type. The mechanism of the reversed heat transfer occurrence in the double-region type is explained and it is shown the mechanism in the double-region type is completely different from that in the single-region type. Based on this finding, a method for preventing the aforementioned heat transfer is suggested. The performance of the four types of the integrated steam generators is assessed. For this assessment, a SG is actually designed for each type and the optimization in the geometric parameters and flow rate are optimized.
Modeling of Liquid Entrainment and Vapor Pull-Through in Header-Feeder Pipes of CANDU
Cho Yong Jin ; Jeun Gyoo Dong ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 142~152
The liquid entrainment and vapor pull-through offtake model of RELAP5/MOD3 had been developed for SBLOCA (Small Break Loss of Coolant Accident). The RELAP5/MOD3 model for horizontal volumes accounts for the phase separation phenomena and computes the flux of mass and energy through a branch when stratified conditions occur in the horizontal pipe. In the case of CANDU reactor, this model should be used in the coolant flow of 95 feeders connected to the reactor header component under the horizontal stratification in header. The current RELAP5 model can treat the only 3 directions junctions; vertical upward, downward, and side oriented junctions, and thus improvements for the liquid entrainment and vapor pull-through model were needed for considering the exact angles. The RELAP5 off-take model was modified and generalized by considering the geometric effect of branching angles. Based on the previous experimental results, the critical height correlation was reconstructed by use of the branch line connection angle and validation analyses were also performed using SET. The new model can be applied to vertical upward, downward and angled branch, and the accuracy of the new correlations is more improved than that of RELAP5.
Numerical Analysis of the Turbulent Flow and Heat Transfer in a Heated Rod Bundle
In Wang-Kee ; Shin Chang-Hwan ; Oh Dong-Seok ; Chun Tae-Hyun ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 153~164
A computational fluid dynamics (CFD) analysis has been performed to investigate the turbulent flow and heat transfer in a triangular rod bundle with pitch-to-diameter ratios (P/D) of 1.06 and 1.12. Anisotropic turbulence models predicted the turbulence-driven secondary flow in a triangular subchannel and the distributions of the time mean velocity and temperature, showing a significantly improved agreement with the measurements from the linear standard
model. The anisotropic turbulence models predicted the turbulence structure for a rod bundle with a large P/D fairly well, but could not predict the very high turbulent intensity of the azimuthal velocity observed in the narrow flow region (gap) for a rod bundle with a small P/D.
Wire-wrap Models for Subchannel Blockage Analysis
Ha K.S. ; Jeong H.Y. ; Chang W.P. ; Kwon Y.M. ; Lee Y.B. ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 165~174
The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.
Implementation of a Dry Process Fuel Cycle Model into the DYMOND Code
Park Joo Hwan ; Jeong Chang Joon ; Choi Hangbok ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 175~183
For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.
ATWS Frequency Quantification Focusing on Digital I&C Failures
Kang Hyun Gook ; Jang Seung-Cheol ; Lim Ho-Gon ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 184~195
The multi-tasking feature of digital I&C equipment could increase risk concentration because the I&C equipment affects the actuation of the safety functions in several ways. Anticipated Transient without Scram (ATWS) is a typical case of safety function failure in nuclear power plants. In a conventional analysis, mechanical failures are treated as the main contributors of the ATWS. This paper quantitatively presents the probability of the ATWS based on a fault tree analysis of a Korea Standard Nuclear Power Plant is also presented. An analysis of the digital equipment in the digital plant protection system. The results show that the digital system severely affects the ATWS frequency. We also present the results of a sensitivity study, which show the effects of the important factors, and discuss the dependency between human operator failure and digital equipment failure.
Effect of pH, Redox Potential (Eh) and Carbonate Concentration on Actinides Solubility in a Deep Groundwater of Korea
Keum Dong-Kwon ; Lee Han-Soo ; Lee Chang-Woo ;
Nuclear Engineering and Technology, volume 36, issue 2, 2004, Pages 196~202
KAERI (Korea Atomic Energy Research Institute) is at present preparing a preliminary performance assessment to set up the HLW disposal concept of Korea. The solubility of the radionuclides contained in HLW is necessary as a source term in order to predict their potential migration in both the near and far fields. The solubility of actinides (Th, Am, U, Np and Pu) for a reference deep groundwater of Korea has been calculated using a geochemical code with thermodynamic data selected by a peer review of existing thermodynamic databases and literature. The solubilities from the experimental study and/or field observations from natural analogue studies are compared. The sensitivity of solubility to the variability of three main parameters of groundwater (pH, Eh, and carbonate concentration) is also investigated. The results of the sensitivity analysis show that the solubility of actinides strongly depends on the parameters considered. Within the range of parameter values studied (pH=7 to 10, Eh=-0.4 to -0.1V, and carbonate concentration=1.E-5 to 1.E-2 mol/L), the solubility of each actinide exists between 1.4E-10 and 1.6E-6 mol/L for Am, 4.9E-9 and 2.8E-6 mol/L for Th, 3.2E-9 and 5.7E-4 mol/L for U, 1.1E-9 and 1.0E-7 mol/L for Np, and 4.0E-11 and 2.8E-6 mol/L for Pu, respectively.