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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 33, Issue 6 - Dec 2001
Volume 33, Issue 5 - Oct 2001
Volume 33, Issue 4 - Aug 2001
Volume 33, Issue 3 - Jun 2001
Volume 33, Issue 2 - Apr 2001
Volume 33, Issue 1 - Feb 2001
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Investigation of the Sensitivity Depletion Laws for Rhodium Self-Powered Neutrorn Detectors (SPNDs)
Kim, Gil-Gon ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 121~131
An investigation of the sensitivity depletion laws for rhodium SPNDs was performed to reduce the uncertainty of the sensitivity depletion laws used in Combustion Engineering (CE) reactors and to develop calculational tools that provide the sensitivity depletion laws to interpret the signal of the newly designed rhodium SPND into the local neutron flux. The calculational tools developed in this work are computer programs for a time-dependent neutron flux distribution in the rhodium emitter during depletion and for a time-dependent beta escape probability that a beta particle generated in the emitter escapes into the collector. These programs provide the sensitivity depletion laws and show the reduction of the uncertainty by about 1 % compared to that of the method employed by CE in interpreting the signal into the local neutron flux. A reduction in the uncertainty by 1 % in interpreting the signal into the local neutron flux reduces the uncertainty tv about 1 % in interpreting the signal into the local power and lengthens the lifetime of the rhodium SPND by about 10% or more.
Treatment of Stainless Steel Cladding in Pressurized Thermal Shock Evaluation: Deterministic Analyses
Changheui Jang ; Jeong, lll-Seok ; Hong, Sung-Yull ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 132~144
Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property, at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined.
Simulation of Water/steam into Sodium Leak Behavior for an Acoustic Noise Generation Mechanism Study
Kim, Tae-Joon ; Hwang, Sung-Tai ; Jeong, Kyung-Chai ; Park, Jong-Hyeun ; Valery S. Yughay ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 145~155
This simulation first allows us to define a transition zone from a bubble to jet mode of the argon out-flow and hereinafter to define a similar area for water-steam leak in the KALIMER SG (Korea Advanced Liquid Metal Reactor Steam Generator) using a water mock-up system, taking into account the KALIMER leak classification and tube bundle design, as a simulation of a real water-steam into sodium leak. in accordance with leak conditions in the KALIMER SG, the transition from bubbling to jetting is studied by means of turbulence regime simulation for argon out-flow through a very small orifice, which has the equivalent diameter of about 0.253 mm. finally the noise generation mechanism is explained from the existing experimental data. We also confirmed the possibility of micro-leak detection from the information of the bubbling mode through simulations and the experiment in this study.
Important Radionuclides and Their Sensitivity for Ground water Pathway of a Hypothetical Near-Surface Disposal Facility
Park, J. W. ; K. Chang ; Kim, C. L. ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 156~165
A radiological safety assessment was performed for a hypothetical near-surface radioactive waste repository as a simple screening calculation to identify important nuclides and to provide insights on the data needs for a successful demonstration of compliance. Individual effective doses were calculated for a conservative ground water pathway scenario considering well drilling near the site boundary. Sensitivity of resulting ingestion dose to input parameter values was also analyzed using Monte Carlo sampling. Considering peak dose rate and assessment time scale, C-14 and T-129 were identified as important nuclides and U-235 and U-238 as potentially important nuclides. For C-14, the dose was most sensitive to Darcy velocity in aquifer The distribution coefficient showed high degree of sensitivity for I-129 release.
An Apparatus for Monitoring Real-time Uranium Concentration Using Fluorescence Intensity at Time Zero
Lee, Sang-Mock ; Shin, Jang-Soo ; Kang, Shin-Won ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 166~174
An apparatus for detecting remote real-time uranium concentration using an optrode was developed. An optrode to detect uranium fluorescence as remote real-time control was designed. Fluorescence intensity at time 2ero was derived by the fluorescence signal processing and the algorithm to exclude the quenching effect of various quenchers and temperature fluctuations. This apparatus employing the above deriving method and the optrode has an error range within 6% in spite of serious fluorescence lifetime changes due to the quenching effect and temperature fluctuations. The detection limit is 0.06 ppm and the linearity is excellent between 0.06 ppm and 2 ppm on the aqueous uranium solution.
Investigation on Nd:YAG Laser Weldability of Zircaloy-4 End Cap Closure for Nuclear Fuel Elements
Kim, Soo-Sung ; Lee, Chul-Yung ; Yang, Myung-Seung ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 175~183
Various welding processes are now available for end cap closure of nuclear fuel element such as TG(Tungsten Inert Gas) welding, magnetic resistance welding and laser welding. Even though the resistance and TIG welding processes are widely used for manufacturing commercial fuel elements, they can not be recommended for the remote seal welding of a fuel element at a hot cell facility due to the complexity of electrode alignment, difficulity in the replacement of parts in the remote manner and a large heat input for a thin sheath. Therefore, the Nd:YAG laser system using optical fiber transmission was selected for Zircaloy-4 end cap welding inside hot cell. The laser welding apparatus was developed using a pulsed Nd:YAG laser of 500 watt average power with optical fiber transmission. The weldability of laser welding was satisfactory with respect to the microstructures and mechanical properties comparing with TIG and resistance welding. The optimum operation processes of laser welding and the optical fiber transmission system for hot cell operation in a remote manner have been developed The effects of irradiation on the properties of the laser apparatus were also being studied.
Characteristics of Heat Shrinkable High Density Polyethylene Crosslinked by
Kang, Phil-Hyun ; Nho, Young-Chang ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 184~191
The effects of
-irradiation on the crosslinking of high density polyethylene (HDPE) was investigated for the purpose of obtaining a suitable formulation for heat shrinkable materials. In this study the HDPE specimens were prepared by blending with cross linking agents and pressed into a 0.2 mm sheet at 18
-irradiation was conducted at 40 to 100 kGy in nitrogen. The heat shrinkable property and thermal mechanical property of the HDPE sheets have been investigated. It was found that the degree of crosslinking of the irradiated HDPE samples were increased with irradiation dose. Compared with the HDPE containing triallylisocyanurate, the HDPE containing trimethylol propane triacrylate shows a slight increase in crosslinking density. The heat transformation and dimension change of HDPE decreased with increasing radiation dose. The heat shrinkage of the samples increased with increasing annealing temperatures. The thermal resistance of HDPE increased upon the crosslinking of HDPE.
Formation and Growth of Hydride Blisters in Zr-2.5Nb Pressure Tubes
Cheong, Yong-Moo ; Gong, Un-Sik ; Choo, Ki-Nam ; Kim, Sung-Soo ; Kim, Young-Suk ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 192~200
Hydride blisters were formed on the outer surface of Zr-2.5Nb pressure tube by a non- uniform steady thermal diffusion process. A thermal gradient was applied to the pressure tube with a heat bath kept at a temperature of 415
and an aluminum cold finger cooled with flowing water of 15
. Optical microscopy and tree-dimensional laser profilometry were used to characterize the hydride blisters with different hydrogen concentrations and thermal diffusion time. Hydride blisters were expected to start at a hydrogen concentration of 30 - 70 ppm and a thermal diffusion time of 4 - 6
sec. The hydride blister size increases with higher hydrogen concentrations and longer thermal diffusion time . Some of the samples revealed cracks on the hydride blisters. The ratio of hydride blister depth to height was estimated as approximately 8: 1.
Development of Liquid Stub and Phase Shifter
Wang, Son-Jong ; Yoon, Jae-Sung ; Hong, Bong-Guen ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 201~208
The high power RF transmission line components are required for transmitting MW level RF power continuously in RF heating and current drive system which heat the plasma and produce plasma current in fusion reactor The liquid stub and phase shifter is proposed as the superior to the conventional stub and phase shifter. Experimental results show that they are reliable and easy to operate compared to the conventional stub and phase shifter. There is no distortion of reflected power during the raising of the liquid level. RF breakdown voltage is over 40kV. Temperature increment of the liquid is expected not to be severe. These results verify that the liquid stub and phase shifter can be used reliably in the high power continuous RF facilities.
Development of a System Analysis Code, SSC-K, for Inherent Safety Evaluation of The Korea Advanced Liquid Metal Reactor
Kwon, Young-Min ; Lee, Yong-Bum ; Chang, Won-Pyo ; Dohee Hahn ; Kim, Kyung-Doo ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 209~224
The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.
Determination of Trace Uranium in Human Hair by Nuclear Track Detection Technique
Chung, Yong-Sam ; Moon, Jong-Hwa ; Zinaida En ; Cho, Seung-Yeon ; Kang, Sang-Hoon ; Lee, Jae-Ki ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 225~230
The aim of this study is to describe a usefulness of nuclear analytical technique in assessing and comparing the concentration levels through the analysis of uranium using human hair sample in the field of environment. A fission track detection technique was applied to determine the uranium concentration in human hair. Hair samples were collected from two groups of people - a) workers not dealing with uranium directly, and b) workers possibly contaminated with uranium. The concentration of
U for the first group varied from <1 to 39 ng/g and the second group can be estimated up to the level of
g/g. Radiographs of heavy-duty work samples contained high dense “hot spots” along a single hair. After washing in acetone and distilled water, external contamination was not totally removed. Insoluble uranium compounds were not completely washed out. The (n, f)- radiography technique, having high sensitivity, and capable of getting information on uranium content at each point of a single hair, is an excellent tool for environmental monitoring.
Selection of Key Radionuclides for P&T Based on Radiological Impact Assessment for the Deep Geological Disposal of Spent PWR/CANDU/DUPIC Fuels
Lee, Dong-Won ; Chung, Chang-Hyun ; Kim, Chang-Lak ; Park, Joo-Wan ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 231~240
When it is assumed that PWR, CANDU and DUPIC spent fuels are disposed of in deep geological repository, consequent annual individual doses are calculated, and it is shown that doses meet the regulatory limit. From these results, the hazardous radionuclides applicable to partitioning and transmutation are selected. These selected radionuclides such as Tc-99, Ⅰ-129, Cs-135 and Np-237 are then reviewed in terms of partitioning and transmutation. Separation of I-129, Np-237 and Tc-99 from spent fuels is considered desirable, and transmutation of these radionuclides results in remarkable hazard reduction. However, it is concluded that separation and transmutation of Cs-135 may be ineffective although it is classified into a hazardous radionuclide.
Effects of Condensation Heat Transfer Model in Calculation for KNGR Containment Pressure and Temperature Response
Eoh, Jae-Hyuk ; Park, Shane ; Jeun, Gyoo-Dong ; Kim, Moo-Hwan ;
Nuclear Engineering and Technology, volume 33, issue 2, 2001, Pages 241~253
Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.