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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 34, Issue 6 - Dec 2002
Volume 34, Issue 5 - Oct 2002
Volume 34, Issue 4 - Aug 2002
Volume 34, Issue 3 - Jun 2002
Volume 34, Issue 2 - Apr 2002
Volume 34, Issue 1 - Feb 2002
Selecting the target year
Evaluation of the KN-12 Spent Fuel Transport Cask by Analysis
Chung, Sung-Hwan ; Lee, Heung-Young ; Song, Myung-Jae ; Rudolf Diersch ; Reiner Laug ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 187~201
The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14
16 and 17
17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.
Electronic States of Uranium Dioxide
Younsuk Yun ; Park, Kwangheon ; Hunhwa Lim ; Song, Kun-Woo ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 202~210
The details of the electronic structure of the perfect crystal provides a critically important foundation for understanding the various defect states in uranium dioxide. In order to understand the local defect and impurity mechanism, the calculation of electronic structure of UO
in the one-electron approximation was carried out, using a semi-empirical tight-binding formalism(LCAO) with and without f-orbitals. The energy band, local and total density of states for both spin states are calculated from the spectral representation of Green’s function. The bonding mechanism in Perfect lattice of UO
is discussed based upon the calculations of band structure, local and total density of states.
Development and High Power RF Test of the Vacuum Feedthrough for KSTAR ICRF Antenna
Bae, Young-Dug ; Hwang, Churl-Kew ; Kwak, Jong-Gu ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 211~217
A 1-MW vacuum feedthrough for the KSTAR ICRF antenna is fabricated and high power RF test is performed. It is designed to have two alumina
ceramic cylinders and O-ring seal instead of a brazed seal for good mechanical and thermal strength, which is important in long pulse or steady state operation. For cooling of the ceramics, dry air is circulated in a space between the two cylinders and the outer conductor. Independent cooling water channels are installed to cool the inner conductor of the feedthrough. RF high voltage test is performed using two kinds of ceramics with the purities of 99.7% and 97%. Stable operation is possible with the RF voltage of 30 kVp at long pulse of 300 sec without any severe damage.
A Control Volume Scheme for Three-Dimensional Transport: Buffer and Matrix Effects on a Decay Chain Transport in the Repository
Lee, Y.M. ; Y.S. Hwang ; Kim, S.G. ; C.H. Kang ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 218~231
Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high- level radioactive waste repository by adopting a finite difference method utilizing the control- volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical harriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three- dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of
Ra, which is the most important chain as far as the human environment is concerned.
A Case Study on the Safety Assessment for Groundwater Pathway in a Near-Surface Radioactive Waste Disposal Facility
Park, Joo-Wan ; Chang, Keun-Moo ; Kim, Chang-Lak ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 232~241
A safety assessment is carried out for the near-surface radioactive waste disposal in the reference engineered vault facility. The analysis is mainly divided into two parts. One deals with the release and transport of radionuclide in the vault and unsaturated zone. The other deals with the transport of radionuclide in the saturated zone and radiological impacts to a human group under well drinking water scenario. The parameters for source-term, geosphere and biosphere models are mainly obtained from the site specific data. The results show that the annual effective doses are dominated by long lived, mobile radionuclides and their associated daughters. And it is found that the total effective dose for drinking water is far below the general criteria of regulatory limit for radioactive waste disposal facility.
Modeling of the Environmental Behavior of Tritium Around the Nuclear Power Plants
Park, Heui-Joo ; Lee, Hansoo ; Kang, Hee-Suk ; Park, Yong-Ho ; Lee, Chang-Woo ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 242~249
The relationship between the tritium release rate from the nuclear power plant and tritium concentration in the environment around the Kori site was modeled. The tritium concentration in the atmosphere was calculated by multiplying the release rates and
/Q values, and the d3V deposition rate at each sector according to the direction and the distance was obtained using a dry deposition velocity. The area around Kori site was divided into 6 zones according to the deposition rate. The six zones were divided into 14 compartments for the numerical simulation. Transfer coefficients between the compartments were derived using site characterization data. Source terms were calculated from the dry deposition rates. Tritium concentration in surface soil water and groundwater was calculated based upon a compartment model. The semi-analytical solution of the compartment model was obtained with a computer program, AMBER. The results showed that most of tritium deposited onto the land released into the atmosphere and the sea. Also, the estimated concentration in the top soil agreed well to that measured. Using the model, tritium concentration was predicted in the case that the tritium release rates were doubled.
Cell Based CMFD Formulation for Acceleration of Whole-core Method of Characteristics Calculations
Cho, Jin-Young ; Joo, Han-Gyu ; Kim, Kang-Seog ; Zee, Sung-Quun ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 250~258
This Paper is to apply the well-established coarse mesh finite difference(CMFD) method to the method of characteristics(MOC) transport calculation as an acceleration scheme. The CMFD problem is first formulated at the pin-cell level with the multi-group structure To solve the cell- based multi-group CMFD problem efficiently, a two-group CMFD formulation is also derived from the multi-group CMFD formulation. The performance of the CMFD acceleration is examined for three test problems with different sizes including a realistic quarter core PWR problem. The CMFD formulation provides a significant reduction in the number of ray tracings and thus only about 9 ray tracing iterations are enough for the realistic problem. In computing time, the CMFD accelerated case is about two or three times faster than the coarse-mesh rebalancing(CMR) accelerated case.
Review on the New Fire Protection Standard for Nuclear Power Plants and Investigation for the Applicability of the Performance-Based Fire Modeling
Jee, Moon-Hak ; Hong, Sung-Yull ; Sung, Chang-Kyung ; Kim, In-Hwang ;
Nuclear Engineering and Technology, volume 34, issue 3, 2002, Pages 259~267
NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.