Go to the main menu
Skip to content
Go to bottom
REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
Journal Basic Information
Journal DOI :
The Korean Radioactive Waste Society
Editor in Chief :
Volume & Issues
Volume 4, Issue 4 - Dec 2006
Volume 4, Issue 3 - Sep 2006
Volume 4, Issue 2 - Jun 2006
Volume 4, Issue 1 - Mar 2006
Selecting the target year
Determination of Pu Oxidation states in the HCl Media Using with UV-Visible Absorption Spectroscopic Techniques
Lee, Myung-Ho ; Suh, Mu-Yeol ; Park, Kyoung-Kyun ; Park, Yeong-Jae ; Kim, Won-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 1~7
The spectroscopic characteristics of Pu (III, IV, V, VI) in the HCl media were investigated by measuring Pu oxidation states using a UV-Vis-NIR spectrophotometer (400-1200 nm) after adjusting Pu oxidation states with oxidation/reduction reagents. Pu in stock solution was reduced to Pu(III) with
HCl, and oxidized to Pu(IV) and Pu(VI) with
, respectively. Also, Pu(V) was adjusted in the Pu(VI) solution with
HCl. The major absorption peaks of Pu (IV) and Pu(III) were measured in the 470 m and 600 nm, respectively. The major absorption peaks of Pu (VI) and Pu(V) were measured in the 830 nm and 1135 nm, respectively. There was not found to be significant changes of UV-Vis absorption spectra for Pu(III), Pu(IV) and Pu(VI) with aging time, except that an unstable Pu(V) immediately reduced to Pu(III).
A Study on the Constructing Discrete Fracture Network in Fractured-Porous Medium with Rectangular Grid
Han, Ji-Woong ; Hwang, Yong-Soo ; Kang, Chul-Hyung ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 9~15
For the accurate safety assessment of potential radioactive waste disposal site which is located in the crystalline rock it is important to simulate the mass transportation through engineered and natural barrier system precisely, characterized by porous and fractured media respectively. In this work the methods to construct discrete fracture network for the analysis of flow and mass transport through fractured-porous medium are described. The probability density function is adopted in generating fracture properties for the realistic representation of real fractured rock. In order to investigate the intersection between a porous and a fractured medium described by a 2 dimensional rectangular and a cuboid grid respectively, an additional imaginary fracture is adopted at the face of a porous medium intersected by a fracture. In order to construct large scale flow paths an effective method to find interconnected fractures and algorithms of swift detecting connectivities between fractures or porous medium and fractures are proposed. These methods are expected to contribute to the development of numerical program for the simulation of radioactive nuclide transport through fractured-porous medium from radioactive waste disposal site.
Structural Safety Analysis of Openable Working Table in ACP Hot Cell for Spent Fuel Treatment
Kwon, Kie-Chan ; Ku, Jeong-Hoe ; Lee, Eun-Pyo ; Choung, Won-Myung ; You, Gil-Sung ; Lee, Won-Kyung ; Cho, Il-Je ; Kuk, Dong-Hak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 17~24
A demonstration facility for advanced spent fuel conditioning process (ACP) is under construction in KAERI. In this hot cell facility, all process equipments and materials are taken in and out only through the rear door. The working table in front of the process rear door is specially designed to be openable for the efficient use of the space. This paper presents the structural safety analysis of the openable working table, for the normal operational load condition and accidential drop condition of heavy object. Both cases are investigated through static and dynamic finite element analyses. The analysis results show that structural safety of the working table is sufficiently assured and the working table is not collapsed even when an object of 500 kg is dropped from the height of 50 cm.
Radio-sensitivity of Dark-striped Field Mice, Apodemus agrarius, as a Biological Dosimeter in Radio-ecological Monitoring System
Kim, Hee-Sun ; Nishimura, Y. ; Kim, Chong-Soon ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 25~32
This study examined the possibility of using dark-striped field mice as a biological indicator for the environmental radio-surveillance. For this study, dark-striped field mice were caught from five areas of Kyonggi, Kyongsang, Chungchong and Cholla provinces. The external morphological characteristics and isoenzymic types of dark-striped field mice were studied after they were captured. Among the external morphological characteristics, the dark-brown coat, dark back stripe, head-to-tail length, tail length, and ear length matched the taxonomical characteristics of dark-striped field mice. The analyses on L-lactate dehydrogenase, aspartate aminotransferase, and malate dehydrogenese revealed that one species of dark-striped field mice, called Apodemus agrarius, was inhabitated throughout a wide range of Korea. On the other hand, the frequency of micronuclei in peripheral polychromatic erythrocytes to survived mice after irradiation also analyzed. The LD50/30 of A. cgrarius and ICR mice were approximately 5 Gy and 7.9Gy, respectively. The results of the study reveal that wild A. asrarius have a high potential as a biological monitoring system to determine the impact of radiation in areas such as those within the vicinity of nuclear power plants.
Melting Characteristics for Radioactive Aluminum Wastes in Electric Arc Furnace
Min, Byung-Youn ; Song, Pyung-Seob ; Ahn, Jun-Hyung ; Choi, Wang-Kyu ; Jung, Chong-Hun ; Oh, Won-Zin ; Kang, Yong ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 33~40
The characteristics of the aluminum waste melting and the distribution of the radioactive nuclides have been investigated for the estimation on the volume reduction and the decontamination of the aluminum wastes from the decommissioning of the TRIGA MARK it and III research reactors at the Korea Atomic Energy Research Institute(KAERI). The aluminum wastes were melted with the use of the fluxes such as flux
, and flux
in the DC graphite arc furnace. For the assessment of the distribution of the radioactive nuclides during the melting of the aluminum, the aluminum materials were contaminated by the surrogate nuclides such as cobalt(Co), cesium(Cs) and strontium(Sr). The fluidity of aluminum melt was increased with the addition of the fluxes, which has slight difference according to the type of fluxes. The formation of the slag during the aluminum melting added the flux type C and D was larger than that with the flux A and B. The rate of the slag formation linearly increased with increasing the flux concentration. The results of the XRD analysis showed that the surrogate nuclide was transferred to the slag, which can be easily separated from the melt and then they combined with aluminum oxide to form a more stable compound. The distribution ratio of cobalt in ingot to that in slag was more than 40% at all types of fluxes. Since vapor pressures of cesium and strontium were higher than those that of the host metals at the melting temperature, their removal efficiency from the ingot phase to the slag and the dust phase was by up to 98%.
Assessment of a Pre-conceptual Design of a Spent PWR Fuel Disposal Container
Choi, Jong-Won ; Cho, Dong-Keun ; Lee, Yang ; Choi, Heui-Joo ; Lee, Jong-Youl ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 41~50
In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by
On-Site Transport and Storage of Spent Nuclear Fuel at Kori NPP by KN-12 Transport Cask
Chung, Sung-Hwan ; Baeg, Chang-Yeal ; Choi, Byung-Il ; Yang, Ke-Hyung ; Lee, Dae-Ki ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 51~58
Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.
The Characterization of Spherical Perticles in Steam Generator Sludge
Pyo, Hyung-Yeal ; Park, Yang-Soon ; Park, Sun-Dal ; Park, Kyoung-Kyun ; Song, Byung-Chul ; Park, Yong-Joon ; Jee, Kwang-Yong ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 59~64
Ion exchange resin particles should not be found in steam generator(S/G) sludge. The suspicious spherical resin particles observed in S/G sludge sample were characterized for particle size distribution under optical microscope using the micro-technique, for element analysis by the electron probe micro analysis (EPMA), and for molecular identification by the IR spectroscopy. The particle sizes are distributed from 1 to
for the sludge, while 40 to
for the spherical resin particles. The results of the elemental analysis showed different major impurities: Si, Al, Mn, Cr, Ni, Zn and Ti for the sludge particles, while Si, Cu, Zn for the spherical resin particles. However, both particles contain Fe as a matrix of magnetite
. IR spectrum of the spherical particles was not quite similar to the IR spectrum of ion exchange resins used in S/G system. These results indicate that the spherical particles are not related to ion exchange resin particles and may be formed by the process of the sludge formation.
Building Transparency on the Total System Performance Assessment of Radioactive Repository through the Development of the Cyber R&D Platform; Application for Development of Scenario and Input of TSPA Data through QA Procedures
Seo, Eun-Jin ; Hwang, Yong-Soo ; Kang, Chul-Hyung ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 65~75
Transparency on the Total System Performance Assessment (TSPA) is the key issue to enhance the public acceptance for a radioactive repository. To approve it, all performances on TSPA through Quality Assurance is necessary. The integrated Cyber R&D Platform is developed by KAERI using the T2R3 principles applicable for five major steps : planning, research work, documentation, and internal & external audits in R&D's. The proposed system is implemented in the web-based system so that all participants in TSPA are able to access the system. It is composed of three sub-systems; FEAS (FEp to Assessment through Scenario development) showing systematic approach from the FEPs to Assessment methods flow chart, PAID (Performance Assessment Input Databases) being designed to easily search and review field data for TSPA and QA system containing the administrative system for QA on five key steps in R&D's in addition to approval and disapproval processes, corrective actions, and permanent record keeping. All information being recorded in QA system through T2R3 principles is integrated into Cyber R&D Platform so that every data in the system can be checked whenever necessary. Throughout the next phase R&D, Cyber R&D Platform will be connected with the assessment tool for TSPA so that it will be expected to search the whole information in one unified system.
Structural Safety Analysis Of Rear Door in ACP Hotcell Facility for Spent Fuel Treatment
Kwon, Kie-Chan ; Ku, Jeong-Hoe ; Lee, Eun-Pyo ; Choung, Won-Myung ; You, Gil-Sung ; Lee, Won-Kyung ; Kuk, Dong-Hak ; Cho, Il-Je ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 77~85
A demonstration facility for an advanced spent fuel conditioning process (ACP) is under construction at KAERI. In this hotcell facility, the rear door is frequently used since all process equipment and materials are taken in and out only through the rear door. Therefore , both the structural safety and stability of the door are essentially required for the safety of ACP facility. In this paper, the finite element analysis has been performed to investigate the structural safety under the impact condition between the rear door and the door frame. Also the possibility of the rear door being tumbled over by the impact force or the inertia force under a sudden stop conditon has been evaluated. The analysis results demonstrate that the structural safety and stability of the rear door are sufficiently assured for both the impact and the accidential stop conditions.
Current Status and Projection of Spent Nuclear Fuel for Geological Disposal System Design
Cho, Dong-Keun ; Choi, Jong-Won ; Hahn, Pil-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 1, 2006, Pages 87~93
Inventories, and characteristics such as dimension, fuel rod array, weight,
enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial
enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed
GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be
GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with
Korean Standard Fuel Assembly, cross section of
, length of 453cm, mass of 672 kg, initial
enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.