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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 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
Pressure Wave Propagation in the Discharge Piping with Water Pool
Bang Young S. ; Seul Kwang W. ; Kim In-Goo ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 285~294
Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined.
Development of a Traceability Analysis Method Based on Case Grammar for NPP Requirement Documents Written in Korean Language
Yoo Yeong Jae ; Seong Poong Hyun ; Kim Man Cheol ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 295~303
Software inspection is widely believed to be an effective method for software verification and validation (V&V). However, software inspection is labor-intensive and, since it uses little technology, software inspection is viewed upon as unsuitable for a more technology-oriented development environment. Nevertheless, software inspection is gaining in popularity. KAIST Nuclear I&C and Information Engineering Laboratory (NICIEL) has developed software management and inspection support tools, collectively named "SIS-RT. "SIS-RT is designed to partially automate the software inspection processes. SIS-RT supports the analyses of traceability between a given set of specification documents. To make SIS-RT compatible for documents written in Korean, certain techniques in natural language processing have been studied . Among the techniques considered, case grammar is most suitable for analyses of the Korean language . In this paper, we propose a methodology that uses a case grammar approach to analyze the traceability between documents written in Korean. A discussion regarding some examples of such an analysis will follow.
An Analysis of the Ageing Effect on the Removal of Cesium and Cobalt from Radioactive Soil by the Electrokinetic Method
Kim Gye-Nam ; Oh Won-Zin ; Won Hui-Zun ; Jung Chong-Hun ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 304~315
The ageing effects of radionuclides in radioactive soil on remediation using the electrokinetic method were analyzed. Comparative experiments were conducted for the reactive soil around a TRIGA research? reactor contaminated with
for 15 years and the non-reactive soil that was intentionally contaminated with
for 3 days. It was observed that because of an aging effect on
, the efficiency of removing it decreased.
used as an additive to increase the removal efficiency showed a higher removal capability than other chemicals for both
. The efficiency of removing radionuclides from the radioactive soil in the column was proportional to the capability of the added chemical to extract radionuclides. It took 10 days to achieve a
from the soil. The volume of the soil wastewater discharged from the soil column by the electrokinetic method was
below that for soil washing.
Optimum Global Failure Prediction Model of Inconel 600 Thin Plate with Two Parallel Through-Wall Cracks
Moon Seong In ; Kim Young Jin ; Lee Jin Ho ; Song Myung Ho ; Choi Young Hwan ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 316~326
$ of wall criterion, which is generally used for the plugging of steam generator tubes, is applied only to a single crack. In a previous study, a total number of 9 failure models were proposed to estimate the local failure of the ligament between cracks, and the optimum coalescence model of multiple collinear cracks was determined among these models. It is, however known that parallel axial cracks are more frequently detected than collinear axial cracks during an in-service inspection. The objective of this study is to determine the plastic collapse model that can be applied to steam generator tubes containing two parallel axial through-wall cracks. Three previously proposed local failure models were selected as the candidates. Subsequently, the interaction effects between two adjacent cracks were evaluated to screen them. Plastic collapse tests for the plate with two parallel through-wall cracks and finite element analyses were performed to determine the optimum plastic collapse model. By comparing the test results with the prediction results obtained from the candidate models, a COD base model was selected as an optimum model.
Structural Integrity Evaluation of Steam Generator Tube with Two Parallel Axial Through-Wall Cracks
Moon Seong In ; Kim Young Jin ; Lee Jin Ho ; Song Myung Ho ; Park Youn Won ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 327~337
It is commonly required that tubes with defects exceeding
of wall thickness in depth should be plugged; however, this criterion is too conservative for some locations and for some types of defects. Many studies have been done with the aim of developing an alternative plugging criteria, and these studies have shown that steam generator tubes with a certain range of axial through-wall cracks could remain in service without any safety or reliability problems. However, these studies have been limited, thus far, to consideration of single cracked tubes, necessitating a study on multiple cracks, which are commonly found. A crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed in the previous study. In this paper, the investigation is extended to the parallel axial cracks spaced in a circumferential direction, because parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks. Interaction effects between two parallel cracks are evaluated by performing elastic and elastic-plastic finite element analyses.
Sensitivity Analysis of Fabrication Parameters for Dry Process Fuel Performance Using Monte Carlo Simulations
Park Chang Je ; Song Kee Chan ; Yang Myung Seung ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 338~345
This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the
fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the
fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the
fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the
Vulnerability Analysis on a VPN for a Remote Monitoring System
Kim Jung Soo ; Kim Jong Soo ; Park Il Jin ; Min Kyung Sik ; Choi Young Myung ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 346~356
14 Pressurized Water Reactors (PWR) in Korea use a remote monitoring system (RMS), which have been used in Korea since 1998. A Memorandum of Understanding on Remote Monitoring, based on Enhanced Cooperation on PWRs, was signed at the 10th Safeguards Review Meeting in October 2001 between the International Atomic Energy Agency (IAEA) and Ministry Of Science and Technology (MOST). Thereafter, all PWR power plants applied for remote monitoring systems. However, the existing method is high cost (involving expensive telephone costs). So, it was eventually applied to an Internet system for Remote Monitoring. According to the Internet-based Virtual Private Network (VPN) applied to Remote Monitoring, the Korea Atomic Energy Research Institute (KAERI) came to an agreement with the IAEA, using a Member State Support Program (MSSP). Phase I is a Lab test. Phase II is to apply it to a target power plant. Phase III is to apply it to all the power plants. This paper reports on the penetration testing of Phase I. Phase I involved both domestic testing and international testing. The target of the testing consisted of a Surveillance Digital Integrated System (SDIS) Server, IAEA Server and TCNC (Technology Center for Nuclear Control) Server. In each system, Virtual Private Network (VPN) system hardware was installed. The penetration of the three systems and the three VPNs was tested. The domestic test involved two hacking scenarios: hacking from the outside and hacking from the inside. The international test involved one scenario from the outside. The results of tests demonstrated that the VPN hardware provided a good defense against hacking. We verified that there was no invasion of the system (SDIS Server and VPN; TCNC Server and VPN; and IAEA Server and VPN) via penetration testing.
Development and Testing of a Prototype Long Pulse Ion Source for the KSTAR Neutral Beam System
Chang Doo-Hee ; Oh Byung-Hoon ; Seo Chang-Seog ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 357~363
A prototype long pulse ion source was developed, and the beam extraction experiments of the ion source were carried out at the Neutral Beam Test Stand (NBTS) of the Korea Superconducting Tokamak Advanced Research (KSTAR). The ion source consists of a magnetic bucket plasma generator, with multi-pole cusp fields, and a set of tetrode accelerators with circular apertures. Design requirements for the ion source were a 120kV/65A deuterium beam and a 300 s pulse length. Arc discharges of the plasma generator were controlled by using the emission-limited mode, in turn controlled by the applied heating voltage of the cathode filaments. Stable and efficient arc plasmas with a maximum arc power of 100 kW were produced using the constant power mode operation of an arc power supply. A maximum ion density of
was obtained by using electrostatic probes, and an optimum arc efficiency of 0.46 A/kW was estimated. The accelerating and decelerating voltages were applied repeatedly, using the re-triggering mode operation of the high voltage switches during a beam pulse, when beam disruptions occurred. The decelerating voltage was always applied prior to the accelerating voltage, to suppress effectively the back-streaming electrons produced at the time of an initial beam formation, by the pre-programmed fast-switch control system. A maximum beam power of 0.9 MW (i.e.
) with hydrogen was measured for a pulse duration of 0.8 s. Optimum beam perveance, deduced from the ratio of the gradient grid current to the total beam current, was
. Stable beams for a long pulse duration of
were tested at low accelerating voltages.
Fluidelastic Instability Characteristics of Helical Steam Generator Tubes
Jo Jong Chull ; Jhung Myung Jo ; Kim Woong Sik ; Choi Young Hwan ; Kim Hho Jung ;
Nuclear Engineering and Technology, volume 36, issue 4, 2004, Pages 364~373
This study investigates the fluidelastic instability characteristics of helical steam generator type tubes used in operating nuclear power plants. To obtain a natural frequency, corresponding mode shape, and participation factor, modal analyses using various conditions are performed for helical type tubes. Investigated are the effects of the number of turns, the number of supports, and the status of the inner fluid on the modal and fluidelastic instability characteristics of the tubes, which are expressed in terms of the natural frequency, the corresponding mode shape, and the stability ratio.