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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 46, Issue 6 - Dec 2014
Volume 46, Issue 5 - Oct 2014
Volume 46, Issue 4 - Aug 2014
Volume 46, Issue 3 - Jun 2014
Volume 46, Issue 2 - Apr 2014
Volume 46, Issue 1 - Feb 2014
Selecting the target year
RADIOACTIVE SOURCE SECURITY: WHY DO WE NOT YET HAVE A GLOBAL PROTECTION SYSTEM?
Englefield, C. ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 461~466
DOI : 10.5516/NET.08.2014.713
Security of radioactive sources has been an issue since the earliest days of safety regulation of such materials. Since the events of September 11 2001, some governments and regulatory bodies have been much more focussed on these issues and have introduced extensive and enhanced security arrangements. International organisations like the IAEA and WINS have worked hard to help States in this regard. However, only a minority of States have implemented statutory security systems for radioactive source security. Why have so many States still to take action? What can be done to encourage and support these changes? This paper will offer some possible explanations for the lack of action in so many States and some potential answers to these questions.
CLARIFYING THE PARADIGM ON RADIATION EFFECTS & SAFETY MANAGEMENT: UNSCEAR REPORT ON ATTRIBUTION OF EFFECTS AND INFERENCE OF RISKS
Gonzalez, Abel J. ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 467~474
DOI : 10.5516/NET.08.2014.714
The aim of this paper is to describe a relatively recent international agreement on the widely debated concepts of: (i) attributing effects to low dose radiation exposure situations that have occurred in the past and, (ii) inferring radiation risk to situations that are planned to occur in the future. An important global consensus has been recently achieved on these fundamental issues at the level of the highest international intergovernmental body: the General Assembly of the United Nations. The General Assembly has welcomed with appreciation a scientific report on attributing health effects to radiation exposure and inferring risks that had been prepared the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) following a formal request by the General Assembly.
A SIMPLE METHOD TO CALCULATE THE DISPLACEMENT DAMAGE CROSS SECTION OF SILICON CARBIDE
Chang, Jonghwa ; Cho, Jin-Young ; Gil, Choong-Sup ; Lee, Won-Jae ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 475~480
DOI : 10.5516/NET.01.2013.051
We developed a simple method to prepare the displacement damage cross section of SiC using NJOY and SRIM/TRIM. The number of displacements per atom (DPA) dependent on primary knock-on atom (PKA) energy was computed using SRIM/TRIM and it is directly used by NJOY/HEATR to compute the neutron energy dependent DPA cross sections which are required to estimate the accumulated DPA of nuclear material. SiC DPA cross section is published as a table in DeCART 47 energy group structure. Proposed methodology can be easily extended to other materials.
ANALYSIS OF UNCERTAINTY QUANTIFICATION METHOD BY COMPARING MONTE-CARLO METHOD AND WILKS' FORMULA
Lee, Seung Wook ; Chung, Bub Dong ; Bang, Young-Seok ; Bae, Sung Won ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 481~488
DOI : 10.5516/NET.02.2013.047
An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.
ERGONOMIC ANALYSIS OF A TELEMANIPULATION TECHNIQUE FOR A PYROPROCESS DEMONSTRATION FACILITY
Yu, Seungnam ; Lee, Jongkwang ; Park, Byungsuk ; Kim, Kiho ; Cho, Ilje ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 489~500
DOI : 10.5516/NET.06.2014.026
In this study, remote handling strategies for a large-scale argon cell facility were considered. The suggested strategies were evaluated by several types of field test. The teleoperation tasks were performed using a developed remote handling system, which enabled traveling over entire cell area using a bridge transport system. Each arm of the system had six DOFs (degrees of freedom), and the bridge transport system had four DOFs. However, despite the dexterous manipulators and redundant monitoring system, many operators, including professionals, experienced difficulties in operating the remote handling system. This was because of the lack of a strategy for handling the installed camera system, and the difficulty in recognizing the gripper pose, which might fall outside the FOV (field of vision) of the system during teleoperation. Hence, in this paper, several considerations for the remote handling tasks performed in the target facility were discussed, and the tasks were analyzed based on ergonomic factors such as the workload. Toward the development of a successful operation strategy, several ergonomic issues, such as active/passive view of the remote handling system, eye/hand alignment, and FOV were considered. Furthermore, using the method for classifying remote handling tasks, several unit tasks were defined and evaluated.
CONTRIBUTION OF HANARO IRRADIATION TECHNOLOGIES TO NATIONAL NUCLEAR R&D
Choo, Kee Nam ; Cho, Man Soon ; Yang, Sung Woo ; Park, Sang Jun ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 501~512
DOI : 10.5516/NET.07.2014.006
HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.
SENSITIVITY ANALYSES OF THE USE OF DIFFERENT NEUTRON ABSORBERS ON THE MAIN SAFETY CORE PARAMETERS IN MTR TYPE RESEARCH REACTOR
Kamyab, Raheleh ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 513~520
DOI : 10.5516/NET.07.2014.002
In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy,
, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the
; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using
and Ag-In-Cd alloy, respectively.
MICROSTRUCTURE AND MECHANICAL STRENGTH OF SURFACE ODS TREATED ZIRCALOY-4 SHEET USING LASER BEAM SCANNING
Kim, Hyun-Gil ; Kim, Il-Hyun ; Jung, Yang-Il ; Park, Dong-Jun ; Park, Jeong-Yong ; Koo, Yang-Hyun ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 521~528
DOI : 10.5516/NET.07.2014.027
The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and
were selected for ODS treatment using LBS. Through the LBS method, the
particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at
was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive
particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.
DROP IMPACT ANALYSIS OF PLATE-TYPE FUEL ASSEMBLY IN RESEARCH REACTOR
Kim, Hyun-Jung ; Yim, Jeong-Sik ; Lee, Byung-Ho ; Oh, Jae-Yong ; Tahk, Young-Wook ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 529~540
DOI : 10.5516/NET.09.2013.103
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.
IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS
Lee, Dong Hyun ; Lim, Ho-Gon ; Yoon, Han Young ; Jeong, Jae Jun ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 541~546
DOI : 10.5516/NET.02.2014.003
Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.
SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS
Ko, Jae-Hun ; Park, Jea-Ho ; Jung, In-Soo ; Lee, Gang-Uk ; Baeg, Chang-Yeal ; Kim, Tae-Man ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 547~556
DOI : 10.5516/NET.08.2013.039
Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding,
pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a
cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the
cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the
cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.
ASSESSMENT OF WIND CHARACTERISTICS AND ATMOSPHERIC DISPERSION MODELING OF
Cs ON THE BARAKAH NPP AREA IN THE UAE
Lee, Jong Kuk ; Kim, Jea Chul ; Lee, Kun Jai ; Belorid, Miloslav ; Beeley, Philip A. ; Yun, Jong-Il ;
Nuclear Engineering and Technology, volume 46, issue 4, 2014, Pages 557~568
DOI : 10.5516/NET.09.2014.029
This paper presents the results of an analysis of wind characteristics and atmosphere dispersion modeling that are based on computational simulation and part of a preliminary study evaluating environmental radiation monitoring system (ERMS) positions within the Barakah nuclear power plant (BNPP). The return period of extreme wind speed was estimated using the Weibull distribution over the life time of the BNPP. In the annual meteorological modeling, the winds from the north and west accounted for more than 90 % of the wind directions. Seasonal effects were not represented. However, a discrepancy in the tendency between daytime and nighttime was observed. Six variations of cesium-137 (
) dispersion test were simulated under severe accident condition. The
dispersion was strongly influenced by the direction and speed of the main wind. A virtual receptor was set and calculated for observation of the
movement and accumulation. The results of the surface roughness effect demonstrated that the deposition of
was affected by surface condition. The results of these studies offer useful information for developing environmental radiation monitoring systems (ERMSs) for the BNPP and can be used to assess the environmental effects of new nuclear power plant.