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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 42, Issue 6 - Dec 2010
Volume 42, Issue 5 - Oct 2010
Volume 42, Issue 4 - Aug 2010
Volume 42, Issue 3 - Jun 2010
Volume 42, Issue 2 - Apr 2010
Volume 42, Issue 1 - Feb 2010
Selecting the target year
INTERNATIONAL COLLABORATION IN ASSESSMENT OF RADIOLOGICAL IMPACTS ARISING FROM RELEASES TO THE BIOSPHERE AFTER DISPOSAL OF RADIOACTIVE WASTE INTO GEOLOGICAL REPOSITORIES
Smith, Graham ; Kato, Tomoko ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 1~8
DOI : 10.5516/NET.2010.42.1.001
Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. These conceptual uncertainties are additional to those that more typically arise, for example, in the assessment of present day situations, but which also have to be addressed. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.
A STUDY OF HYDRAULIC PROPERTIES IN A SINGLE FRACTURE WITH IN-PLANE HETEROGENEITY: AN EVALUATION USING OPTICAL MEASUREMENTS OF A TRANSPARENT REPLICA
Sawada, Atsushi ; Sato, Hisashi ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 9~16
DOI : 10.5516/NET.2010.42.1.009
Experimental examinations for evaluating fracutres were conducted by using transparent replicas of a single fracture in order to obtain the fracture data to contribute to the methodlogy on how to improve the definitaion of representative parameter values used for a parallel plate fracture model. Quantitative aperture distribution and quantitative tracer concentration data at each point in time were obtained by measuring the attenuation of transmitted light through the fracture in high spatial resolution. the representative aperture values evaluated from the multiple different measurement methods, such as arithmetic mean of aperture distribution measured by the optical method, transport aperture evaluated from the tracer test, and average aperture evaluated from the fracture void volume measurement converged to a unique value that indicates the accuracy of this experimental study. The aperture data was employed for verifying the numerical simulation under the assuption of Local Cubic Law and showed that the calculated flow rate through the fracture is 10%-100% larger than hydraulic test results. The quantitative tracer concentration data is also very valuable for validating existing numerical code for advection dispersion transport in-plane heterogeneous fractures.
A SYSTEMS ASSESSMENT FOR THE KOREAN ADVANCED NUCLEAR FUEL CYCLE CONCEPT FROM THE PERSPECTIVE OF RADIOLOGICAL IMPACT
Yoon, Ji-Hae ; Ahn, Joon-Hong ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 17~36
DOI : 10.5516/NET.2010.42.1.017
In this study, we compare the mass release rates of radionuclides(1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. the total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fission-product radionuclides,
, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would by sited. the footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal.
THE DEVELOPMENT OF A SAFETY ASSESSMENT APPROACH AND ITS IMPLICATION ON THE ADVANCED NUCLEAR FUEL CYCLE
Hwang, Yong-Soo ; Kang, Chul-Hyung ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 37~46
DOI : 10.5516/NET.2010.42.1.037
The development of advanced nuclear fuel cycle(ANFC) technology is essential to meet the national mission for energy independence via a nuclear option in Korea. The action target is to develop environmentally friendly, cost-effective measures to reduce the burden of long term disposal. The proper scenarios regarding potential radionuclide release from a repository have been developed in this study based on the advanced korean Reference Disposal System(A-KRS). To predict safety for the various scenarios, a new assessment code based on the GoldSim software has also been developed. Deterministic analysis indicates an environmental benefit from the ANFC as long as the solid waster from the ANFC act as a proper barrier.
THREE-DIMENSIONAL CORE DESIGN OF A SUPER FAST REACTOR WITH A HIGH POWER DENSITY
Cao, Liangzhi ; Oka, Yoshiaki ; Ishiwatari, Yuki ; Ikejiri, Satoshi ; Ju, Haitao ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 47~54
DOI : 10.5516/NET.2010.42.1.047
The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/
. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.
A STUDY OF THE HISTORICAL EARTHQUAKE CATALOG AND GUTENBERG-RICHTER PARAMETER VALUES OF THE KOREAN PENINSULA
Seo, Jeong-Moon ; Choi, In-Kil ; Rhee, Hyun-Me ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 55~64
DOI : 10.5516/NET.2010.42.1.055
The KIER's Korean historical earthquake catalog was revised for MMI
VI events recorded from the years 27 A.D. to 1904. the magnitude of each event was directly determined from the criteria suggested by Seo. The criteria incorporated the damage phenomena of the Japanese historical earthquake catalog, recent seismological studies, and the results of tests performed on ancient structures in Korea. Thus, the uncertainty of the magnitudes of the Korean historical earthquakes can be reduced. Also, the Gutenberg-Richter parameter values were estimated based on the revised catalog of this study. It was determined that the magnitudes of a maximum inland and minimum offshore event were approximately 6.3 and 6.5, respectively. The Gutenberg-Richter parameter pairs of the historical earthquake catalog were estimated to be a=5.32
0.19, which were somewhat lower than those obtained from recent complete instrumental earthquakes. No apparent change in the Gutenberg-Richter parameter is observed for the
centuries of the seismically active period.
MONTE CARLO METHOD EXTENDED TO HEAT TRANSFER PROBLEMS WITH NON-CONSTANT TEMPERATURE AND CONVECTION BOUNDARY CONDITIONS
Cho, Bum-Hee ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 65~72
DOI : 10.5516/NET.2010.42.1.065
The Monte Carlo method for solving heat conduction problems [1-3] is extended for non-constant temperature boundary conditions in this study. The new method can treat problems with any given non-constant boundary temperatures, including heat convection problems with non-constant fluid bulk temperature. A set of problems, particularly the heat transfer problem in a pebble fuel, is analyzed by this new method. In addition, a new method to reduce the statistical errors in kernel fuel regions is introduced when the Monte Carlo method is applied to a pebble fuel.
MOLTEN SALT VAPORIZATION DURING ELECTROLYTIC REDUCTION
Hur, Jin-Mok ; Jeong, Sang-Moon ; Lee, Han-Soo ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 73~78
DOI : 10.5516/NET.2010.42.1.073
The suppression of molten salt vaporization is one of the key technical issues in the electrolytic reduction process developed for recycling spent nuclear fuel from light-water reactors Since the Hertz-Langmuir relation previously applied to molten salt vaporization is valid only for vaporization into a vacuum, a diffusion model was derived to quantitatively assess the vaporization of LiCl,
and Li from an electrolytic reducer operating under atmospheric pressure. Vaporization rates as a function of operation variables were calculated and shown to be in reasonable agreement with the experimental data obtained from thermogravimetry.
LOCAL BURNUP CHARACTERISTICS OF PWR SPENT NUCLEAR FUELS DISCHARGED FROM YEONGGWANG-2 NUCLEAR POWER PLANT
Ha, Yeong-Keong ; Kim, Jung-Suck ; Jeon, Young-Shin ; Han, Sun-Ho ; Seo, Hang-Seok ; Song, Kyu-Seok ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 79~88
DOI : 10.5516/NET.2010.42.1.079
nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by
in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500
in the center to 100
in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the
ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.
ABSORBED INTERNAL DOSE CONVERSION COEFFICIENTS FOR DOMESTIC REFERENCE ANIMALS AND PLANT
Keum, Dong-Kwon ; Jun, In ; Lim, Kwang-Muk ; Choi, Yong-Ho ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 89~96
DOI : 10.5516/NET.2010.42.1.089
This paper describes the methodology of calculating the internal dose conversion coefficient in order to assess the radiological impact on non-human species. This paper also presents the internal dose conversion coefficients of 25 radionuclides (
) for domestic seven reference animals (roe deer, rat, frog, snake, Chinese minnow, bee, and earthworm) and one reference plant (pine tree). The uniform isotropic model was applied in order to calculate the internal dose conversion coefficients. The calculated internal dose conversion coefficient (
) ranged from
according to the type of radionuclides and organisms studied. It turns out that the internal does conversion coefficient was higher for alpha radionuclides, such as
, and for large organisms, such as roe deer and pine tree. The internal dose conversion coefficients of
were independent of the organism.
EFFECTS OF SUPPORT STRUCTURE CHANGES ON FLOW-INDUCED VIBRATION CHARACTERISTICS OF STEAM GENERATOR TUBES
Ryu, Ki-Wahn ; Park, Chi-Yong ; Rhee, Hui-Nam ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 97~108
DOI : 10.5516/NET.2010.42.1.097
Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.
THE BIDAS-2007: BIOASSAY DATA ANALYSIS SOFTWARE FOR EVALUATING A RADIONUCLIDE INTAKE AND DOSE
Lee, Jong-Il ; Lee, Tae-Young ; Kim, Bong-Whan ; Kim, Jang-Lyul ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 109~114
DOI : 10.5516/NET.2010.42.1.109
Bioassay data analysis software (BiDAS-2007) has been developed by KAERI, which adds several new functions to its previous version. New functions of the BiDAS-2007 computer code enable the user not only to do a simultaneous analysis by using two or more types of bioassay for the best internal dose evaluation, but also to do a continual internal dose evaluation from a change of the internal exposure conditions such as an intake type (acute, chronic), an intake pathway (inhalation, ingestion), an absorption type (Type F, M, S), and a particle size (AMAD, activity median aerodynamic diameter), and also to estimate the intakes in various conditions of an internal exposure at a time. The values calculated by the BiDAS-2007 code are consistent and in good agreement with those values by IMIE-2004 code by Berkovski and IMBA code by Birchall. The BiDAS-2007 computer code is very useful and user-friendly to estimate the radionuclide intakes and committed effective doses of a radiation worker.
INVESTIGATING THE APPROPRIATENESS OF THE TACOM MEASURE - APPLICATION TO THE COMPLEXITY OF PROCEDURALIZED TASKS FOR HIGH SPEED TRAIN DRIVERS
Park, Jin-Kyun ; Jung, Won-Dea ; Ko, Jong-Hyun ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 115~124
DOI : 10.5516/NET.2010.42.1.115
According to wide-spread experience in many industries, a procedure is one of the most effective countermeasures to reduce the possibility of human related problems. Unfortunately, a systematic framework to evaluate the complexity of procedural tasks seems to be very scant. For this reason, the TACOM measure, which can quantify the complexity of procedural tasks, has been developed. In this study, the appropriateness of the TACOM measure is investigated by comparing TACOM scores regarding the procedural tasks of high speed train drivers with the associated workload scores measured by the NASA-TLX technique. As a result, it is observed that there is a meaningful correlation between the TACOM scores and the associated NASA-TLX scores. Therefore, it is expected that the TACOM measure can properly quantify the complexity of procedural tasks.
SIPPING TEST: CHECKING FOR FAILURE OF FUEL ELEMENTS AT THE OPAL REACTOR
Smith, Michael Leslie ; Bignell, Lindsey Jorden ; Alexiev, Dimitri ; Mo, Li ;
Nuclear Engineering and Technology, volume 42, issue 1, 2010, Pages 125~130
DOI : 10.5516/NET.2010.42.1.125
Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.