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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 45, Issue 7 - Dec 2013
Volume 45, Issue 6 - Nov 2013
Volume 45, Issue 5 - Oct 2013
Volume 45, Issue 4 - Aug 2013
Volume 45, Issue 3 - Jun 2013
Volume 45, Issue 2 - Apr 2013
Volume 45, Issue 1 - Feb 2013
Selecting the target year
RADIATION DOSE TO HUMAN AND NON-HUMAN BIOTA IN THE REPUBLIC OF KOREA RESULTING FROM THE FUKUSHIMA NUCLEAR ACCIDENT
Keum, Dong-Kwon ; Jun, In ; Lim, Kwang-Muk ; Choi, Yong-Ho ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 1~12
DOI : 10.5516/NET.03.2011.063
This paper describes the radiation doses to human and non-human biota in the Republic of Korea, as a result of the Fukushima nuclear accident. By using the measured airborne activity and ground deposition, the effective and thyroid doses of five human age groups (infant, 5 years, 10 years, 15 years and adult) were estimated by the ECOSYS code, and the whole body absorbed dose rate of the eight Korean reference animals and plants (RAPs) was estimated by the K-BIOTA (the Korean computer code to assess the risk of radioactivity to wildlife). The first-year effective and thyroid human doses ranged from 5.7E-5 mSv in the infant group to 2.0E-4 mSv in the 5 years group, and from 5.0E-4 mSv in the infant group to 3.4E-3 mSv in the 5 years group, respectively. The life-time (70 years) effective and thyroid human doses ranged from 1.5E-4 mSv in the infant group to 3.0E-4 mSv in the 5 years group, and from 6.0E-4 mSv in the infant group to 3.5E-3 mSv in the 5 years group, respectively. The estimated maximum whole body absorbed dose rate to the Korean RAPs was 6.7E-7 mGy/d for a snake living in soil (terrestrial biota), and 2.0E-5 mGy/d for freshwater fish (aquatic biota), both of which were far less than the generic dose criteria to protect biota from ionizing radiation. Also, the screening level assessment for ERICA's (Environmental Risks from Ionizing Contaminants: Assessments and management) limiting organisms showed that the risk quotient (RQ) for the estimated maximum soil and water activity was significantly less than unity for both the terrestrial and freshwater organisms. Conclusively, the radiological risk of the radioactivity released into the environment by the Fukushima nuclear accident to the public and the non-human biota in the republic of Korea is considered negligible.
THE FUKUSHIMA DISASTER - SYSTEMIC FAILURES AS THE LACK OF RESILIENCE
Hollnagel, Erik ; Fujita, Yushi ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 13~20
DOI : 10.5516/NET.03.2011.078
This paper looks at the Fukushima disaster from the perspective of resilience engineering, which replaces a search for causes with an understanding of how the system failed in its performance. Referring to the four resilience abilities of responding, monitoring, learning, and anticipating, the paper focuses on how inadequate engineering anticipation or risk assessment during the design, in combination with inadequate response capabilities, precipitated the disaster. One lesson is that systems such as nuclear power plants are complicated, not only in how they function during everyday or exceptional conditions, but also during their whole life cycle. System functions are intrinsically coupled synchronically and diachronically in ways that may affect the ability to respond to extreme conditions.
NUMERICAL INVESTIGATION OF THE SPREADING AND HEAT TRANSFER CHARACTERISTICS OF EX-VESSEL CORE MELT
Ye, In-Soo ; Kim, Jeongeun Alice ; Ryu, Changkook ; Ha, Kwang Soon ; Kim, Hwan Yeol ; Song, Jinho ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 21~28
DOI : 10.5516/NET.03.2012.011
The flow and heat transfer characteristics of the ex-vessel core melt (corium) were investigated using a commercial CFD code along with the experimental data on the spreading of corium available in the literature (VULCANO VE-U7 test). In the numerical simulation of the unsteady two-phase flow, the volume-of-fluid model was applied for the spreading and interfacial surface formation of corium with the surrounding air. The effects of the key parameters were evaluated for the corium spreading, including the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The results showed a reasonable trend of corium progression influenced by the changes in the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The modeling of the viscosity appropriate for corium and the radiative heat transfer was critical, since the front progression and temperature profiles were strongly dependent on the models. Further development is required for the code to consider the formation of crust on the surfaces of corium and the interaction with the substrate.
DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA
Choi, Heui-Joo ; Lee, Jong Youl ; Choi, Jongwon ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 29~40
DOI : 10.5516/NET.06.2012.006
Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.
AN ANALYSIS OF THE THERMAL AND MECHANICAL BEHAVIOR OF ENGINEERED BARRIERS IN A HIGH-LEVEL RADIOACTIVE WASTE REPOSITORY
Kwon, S. ; Cho, W.J. ; Lee, J.O. ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 41~52
DOI : 10.5516/NET.06.2012.015
Adequate design of engineered barriers, including canister, buffer and backfill, is important for the safe disposal of high-level radioactive waste. Three-dimensional computer simulations were carried out under different condition to examine the thermal and mechanical behavior of engineered barriers and rock mass. The research looked at five areas of importance, the effect of the swelling pressure, water content of buffer, density of compacted bentonite, emplacement type and the selection of failure criteria. The results highlighted the need to consider tensile stress in the outer shell of a canister due to thermal expansion of the canister and the swelling pressure from the buffer for a more reliable design of an underground repository system. In addition, an adequate failure criterion should be used for the buffer and backfill.
A STUDY OF THE PRESSURE SOLUTION AND DEFORMATION OF QUARTZ CRYSTALS AT HIGH pH AND UNDER HIGH STRESS
Choi, Jung-Hae ; Seo, Yong-Seok ; Chae, Byung-Gon ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 53~60
DOI : 10.5516/NET.06.2012.024
Bentonite is generally used as a buffer material in high-level radioactive waste disposal facilities and consists of 50% quartz by weight. Quartz strongly affects the behavior of bentonite over very long periods. For this reason, quartz dissolution experiment was performed under high-pressure and high-alkalinity conditions based on the conditions found in a high-level radioactive waste disposal facility located deep underground. In this study, two quartz dissolution experiments were conducted on 1) quartz beads under low-pressure and high-alkalinity conditions and 2) a single quartz crystal under high-pressure and high-alkalinity conditions. Following the experiments, a confocal laser scanning microscope (CLSM) was used to observe the surfaces of experimental samples. Numerical analyses using the finite element method (FEM) were also performed to quantify the deformation of contact area. Quartz dissolution was observed in both experiments. This deformation was due to a concentrated compressive stress field, as indicated by the quartz deformation of the contact area through the FEM analysis. According to the numerical results, a high compressive stress field acted upon the neighboring contact area, which showed a rapid dissolution rate compared to other areas of the sample.
DETERMINATION OF THE
I IN PRIMARY COOLANT OF PWR
Choi, Ke Chon ; Park, Yong Joon ; Song, Kyuseok ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 61~66
DOI : 10.5516/NET.06.2012.040
Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as
is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for
as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding
as a radio-isotopic tracer (
= 60.14 d) to the simulation sample, in order to measure the activity concentration of
in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh,
form) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of
was examined, as was the effect of
on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous
presence was found with activity concentrations of
lower than 50 Bq/mL, and with a boron concentration of less than 2,000
SUSCEPTIBILITY OF ALLOY 690 TO STRESS CORROSION CRACKING IN CAUSTIC AQUEOUS SOLUTIONS
Kim, Dong-Jin ; Kim, Hong Pyo ; Hwang, Seong Sik ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 67~72
DOI : 10.5516/NET.07.2012.021
Stress corrosion cracking (SCC) behaviors of Alloy 690 were studied in lead-containing aqueous alkaline solutions using the slow strain rate tension (SSRT) tests in 0.1M and 2.5M NaOH with and without PbO at
. The side and fracture surfaces of the alloy were then examined using scanning electron microscopy after the SSRT test. Microstructure and composition of the surface oxide layer were analyzed by using a field emission transmission electron microscopy, equipped with an energy dispersive X-ray spectroscopy. Even though Alloy 690 was almost immune to SCC in 0.1M NaOH solution, irrespective of PbO addition, the SCC resistance of Alloy 690 decreased in a 2.5M NaOH solution and further decreased by the addition of PbO. Based on thermodynamic stability and solubility of oxide, high Cr of 30wt% in the Alloy 690 is favorable to SCC in mild alkaline and acidic solutions whereas the SCC resistance of high Cr Alloy 690 is weakened drastically in the strong alkaline solution where the oxide is not stable any longer and solubility is too high to form a passive oxide locally.
ROLE OF GRAIN BOUNDARY CARBIDES IN CRACKING BEHAVIOR OF Ni BASE ALLOYS
Hwang, Seong Sik ; Lim, Yun Soo ; Kim, Sung Woo ; Kim, Dong Jin ; Kim, Hong Pyo ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 73~80
DOI : 10.5516/NET.07.2012.013
The primary water stress corrosion cracking (PWSCC) of Alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and the pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of steam generators (SG) in Korea. The PWSCC resistance of Ni base alloys which have intergranular carbides is higher than those which have intragranular carbides. Conversely, in oxidized acidic solutions like sodium sulfate or sodium tetrathionate solutions, the Ni base alloys with a lot of carbides at the grain boundaries and shows less stress corrosion cracking (SCC) resistance. The role of grain boundary carbides in SCC behavior of Ni base alloys was evaluated and effect of intergranular carbides on the SCC susceptibility were reviewed from the literature.
EXPERIMENTAL ANALYSES OF SPALLATION NEUTRONS GENERATED BY 100 MEV PROTONS AT THE KYOTO UNIVERSITY CRITICAL ASSEMBLY
Pyeon, Cheol Ho ; Azuma, Tetsushi ; Takemoto, Yuki ; Yagi, Takahiro ; Misawa, Tsuyoshi ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 81~88
DOI : 10.5516/NET.08.2012.003
Neutron spectrum analyses of spallation neutrons are conducted in the accelerator-driven system (ADS) facility at the Kyoto University Critical Assembly (KUCA). High-energy protons (100 MeV) obtained from the fixed field alternating gradient accelerator are injected onto a tungsten target, whereby the spallation neutrons are generated. For neutronic characteristics of spallation neutrons, the reaction rates and the continuous energy distribution of spallation neutrons are measured by the foil activation method and by an organic liquid scintillator, respectively. Numerical calculations are executed by MCNPX with JENDL/HE-2007 and ENDF/B-VI libraries to evaluate the reaction rates of activation foils (bismuth and indium) set at the target and the continuous energy distribution of spallation neutrons set in front of the target. For the reaction rates by the foil activation method, the C/E values between the experiments and the calculations are found around a relative difference of 10%, except for some reactions. For continuous energy distribution by the organic liquid scintillator, the spallation neutrons are observed up to 45 MeV. From these results, the neutron spectrum information on the spallation neutrons generated at the target are attained successfully in injecting 100 MeV protons onto the tungsten target.
MODELING OF NONLINEAR CYCLIC LOAD BEHAVIOR OF I-SHAPED COMPOSITE STEEL-CONCRETE SHEAR WALLS OF NUCLEAR POWER PLANTS
Ali, Ahmer ; Kim, Dookie ; Cho, Sung Gook ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 89~98
DOI : 10.5516/NET.09.2011.055
In recent years steel-concrete composite shear walls have been widely used in enormous high-rise buildings. Due to high strength and ductility, enhanced stiffness, stable cycle characteristics and large energy absorption, such walls can be adopted in the auxiliary building; surrounding the reactor containment structure of nuclear power plants to resist lateral forces induced by heavy winds and severe earthquakes. This paper demonstrates a set of nonlinear numerical studies on I-shaped composite steel-concrete shear walls of the nuclear power plants subjected to reverse cyclic loading. A three-dimensional finite element model is developed using ABAQUS by emphasizing on constitutive material modeling and element type to represent the real physical behavior of complex shear wall structures. The analysis escalates with parametric variation in steel thickness sandwiching the stipulated amount of concrete panels. Modeling details of structural components, contact conditions between steel and concrete, associated boundary conditions and constitutive relationships for the cyclic loading are explained. Later, the load versus displacement curves, peak load and ultimate strength values, hysteretic characteristics and deflection profiles are verified with experimental data. The convergence of the numerical outcomes has been discussed to conclude the remarks.
ASSESSMENT OF THERMAL FATIGUE IN MIXING TEE BY FSI ANALYSIS
Jhung, Myung Jo ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 99~106
DOI : 10.5516/NET.09.2012.026
Thermal fatigue is a significant long-term degradation mechanism in nuclear power plants. In particular, as operating plants become older and life time extension activities are initiated, operators and regulators need screening criteria to exclude risks of thermal fatigue and methods to determine significant fatigue relevance. In general, the common thermal fatigue issues are well understood and controlled by plant instrumentation at fatigue susceptible locations. However, incidents indicate that certain piping system Tee connections are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentations. Therefore, in this study thermal fatigue evaluation of piping system Tee-connections is performed using the fluid-structure interaction (FSI) analysis. From the thermal hydraulic analysis, the temperature distributions are determined and their results are applied to the structural model of the piping system to determine the thermal stress. Using the rain-flow method the fatigue analysis is performed to generate fatigue usage factors. The procedure for improved load thermal fatigue assessment using FSI analysis shown in this study will supply valuable information for establishing a methodology on thermal fatigue.
THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR
Yoon, Hyoungju ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 107~114
DOI : 10.5516/NET.03.2011.051
It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl,
, and Cs are very low.
REVIEW OF SPENT FUEL INTEGRITY EVALUATION FOR DRY STORAGE
Kook, Donghak ; Choi, Jongwon ; Kim, Juseong ; Kim, Yongsoo ;
Nuclear Engineering and Technology, volume 45, issue 1, 2013, Pages 115~124
DOI : 10.5516/NET.06.2012.016
Among the several options to solve PWR spent fuel accumulation problem in Korea, the dry storage method could be the most realistic and applicable solution in the near future. As the basic objectives of dry storage are to prevent a gross rupture of spent fuel during operation and to keep its retrievability until transportation, at the same time the importance of a spent fuel integrity evaluation that can estimate its condition at the final stage of dry storage is very high. According to the national need and technology progress, two representative nations of spent fuel dry storage, the USA and Japan, have established different system temperature criteria, which is the only controllable factor in a dry storage system. However, there are no technical criteria for this evaluation in Korea yet, it is necessary to review the previously well-organized methodologies of advanced countries and to set up our own domestic evaluation direction due to the nation's need for dry storage. To satisfy this necessity, building a domestic spent fuel test database should be the first step. Based on those data, it is highly recommended to compare domestic data range with foreign results, to build our own criteria, and to expand on evaluation work into recently issued integrity problems by using a comprehensive integrity evaluation code.