• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2001.05c
• /
• pp.101-105
• /
• 2001

For high voltage XLPE power cables, semiconducting layers have been applied to prevent discharge at the interface between conductor and insulation, and/or insulation and external shielding layer. The semiconducting layers may be also effective to release electrical stress in the interface. The property of semiconducting layers are significantly related to the quality and reliability of power cables. Generally, these semiconducting layers are formed by extruding, the conductibility of the material is given by the carbon black mixed with base polymer. The life of power cables is depended on the smoothness of the interface between insulation and semiconducting layer. If the smoothness is no good, the life of power cables is shorter because the electrical stress and water tree is increased. The causes of no good smoothness are the void of the interface, the protrusions, the contaminants and impurities of the semiconducting layer. The selection and dispersion of the Carbon Black is the significant factor to determine the life of power cable in the manufacturing of semiconducting compound.

• Nuclear Engineering and Technology
• /
• v.14 no.4
• /
• pp.196-203
• /
• 1982

To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k$_{eff}$ was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.

• Journal of Radiation Protection and Research
• /
• v.28 no.2
• /
• pp.127-135
• /
• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Journal of the Korean Solar Energy Society
• /
• v.39 no.6
• /
• pp.127-135
• /
• 2019

In this study, we analyzed the relationship between five factors: long-wave radiation, short-wave radiation, cloudiness, SVF and summer shelters. In the previous study, we recognized the correlation between single building SVF and long-wave radiation. Furthermore, this study attempted to confirm the relationship at the summer shelter with high solar radiation blocking rate. The observations are as follows. ① Cooling in summer shelters was not the effect of temperature but the effect of radiation reduction due to short-wave radiation shielding. ② In the case of the canopy tent with low heat capacity, the long-wave radiation was observed to be 16.7% higher per hour than the comparison control point due to the increase in surface temperature. ③ The long-wave radiation increase rate was different according to SVF, but showed very similar pattern according to the material characteristics of the summer shelters. ④ Passive Cooling effect on the type of summer shelters are determined by the size of the total long and short-wave radiation at that point.

• Journal of radiological science and technology
• /
• v.11 no.2
• /
• pp.65-71
• /
• 1988

Radiation dose outside the radiotherapy treatment field can be significant and therefore is of clinical interest estimating organ dose. We have made measurements of dose at distances up to 70 cm from the central axis of $5{\times}5$, $10{\times}10$, $15{\times}15$, and $25{\times}25$ cm radiation fields of Co-60 ${\gamma}-ray$, at 5 cm depth in water. Contributions to the total secondary radiation dose from water scatter, machine (collimator) scatter and leakage radiation have been seperated. We have found that the component of dose from water scatter can be described by simple exponential function of distance from the central axis of the radiation field for all field sizes. Machine scatter contributes 20 to 60% of the total secondary dose depending on field size and distance from the field. Leakage radiation contributes very little dose, but becomes the dominant componant at distance beyond 40 cm from the central axis. Then, wedges can cause a factor 2 to 3 increase in dose at any point outside the field compared with the dose when no wedge is used. Adding blocks to a treatment field can cause an increase in dose at points outside the field, but the effect is much smaller than the effect of a wedge. From the results of these measurements, doses to selected organs outside the field for specified treatment geometries were estimated, and the potential for reducing these organ doses by additional shielding was assessed.

• Nuclear Engineering and Technology
• /
• v.49 no.3
• /
• pp.549-555
• /
• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• Wind and Structures
• /
• v.28 no.4
• /
• pp.215-224
• /
• 2019

Because of the particularity and complexity of direct air-cooling structures (ACS), wind parameters given in the general load codes are not suitable for the wind-resistant design. In order to investigate the wind loads of ACS, two 1/150 scaled three-span models were designed and fabricated, corresponding to a rigid model and an aero-elastic model, and wind tunnel tests were then carried out. The model used for testing the wind pressure distribution of the ACS was defined as the rigid model in this paper, and the stiffness of which was higher than that of the aero-elastic model. By testing the rigid model, the wind pressure distribution of the ACS model was studied, the shape coefficients of "A" shaped frame and windbreak walls, and the gust factor of the windbreak walls were determined. Through testing the aero-elastic model, the wind-induced dynamic responses of the ACS model was studied, and the wind vibration coefficients of ACS were determined based on the experimental displacement responses. The factors including wind direction angle and rotation of fan were taken into account in this test. The results indicated that the influence of running fans could be ignored in the structural design of ACS, and the wind direction angle had a certain effect on the parameters. Moreover, the shielding effect of windbreak walls induced that wind loads of the "A" shaped frame were all suction. Subsequently, based on the design formula of wind loads in accordance with the Chinese load code, the corresponding parameters were presented as a reference for wind-resistant design and wind load calculation of air-cooling structures.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.2000-2010
• /
• 2021

Mechanical moduli, such as Young's modulus (E), Bulks modulus (B), Shear modulus (S), longitudinal modulus (L), Poisson's ratio (σ) and micro Hardness (H) were theoretically calculated for (100-x)TeO₂+x MgO glasses, where x = 10, 20, 30, 40 and 45 mol%, based on the Makishima-Mackenzie model. The estimated results showed that the mechanical moduli and the microhardness of the glasses were improved with the increase of the MgO contents in the TM glasses, while Poisson's ratio decreased with the increase in MgO content. Moreover, the radiation shielding capacity was evaluated for the studied TM glasses. Thus, the linear attenuation coefficient (LAC), mass attenuation coefficient (MAC), transmission factor (TF) and half-value thickness (𝚫_0.5) were simulated for gamma photon energies between 0.344 and 1.406 MeV. The simulated results showed that glass TM10 with 10 mol % MgO possess the highest LAC and varied in the range between 0.259 and 0.711 cm^-1, while TM45 glass with 45 mol % MgO possess the lowest LAC and vary in the range between 0.223 and 0.587 cm^-1 at gamma photon energies between 0.344 and 1.406 MeV. Furthermore, the BXCOM program was applied to calculate the effective atomic number (Z_eff), equivalent atomic number (Z_eq) and buildup factors (EBF and EABF) of the glasses. The effective removal cross-section for the fast neutrons (ERCSFN, ∑_R) was also calculated theoretically. The received data depicts that the lowest ∑_R was achieved for TM10 glasses, where ∑_R = 0.0193 cm² g^-1, while TM45 possesses the highest ERCSFN where ∑_R = 0.0215 cm² g^-1.

• Journal of the Society of Disaster Information
• /
• v.19 no.3
• /
• pp.582-596
• /
• 2023

Purpose: This study aims to explore the ways to improve the security of temporary shelters in case of nuclear power plant accidents. Method: In this study, we mainly rely on the case studies on previous nuclear power plant accidents-Chernobyl, Fukushima, and Three Mile Island (TMI) cases. Result: The current radiation emergency response plans for nuclear power plant accidents center around the evacuation procedure. As a result, the concept of "shelter in place" has been understood as a means of assisting resident evacuation. However, based on the case studies, we find that encouraging shelter in place, rather than simply emphasizing evacuation, would help minimize unnecessary casualties, especially in case of the accidents rated greater than or equal to INES 5. To facilitate better shelter in place, we recommend utilize apartments as temporary shelters and suggest some possible improvements to ensure those apartments could be equipped with technologies for high radiation protection. Conclusion: To ensure better shelter in place, we recommend using apartments as temporary shelters, and we seek to supplement the function of apartments by using shielding, positive pressure, and sealing technologies.

• Wind and Structures
• /
• v.37 no.6
• /
• pp.425-444
• /
• 2023

This study is to investigate the effect of interference between two C-shaped high-rise buildings by computational fluid dynamics (CFD), focusing on the variation of the local pressure coefficient (C_P) and the mean pressure coefficient (C_PMEAN). Sixteen building position cases are considered for the present study. These cases were based on the position and height of the interference building (IB). The pressure coefficient (C_P) is calculated on the principal building (PB) and is compared with an isolated building identical in shape and size. The interference effect on PB has also been presented in reference for the interference factor (IF). According to the findings, the maximum force coefficient on the PB is 0.971 and it is 10.97% more than the isolated PB when IB is located at position 2b (two times the width of the building), and the interfering height of 13H/15 mm. The moment coefficient on PB is 1.27, which is 27.36% less than the isolated case in which IB pushed 2b to 3b in the y direction with 750 mm height. In most of the cases, because of the shielding effect of the IB, the value of force coefficient (C_F) on PB has been reduced. On the face of the PB, there are also considerable differences in the mean pressure coefficient C_PMEAN. When IB was positioned at a location of 2b in Y direction and an interfering height of 13H/15 mm, the maximum C_PMEAN (1.58) was observed on the leeward face of PB.