• Nuclear Engineering and Technology
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• 제14권4호
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• pp.196-203
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• 1982

본 논문에서는 고리 1호기의 기사용 핵연료 집합체를 수송하기 위한 Cask를 설계하였다. 이를 위하여 고리 1호기의 기사용 핵연료 집합체로부터 방출되는 감마선과 중성자를 계산하여 MORSE 및 ANISN전산 코드로써 차폐 계산을 수행하였다. 그 결과, 9개의 집합체를 동시에 수송할 수 있는 Steel Cask가 가장 적합하다는 것을 밝혔다. 이 Steel Cask에 대한 안전성을 평가하기 위하여 연료봉의 중심 온도와 복재온도를 계산하여 핵연료의 용융점보다 훨씬 낮음을 증명하였다. 또한 KENO와 MORSE전산 코드를 사용하여 임계도 계산을 수행하여 미임계 상태임을 증명하였다. 이로써 9개의 기사용 핵연료 집합체를 동시에 수송할 수 있는 Steel Cask를 간단히 설계하였다.

• Journal of Radiation Protection and Research
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• 제28권2호
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• pp.127-135
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• 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.

• 한국태양에너지학회 논문집
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• 제39권6호
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• pp.127-135
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• 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.

• 대한방사선기술학회지:방사선기술과학
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• 제11권2호
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• pp.65-71
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• 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
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• 제49권3호
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• pp.549-555
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• 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
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• 제28권4호
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• pp.215-224
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• 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
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• 제53권6호
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• pp.2000-2010
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• 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.

• 한국재난정보학회 논문집
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• 제19권3호
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• pp.582-596
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• 2023

연구목적: 본 연구는 원전사고 발생 시 임시 대피시설의 안전성을 강화할 수 있는 방안을 연구하였다. 연구방법: 본 연구에서는 과거의 체르노빌, 후쿠시마, 쓰리마일섬(TMI) 원전사고에 대한 사례연구를 활용하였다. 연구결과: 현재의 방사선 비상 대응 계획은 소개에 중점을 두고 있으며, 실내대피 개념은 단순히 주민 대피를 지원하는 수단으로 활용되고 있다. 그러나 사례연구를 통해 INES-5 이상의 사고의 경우, 소개보다 실내대피를 장려하는 것이 사상자 최소화에 효과적임을 확인하였다. 더 나은 실내대피를 보장하기 위해 아파트를 임시 대피소로 활용할 것을 권장하며, 아파트가 방사선 방호 기술을 갖추도록 개선사항을 제안한다. 결론: 더 나은 실내대피를 보장하기 위해 아파트를 임시 대피소로 활용할 것을 권장하며, 차폐, 양압, 밀폐기술을 이용해서 아파트의 성능을 보완하고자 한다.

• Wind and Structures
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• 제37권6호
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• pp.425-444
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• 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.

• 방사성폐기물학회지
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• 제4권1호
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• pp.41-50
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• 2006

본 연구에서는 사전연구로부터 사용후핵연료의 처분용기 원형모델로 제안된 처분용기의 전체 크기와 배열을 평가하기 위하여 일련의 공학적 분석을 수행하였다. 그러한 노력의 결과 용기 내부 저장통의 배열형태와 외곽쉘과 상하부 뚜껑의 두께와 같은 새로운 설계변수를 도출하였다. 공학적 분석 작업에는 처분용기의 기계구조 해석 결과를 근거로 도출된 용기의 규격자료에 대한 방사선 안전성 측면에서의 타당성을 검토하기 위하여 방사선차폐 해석과 핵 임계 해석 등이 수행되었다. 처분용기 내부 삽입체의 직경 변화에 따른 구조안정성 해석 결과에 따르면, 직경 102cm 일 때 극한 외압조건은 물론 정상적인 외압조건 하에서도 최대 Von Mises 응력이 안전계수 2.0을 만족하는 것으로 나타났다. 이 경우에서도 핵 임계 및 방사선차폐 해석 결과 안전기준치를 만족시키며, 무게는 20톤 가량 줄어드는 효과가 있는 것으로 나타났다.