• 에너지공학
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• 제24권2호
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• pp.40-44
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• 2015

중수로 일차냉각재계통 액체방출밸브의 개방압력에 대한 안전여유 및 시간지연을 반영하여 열수력코드로 경년열화가 반영된 노심에 대해 민감도를 평가하였다. 과거에는 안전해석을 수행할 때 안전여유와 시간지연을 반영하여 평가하지 않았으나, 월성1호기 안전해석 인허가 심사과정중 반영 평가하였다. 중수로 안전해석에서 압력경계는 일차냉각재계통 액체방출밸브이다. 따라서 액체방출밸브 응동이 안전해석에 직접적인 영향을 주므로 안전여유와 시간지연 부가가 안전해석 결과에 미치는 영향을 파악하고 해석에 반영하기 위해 일차냉각재계통 과압이 걸리는 사고들에 대해 평가하였다.

• 대한안전경영과학회:학술대회논문집
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• 대한안전경영과학회 2008년도 춘계학술대회
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• pp.295-299
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• 2008

This paper discusses tests of factor effect or contrast by the use of saturated design $k^n$ factorial design. The nine nonparametric rank measures in normality test using normal probability pot are proposed. Length's PSE(Pseduo Standard Error) test [4] which relies on the concept of effect sparsity is also introduced and extended to the margin of error(ME) and Simultaneous margin of error(SME).

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1924-1934
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• 2023

In this study, the probabilistic fatigue life model for Ni-base alloys was developed based on the Weibull distribution using statistical analysis of fatigue data reported in NUREG/CR-6909 and the new fatigue data of Alloy 52M/152 and 82/182. The developed Weibull model can consider right-censored data (i.e., non-failed data) and quantify the improved safety (or reliability) based on the level of failure probability. The overall margin in the current fatigue design limit model (ASME design curve + NUREG/CR-6909 F_en model) is similar to that of the Weibull model with a cumulative failure probability of approximately 2.5%. The margin in the current fatigue design limit model demonstrated inconsistencies for the Ni-base alloy weld data, whereas the Weibull model showed a consistent margin. Therefore, the Weibull model can systematically mitigate the excessive safety margin.

• 대한두개안면성형외과학회지
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• 제25권2호
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• pp.90-94
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• 2024

Primary cutaneous mucinous carcinoma (PCMC) is a rare malignancy of the sweat glands that most commonly affects the periorbital area. It is characterized by slow growth over a prolonged period, and its morphology can be easily confused with a benign tumor, such as an epidermal cyst. Consequently, many patients experience recurrence after undergoing multiple resections. However, there are few reports concerning the surgical management of PCMC. We present two cases of PCMC originating in the periorbital area. The first case involved a 76-year-old man with a mass measuring 3.0×1.5 cm that had been increasing in size. The second case was a 61-year-old man with two masses, each measuring 1.0×1.0 cm, that were also growing. Both patients underwent wide excision with a 5-mm safety margin, which was determined based on the widest view of the cross-section of the mass on the magnetic resonance imaging. Subsequently, based on the intraoperative frozen biopsy results, both patients underwent additional excision with a 5-mm safety margin in only one direction. This report shows that, when determining the surgical margin of PCMC in periorbital area, employing imaging modalities and intraoperative frozen biopsies can be helpful for narrowing the surgical margin.

• 기술사
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• 제30권5호
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• 대한기계학회논문집A
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• 제29권1호
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• pp.107-114
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• 2005

In order to analyze the elastic-plastic fracture behavior of a structure, the fracture resistance curve of the material should be known first. The standard CT specimen was used to obtain the fracture resistance curves of a piping system. However, it is known that the fracture resistance curve by the standard CT specimen is very conservative to evaluate the integrity of a structure. Also the fracture resistance curve is effected by the specimen geometry and the dimensions because of the constraint effect. The objective of this paper is to be certain the conservativeness of the fracture resistance curve by the standard CT specimen and to provide an additional safety margin. For these, the fracture tests using a real pipe specimen and the standard CT specimen test were performed. A 4-point bending jig was manufactured for the pipe test and the direct current potential drop method was used to measure the crack extension and the length for the pipe test. Also finite element analyses were performed with a CT specimen and a pipe in order to prove the additional safety margin. From the result of tests and analyses of the pipe and the standard CT specimen, it was observed that the fracture analysis with the standard CT specimen is conservative and the additional safety margin was proved.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, 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 $B_4C$; 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 $B_4C$ and Ag-In-Cd alloy, respectively.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.45-57
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• 1990

아날로그와 디지탈 보호계통의 정상상태 여유도를 비교하였다. 비교된 예는 웨스팅하우스사의 OP Delta T 및 OT Delta T 계통과 CE사의 CPCS 계통이다. 안전해석 방법의 차이에 의한 여유도 영향을 제거하기 위해 Dynamic Offset은 고려하지 않았다. 핵연료봉 중심선의 용융을 방지하는데 있어서 디지탈 보호계통이 아날로그 보호계통보다 약 30% 출력 정도의 운전 여유도를 더 가졌다. DNB를 방지하는데 있어서는 주기말에서는 양 보호계통이 비슷한 여유를 가졌으나 주기초에서는 디지탈 보호계통이 약 10%의 더 많은 운전여유를 가지는 것으로 판단된다.

• 한국정밀공학회지
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• 제24권6호
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• pp.78-85
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• 2007

In this study, we performed finite element analysis for the design of a medical centrifuge and two-types of centrifuge were compared with each other. The types of centrifuge are 2-arm straight type and 3-arm type. Structural analysis was done with respect to the change of the rotational speed of the rotor of a centrifuge. When the rotor of centrifuge was rotated, the von Mises stress of 2-arm straight type-rotor was compared with the von Mises stress of 3-arm type. The margin of safety was estimated from the result. We found the critical speed of centrifuge from the campbell diagram by modal analysis.

• Structural Engineering and Mechanics
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• 제37권3호
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.