• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3643-3652
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• 2021

Sizing the tube inlet orifice of a Once-Through Steam Generator (OTSG) is important to protect the integrity of the tubes from thermal cycling and vibration wear. In this study, a new sizing criterion is proposed for the tube inlet orifice to suppress the parallel channel instability in an OTSG. A perturbation method is used to capture the essential parts of the thermal-hydraulic phenomena of the parallel channel instability. The perturbation model of the heat transfer regime boundaries is identified as a missing part in existing models for sizing the OTSG tube inlet orifice. Limitations and deficiency of the existing models are identified and the reasons for the limitations are explained. The newly proposed model can be utilized to size the tube inlet orifice to suppress the parallel channel instability without excessive engineering margin.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.166-172
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• 1996

The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for Westinghouse type PWR plants as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increase approximately by 1.2%. The resulting DNBR penalties can be covered within the current DNBR margin. Accident analyses are also investigated and the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation.

• 한국항공우주학회지
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• 제42권9호
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• pp.789-794
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• 2014

정지궤도급 차세대 통신위성에 탑재될 디지털신호처리기에는 디지털 고속통신을 위한 FPGA가 사용된다. 적용된 FPGA는 높은 열소산량을 가지고 있으며, 이로 인한 접합온도의 상승은 부하경감 요구조건을 만족하기 어렵고 장비의 수명과 신뢰도 저하의 주요 원인이다. 지상과는 달리 우주환경에서의 전장품의 열제어는 대부분 열전도를 통하여 이루어지고 있다. CCGA 또는 BGA 형태의 FPGA는 인쇄회로기판에 장착되지만, 인쇄회로기판의 열전도율은 FPGA의 열제어에 효율적이지 못하다. FPGA의 열제어를 위하여 부품 리드와 하우징을 직접 연결하는 히트싱크를 제작하였으며, 우주인증레벨의 열진공시험을 통하여 그 성능을 확인하였다. 높은 전력소모량을 가진 FPGA는 우주환경에 적용하기 어려웠으나, 히트싱크를 적용함으로써 부하경감 온도 마진을 확보하였다.

• 대한전자공학회논문지SD
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• 제38권4호
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• pp.243-250
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• 2001

본 논문에서는 μspring 패키지의 구조와 제조공정을 소개하고, 전기적 특성을 μBGA와 비교 분석한 결과를 제시하였다. μBGA에서와 같이 μSpring 패키지의 연결선 인덕턴스 값은 기존의 TSOP 패키지의 반 이하로서 월등한 고속 신호 전달 특성을 제공하게 된다. 또한 μSpring CSP 패키지의 경우 가장 열악한 substrate trace를 가진 핀에서도 2.9nH로 평가되어, Rambus DRAM module의 인덕턴스 규격 상한 값 4nH에 비하여, 약 25% 정도의 margin을 제공한다. μSpring CSP패키지는 μBGA의 약 50%의 제조 비용으로서 μBGA가 만족시키지 못하는 JEDEC Level 1 규격을 충족시킬 뿐만 아니라, thermal cycle 1000회를 통과하는 높은 신뢰성을 제공하여 강력한 경쟁력을 가진다.

• 에너지공학
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• 제12권3호
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• pp.177-183
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• 2003

본 논문은 방사능조사의 조건하에서의 설계조건을 만족해야 하는 fuel rod의 설계응력값을 검증하는데 그 목적이 있다. 이 경우에, 안전성에 있어서 열에 의한 영향이 가장 주된 고려조건이 된다. 그러나 이러한 열영향이 고려된 해석은 구조물의 안전성해석에서 비교적 간단히 해결되는 문제가 아니다. 여기서는 이론적 해석을 통한 접근방식으로 보수적인 관점에서 fuel rod의 설계에 적용되는 설계 응력값을 검증하고자 하였다. 추후에 시도하는 fuel rod의 설계에 있어서 본 해석방법을 이용하면 안전설계의 검증을 위한 이론적 접근방법의 하나로 이용할 수 있을 것으로 사료된다.

• 한국안전학회지
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• 제33권5호
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• pp.164-169
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• 2018

A detailed fatigue evaluation procedure was developed to mitigate the excessive conservativeness of the conventional environmental fatigue evaluation method for the pressurizer spray line elbow of domestic new nuclear power plants. The pressurizer spray line is made of austenitic stainless steel, which is relatively sensitive to the environmentally assisted fatigue, and has a low degree of design margin in terms of environmentally assisted fatigue due to the thermal stratification phenomenon on the pipe cross section as a whole or locally. In this study, to meet the environmental fatigue design requirements of the pressurizer spray line elbow, the new environmental fatigue evaluation has been performed, which used the ASME Code NB-3200-based detailed fatigue analysis and the environmental fatigue correction factor instead of the existing NB-3600 evaluation method. As a result, the design requirements for environmentally assisted fatigue were met in all parts of the pressurizer spray line elbow including the fatigue weakened zones by thermal stratification.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.481-488
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• 2014

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• 한국압력기기공학회 논문집
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• 제20권2호
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• pp.107-114
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• 2024

This study analyzes the effect of solid-state phase transformation (SSPT) on finite element (FE) residual stress analysis. The SA508 Grade 3 Class 1 plate was used for the analysis. A cylindrical 3D heat source was applied for the thermal simulation, and the thermal model variables were determined using experimental temperature variation over time and the shape of fusion zone after welding. Stress analysis was performed based on the temperature gradient obtained from the thermal simulation, and the experimental residual stress of electron beam-welded SA508 was accurately reproduced with consideration of SSPT during simulation. The validated model was then used to investigate the effect of SSPT on welding residual stress. The results showed that longitudinal residual stress at the weld centerline was significantly overestimated, however, the transverse residual stress was acceptably reproduced. Consequently, the transverse residual stress might be estimated without consideration of SSPT, and the longitudinal stress should be analyzed with consideration of SSPT to improve the design margin for electron beam-welded structures.