• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.

• Investigative Magnetic Resonance Imaging
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• v.3 no.1
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• pp.73-77
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• 1999

Purpose : The purpose of the study was to evaluate the sensitivity and specificity of proton-weighted fast spin-echo MR imaging in diagnosing the meniscal tear of knee as an reasonable substitute for conventional spin-echo imaging. Materials and Methods : 102 consecutive patients, proved by surgery, proved by surgery, were participated in this study. All of them were suspected internal derangement of knee, examined by fast spin-echo MR imaging including sagittal and coronal images on a 1.5T MR imager and underwent arthroscopic or open surgery of knee joint within 2 months. These images were reviewed retrospectively by three radiologists. The sensitivity and specificity of meniscal tear were calculated. Results : The sensitivity and specificity of meniscal tear using proton-weighted fast spin-echo MR imaging were 94%, 93% inmedialmeniscus and 92%, 88% in lateral meniscus. Conclusion : The sensitivity and specificity of meniscal tear using proton weighted fast-spin echo MR images were as high as those using conventional spin-echo images. The proton-weighted fast-spin echo MR imaging can be an alternative to conventional spin-echo MR imaging in diagnosing meniscal tear of the knee.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.23 no.5
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• pp.535-541
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• 2010

In this paper, we have derived a continuum-based adjoint design sensitivity of general performance functionals with respect to Young' modulus and heat conduction coefficient for steady-state nonlinear thermoelastic problems. An adjoint equation for temperature and displacement fields is defined for the efficient computation of the coupled field design sensitivity. Through numerical examples, we investigated the mesh dependency of the topology optimization method in the thermoelastic problems. Also, comparing the dominant loading cases of thermal and mechanical ones, the loading dependency of topology design optimization in coupled multi-physics problems is investigated.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.1855-1856
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• 2008

본 연구에서는 공정 중에 민감하게 반응하는 플라즈마로부터 수집되는 이온에너지 분포(IED : Ion Energy Distribution)와 시계열 신경망 모델링을 결합한 플라즈마 감시 기술을 개발하였다. NIEA(Non-invasive ion analyzer)를 이용하여 IED를 측정하였으며, 모델링에 사용된 신경망은 자기 상관 시계열 신경망(A-NTS : Auto-Correlated Neural Time-Series)이다. 모델 개발을 위한 학습과 테스트 데이터로는 Duty ratio 100%에서 수집한 IED를 이용하였으며, 개발된 모델의 감시 성능은 60%에서 수집된 IED로 평가하였다. 학습인자 k와 m의 범위는 각각 1-3 으로 총 9종류의 (k, m) 조합에 대해서 모델 성능을 평가하였다. 신경망 은닉층 뉴런수는 2-9의 범위에서 최적화하였다. 최적화된 모델은 (2, 3)과 뉴런수 2에서 구해졌으며, 0.335의 예측 에러를 보였다. 60% IED 데이터로 평가한 결과 플라즈마 고장에의 민감도는 62% 이상이었다. 이는 IED의 A-NTS 모델이 플라즈마 고장의 감시에 효과적으로 적용될 수 있음을 의미한다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.40-49
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• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• Journal of the Korean Institute of Landscape Architecture
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• v.44 no.4
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• pp.109-120
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• 2016

This study aims to assess vulnerability in urban thermal environments of Seoul by using indicators frequently used in previous studies as well as effective indicators for spatial planning like urban and environmental planning. For this purpose, fifteen indicators that influence urban thermal environments such as heat waves, urban heat island effects, and urban micro-climates were identified based on literature reviews. Indicators for presenting urban structure and spatial properties were included; for example, building volume as 'exposure to climate', buildings completed before 1980 as 'sensitivity', and green space areas as 'adaptive capacity'. Among them, twelve indicators were applied to assess vulnerability in urban thermal environments of Seoul by using a GIS spatial analysis combined with fuzzy logic. The results show that the Gangnam area is identified as more vulnerable to a heat environment as compared to the Gangbuk area. In the Gangnam area, Seocho-gu, Gangnam-gu, Dongjak-gu, Yeongdeungpo-gu, Gangseo-gu were relatively high in vulnerability, while Dongdaemun-gu, Gangbuk-gu, Gwangjin-gu, Jungrang-gu were relatively high in the Gangbuk area. Gwanak-gu, Dobong-gu, Eunpyeong-gu, and Nowon-gu, which include forested areas, have low vulnerability in the sectors of 'exposure to climate' and 'sensitivity' due to the impact of Gwanaksan and Bukhansan. However, some areas with high vulnerability like Seocho-gu and Gangnam-gu may have lower vulnerability if the indicator 'status of air conditioning' from the sector of 'adaptive capacity' is used. This study could support the establishment of a practicable thermal environment policy and spatial planning to reduce heat-related risks in the field of urban and environmental planning.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.243-244
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• 2015

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.526-531
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• 1998

본 논문은 농축도 0.9%의 순환우라늄 핵연료(CANFLEX-RU)에 대한 축방향 출력분포(AFD) 및 반경방향 출력분포(RFD) 특성을 조사하고 CANFLEX-RU 다발이 장전된 CANDU줄 채널의 예비 열수력 해석을 수행하였다. CANFLEX-RU 다발의 4 bundle shift 핵연료 교체 방법에 따라 AFD 분포 특성은 정점(Peak) 열속이 채널 상류쪽으로 이동하였고 채널 중심 부근에서 평탄하거나 다소 오목한 형상을 보여주었다. RFD 분포를 표현하는 적절한 변수로서 국부 다발열유속비를 정의하고, 이 비와 국부 표면열유속비의 상호 관계식을 도출하였다. 연소도에 따른 최외환봉의 국부 다발열유속비 변화를 조사한 결과로서, CANFLEX-RU 다발의 최대 국부 다발열유속비는 초기 연소도에서 발생되었고 이 값 CANFLEX-NU 다발 보다는 크고 37-핵연료봉다발 보다는 작았다. CCP 계산시에 RFD 분포 효과를 고려하는 방안으로서 최외환봉 열유속을 다발의 국부 열유속으로 가정하였다 이는 임계열유속이 -10.2% 감소한 조건을 사용하여 CCP를 계산하는 결과가 되었다. 다발-블균형 계수를 이용한 CCP 민감도 결과와 본 계산에서 얻은 CCP 결과에 의하면, CANFLEX-RU의 CCP 는 CANFLEX-NU에 비교해서 土1.0% 이내로 근사한 분포가 예상되었으며 이는 AFD 분포 효과가 RFD 분포에 의한 CCP 감소를 보상하기 때문이다. 결론적으로, CANFLEX-RU는 열수력적 설계 관점에서 CANFLEX-NU에 비교해서 열적 성능이 저하되지 않았고 따라서 기존 37-핵연료봉다발에 대한 CANFLEX-NU의 열여유도 증가와 같은 장점을 유지할 것으로 예상되었다.

• Fire Science and Engineering
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• v.24 no.5
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• pp.60-67
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• 2010

Fire model shall be verified and validated to reliably predict the consequences of fires within its limitations. Generally the verification and validation procedures are conducted by comparison with experimental test data. This study aims to evaluate predictive capabilities of FDS in the pool fire with two rooms and the sensitivity between input parameters such as heat release rate and ventilation rate and the output values like temperature, concentration, and heat flux. The predictive capabilities of FDS will be evaluated by comparing FDS simulation results with PRISME experimental data which result from the international fire test project. The sensitivity analysis will be conducted to decide which one of input parameters affects outcomes by comparison of FDS results with ${\pm}$ 10% changes of input parameter. From this study, the FDS predictive capabilities are within 20% error range. Heat release rate as input parameter affects most of outcomes and flow rate only has relation with concentration of oxygen and combustion products.