• Nuclear Engineering and Technology
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• 제52권3호
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• 전기화학회지
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• 제4권2호
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• pp.70-76
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• 2001

본 연구는 Pd박막 전극에서 수소 확산시 발생되는 응력해석에 대한 레이저 빔 디플렉션 방법의 응용에 대해 기술하였다. 우선, 탄성에 의한 확산 (고스키 효과) 및 확산에 의한 탄성 현상에 대해 간략히 설명하였고, 주어진 초기 및 경계 조건하에서 Fick 방정식의 해와 Vegard 및 Hooke의 법칙으로부터 확산에 의한 탄성 현상의 모델을 이론적으로 유도하였다. 다음으로 레이저 빔 디플렉션 방법이 수소 확산으로 인해 발생되는 응력해석에 어떻게 사용될 수 있는지 실험 장치 및 시편에 대해 소개하였다 마지막으로, 수학적으로 계산된 디플렉션 시간 추이 곡선과 실험적으로 얻어진 곡선의 비교로부터, 시간에 따른 인장 디플렉션의 변화를 시간에 따른 전극 내부의 수소 농도 구배의 변화 및 수소 확산계수의 차이로 설명하였다.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Journal of Radiation Protection and Research
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• 제27권3호
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• pp.189-198
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• 2002

방사선 비상사고시 예상되는 주민행동특성 조사 및 교통분석을 통해 실제적인 가정에 기초한 울진원전 비상계획구역내 주민들에 대한 소개시간평가를 수행하였다. 본 연구에서 소개시간은 주민통보, 소개준비 및 차량소개시간으로 구성되었다. 소개대상인구는 비상계획구역내 인구밀도 행정구역 및 일시체류인구 등을 고려해 4개의 그룹으로 분류하였다. 주민행동특성 조사를 위해 비상계획구역내 200가구에 대해 설문조사를 실시하였으며, 설문조사에는 가상사고상황을 설명하는 시나리오를 포함하여, 거주지, 소개준비 소요시간, 소개시 교통수단, 대피장소, 소개방향 등에 대한 질의를 포함하였다. 계산된 소개시작 시간분포 및 미시적 교통분석모델인 CORSIM을 이용하여 도로상에서 소개하는 각 차량들의 거동을 모사하였다. 본 연구결과에서 모든 소개대상차량이 비상계획구역 외부로 소개하는 데 있어서는 밤보다는 낮에 소개하는 경우에 더 오랜 시간이 소요되며 반면에 교차로에서의 지체시간은 낮보다는 밤이 더 장시간 지체되는 것을 확인할 수 있었다. 이것은 차량소개 시작분포에 의한 영향에 기인하는 것으로 분석되었다. CORSIM 모델이 비상사고시 나타날 수 있는 혼잡한 교통현상을 적절히 모사할 수 있는 가를 검증하기 위해 오전 출근시간대에 울진원전 주변의 신호등이 없는 교차로에서 Benchmark Test를 수행하였다. 이 시험에서 CORSIM 모델의 예측치는 관찰된 통과차량 수와 잘 일치하여 본 연구목적을 만족시키고 있음을 확인할 수 있었다.