• 방사성폐기물학회지
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• 제12권3호
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• pp.235-243
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• 2014

본 논문에서는 X선 광전자 분광법(X-ray photoelectron spectroscopy, XPS)을 이용하여 ZIRLO (ZIRconium Low Oxidation) 피복관의 표면 산화 거동을 연구하였다. 산화 시간 (10-336 시간, $500^{\circ}C$) 및 산화 온도 ($400-700^{\circ}C$, 10 시간)에 따른 산화 특성변화를 관찰하였다. XPS peak 분석 결과, $500^{\circ}C$에서 산화된 피복관의 산화 시간이 24 시간이 될 때 $ZrO_2$ peak가 11.86% 관찰되었으며, 이후 산화 시간이 길어질수록 $ZrO_2$의 비율이 17.93%까지 (336 시간) 증가하는 것이 확인되었다. 반면, 10 시간 산화된 피복관에서 5.68% 존재하던 ZrO 상은 산화 시간이 24 시간으로 늘어남에 따라 사라지는 것이 관찰되었다. 산화 온도 증가에 의한 영향 분석 결과에서는 산화 온도가 $400^{\circ}C$에서 500, 600, $700^{\circ}C$로 증가할 때 ZrO 상의 비율이 0% 부터 5.68, 8.31, 9.16%로 증가하는 것이 확인되었다. 이 때, $ZrO_2$ 상은 $700^{\circ}C$에서 산화된 시료에서만 관찰되었다. ZrO 상의 형성 메커니즘은 불명확하지만, 고온에서 공기 중의 수분과 Zr의 반응으로 인해 $Zr(OH)_4$ 상의 형성이 가속되는 것으로 예상된다. 본 논문에는 $500^{\circ}C$에서 산화된 ZIRLO 피복관과 이들의 염소화 반응 특성에 대한 논의도 포함되었으며, 염소화 반응 진행 가능성에 있어서 산화막의 두께가 중요한 역할을 하는 것으로 보여진다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• 한국결정성장학회지
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• 제19권1호
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• pp.6-10
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• 2009

본 논문에서는 혼합소스(mixed-source) HVPE(hydride vapor phase epitaxy)방법으로 선택성장(SAC: selective area growth) GaN/AlGaN 이종접합구조의 발광다이오드를 r-plane 사파이어 기판 위에 제작하였다. SAG-GaN/AlGaN DH(double heterostructure)는 고온 GaN 버퍼층, Te 도핑된 AlGaN n-클래딩층. Gan 활성층. Mg 도핑된 AlGaN p-클래딩층. Mg 도핑된 GaN p-캡층으로 구성되어있다. GaN/AlGaN 이종접합구조의 발광다이오드의 특성을 알아보기 위해 SEM을 통한 구조적 분석과 전류-전압 측정(I-V: current-voltage measurement), 전류-광출력(EL: electroluminescence) 측정을 통하여 전기적, 광학적 특성을 평가하였다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Journal of the Optical Society of Korea
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• 제13권4호
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• pp.428-433
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• 2009

A photonic crystal fiber with two large air holes outside the holey cladding region is fabricated to induce an effective long periodic grating (LPG) in the core by an electric arc discharge. We believe that the two large air holes lead to the asymmetric perturbation in the core under the electric arc discharge, thereby introducing the coupling to the first higher-order mode. The transmission characteristics of the PCF with the LPG for the external perturbation such as strain, curvature, and temperature are also investigated. It was found that the shift of resonance peak in the transmission spectrum depends on the bending direction. The curvature of 8.55 $m^{-1}$ results in the center wavelength shifts of 1.8, 4.3, and 11 nm for a vertical, diagonal, and horizontal direction of the curvature to the large air-hole alignment, respectively.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.97-109
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• 2020

On March 11, 2011, an earthquake followed by a tsunami caused an extended station blackout (SBO) at the Fukushima Dai-ichi NPP Units. The accident was initiated by a total loss of both onsite and offsite electrical power resulting in the loss of the ultimate heat sink for several days, and a consequent core melt in some units where proper mitigation strategies could not be implemented in a timely fashion. To enhance the plant's coping capability, the Diverse and Flexible Strategies (FLEX) were proposed to append the Emergency Operation Procedures (EOPs) by relying on portable equipment as an additional line of defense. To assess the success window of FLEX strategies, all sources of uncertainties need to be considered, using a physics-based model or system code. This necessitates conducting a large number of simulations to reflect all potential variations in initial, boundary, and design conditions as well as thermophysical properties, empirical models, and scenario uncertainties. Alternatively, data-driven models may provide a fast tool to predict the success window of FLEX strategies given the underlying uncertainties. This paper explores the applicability of Artificial Intelligence (AI) to identify the success window of FLEX strategy for extended SBO. The developed model can be trained and validated using data produced by the lumped parameter thermal-hydraulic code, MARS-KS, as best estimate system code loosely coupled with Dakota for uncertainty quantification. A Systems Engineering (SE) approach is used to plan and manage the process of using AI to predict the success window of FLEX strategies under extended SBO conditions.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.