• Journal of Magnetics
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• 제22권1호
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• pp.34-42
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• 2017

Material defects in the Steam Generator Tubes (SGT) of sodium cooled fast breeder reactor (PFBR) can lead to leakage of water into sodium. The water and sodium reaction will lead to major accidents. Therefore, the examination of steam generator tubes for the early detection of defects is an important requirement for safety and economic considerations. In this work, the Magnetic Flux Leakage (MFL) based Non Destructive Testing (NDT) technique is used to perform the defect detection process. The rectangular notch defects on the outer surface of steam generator tubes are modeled using COMSOL multiphysics 4.3a software. The obtained MFL images are de-noised to improve the integrity of flaw related information. Grey Level Co-occurrence Matrix (GLCM) features are extracted from MFL images and taken as input parameter to train the neural network. A comparative study on characterization have been carried out using feed-forward back propagation (FFBP) and cascade-forward back propagation (CFBP) algorithms. The results of both algorithms are evaluated with Mean Square Error (MSE) as a prediction performance measure. The average percentage error for length, depth and width are also computed. The result shows that the feed-forward back propagation network model performs better in characterizing the defects.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.785-796
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• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.

• 대한기계학회논문집A
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• 제38권12호
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• pp.1421-1426
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• 2014

소듐냉각고속로(Sodium-cooled Fast Reactor, SFR)의 안전성 향상을 위해 고온 증기 Rankine 싸이클 대신 초임계 이산화탄소(Supercritical $CO_2$, $S-CO_2$) Brayton 싸이클을 전력변환 시스템에 사용하는 방안이 제시되고 있다. 이 경우, 중간 열교환기로는 확산 접합(Diffusion Bonding)에 의해 제작되는 미소채널형 열교환기인 PCHE(Printed Circuit Heat Exchanger)가 고려되고 있다. 따라서 본 연구에서는 PCHE 형 열교환기 후보재료인 다양한 오스테나이트계 합금의 확산접합 특성을 평가하였다. 후보재료별로 다양한 조건에서 확산접합부를 제작하고 상온에서 $650^{\circ}C$까지의 인장 특성을 평가하였다. 평가 결과 SS 316H와 SS 347H는 $550^{\circ}C$까지 모재와 유사한 특성을 보였지만 Fe-Ni-Cr 합금인 Incoloy 800HT는 모든 온도에서 인장특성이 감소하였다. 연신율 저하의 원인을 이해하기 위해 접합부 부근의 미세조직을 분석하였다.

• 대한기계학회논문집A
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• 제38권12호
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• pp.1415-1420
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• 2014

소듐냉각고속로(Sodium-cooled Fast Reactor, SFR)의 증기 Rankine 싸이클 발전시스템을 높은 열효율과 안전성을 가지는 초임계 이산화탄소(Supercritical carbon dioxide, $S-CO_2$) Brayton 싸이클로 대체하는 방안이 고려되고 있다. 다양한 오스테나이트계 합금이 고온 $S-CO_2$ 환경 열교환시스템 구조재료로 제시되고 있다. 구조재료는 장시간 고온 $S-CO_2$ 환경에 노출됨에 따라 미세구조에 변화가 일어나고, 나아가 기계적 특성의 저하가 발생할 수 있다. 본 연구에서는 오스테니틱 스테인리스강들과 Alloy 800HT를 고온 $S-CO_2$ 환경에 노출시키고, 그에 따른 부식특성 및 인장특성을 평가하였다. 그 결과 $650^{\circ}C$, 250시간 노출 후 316H SS와 800HT에서 큰 연신율 감소를 보였다. $S-CO_2$ 환경이 인장특성 변화에 미치는 영향을 표면 산화막 및 탄화거동을 통해 분석한 결과, 316H 와 800H 의 거동이 큰 차이를 보였다.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.