• 대한환경공학회지
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• 제27권2호
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• pp.215-223
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• 2005

본 연구는 세 가지 다른 형태의 플라즈마 반응기 (SD, DBD, PDC)를 이용한 기상의 스타이렌 분해실험을 통하여 최적 플라즈마 반응기에 대하여 고찰하였다. 각 플라즈마 반응기에 대한 비교평가를 위하여 스타이렌 분해효율, 탄소수지, 반응 생성물의 동정, 생성물의 수율 및 선택성 등의 항목을 평가하였다. 스타이렌의 전환과정은 오존과의 반응이 중요하며, PDC 반응기보다 오존생성량이 많은 SD와 DBD 반응기가 스타이렌의 전환율이 더 높은 것으로 나타났다. 한편, PDC 반응기는 탄소수지, COx ($CO+CO_2$)의 수율 및 선택성에 있어서, SD와 DBD 반응기 보다 훨씬 더 뛰어난 것으로 판명되었다. 스타이렌 초기농도를 100ppmv로 하였을 때, PDC 반응기와 플라즈마 단독공정에서 탄소수지 100%를 달성하기 위해 필요한 비투입 에너지는 각각 110 J/L와 420 J/L로, PDC 반응기가 훨씬 더 낮은 에너지로 스타이렌의 완전분해가 가능하였다. 스타이렌의 분해과정에서 생성되는 주된 생성물로는 CO와 $CO_2$가 있으며 HCOOH가 미량 성분으로 관찰되었다. 이러한 반응 생성물의 수율에는 차이가 나타나지만 이들의 분포는 플라즈마 반응기의 종류와 관계없이 거의 동일한 것으로 나타났다. 이상적인 스타이렌 분해 성성물인 $CO_2$의 선택성에 있어서 SD와 DBD 반응기는 $39.5{\sim}60%$ 정도를 나타내었으나 PDC 반응기에서는 $68.5{\sim}75.5%$ 정도로 훨씬 더 높은 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2047-2052
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• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.153-153
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• 1998

• 대한전기학회:학술대회논문집
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• 대한전기학회 1988년도 추계학술대회 논문집 학회본부
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• pp.388-393
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• 1988

New circuit D-Q transformation concept is introduced to analyze AC converters such as inverters, rectifiers and cyclo-converters with ease. The equivalent linear time invariant circuit is obtained by substituting switches with equivalent turn-ratio variable transformers and changing balanced AC reactors into equivalent DC reactors combined by gyrators. This circuit enables us to utilize the powerful linear system analysis techniques such as Laplace transform otherwise which could not be applied to the time varying switching systems. Direct substitution of switches of DC converters with transformers is shown as a preliminary. Then the modeling procedure is shown for a controlled rectifier-inverter circuit. Finally an analysis example is proposed for a buck-boost inverter and the result is compared with the conventional approach. This approach is applicable to all AC converter families to determine the AC transfer functions and the DC operating points. It is identified that the switching systems are equivalent to the RLC filter circuits with transformers and gyrators.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.501-512
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• 2014

HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.660-665
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• 1998

The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scale and verified through the methodology in this paper, which is referred to advanced Liquid Metal Reactor (ALMR). a Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the intergrated code system. Intergrated Code System (ICS) consist of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related are analyzed once-through. Results of conceptual design are attached in this paper.

• Nuclear Engineering and Technology
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• 제38권8호
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• pp.753-762
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• 2006

In this paper, works related to enhancement of the CHF are reviewed in terms of fundamental mechanisms and practical applications. Studies on CHF enhancement in forced convection are divided into two categories, CHF enhancement of internal flow in tubes and enhancement of CHF in the nuclear fuel bundle. Methods of enhancing the CHF of internal flows in tubes include enhancement of the swirl flow using twisted tapes, a helical coil, and a grooved surface; promotion of flow mixing using a hypervapotron; altering the characteristics of the heated surface using porous coatings and nano-fluids; and changing the surface tension of the fluid using additives such as surfactants. In the fuel bundle, mixing vanes or wire wrapped rods can be employed to enhance the CHF by changing the flow distributions. These methods can be applied to practical heat exchange systems such as nuclear reactors, fossil boilers, fusion reactors, etc.

• Nuclear Engineering and Technology
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• 제39권1호
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• pp.21-30
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• 2007

Current high-temperature gas-cooled reactors (HTGRs) are based on a closed Brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.