• 상하수도학회지
• /
• 제10권4호
• /
• pp.119-126
• /
• 1996

Influence factors and efficiency characteristics for treatment of wastewater containing phenol were studied with using Pseudomonas sp. B3. It took 130 hours to remove phenol, when only activated sludge of terminal disposal palnt of sewage was innoculated in batch culture, but it was required just 36 hours, when bacteria degrading phenol and activated sludge were simultaneously innoculated. If only phenol an carbon source was used, it necessary 36 hours for biodegradation of phenol, while glucose was added to medium, it took 73 hours. It was revealed as excellent effluent and SVI, when the F/M ratio, COD and phenol concentration were 53mg/l and 1.2mg/l, respectively, and optimum F/M ratio was revealed 0.31. The reactor were seriously shocked as reducing hydraulic retention time at constant phenol concentration more than increasing phenol concentration at constant hydraulic retention time, when volumetric loading rate was increased to $0.8kg\;phenol/m^3{\codt}d$ from $1.6kg\;phenol/m^3{\codt}d$. And also the effluent phenol concentration was 34mg/l after starting 12 hours of shocking and reactor was recovered as steady state after 65 hours of changing in the former test. Although the effluent phenol concentration was maximum value with 12mg/l after starting 20 hours of shocking and reactor was recovered as steady state after 54 hours of changing in the later test.

• Nuclear Engineering and Technology
• /
• 제40권6호
• /
• pp.517-526
• /
• 2008

The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.

• 전기학회논문지P
• /
• 제63권3호
• /
• pp.138-142
• /
• 2014

In general, the voltage stability of induction generator is lower than synchronous generator. However, induction generator has many advantages rather than a synchronous generator in terms of price and maintenance. So Induction generator is used little by little in small hydroelectric power station rather than 1000kW recently. Squirrel cage induction generator generates a high inrush current at the grid-connection. This high inrush current causes a voltage drop on the grid. In order to increase the penetration of the induction generator, it is necessary to present a method of reducing inrush current. In this study, we suggested that it is possible to present a reactor startup method, by applying the parameter to reduce the voltage drop.

• Nuclear Engineering and Technology
• /
• 제45권4호
• /
• pp.477-488
• /
• 2013

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

• Nuclear Engineering and Technology
• /
• 제45권6호
• /
• pp.767-790
• /
• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Journal of Radiation Protection and Research
• /
• 제37권2호
• /
• pp.63-69
• /
• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• 한국대기환경학회지
• /
• 제16권4호
• /
• pp.399-408
• /
• 2000

This study evaluated the SO2 adsorption characteristics using a continous moving bed system. Natural manganese oxide (NMO) reaction condition such as L/D the starting time of the NMO feed, feed rate, and flow rate of simulated flue gas, and NMO size were tested. The results showed that optimum L/D was 1.0 in this moving bed system. The higher the feeding rate was the higher the SO2 removal efficiency was and the higher the flow rate of simulated flue gas was the shorter the time to reach the euqilibirum concentration was. The final SO2 con-centration when it reaches the equilibrium concentration was not affected by the starting time of the NMO feed.

• 한국정밀공학회지
• /
• 제28권2호
• /
• pp.233-238
• /
• 2011

In piping design of nuclear power plant facilities, the load stress according to self-weight is important for design values in test run(shutdown and starting). But sometimes it needs more studies, such as seismic analysis of an earthquake of power plant area and fatigue life and stress of thermal expansion and anchor displacement in operating run. In this paper, seismic evaluations were performed to nuclear piping system of Shin-Kori NO. 3&4 being built in Pusan lately. Results of seismic analysis are evaluated on basis of KEPIC MN code. The structural integrity on RCB piping system was proved.

• 대한화학회지
• /
• 제32권2호
• /
• pp.149-155
• /
• 1988

ZSM-5, ZSM-8, ZSM-11 및 실리카라이트를 포함하는 펜타실 제올라이트들을 여러가지 유기양이온과 실리카졸을 사용하여 합성하고 반응온도, 교반, 숙성시간 등 합성조건을 변화시키므로서 균일한 크기와 모양을 갖는 결정을 얻었다. TEA-OH(수산화 사에틸암모늄), TPA-OH, TBA-OH 및 Choline이 유기양이온으로 사용되었으며 콜로이달 실리카(Snowtex)를 실리카원으로 사용했다. 합성 실험은 $2{\ell}$ 압력 반응기 (parr제품)와 자체 제작된 자석식 반응기를 사용하였으며, 반응물 조성(실리카/알루미나 비), 반응온도 등이 펜타실 제올라이트 합성의 중요한 인자임을 알았다.

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