• Journal of Microbiology and Biotechnology
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• 제16권11호
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• pp.1720-1727
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• 2006

To set up a kinetic model that can provide a theoretical basis for developing a new mathematical model of the Cr(VI) biosorption column using brown seaweed Ecklonia biomass, a differential reactor system was used in this study. Based on the fact that the removal process followed a redox reaction between Cr(VI) and the biomass, with no dispersion effect in the differential reactor, a new mathematical model was proposed to describe the removal of Cr(VI) from a liquid stream passing through the differential reactor. The reduction model of Cr(VI) by the differential reactor was zero order with respect to influent Cr(IlI) concentration, and first order with respect to both the biomass and influent Cr(VI) concentrations. The developed model described well the dynamics of Cr(VI) in the effluent. In conclusion, the developed model may be used for the design and performance prediction of the biosorption column process for Cr(VI) detoxification.

• International Journal of Control, Automation, and Systems
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• 제1권2호
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• pp.210-214
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• 2003

This paper presents a residence time distribution (RTD) measurement method using a radioisotope tracer and the estimation method of RTD model parameters to analyze a submerged bio-reactor. The mathematical RTD models have been investigated to represent the flow behavior and the existence of stagnant regions in the reactor. Knowing the parameters of the RTD model is important for understanding the mixing characteristics of a reactor The radioisotope tracer experiment was carried out by injecting a radioisotope tracer as a pulse into the inlet of the reactor and recording the change of its concentration at the outlet of the reactor to obtain the experimental RTD response. The parameter estimation was performed by the Levenberg-Marquardt optimization algorithm. The proposed scheme allowed the parameter estimation of RTD model suggested by Adler-Hovorka with very low deviations. The estimation procedure is shown to lead to accurate estimation of the RTD parameters and to a good agreement between experimental and simulated response.

• 한국물환경학회지
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• 제22권6호
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• pp.1137-1143
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• 2006

In this study we investigated the removal characteristics of NOM and bromate formation characteristics in SOMT reactor. The system was recently developed as a novel ozone reactor and installed in SJ pilot plant. DOC values were decreased within 3% after treatment of 0.5~2.0 mg/L ozone dosage in SOMT reactor while the $UV_{254}$ value was 69% decreased at 2.0 mg/L ozone dosage. The composition of NOM was analysed by LC-OCD (Organic Cabon Detector) after ozone treatment in SOMT reactor to elucidate the variation of NOM character. Polysaccharide (more than 20,000 g/mol) fraction of NOM was decomposed while building blocks (350~500 g/mol) and neutral (less than 350 g/mol) fraction increased. Spiked bromide reacted with 0.5~2.0 mg/L ozone dosage in the SOMT reactor. The bromate formation was proportional to the ozone dosage ($R^2=0.978$) but not proportional to reaction time. The maximum concentration of formated bromate was not exceeded to 10% of spiked bromide concentration.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.185-191
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• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3824-3832
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• 2022

Prompt capture gammas are an important part of the fission reactor gamma field. Because some of the structural materials after neutron capture can emit photons with high energies forming the dominant component of the gamma spectrum in the high energy region, the following study of the high energy capture gamma was carried out. High energy gamma radiation may play a major role in areas of the radiation sciences as reactor dosimetry. The HPGe measurements and calculations of the high-energy aluminum capture gamma were performed at two moderator levels in the VR-1 pool-type reactor. The result comparison for nominal levels was within two sigma uncertainties for the major 7.724 MeV peak. A larger discrepancy of 60% was found for the 7.693 MeV peak. The spectra were also measured using a stilbene detector, and a good agreement between HPGe and stilbene was observed. This confirms the validity of stilbene measurements of gamma flux. Additionally, agreement of the wide peak measurement in 7-9.2 MeV by stilbene detector shows the possibility of using the organic scintillators as an independent power monitor. This fact is valid in these reactor types because power is proportional to the thermal neutron flux, which is also proportional to the production of capture gammas forming the wide peak.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• 한국전산구조공학회논문집
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• 제31권1호
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• pp.31-38
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• 2018

설계기준을 초과하는 지진 재해는 원자력 시설물에 상당한 위험을 유발할 수 있다. 이러한 위험성을 확률론적으로 정량화하는 방법이 확률론적 지진 안전성 평가(seismic probabilistic safety assessment)이다. 이에 따라 지진 PSA는 국내외 다수의 원자력 발전소에 적용되어 지진 재해에 대한 원전의 안전성을 확률론적으로 평가하고 이에 대비토록 하고 있다. 그러나 원전에 비해 상대적으로 규모가 작은 연구용 원자로와 같은 경우에는 지진 PSA가 적용된 예가 거의 없다. 따라서, 본 연구에서는 지진 PSA기법을 실제 완공된 연구로에 적용하여 안전성을 분석하였다. 또한, 이를 바탕으로 연구로를 구성하는 시스템의 지진 내력에 대한 최적화 연구를 수행하였다. 그 결과, 지진 재해 하에서 연구로에 발생할 수 있는 노심 손상 가능성을 정량화하였고, 현재 설계안과 비교하여 적은 비용으로 최대의 안전성을 확보하는 최적 지진 내력 분포를 도출하였다. 이러한 결과는 향후 지진에 대비하여 연구로 안전성을 효과적으로 제고할 수 있는 정량적 지표로 활용할 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.546-555
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• 2019

Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2003년도 ICCAS
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• pp.2063-2065
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• 2003

The control rod drive mechanisms in a reactor head are arranged too narrow for a human worker to approach. Moreover, the working environment is in high radiation area. In order to inspect defections in the surfaces of the reactor head and welding parts, a visual inspection device that can approach such a narrow and high radiation area is required. This paper introduces a tele-operated mobile robot for visual inspection of a reactor head, which has pan/tilt camera, fixed rear camera, ultrasonic collision detection system, and so on. Moreover, the host controller and digital video logging system are developed and integrated control software is also developed. The robot is operated by a wireless control, which gives flexibility for the inspection.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2005년도 연구개발 발표회 논문집
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• pp.503-508
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• 2005

An advanced Research Reactor (ARR) consists of an open-tank-type reactor assembly within a light water pool and generates thermal power of 20 MW. The thermal power is including a fission heat in the core, a fuel generated heat temporary stored in the pool, a circulating pumps generated heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the primary cooling system will be installed. In this study, the conceptual design of the primary cooling system has been carried out using a design methodology of HANARO within a permissible range of safety. As results, it has been established that the conceptual design of the primary cooling system including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system functions.