• 한국전산유체공학회지
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• 제13권3호
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• pp.21-27
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• 2008

The reactor head of a sodium-cooled fast reactor KALIMER-600 should be cooled during the reactor operation in order to maintain the integrity of sealing material and to prevent a creep fatigue. Analyzing turbulent natural convection flow in the cover gas region of reactor vessel with the commercial CFD code CFX10.0, the cooling requirement for the reactor head and the performance of the insulation plate were assessed. The results showed that the high temperature region around reactor vessel was caused by the convective heat transfer of Helium gas flow ascending the gap between the insulation plate and the reactor vessel inner wall. The insulation plate was shown to sufficiently block the radiative heat transfer from pool surface to reactor head to a satisfactory degree. More than $32.5m^3$/sec of cooling air flow rate was predicted to maintain the required temperature of reactor head.

• Transactions on Electrical and Electronic Materials
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• 제4권4호
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• pp.30-35
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• 2003

In this paper, the $CF_4$ decomposition efficiency is investigated for three simulated plasma reactors that are needle plate reactor, metal particle reactor, and dielectric barrier reactor with hole (DBH). The$CF_4$ decomposition efficiency by DBH is much better than that by needle plate reactor or metal particle reactor. When applied voltage is increased up to the critical voltage for spark formation in the all reactors, the $CF_4$ decomposition efficiency is increased. The $CF_4$ decomposition efficiency in needle plate reactor and metal particle reactor is about $12\%$ and $22\%$ respectively at applied voltage of 23 kV (consumption power: 110 W) and $CF_4$ concentration of 500 ppm, however, the $CF_4$ decomposition efficiency is more than $95\%$ in case of DBH. DBH should be much better than two reactors investigated for $CF_4$ decomposition.

• Korean Journal of Chemical Engineering
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• 제35권11호
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• pp.2157-2163
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• 2018

Catalytic reactors have been essential for chemical engineering process, and different designs of reactors in multi-scales have been previously studied. Computational fluid dynamics (CFD) utilized in reactor designs have been gaining interest due to its cost-effective advantage in designing the actual reactors before its construction. In this work, butadiene synthesis via oxidative dehydrogenation (ODH) of n-butene using tubular reactor was used as a case study in the CFD model. The effects of coolant and reactor diameter were investigated in assessing the reactor performance. Based on the results of the CFD model, the conversion and selectivity were 86.5% and 59.5% respectively in a fixed bed reactor under adiabatic condition. When coolants were used in a tubular reactor, reactor temperature profiles showed that solar salt had lower temperature gradients inside the reactor than the cooling water. Furthermore, higher conversion (90.9%) and selectivity (90.5%) were observed for solar salt as compared to the cooling water (88.4% for conversion and 86.3% for selectivity). Meanwhile, reducing the reactor diameter resulted in smaller temperature gradients with higher conversion and selectivity.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.425-425
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• 2002

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.193-194
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• 2019

• 환경위생공학
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• 제18권1호
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• pp.15-22
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• 2003

Treatability tests were conducted using EMC process to study the feasibility of applying this process as recycling-water treatment system in high density seawater aquaculture farm. To study the effect of organic and ammonia nitrogen loading on system performance, hydraulic retention time of reactor was reduced gradually from 12hr to 10min. The conclusions are can be summarized as follows. When the system HRT was reduced from 12hr to 10 min gradually, there was little noticeable change(reduction) in ammonia nitrogen removal efficiencies until 2hr of HRT, however, removal efficiencies were decreased dramatically when the system was operated under the HRT of less than 2hr. In case of organics(COD), there was no dramatic deterioration in removal efficiencies depending on HRT reduction. More than 90% of removal efficiencies were maintained successfully when the system was operated at the HRT of 10 min. In case of system performance depending on media packing ratio in reactor, there was little difference in each reactor performance depending on media packing ratio in reactor when the reactors were operated under the HRT of longer than 1hr, however, differences in reactor performances were considerably evident when the reactors were operated under the HRT of shorter than 1hr. That is, the more reactor was packed, the better reactor performed. When comparing reactor performance among 25%, 50%, 75% packed reactor, it can be judged that media packing ratio more than 50% plays no significant role in increasing reactor performance. For this reason, packing the media less than 50% is more reasonable way in view of economic. Such a tendency well agreed with the variation of ammonia-nitrogen removal efficiencies according to the media packing ratio in reactors at each HRT. Difference in effluent ammonia-nitrogen concentration between 50% media packing reactor and 75% media packing reactor was negligible. When comparing with the results of 25% packing reactor, difference was not so great.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• 한국유체기계학회 논문집
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• 제13권5호
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• pp.5-10
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• 2010

5MW, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.