• Journal of computational fluids engineering
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• v.13 no.3
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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.

• Korean Journal of Chemical Engineering
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• v.35 no.11
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1353-1354
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• 2004

It was estimated by an analysis method thai the inner shell of HANARO reactor will be deformed due to pressure, loads, creep and growth during reactor operation. To confirm the analysis validity and safe operation of reactor, we developed tools to remotely measure the straightness of the inner shell located 12m below the pool top. The performance and the accuracy of the measurement tools have been verified through tests using a dummy inner shell and steel straight edge. The accuracy of the measurement shows very good results with a maximum error of 0.06mm by steel straight edge. The technical experiences described in this paper will be a good reference not only for the operation and maintenance of HANARO but also for the next performance of the measurement in the future.

• Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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• 2000.05a
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• pp.92-95
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• 2000

A mixed bacterial culture capable of mineralizing methyl tort-butyl ether (MTBE), other fuel oxygenates ethers, tertiary carbon alcohols, benzene and toluene was used to inoculate batch reactor and sequence batch reactor (SBR) to treat gasoline contaminated ground water containing about 60 mg/L MTBE, 5 mg/L benzene, 5 mg/L toluene, and low concentrations of several other aromatic and aliphatic hydrocarbons. Respirometery studies showed that MTBE degrading mixed culture could treat MTBE contaminated ground water with addition of nitrogen and phosphate. SBR was operated to demonstrate the feasibility of using suspended growth activated system for the treatment of ground water and to confirm that the respirometry derived kinetics and stoichiometric coefficients were useful for predicting reactor performance. Theoretical performance of the reactor was predicted using mathematical models calibrated with biokinetic parameters derived from respirometry studies.

• Journal of the Korean Society of Visualization
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• v.9 no.4
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• pp.54-62
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• 2011

KAERI has developed a KALIMER-600 which is a pool-type sodium-cooled fast reactor with a 600MWe electric generation capacity. For a SFR development, one of the main topics is an enhancement of the reactor system safety. Therefore, we have a long-term plan to design the large sodium experimental facility to evaluate the reactor safety and component performance. In order to extrapolate a thermal hydraulic phenomena in a large sodium reactor, the thermal hydraulics phenomena is under investigation in a 1/$10^{th}$ water-simulant facility for the KALIMER-600. In this paper, we shortly described the experimental facility setup and the measurement of the isothermal global flow behavior. For the flow field measurement, the PIV method was used in a transparent Plexiglas reactor vessel model at around $20^{\circ}C$ water condition.

• Proceedings of the Korean Vacuum Society Conference
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• 2013.02a
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• pp.670-670
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• 2013

The concept of a fusion-driven transmutation reactor based on LAR (Low Aspect Ratio) tokamak as a neutron source is studied based on ITER physics and technology. The radial build of transmutation reactor components are self-consistently determined by coupling the systems analysis with radiation transport analysis and an optimal configuration of a transmutation reactor for aspect ratio, A in the range of 1.5 to 2.0 is found. The performance of a transmutation reactor is investigated and shows that a transmutation reactor with a neutron source producing fusion power less than 150 MW can destroy the transuranic actinides contained in the spent fuels produced from more than two 1 GWe PWRs with production of the fission power being greater than 2 GW.

• 제어로봇시스템학회:학술대회논문집
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• 1993.10a
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• pp.766-771
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• 1993

A mathematical model is developed to represent the batch reactor for free radical polymerization of PMMA The model is validated by comparing the simulation result with the experimental data. A computational scheme is proposed to determine the trajectory of the reactor temperature that will produce polymer product having the desired molecular weight distribution. For this instantaneous number average chain length and polydispersity are introduced to calculate the reactor temperature. The former is assumed to be a second order polynomial of the sum of the living and dead polymer concentrations. Various reactor temperature, trajectories are observed depending on the reactor conditions and prescribed molecular weight distributions. Fuzzy and PID control algorithms are applied to trace the reactor temperature trajectory. While the PID control gives rise to an overshoot when the trajectory changes its direction, the fuzzy controller yields a more satisfactory performance because it secures information on the trajectory pattern.

• Proceedings of the KSME Conference
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• 2000.11b
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.551-560
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• 2006

In this study, a controller for a nuclear reactor power is designed. The reactor is modeled using the three dimensional reactor design code MASTER. From the relationship of the input and output of the reactor code, a reactor dynamic model is derived by the system identification method. This model is more realistic than the one based on mathematical theories. With this model, a robust controller is designed by the extended frequency response method. As this method has the same theoretical background as the classical method, all of the existing design techniques of the classical method can be used directly. Furthermore, by introducing the real part of a Laplacian operator into the frequency response, the control design specification can be considered at the initial stage of design. The designed controller is simple, and gives a sufficient robustness with good performance.

• Nuclear Engineering and Technology
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• v.46 no.3
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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.