• Biotechnology and Bioprocess Engineering:BBE
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• v.11 no.3
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• pp.199-204
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• 2006

The purpose of this work was to evaluate the development of the anammox process by the use of granular sludge selected from a digestion reactor as a potential seed source in a lab-scale UASB (upflow anaerobic sludge blanket) reactor system. The reactor was operated for approximately 11 months and was fed by synthetic wastewater. After 200 days of feeding with $NH_4^+\;and\;NO_2^-$ as the main substrates, the biomass showed steady signs of ammonium consumption, resulting in over 60% of ammonium nitrogen removal. This report aims to present the results and to more closely examine what occurs after the onset of anammox activity, while the previous work described the start-up experiment and the presence of anammox bacteria in the enriched community using the fluorescence in situ hybridization (FISH) technique. By the last month of operation, the consumed $NO_2^--N/NH_4^+-N$ ratio in the UASB reactor was close to 1.32, the stoichiometric ratio of the anammox reaction. The obtained results from the influent-shutdown test suggested that nitrite concentration would be one key parameter that promotes the anammox reaction during the start-up enrichment of anammox bacteria from granular sludge. During the study period, the sludge color gradually changed from black to red-brownish.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.104-106
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• 2005

In this work, a PWR reactor core dynamics is identified online by a recursive least squares method. Based on this identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to design an automatic controller for thermal power control in PWRs. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired one, and the variation of the control rod positions. Also, the objectives are subject to maximum and minimum control rod positions and maximum control rod speed. Therefore, the genetic algorithm that is appropriate to accomplish multiple objectives is used to optimize the model predictive controller. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), is used to verify the proposed controller for a nuclear reactor. From results of numerical simulation to check the performance of the proposed controller at the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are design requirements, it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Applied Chemistry for Engineering
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• v.23 no.2
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• pp.204-209
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• 2012

In this study, kinetics data was obtained for steam reforming reaction of ethane over the commercial ruthenium catalyst. The variables of ethane steam reforming were the reaction temperature, partial pressure of ethane, and steam/ethane mole ratio. Parameters for the power rate law kinetic model and the Langmuir-Hinshelwood model were obtained from the kinetic data. Also, sizing of steam reforming reactor was performed by using PRO/II simulator. The reactor size calculated by the power rate law kinetic model was bigger than that of using the Langmuir-Hinshelwood model for the same conversion of ethane. Reactor size calculated by the Langmuir-Hinshelwood model seems to be more suitable for the reactor design because the Langmuir-Hinshelwood model was more consistent with the experimental results.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.19-24
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• 2004

A number of creep data were collected and filed for type 316LN stainless steels through literature survey and experimental data produced in KAERI. Using these data, polynomial equations for predicting creep life were obtained for Larson Miller (L-M), Qrr-Sherby-Dorn (O-S-D) and Manson-Haferd (M-H) parametric methods. In order to find out the suitability for them, the relative standard error (RSE) and standard error of estimate (SEE) values were obtained by statistical process of creep data. The O-S-D parameter showed better fitting to creep-rupture data than the L-M or the M-H parameters, and the three parametric methods did not generate the large difference in the SEE and the RSE values.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.247-255
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• 1985

This paper describes a new method for the design of variable structure model-following control systems(VSMFC). This design concept is developed using the theory of variable structure systems (VSS) and slide mode. The new results are presented on the sliding control methodology to achieve accurate tracking for a class of nonlinear, multi-input multi-output(MIMO), time varying systems in the presence of parameter variations. The design requires little computational effort. The dynamic response is insensitive to parameter variations. The feasibility and the advantages of the method are illustrated by applying it to a 1000 MWe boiling water reactor(BWR). The control is studied in the range of 85%∼90% of rated power for load-following control. A set of 12 nonlinear differential equations is used to simulate the total plant. A 6-th order linear model has been developed from these equations at 85% of rated power. The obtained controller is shown by simulations to be able to compensate for a plant parameter variation over a wide power range.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.81-92
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• 2006

In this study, the core dynamics of a PWR reactor is identified online by a recursive least-squares method. Based on the identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to designing an automatic controller for the thermal power control of PWR reactors. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, this procedure for solving the optimization problem is repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired temperature, as well as minimizing the variation of the control rod positions. In addition, the objectives are subject to the maximum and minimum control rod positions as well as the maximum control rod speed. Therefore, a genetic algorithm that is appropriate for the accomplishment of multiple objectives is utilized in order to optimize the model predictive controller. A three-dimensional nuclear reactor analysis code, MASTER that was developed by the Korea Atomic Energy Research Institute (KAERI) , is used to verify the proposed controller for a nuclear reactor. From the results of a numerical simulation that was carried out in order to verify the performance of the proposed controller with a $5\%/min$ ramp increase or decrease of a desired load and a $10\%$ step increase or decrease (which were design requirements), it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.585-590
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• 2006

This paper describes the results of an irradiation test and the specifications of the pneumatic transfer system (PTS) in the NAA #3 irradiation hole at the HANARO research reactor, which was reinstalled after some modifications of the operation mode at the end of 2004. The outer and inner diameters of the PE transfer tube are 34.1 and 27.5 mm, respectively. PE rabbit was used for sample irradiation. The $N_2$ gas pressure of the PTS lines was adjusted to 0.75 bar. The average sending time to the reactor was $8.5{\pm}0.3$ s and the average receiving time back to the receiver was $3.2{\pm}0.2$ s. The internal and external temperature of the irradiation tube was measured in a range of 50 to $80^{\circ}C$ for a 40 s to 80 s irradiation time, respectively. The optimum irradiation time was estimated to be less than 80 s. The thermal, epithermal and fast neutron flux at 30 MW thermal power were $1.42{\pm}0.01{\times}10^{14},\;1.51{\pm}0.04{\times}10^{13}$ and $9.48{\pm}0.69{\times}10^{11} n{\cdot}cm^{-2}{\codt}s^{1-}$, respectively. The cadmium ratio was approximately 9.40. The data obtained will be applied to supplement user information and for reactor management.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1610-1616
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• 2017

To validate the new Evaluated Nuclear Data File $(ENDF)/B-VIII.0{\beta}4$ library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the $ENDF/B-VIII.0{\beta}4$ library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

• Journal of Korean Society of Water and Wastewater
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• v.12 no.4
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• pp.86-96
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• 1998

In this study, an intermittently aerated activated sludge process, modified process from conventional activated sludge process, was developed to treat high strength swine wastewater, which has been blamed as major pollutant for stream pollution. Therefore, the optimum cycle for oxic and anoxic period, SRT, and OLR were studied as design parameters. The effects of different time interval for oxic and anoxic period on nitrification and denitrification were examined by operating two reactors with 60/60min and 60/90min as oxic/anoxic period. Although the reactor with 60/60min showed complete denitrification of $NO_x-N$ generated during oxic period, the reactor with 60/90min showed incomplete nitrification due to the inactivity of nitrifier by accumulated $NH_3-N$ toxicity during anoxic period. Therefore, it is recommended to operate same interval for oxic and anoxic period. In order to determine the optimum cycle for oxic/anoxic period, four different reactors with 30/30, 60/60, 90/90 and 120/120min were examined. The reactor operation with 90/90min was optimum to get the most stable results in this study. However, the optimum cycle for oxic and anoxic period should be changed with characteristics of influent wastewater and operating conditions. According to lie operation results of three reactors with SRT of 15, 20 and 30days. The reactor with 2Odays SRT showed best removal efficiency of T-N. The optimum OLR would be $2.5Kg\;COD/m^3/day$ which showed the most stable nitrification and denitrification. Since characteristics of influent wastewater in the real system has a severe fluctuation, so it is very difficult to determine each interval for oxic and anoxic period. Therefore, ORP curves, describing the change of oxidation/reduction potential in reactor, can be used as a control parameter for automatic control of oxic and anoxic period. In other words, bending point (Nitrate Knee) of ORP curve during anoxic period could be used as a starting point of oxic period.