• Membrane and Water Treatment
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• v.11 no.1
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• pp.79-85
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• 2020

A single-stage deammonification with a sequencing batch reactor (SBR) that simultaneous nitritation, anaerobic ammonia oxidation (anammox), and denitrification (SNAD) occur in one reactor has been widely applied for sidestream of wastewater treatment plant. For the stable and well-balanced SNAD, a feeding strategy of influent wastewater is one of the most important operating factors in the single-stage deammonification SBR. In this study, single-stage deammonification SBR (working volume 30L) was operated to treat a high-strength ammonium wastewater (1200 mg NH₄⁺-N/L) with different feeding strategies (single feeding and nine-step feeding) under the condition without COD. Each cycle of the step feeding involved 6 sub-cycles consisted of aerobic and anoxic periods for partial nitritation (PN) and anammox, respectively. Contrary to unstable performance in the single feeding, the step feeding showed better deammonification performance (0.565 kg-N/m³/day). Under the condition with COD, however, the nitrogen removal rate (NRR) decreased to 0.403 kg-N/m³/day when the Nine-step feeding strategies had an additional denitrification period before sub-cycles for PN and anammox. The NRR was recovered to 0.518 kg-N/m³/day by introducing an enhanced multiple-step feeding strategy. The strategy had 50 cycles consisted of feed, denitrification, PN, and anammox, instead of repeated sub-cycles for PN and anammox. The multiple-step feeding strategy without sub-cycle showed the most stable and excellent deammonification performance: high nitrogen removal efficiency (98.6%), COD removal rate (0.131 kg-COD/m³/day), and COD removal efficiency (78.8%). This seemed to be caused by that the elimination of the sub-cycles might reduce COD oxidation during aerobic condition but increase the COD utilization for denitrification period. In addition, among various sensor values, the ORP pattern appeared to be applicable to monitor and control each reaction step for deammonification in the multiple-step feeding strategy without sub-cycle. Further study to optimize the number of multiple-step feeding is still needed but these results show that the multiple-step feeding strategy can contribute to a well-balanced SNAD for deammonification when treating high-strength ammonium wastewater with COD in the single-stage deammonification SBR.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Journal of KSNVE
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• v.7 no.5
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• pp.837-845
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• 1997

The dynamic interactions between the reactor vessel and the control element drive mechanisms (CEDMs) of a pressurized water reactor are studied with the simplified mathematical model. The CEDMs are modeled as multiple substructures having different masses and the reactor vessel as a single degree of freedom system. The explicit equation for the frequency responses of the multiple substructure system are presented for the case of harmonic base excitations. The optimum dynamic characteristics of the CEDMs are presented to reduce the dynamic responses of the reactor vessel. The mathematical model and its response equations are verified by finite element analysis for the detailed model of the reactor vessel and the CEDMs for the harmonic base excitations. It is finally shown that the optimal dynamic characteristics of the CEDM presented can be applicable for the aseismic design.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.02a
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• pp.43-46
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• 2003

The dc reactor type high-Tc superconducting fault current limiter(SFCL) is composed of three parts, a power converter, a magnetic core reactor(MCR) and a dc reactor. This study concerned with the power converter of the DC reactor type high-Tc SFCL. The rectifying devices which power converter of 6.6kV/200A SFCL consists of have to endure high voltage. We propose the dual mode power converter to reduce the voltage which each rectifying device endures. In the single phase the experiment and simulation of dual mode power converter and the simulation of power converter with one bridge rectifier are performed. The current of each system with different power converter has a same tendency and the voltage which rectifying device of dual mode power converter endures is reduced in half by comparison with that of power converter with one bridge rectifier. We found dual mode power converter can be applied to SFCL.

• Environmental Engineering Research
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• v.10 no.3
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• pp.122-130
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• 2005

Gasification of carbonaceous wastes such as shredded tire, waste lubricating oil, plastics, and powdered coal initiates a single-stage reforming reactor(reformer) Without catalyst and a syngas burner. Syngas is combusted with $O_2$ gas in the syngas burner to produce $H_2O\;{and}\;CO_2$ gas with exothermic heat. Reaction products are introduced into the reforming reactor, reaction heat from syngas burner elevates the temperature of reactor above $1,200^{\circ}C$, and hydrogen gas fraction reaches 65% of the product gas output. Reactants and heat necessary for the reaction are provided through the syngas burner only. Neither $O_2$ gas nor steam is injected into the reforming reactor. Multiple syngas burners may be connected to the reforming reactor in order to increase the syngas output, and the product syngas is recycled into syngas burner.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• v.52 no.1
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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.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.85-91
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• 1986

A single comparator method with its accuracy has been studied for determining multielement by reactor neutron activation analysis. Spectral index at the irradiation position of each sample was determined using two flux monitors of Au and Co, one of which was used as a single comparator. The uncertainties of nuclear data related to the method were investigated for 18 elements and the error of the analytical result due to the uncertainties of nuclear data related is found to be less than 6%. The analytical results of 4 USGS reference samples agree well within 15% deviation with the results evaluated by USGS.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Journal of the Korean Society of Safety
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• v.34 no.3
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.