• Journal of the Korean Applied Science and Technology
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• v.23 no.2
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• pp.130-136
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• 2006

Theoretical total nitrogen removal efficiency and reactor volume ratio in oxic-anoxic-oxic system can be found by influent water quality in this study. The influent water quality items for calculation were ammonia, nitrite, nitrate, alkalinity, and COD which can affect nitrification and denitrification reaction. Total nitrogen removal efficiency depends on influent allocation ratio. The total nitrogen removal follows the equation of 1/(1+b). Optimal reactor volume ratio for maximum TN removal efficiency was expressed by those influent water quality and nitrification/denitrification rate constants. It was possible to expect optimal reactor volume ratio by the calculation with the standard deviation of ${\pm}14.2$.

• Journal of the Korean Institute of Intelligent Systems
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• v.5 no.3
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• pp.52-63
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• 1995

A neuro-fuzzy algorithm is presented for nuclear reactor power control in a pressurized water reactor. Automatic reacotr power control is complicated by the use of control rods because of highly nonlinear dynamics in the axial power shape. Thus, manual shaped controls are usually employed even for the limited capability during the power maneuvers. In an attempt to achieve automatic shape control, a neuro-fuzzy approach is considered because fuzzy algorithms are good at various aspects of operator's knowledge representation while neural networks are efficinet structures capable of learning from experience and adaptation to a changing nuclear core state. In the proposed neuro-fuzzy control scheme, the rule base is formulated based ona multi-input multi-output system and the dynamic back-propagation is used for learning. The neuro-fuzzy powere control algorithm has been tested using simulation fesponses of a Korean standard pressurized water reactor. The results illustrate that the proposed control algorithm would be a parctical strategy for automatic nuclear reactor power control.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1338-1348
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• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Magazine of the Korea Concrete Institute
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• v.9 no.4
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• pp.155-165
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• 1997

Since the biggest time-dependent prestress loss of reactor containment structure is due to creep of concrete. the creep is one of important structural factors to be considered for the safety maintenance in the containment structure during design. construction and main enance. This paper is about the creep charactoristies of concrete for the reactor containment structure. In this paper, creep test was performed to show the creep characteristics of reactor containment concrete structure made of the type-V cement. Then, in order to evaluate the applicability of creep prediction equations of recently revised Korean Concrete Standard Specification(KSCE-96) and Japanes Concrete Standard Specification. ACI-209. CEB/FIP-90. and HANSEN, creep test results were compared with prediction results obtained from he equations. From the comparisons, it was shown that the equation of th KSCE-96 predicts creep for younger concrete than 1 year, better than the other equations and that all of the equations predicts creep, for older concrete than 1 year, smaller than test. From regression analysis. a creep prediction equation which effectively predicts creep of concrete due to loading after 1year was proposed.

• Journal of the Korean Society of Industry Convergence
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• v.25 no.2_1
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• pp.149-159
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• 2022

Biomass is material that is comprehensive of carbonaceous materials from plants, crops, animals, and algae. It has been used as one of heating fuel since the beginning the emergence of human beings. Since biomass is regarded as carbon-neutral energy source, it has recently been attracting attention as an energy source that can replace fossil fuels. The most widely applied field is distributed power generation, and a method of generating electric power by driving an internal combustion engine with syngas produced by gasifier is chosen. While the composition of the syngas produced in gasifiers changes depending on the air flowing into the reactor, commercialized gasifiers so far do not control the air flowing into the reactor. When the inner pressure in reactor increases, the air sucked into the reactor is reduced. That change of amount of air makes the composition of syngas varied. Those variations of composition of syngas cause the incomplete combustion hence the power output of engine drops, which is a critical weakness of the gasification technology. In this paper, to produce the uniformly composed syngas, PID control is applied. The result was shown when the amount of air into the reactor is supplied with the constant amount using PID control, the standard deviation of caloric values of syngas is around 2[%] of its average value. Meanwhile the gasifier without PID control has the standard deviation of caloric values is around 7[%]. Therefore, Adopting PID control to supply constant air to the gasifier is highly desirable.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.13 no.7
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• pp.637-646
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• 2001

VOC(Volatile Organic Compound) removal characteristics in continuous flow reactors have been numerically investigated. The photocatalytic reaction have been simulated with the binding constant and the reaction rate constant obtained from experimental data for the constant-volume batch reactor, and then VOC abatement in continuous flow reactors with the same conditions as those of batch reactor has been analyzed. The standard 4\kappa-\varepsilon$ model and mass conservation equation have been employed for numerical calculation, and heterogeneous reaction rate has been used in terms of the boundary condition of the conservation equation. in the case of the continuous flow reactor, reaction characteristics have been estimated with various inlet velocities and with different number of baffles. The result shows that the concentration distribution and flow patterns are strongly affected by the inlet velocity, and that with the increased inlet velocity, VOC removal rate is increased, while removal efficiency is decreased. This result may be useful in the design of reactors with improved VOC removal efficiency.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• v.55 no.9
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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.