• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.397-408
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• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• Nuclear Engineering and Technology
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• v.32 no.6
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.

• KSBB Journal
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• v.10 no.3
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• pp.335-342
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• 1995

Solid and liquid phase peroxidases were extracted from Chinese cabbage roots by using commercial juicer in order to use peroxidases from agricultural waste for industrial applications. Since peroxidases are distributed into 66% in liquid (juice) and 34% in solid phase (pulp), enzymes from both phases were applied to investigate the enzymatic removal of phenol from waste water. After contacting 150 ppm Phenol solution with liquid phase enzyme (1,800 unit/$\ell$) for 3 hours in a batch stirred reactor, 96% of phenol could be removed through polymerization and precipitation. Also, phenol could be removed from initial 120ppm to 5ppm by applying solid phase enzyme in an air lift reactor ($600 unit/\ell$). Almost equivalent efficiencies of phenol removal were observed between two systems, even though only one third of the enzymes in batch stirred reactor was applied in airlift reactor. The possible reason for this phenomenon is because peroxidases exist as immobilized forms in solid phase.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.922-929
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• 2019

This paper presents the drop performance test of the control rod assembly which is one of the main components strongly related to the safety of the prototype generation IV sodium-cooled fast reactor. To investigate the drop performance, a real-sized control rod assembly that was recently modified based on the drop analysis results was newly fabricated, and several free drop tests under different flow rate conditions were carried out. Then the results were compared with those obtained from the previous tests conducted on the conceptually designed control rod assembly to demonstrate the improvement in performance. Moreover, the drop performance tests under several types and magnitudes of seismic loadings were also conducted to investigate the effect of the seismic loading on the drop performance of the modified control rod assembly. The results showed that the effects of the type and magnitude of the seismic loading on the drop performance of the modified control rod assembly were not significant. Also, the drop time requirement was successfully satisfied, even under the seismic loading conditions.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Korean Chemical Engineering Research
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• v.59 no.3
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• pp.467-479
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• 2021

Abstract: A kinetic model was developed for the dehydration of ethanol to ethylene based on two parallel reaction pathways. Kinetic parameters were estimated by fitting experimental data of powder catalysts in a lab-scale test, and the effectiveness factor was determined using data from pellet-type catalysts in bench-scale experiments. The developed model was used to design a multitubular fixed-bed reactor (MTR) and an adiabatic reactor (AR) at a 10 ton per day scale. The two different reactor types resulted in different process configurations: the MTR consumed the ethanol completely and did not produce the reaction intermediate, diethyl ether (DEE), resulting in simple separation trains at the expense of high equipment cost for the reactor, whereas the AR required azeotropic distillation and cryogenic distillation to recycle the unreacted ethanol and to separate the undesired DEE, respectively. Quantitative analysis based on the equipment and annual energy costs showed that, despite high equipment cost of the reactor, the MTR process had the advantages of high productivity and simple separation trains, whereas the use of additional separation trains in the AR process increased both the total equipment cost and the annual energy cost per unit production rate.

• Journal of the Korean Society of Safety
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• v.37 no.2
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• pp.43-53
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• 2022

This paper describes an attempt to improve and optimize the operational safety level of a domestic research reactor by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety at an operating research reactor in Korea, to evaluate whether it is probabilistically safe and reliable to operate, and to obtain insights regarding the requisite procedural and design improvements for achieving safer operation. The technical objectives were to use the PSA to identify the accident sequences leading to core damage, and to conduct sensitivity analyses based thereon to derive insights regarding potential design and procedural improvements. Based on the dominant accident sequences identified by the PSA, eight types of sensitivity analysis were performed, and relevant insights for achieving safer operation were derived. When these insights were applied to the reactor design and operating procedure, the risk was found to be reduced by approximately ten times, and the safety was significantly improved. The results demonstrate that the PSA methodology is very effective for improving reactor safety in the full-power operating phase. In particular, it is a highly suitable approach for identifying the deficiencies of a reactor operating at full power, and for improving the reactor safety by overcoming those deficiencies.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4373-4391
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• 2022

Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate the hydrodynamic phenomena in the reactor under motion conditions; however, its applicability does not cover the MULTID component used in multidimensional flow analyses. In this study, a moving reactor model is implemented for the MULTID component to address the importance of multidimensional flow effects under dynamic motion. The concept of the volume connection is generalized to facilitate the handling of the junction of MULTID. Further, the accuracy in calculating the pressure head between volumes is enhanced to precisely evaluate the additional body force. Finally, the Coriolis force is modeled in the momentum equations in an acceleration form. The improvements are verified with conceptual problems; the modified model shows good agreement with the analytical solutions and the computational fluid dynamic (CFD) simulation results. Moreover, a simplified gravity-driven injection is simulated, and the model is validated against a ship flooding experiment. Throughout the verifications and validations, the model showed that the modification was well implemented to determine the capability of multidimensional flow analysis under ocean conditions.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.28 no.3
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• pp.265-270
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• 2004

Experiment study on a down scaled two-phase catalytic reactor is presented. As a preliminary step for the development of catalytic reactor, nano-particulate catalyst was prepared. Perovskite La$\_$0.8/Sr$\_$0.2/CoO$_3$is chosen and synthesized as a catalyst considering superior catalytic performance in reduction and oxidation process where oxygen is involved among the reagent. Reactor that has a scale of 2${\times}$10${\times}$25mm was made by machining of A1 block as a layered structure considering further extension to micro-machining. Hydrogen peroxide of 70wt% was adopted as reactant and was provided to the reactor loaded with 1.5 g of catalyst. Reactant flow rate was varied by precision pump with a range of 0.15cc/min to 17.2cc/min. Temperature distribution within reactor was recorded by 3 thermocouples and total amount of liquid product was measured. Temperature distribution and factors that affect temperature were observed and relation between temperature distribution and production rate was also analyzed. Relative time scale plays a significant role in the performance of the reactor. To obtain steady state operation, appropriate ratio of flow rate, catalyst mass and reactor geometry is required and furthermore to get more efficient production rate temperature distribution should be evenly distributed. The database obtained by the experiment will be used as a design parameter for micro reactor.

• Journal of Environmental Science International
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• v.18 no.3
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• pp.325-331
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• 2009

This study was aimed to behaviour of organics and nitrogen on the upflow anaerobic reactor when a acid fermenter is added. Up flow anaerobic reactor (UAR) reaction will result which operates, COD removal efficiencies of reactor with nitrate loading rate 0.11, 0.66g/L/d were over 77%, but one with 1.0g/L/d was 73.5%. Especially, on NLR 0.11g/L/d, COD removal was 77% and nitrate removal efficiency was 93% simultaneously. The other side upflow anaerobic reactor and acid fermenter (UAR+AF) reaction will result witch operates, COD removal efficiencies of reactor with nitrate loading rate 0.11, 0.66g/L/d were over 85%, but one with 1.0g/L/d was 80%. Especially, on NLR 0.11g/L/d, COD removal was 85% and nitrate removal efficiency was 98% simultaneously. Also, without in reaction condition increase of influent nitrate concentration resulted in the linear decrease of nitrate removal efficiency and nitrate removal efficiency at influent nitrate-nitrogen 800mg/L was 50%. Alkalinity was increased theoretically by denitrification at low nitrate-N concentration, however, it was not increased theoretically at high nitrate-No 40% nitrate-N of UAR was denitrified until 70% height of reactor and 90% nitrate-N of UAR+AF was denitrified until 30% height of reactor Upflow anaerobic reactor was to occur accumulate acid, which TVA/Alkalinity is 0.3$\sim$0.47. Increase of NLR resulted increase of effluent alkalinity and TVA production