• Environmental Engineering Research
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• v.23 no.2
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• pp.195-204
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• 2018

A pilot scale study was conducted on a down-flow hanging sponge (DHS) reactor installed at a sewage treatment plant in Banda Aceh, Indonesia for treatment of desludging septic tank wastewater. Raw wastewater with an average biochemical oxygen demand (BOD) and total suspended solids of 139 mg/L and 191 mg/L, respectively, was pumped into the reactor. Two different hydraulic retention times (HRTs, 3 h and 4 h) were investigated, equivalent to organic loadings of 1.11 and $0.78kg\;BOD/m^3/d$, respectively. The average BOD concentration in the final effluent was 46 and 26 mg/L at HRTs of 3 and 4 h, respectively. The concentration of retained sludge along the reactor height was 10.2-18.7 g VSS/L-sponge, and the sludge activities were 0.24-0.32 and 0.04-0.40 mg/g VSS/h for heterotrophs and nitrification, respectively. Values of water hold-up volume, dispersion coefficient, and number of tank in-series found from tracer studies of clean sponge and biomass-loaded sponge confirmed that growth of retained sludge on the sponge module improved hydraulic performance of the reactor. Adoption of the DHS reactor by this Indonesian sewage treatment plant would enhance the role of the current desludging septic tank wastewater treatment system.

• 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.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• Resources Recycling
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• v.13 no.1
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• pp.59-67
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• 2004

A universal plasma-chemical module (PChM) for the industrial processing of different hydrocarbon raw material pyrolysis was designed and tested. Laboratory investigations for the plasma-chemical method of acetylene production from natural gas and different coals were made. Similar laboratory tests on the industrial production of acetylene as a raw material for organic syn-thesis were developed using the PChM. A comparison of the suggested plasma-chemical method with the traditional process of acetylene production were carried out. The outlook of the plasma-chemical method was shown.

• Journal of Environmental Science International
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• v.15 no.10
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• pp.945-949
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• 2006

For a membrane bio-reactor, it is possible to fillet and separate activated sludge and effluent by head loss of centimeters, if non-woven fabric material is used as titration media. However, if non-woven fabric material is used to thicken high-concentration sludge, excessive sludge attachment causes the rapid decrease of flux. Mesh with fore sizes of $100{\mu}m,\;150{\mu}m,\;and\;200{\mu}m$ allows for easy separation of attached sludge. This study examined the possibility of mesh as filtration media. Existing close-flow filtration process, which requires maintaining sludge movement, makes It difficult to obtain high thickening rate. With a view of complementing this weakness, this study has made an experimental examination on how high-concentration sludge (about 3,000mg/L to 10,000mg/L) will be filtered and thickened when mesh module is submersed in the bio-reactor. Effluent flowed from the bottom of the bio-reactor by head loss of 65cm. In case of pore size of $100{\mu}m$, SS showed high recovery of 80% to 96%; therefore, it has been decided that mesh can be used as filtration media. Filtration lasted for more than 9 hours, until sludge with 9,000mg/L in MLSS concentration was thickened 9 times as dense. In the range from 3,610mg/L to 9,060mg/L in MLSS concentration, it was possible to obtain effluent with less than 2mg/L in MLSS concentration within 10 minutes.

• Membrane Journal
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• v.8 no.3
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• pp.170-176
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• 1998

A study on the bioconversion of soluble starch to cyclodextrin(CD) homologue by CGTase was performed in the membrane-enzyme reactor equipped with a dead-end type membrane module. in the batch reactor, the total conversion of soluble starch to CD homologue was decreased rapidly from a maximum value of 45 % with increasing reaction time due to the product inhibition and breakdown of CD homologue to the reducing sugars. However, in the membrane-enzyme reactor, the total conversion of soluble starch was maintained at a constant value of 35 % throughout the reaction, since the membrane(MWCO = 10,000) promptly separated CD homologue from the reaction mixture. After the macdon for 24 hr in the membrane-enzyme reactor using a 10 % soluble starch solution, the cumulative production amount of CD homologue was about 3.7 kg/m$^2$ at the operating pressure of 2 atm.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.175-183
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• 2004

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.456-457
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• 2005

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.138-151
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• 2022

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.203-205
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• 2005

A reactor protection system (RPS) plays the roles of generating the reactor trip signal and the engineered safety features (ESF) actuation signal when the monitored plant processes reach predefined limits. A Korean project group is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS which is used for the RPS integrity testing and monitoring by an equipment operator. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS operation. To support the RPS COM user interface design, actually the FPD screen design, we developed a user interface style guide because the system designer could not properly deal with the many general human factors design guidelines. To develop the user interface style guide, various design guideline gatherings, a walk-though with a video recorder, guideline selection with respect to user interface design elements, determination of the properties of the design elements, discussion with the system designers, and a conversion of the properties into a screen design were carried out. This paper describes the process in detail and the findings in the course of the style guide development.