• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.673-684
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• 2024

STREAM, developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST), is a deterministic neutron- and photon-transport code primarily designed for light water reactor (LWR) analysis. Initially, the photon module in STREAM did not account for fluorescence and bremsstrahlung photons. This article presents recent developments regarding the integration of atomic relaxation and bremsstrahlung models into the existing photon module, thus allowing for the transport of secondary photons. The photon flux and photon heating computed with the newly incorporated models is compared to results obtained with the Monte Carlo code MCS. The incorporation of secondary photons has substantially improved the accuracy of photon flux calculations, particularly in scenarios involving strong gamma emitters. However, it is essential to note that despite the consideration of secondary photon sources, there is no noticeable improvement in the photon heating for LWR problems when compared to the photon heating obtained with the previous version of STREAM.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3324-3335
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• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.286-298
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• 2003

A steam explosion module, STX, has been developed using the mechanistic steam explosion analysis code, TEXAS-V, in order to estimate the dynamic load with steam explosion by implementing the module to the integrated safety analysis code, MELCOR. One of the difficulties in using mechanistic steam explosion codes is that they do not have any obvious criteria for defining some uncertain parameters such as triggering timing, triggering magnitude, mesh axial length and mesh cross-sectional area. These parameters have been user decision parts in the past. Steam explosion sample calculations and sensitivity studies on uncertain parameters were conducted to investigate those uncertain parameters. The TEXAS-V simulations were summarized in the format of a look-up table and a linear interpolation technique was adopted to calculate the steam explosion load between the data points in the table. The STX-module merged with MELCOR showed the same results as the original MELCOR and additionally it could estimate the steam explosion load in the reactor cavity.

• Nuclear industry
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• v.7 no.12 s.58
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• pp.78-83
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• 1987

GE nuclear Energy는 미국에너지성(DOE)의 개량형액체금속원자로계획의 일환으로 표준화된 멀티플블록 발전소의 기본발전유니트인 PRISM(Power Reactor, Inherently Safe Module)을 개발하고 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1065-1066
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• 2005

• Proceedings of the KIEE Conference
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• 2015.07a
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• pp.1114-1115
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• 2015

This paper describes the design of toroid-type HTS DC reactor magnet. Target operating current and inductance of the HTS DC reactor are 1,500 A and 400 mH, respectively. The HTS DC reactors were designed through electromagnetic analysis and 3D CAD program. And, we analyze the operating performance of the Double Pancake Coil module for the 1,500 A, 400 mH HTS DC reactor magnet under the liquid nitrogen condition.

• Proceedings of the KIEE Conference
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• 2004.05a
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• pp.29-32
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• 2004

Reliability is an important issue on the digital reactor protection system. This paper presents a Quantitative reliability evaluation method to find out an improvement effect of availability for the digital control module with a fault detection function. It is a reliability evaluation model which considers only the electronics parts ocurring a spurious reactor trip by the FMEA(Failure Mode Effect Analysis). Applying the previous and present methods to the reactor protection system, the availability factors are evaluated and compared.

• Proceedings of the Membrane Society of Korea Conference
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• 2003.07a
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• pp.7-10
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• 2003

Factors affecting filtration performance were investigated in a Sequencing Batch Reactor (SBR) coupled with a submerged microfiltration module. Special bioreactors for aerobic and anoxic phases, respectively, were specifically designed in order to differentiate tile effect of Dissolved oxygen (DO) from that of mixing intensity on membrane filterability. DO concentration as well as mixing intensity proved to have a major influence on the membrane performance regardless of the SBR phase. A higher DO concentration resulted in a slower rise in TMP, corresponding to less membrane fouling.

• 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.