• Journal of Korean Society of Environmental Engineers
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• v.38 no.6
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• pp.323-328
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• 2016

Pressurized membrane used for side-stream MBR process requires fouling control strategy both for normal and abnormal operation conditions for stable operation of the facilities. In this study, $85m^3/day$ of pilot-scale side-stream MBR process was constructed for the evaluation of fouling mitigation by air bubble injection into the membrane module. In addition, fouling phenomena at abnormal operation conditions of low influent and/or loading rate were also investigated. Injection of air bubble was found to be effective in delaying transmembrane pressure (TMP) increase mainly due to scouring effect on the membrane surface, resulting in expanded filtration cycle at a high flux of $40L/m^2{\cdot}h$ (LMH). At abnormal operation condition, injection of PACl (53 mg/L as Al) into the bioreactor showed 19% reduction of TMP increase. However, inhibition of nitrifying bacteria by continuous PACl injection was observed from batch experiments. In contrast, injection of powdered activated carbon (PAC, 0.6 g/L) was able to maintain the initial TMP of $0.2kg/cm^2$ for 5 days at the abnormal conditions. It may have been caused from the adsorption of extracellular polymeric substances (EPS), which was known to be excessively released during growth inhibition condition and act as the major foulants in MBR operations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.59-73
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• 2019

A decontamination technology for radioactive liquid wastes was newly developed and hypothetically applied to the liquid waste management system (LWMS) of the nuclear power plant (NPP) to evaluate its decontamination efficacy for the purpose of the fundamental reduction of spent resins. The basic principle of the developed technology is to convert major radionuclide ions in the liquid wastes into inorganic crystal minerals via chemical or biological techniques. In a laboratory batch experiment, the biological method selectively removed more than 80% of cesium within 24 hours, and the chemical method removed more than 95% of cesium. Other major nuclides (Co, Ni, Fe, Cr, Mn, Eu), which are commonly present in nuclear radioactive liquid wastes, were effectively scavenged by more than 99%. We have designed a module including the new technology that could be hypothetically installed between the reverse osmosis (R/O) package and the organic ion-exchange resin in the LWMS of the APR1400 reactor. From a technical evaluation for the virtual installation, we found that more than 90% of major radionuclides in the radioactive liquid wastes were selectively removed, resulting in a large volume reduction of spent resins. This means that if the new technology is commercialized in the future, it could possibly provide drastic cost reduction and significant extension of the life of resins in the management of spent resins, consequently leading to delay the saturation time of the Wolsong repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.