• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.138-145
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• 2012

As of the end of 2010, 21 nuclear power reactors were operating in Korea. Radioactive effluents from nuclear power plants (NPPs) had been increased continuously and the radioactivity of effluents released in 2010 was 547.12 TBq. From 2001 to 2010, the annual average radioactivity of gaseous and liquid effluents per reactor was 11.61 TBq for pressurized water reactor (PWR) plants and 118.12 TBq for PHWR (pressurized heavy water reactor) plants. Most of the radioactivity from gaseous and liquid effluents was came from $^3H$. Based on the results of release trends and analysis, effluents characteristics was suggested for the management of radioactive effluents from NPPs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.345-352
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• 2006

The cemented and paraffin wastes form which are incorporated the concentrated wastes, the cemented waste form which is incorporated the spent ion-exchange resins, and the miscellaneous waste(decontamination paper) were irradiated up to $10^{+8}$ rads at $5.43{\times}10^{+5}$ rads/hr with Co-60(72,023.9 Ci) as an external irradiation source. As a result, the radiolysis gases such as $H_2,\;CH_4,\;N_2,\;C_2H_6,\;O_2,\;CO\;and\;CO_2$, were measured in all the wastes. The major gas which was generated in all the wastes was hydrogen($H_2$). The volume of the generated gases showed a difference from $0.029{\sim}0.788\;cm^3.atm/1.1g$ according to the type of wastes, and more was generated in the cemented waste form incorporated a spent ion-exchange resin than in the other wastes. More hydrogen($H_2$) gas was generated in the decontamination paper waste than in the other wastes, and the G($H_2$) value was 0.12.

• The Journal of Engineering Geology
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• v.24 no.4
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• pp.535-550
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• 2014

The origins and varieties of natural radioactive materials, including uranium and radon-222, were examined in a drilled borehole extending to a depth of 120 m below the surface in the Goesan area. In addition to core samples, eight groundwater samples were collected at different depths, using a double packer system and bailer, and their geochemical characteristics were determined. Most of the rock samples from the drilled core consisted of granite porphyry, with sedimentary rocks (slate, carbonate, or lime-silicates) and pegmatite occurring in certain sections. The pH of samples varied from 7.8 to 8.4, and the groundwater was of a Na-$HCO_3$type. Uranium and thorium concentrations in the core were < 0.2-14.8 ppm and 0.56-45.0 ppm, respectively. Observations by microscope and an electron probe microanalyzer (EPMA) showed that the mineral containing the natural radioactive materials was monazite contained in biotite crystals. The uranium, which substituted for major elements in the monazite, appeared to have dissolved and been released into the groundwater in a shear zone. Concentrations of Radon-222 in the borehole showed no close relationship with levels of uranium. The isotopes of noble gases, such as helium and neon, would be useful for analyzing the origins and characteristics of the natural radioactive materials.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.315-319
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• 2001

It is shown, that both cosmic radiation (external source) and natural radioactive gases (inner source) are sources of neutrons near the Earth crust. Correlation between the Earth crust dynamics and variations of thermal and slow neutron flux near the Earth surface is studied. It is shown, that variations of neutron flux near the Earth crust can be used for short-term predicting of natural hazards.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3310-3316
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• 2022

Owing to the rising concerns about the safety of nuclear power plants (NPP), it is essential to study the venturi scrubber in detail, which is a key component of the filtered containment venting system (FCVS). FCVS alleviates the pressurein containment byfiltering and venting out the contaminated air. Themain objective of this research was to perform a CFD investigation of different configurations of a circular, non-submerged, self-priming venturi scrubber to estimate and improve the performance in the removal of elemental iodine from the air. For benchmarking, a mass transfer model which is based on two-film theory was selected and validated by experimental data where an alkaline solution was considered as the scrubbing solution. This mass transfer model was modified and implemented on a unique formation of two self-priming venturi scrubbers in series. Euler-Euler method was used for two-phase modeling and the realizable K-ε model was used for capturing the turbulence. The obtained results showed a remarkable improvement in the removal of radioactive iodine from the air using a series combination of venturi scrubbers. The removal efficiency was improved at every single data point.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.39-52
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• 2007

Quantitative analysis on release behavior of the $^{85}Kr\;and\;^{14}C$ fission gases from the spent fuel material during the voloxidation and OREOX process has been performed. This thermal treatment step in a remote fabrication process to fabricate the dry-processed fuel from spent fuel has been used to obtain a fine powder The fractional release percent of fission gases from spent fuel materials with burn-up ranges from 27,000 MWd/tU to 65,000 MWd/tU have been evaluated by comparing the measured data with these initial inventories calculated by ORIGEN code. The release characteristics of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation process at $500^{\circ}C$ seem to be closely linked to the degree of conversion efficiency of $UO_2\;to\;U_3O_8$ powder, and it is thus interpreted that the release from grain-boundary would be dominated during this step. The high release fraction of the fission gas from an oxidized powder during the OREOX process would be due to increase both in the gas diffusion at a temperature of $500^{\circ}C$ in a reduction step and in U atom mobility by the reduction. Therefore, it is believed that the fission gases release inventories in the OREOX step come from the inter-grain and inter-grain on $UO_2$ matrix. It is shown that the release fraction of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation step would be increased as fuel burn-up increases, ranging from 6 to 12%, and a residual fission gas would completely be removed during the OREOX step. It seems that more effective treatment conditions for a removal of volatile fission gas are of powder formation by the oxidation in advance than the reduction of spent fuel at the higher temperature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.169-177
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• 2020

Here, the stability of stainless steel 316 (SS-316) was investigated to identify its applicability in the oxide reduction process, as a component in related equipment, to produce a complicated gas mixture composed of O₂ and Cl₂ under an argon (Ar) atmosphere. The effects of the mixed gas composition were investigated at flow rates of 30 mL/min O₂, 20 mL/min O₂ + 10 mL/min Cl₂, 10 mL/min O₂ + 20 mL/min Cl₂, and 30 mL/min Cl₂, each at 600℃, during a constant argon flow rate of 170 mL/min. It was found that the corrosion of SS-316 by the chlorine gas was suppressed by the presence of oxygen, while the reaction proceeded linearly with the reaction time regardless of gas composition. Surface observation results revealed an uneven surface with circular pits in the samples that were fed mixed gases. Thermodynamic calculations proposed the combination of Fe and Ni chlorination reactions as an explanation for this pit formation phenomenon. An exponential increase in the corrosion rate was observed with an increase in the reaction temperature in a range of 300 ~ 600℃ under a flow of 30 mL/min Cl₂ + 170 mL/min Ar.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1176-1179
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• 2019

Measurements of activity concentrations of radionuclides in atmospheric releases were performed in 2017-2018 at vent stacks of seven Russian nuclear power plants. The selected instruments and research methods, with detection limits significantly lower than the existing detection limit of Russian NPPs routine control, allowed to reliably determine up to 26 radionuclides. Analysis of experimental data allows to determine the list of radionuclides for calculation the effective dose rates to public and the permissible annual discharge levels for each Russian NPP. Radiocarbon is determined as major contributor for the dose from the atmospheric releases of LWGR reactors - up to 98% for EGP-6 and RBMK-1000 (Smolensk NPP) reactors. For PWR reactors (VVER) radionuclides contribution to the annual dose from atmospheric releases is more complicated, but, in general, dose is formed by tritium, $^{14}C$ and noble gases. The special monitoring results with ranking of measured radionuclides according to their contribution to the effective dose makes it possible to optimize the list of controlled radionuclides in airborne releases of Russian NPPs from 94 to 8-16 for different NPPs.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.386-392
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• 2021

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.