• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.299-309
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• 2003

In an accidental release of radioactive materials to the environment the contaminative influence of animal products due to the inhalation of air and the ingestion of soil of livestock, both of which are dealt with as minor contaminative pathways in most radioecological models but may not be neglected, was investigated with the improvement of the Korean dynamic food chain model DYNACON Although mathematical models for both contaminative pathways have been established for considering all animal products and incorporated into the model, investigation was limited to milk. As a result, it was found that both pathways are influential in the contamination of milk in the case of an accidental release during the non-grazing period of dairy cows. In the case of an accidental release during the non-grazing period, the inhalation of air was more influential than the ingestion of soil in the early days following an accidental release. While, it was the opposite with the lapse of time. If precipitation is encountered during an accidental release, contaminative influence due to the ingestion of soil was greater compared with the cases of no precipitation, in general, because of a stealer deposition of radionuclides onto the ground. Precipitation during an accidental release was a less influential factor in $^{131}I$ (elemental iodine) contamination compared with the $^{137}Cs\;and\;^{90}Sr$ contaminations. In the case of an accidental release during the grazing period of dairy cows, the contaminative influence due to the inhalation of air was negligible.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1830-1837
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• 2023

The thermal cutting of contaminated or activated metals during decommissioning nuclear power plants inevitably results in the release of radioactive aerosol. Since radioactive aerosols are pernicious particles that contribute to the internal dose of workers, air conditioning units with a HEPA filter are used to remove radioactive aerosols. However, a HEPA filter cannot be used permanently. This study evaluates the efficiency and lifetime of filters in actual metal cutting condition using a plasma arc cutter and a high-resolution aerosol detector. The number concentration and size distribution of aerosols from 6 nm to 10 ㎛ were measured on both the upstream and downstream sides of the filter. The total aerosol removal efficiency of HEPA filter satisfies the standard of removing at least 99.97% of 0.3 ㎛ airborne particles, even if the pressure drop increases due to dust feeding load. The pressure drop and particle size removal efficiency at 0.3 ㎛ of the HEPA filter were found to increase with repeated cutting experiments. By contrast, the efficiency of used HEPA filter reduced in removing nano-sized aerosols by up to 79.26%. Altogether, these results can be used to determine the performance guidance and replacement frequency of HEPA filters used in nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.247-253
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• 2023

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.793-794
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• 2004

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4715-4721
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• 2022

A comprehensive approach for modeling the pulse height spectra of gamma-ray detectors from passing radioactive cloud in a case of accident at NPP has been developed. It involves modeling the transport of radionuclides in the atmosphere using Lagrangian stochastic model, WRF meteorological processor with an ARW core and GFS data to obtain spatial distribution of radionuclides in the air at a given moment of time. Applying representation of the cloud as superposition of elementary sources of gamma radiation the pulse height spectra are calculated based on data on flux density from point isotropic sources and detector response function. The proposed approach allows us to obtain time-dependent spectra for any complex radionuclide composition of the release. The results of modeling the pulse height spectra of the scintillator detector NaI(Tl) Ø63×63 mm for a hypothetical severe accident at a NPP are presented.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.291-298
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• 2001

The designed release rate of liquid effluents from radwaste treatment system should be calculated and evaluated during normal operation, including anticipated operational occurrence and be assured that the release concentration and off-site dose at unrestricted area do not exceed the limits of regulation. The expected annual release rate and off-site dose for the currently operating nuclear power plants in Korea had been calculated and evaluated using PWR-GALE and LADTAP-II which was based on USNRC Regulatory Guide 1.109. Recently, the MOST Notice 2001-2 related to release concentration and off-site dose at unrestricted area was revised to reflect the concept of ICRP-60. It is necessary for KORI 3&4 to re-calculate the release concentration and off-site dose and to compare these results with the limits of regulation. As the results of assessment, we confirmed that the release concentrations were less than its limits of MOST Notice 2001-2 and the off-site dose at unrestricted area using K-DOSE60 was 3.61E-03 mSv/yr to the age of five for the effective dose, and 4.10E-2 mSv/yr to thyroid of the age of five for the organ equivalent dose. We also confirmed the off-site dose was within the limits of MOST Notice 2001-2. Therefore, the release concentration and off-site dose re-evaluated at unrestricted area in KORI 3&4 were well below the regulation limits of MOST Notice 2001-2.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.45-59
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• 2023

The transport calculation for a wide variety of radionuclides contained in high-level radioactive waste, especially spent nuclear fuel, is computationally difficult, and input data collection for this also take a considerable amount of time. Accordingly, considering limited resources, it is possible to reduce the calculation time while minimizing impact on accuracy by including only radionuclides important to calculation result through applying some criteria among potential radiation source terms that may release into environment. In this paper, therefore, we reviewed and analyzed the screening process performed to select radionuclides to be considered in the safety assessment for the KBS-3 type repository in Sweden and Finland. In both countries, it was confirmed that a list of radionuclides was selected by comprehensively considering screening criteria such as radioactivity inventory, half-life, radiotoxicity, risk quotient, and transport properties, and etc. A comparison of radionuclides included in the radiological safety assessment in both countries suggests that most of nuclides are considered in common, and a few nuclides considered only in one country are due to differences in decay chain treatment or spent fuel types. As of now, since most of information on the disposal facility in Korea has not been determined, it is necessary to comprehensively model release and transport of all radionuclides considered in Sweden and Finland when performing the radiological safety assessment. Based on these results, we derived the screening concept of selecting a list of radionuclides to be considered in the radiological safety assessment for the domestic KBS-3 type geological disposal facility, and this result is expected to be used as technical basis for confirming conformity with the safety objective. In a more detailed evaluation reflecting domestic characteristics in the future, it would be desirable to consider only radionuclides selected in accordance with the screening procedure. However, further research should be conducted to determine the quantitative limit for each criteria.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.263-272
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• 2012

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.