• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.177-186
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• 2021

Radioactive materials from nuclear power facilities can be released into the atmosphere through various channels. Recently, the dispersion of radioactive materials has become critical issue in Korea after Kori Unit 1 and Wolsong Unit 1 were permanently shut down. In this study, annual atmospheric dispersion factors were compared based on the continuous release and purge release using the XOQDOQ computer program, a method for calculating atmospheric dispersion factors at commercial nuclear power stations. The meteorological data analyzed in this study was based on the Shin Kori nuclear power meteorological tower which has the largest operating nuclear power plants in Korea, for three years (from 2008 to 2010). The analysis results of the dispersion factor of the radioactive material release obtained using the XOQDOQ program showed that the difference between the continuous release and purge release was within two times. This study will be valuable helpful for revealing the uncertainty of the predictive atmospheric dispersion factor to achieve regulation.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• Journal of Radiation Protection and Research
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• v.35 no.4
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• pp.172-177
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• 2010

The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.321-338
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• 2022

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three-dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of ¹³⁷Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.188-188
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• 2004

In the present study, a multicriteria decision-making (MCDM) problem of ranking of important radionuclide release scenarios in a low and intermediate radioactive waste repository is to treat on the basis of non-additive fuzzy measures and fuzzy integral theory. Ranking of important scenarios can lead to the provision of more effective safety measure in a design stage of the repository. The ranking is determined by a relative degree of appropriateness of scenario alternatives.(omitted)

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.185-186
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• 2017

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.299-311
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• 2019

Decommissioning is a critical issue in Korea. Although compared with the operation of nuclear power plants the release of radioactive materials during decommissioning is not expected to be significant, residents should always be protected from radiation exposure. To manage this effectively, Annual Release Objectives (ARO) and Annual Release Limits (ARL) were derived from dose standards in the NSSC Notice and dose limit for the public. Based on meteorological data for the three years from 2008 to 2010 in the Shin Kori nuclear power plant site, atmospheric dispersion and ground deposition factors of gaseous effluent were evaluated using the XOQDOQ computer code. The exposure dose was evaluated using the ENDOS-G computer code. Because of differences in radiological sensitivity according to age groups, the results of Annual Release Objectives (ARO) and Annual Release Limits (ARL) showed significant differences depending on the radionuclides. The evaluation methodology of this study will provide meaningful information for radioactive effluent management for decommissioning of nuclear power plants.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2690-2701
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• 2022

In this paper, we select important fission products for the estimation of the source term during a severe accident of a PWR. The selection is based on the numerical results obtained from depletion calculations for the typical PWR fuel via the in-house code named DEGETION (Depletion, Generation, and Transmutation of Isotopes on Nuclear Application), release fractions of the fission products derived from NUREG-1465, and effective dose conversion coefficients from ICRP 119. Then, for the selected fission products, we obtain the adjoint solutions of the Bateman equations for radioactive decay in order to determine the importance of precursors producing the aforementioned fission products via radioactive decay, which would provide insights into the assumption used in MACCS 2 for a level 3 PSA analysis in which up to six precursors are considered in the calculations of radioactive decays for the fission product after release from the reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.833-838
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• 1995

This study presents a method to predict plant-specific and operational history dependent scaling factors. Realistic and detailed approaches are taken to find scaling factors at reactor coolant. This approach begins with fission product release mechanisms and fundamental release properties of fuel-source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the use of radionuclide retention and build up model. This model makes use of radioactive material balance within the radioactive waste processing systems. According to input parameters of plant operation history, scaling factors predicted at reactor coolant and waste streams are well brought out the effects of plant operation history.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.