• 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 Radiation Industry
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• v.17 no.1
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• pp.33-44
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• 2023

The deep geological repository of high-level radioactive waste shall be designed to meet the safety objective set in the form of radiation dose or corresponding risk to protect human and the environment from radiation exposure. Engineering feasibility and conformity with the safety objective of the facility conceptual design can be demonstrated by comparing the assessment result using the computational model for scenario(s) describing the radionuclide release and transport from repository to biosphere system. In this study, as the preliminary study for developing the high-level radioactive waste disposal facility in Korea, we reviewed and analyzed the entire list of FEPs and how to handle each FEP from a general point of view, which are selected for the geosphere region in the radiological safety assessment performed for the license application of the KBS-3 type deep geological repository in Finland and Sweden. In Finland, five FEPs (i.e., stress redistribution, creep, stress redistribution, erosion and sedimentation in fractures, methane hydrate formation, and salt exclusion) were excluded or ignored in the radionuclide release and transport assessment. And, in Sweden, six FEPs (i.e., creep, surface weathering and erosion, erosion/sedimentation in fractures, methane hydrate formation, radiation effects (rock and grout), and earth current) were not considered for all time frames and earthquake out of a total of 25 FEPs for the geosphere. Based on these results, an FEP list (draft) for the geosphere was derived, and the relative importance of each item was evaluated for conducting the radiological safety assessment of the domestic deep geological disposal facility. Since most of information on the disposal facility in Korea has not been determined as of now, it is judged that all FEP items presented in Table 3 should be considered for the radiological safety assessment, and the relative importance derived from this study can be used in determining whether to apply each item in the future.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.457-463
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• 2005

KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.9-27
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• 2021

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

• 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.

• Proceedings of the KSEEG Conference
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• 2002.06a
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• pp.81-100
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• 2002

While most of regulatory communities in abroad recognize ' 'natural attenuation " to include degradation, dispersion, dilution, sorption (including precipitation and transformation), and volatilization as governing Processes, regulators prefer "degradation" because this mechanism destroys the contaminant of concern. Unfortunately, true degradation only applies to organic contaminants and short- lived radionuclides, and leaves most metals and long-lived radionuclides. The natural attenuation Processes may reduce the potential risk Posed by site contaminants in three ways: (i)contaminants could be converted to a less toxic form througy destructive processes such as biodegradation or abiotic transformations; (ii) potential exposure levels may be reduced by lowering concentrations (dilution and dispersion); and (iii) contaminant mobility and bioavailability may be reduced by sorption to geomedia. In this review, authors will focus will focul on "sorption" among the natural attenuation processes of hazardous inorganic contaminants including radionuclides. Note though that sorption and transformation processes of inorganic contaminants in the natural setting could be influenced by biotic activities but our discussion would limit only to geochemical reactions involved in the natural attenuation. All of the geochemical reactions have been studied in-depth by numerous researchers for many years to understand "retardation" process of contaminants in the geomedia. The most common approach for estimating retardation is the determination of distrubution coefficiendts ($K_{d}$) of contaminants using parametric or mechanistic models. As typocally used in fate and contaminant transport calculations such as predictive models of the natural attenuation, the $K_{d}$ is defined as the ratio of the contaminant concentration in the surrounding aqueous solution when the system is at equilibrium. Unfortunately, generic or default $K_{d}$ values can result in significant error when used to predict contaminant migration rate and to select a site remediation alternative. Thus, to input the best $K_{d}$ value in the contaminant transport model, it is essential that important geochemical processes affecting the transport should be identified and understood. Precipitation/dissolution and adsorption/desorption are considered the most important geochemical processes affecting the interaction of inorganic and radionuclide contaminants with geomedia at the near and far field, respectively. Most of contaminants to be discussed in this presentation are relatively immobile, i.e., have very high $K_{d}$ values under natural geochemical environments. Unfortunately, the obvious containment in a source area may not be good enough to qualify as monitored natural attenuation site unless owner demonstrate the efficacy if institutional controls that were put in place to protect potential receptors. In this view, natural attenuation as a remedial alternative for some of sites contaminated by hazardous-inorganic components is regulatory and public acceptance issues rather than scientific issue.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.281-291
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• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

• Tunnel and Underground Space
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• v.11 no.2
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• pp.132-145
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• 2001

The purpose and need of the study is to quantify the advantage or disadvantage of the environmental friendliness of the partitioning of nuclear fuel cycle. To this end, a preliminary study on the quantitative effect of the partition on the permanent disposal of spent PWR and CANDU fuel (HLW) was carried out. Before any analysis, the so-called reference radionuclide release scenario from a potential repository embedded into a crystalline rock was developed. Firstly, the feature, event and processes (FEPs) which lead to the release of nuclides from waste disposed of in a repository and the transport to and through the biosphere were identified. Based on the selected FEPs, the ‘Well Scenario’which might be the worst case scenario was set up. For the given scenario, annual individual doses to a local resident exposed to radioactive hazard were estimated and compared to that from direct disposal. Even though partitioning and transmutation could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and also minimize the repository area through the proper handling of nuclides, it should overcome major disadvantages such as echnical issues on the partitioning and transmutation system, cost, and public acceptance, and environment friendly issues. In this regard, some relevant issues are also discussed to show the direction for further studies.

• Tunnel and Underground Space
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• v.18 no.5
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• pp.328-337
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• 2008

Quantifying the extent and characteristics of the excavation damage zone(EDZ) is important for the nuclear waste industry which relies on the sealing of underground openings to minimize the risk for radionuclide transport. At AECL's Underground Research Laboratory(URL) the Tunnel Sealing Experiment(TSX) was conducted and the tunnel geometry and orientation relative to the stress field had been selected to minimize the potential for the development of an EDZ. The extent and characteristics of the EDZ was measured using velocity profiling and permeability measurements in radial boreholes. The results from this EDZ characterization are used in this paper to evaluate a modeling fir estimating the extent of the EDZ. The methodology used a damage model formulated in the Universal Distinct Element Code and calibrated to laboratory properties. This model was then used to predict the extent of crack initiation and growth around the TSX tunnel and the results compared to the measured damage. The development of the damage zone in the numerical model was found to be in good agreement with the field measurements.