• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.129-139
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• 2008

In this study, a sorption experiment of multivalent nuclides such as Eu(III) and Th(IV) relatively stable for redox reactions was carried out for bentonite colloids which had been prepared from the domestic Gyeongju bentonite. The amounts of the nuclides lost by an attachment to bottle walls, by a precipitation, and by a colloid formation were estimated by performing blank tests for the sorption experiments. Sorption coefficients, $K_d's$, reflecting the mass losses were obtained and investigated for the sorption of Eu(III) and Th(IV) onto the bentonite colloids. The net sorption coefficients $K_d's$ considering all the three mass losses were measured as about $10^6-10^7\;mL/g$ and $7{\times}10^6-10^7\;mL/g$ for Eu(III) and Th(IV), respectively, depending on pH. In particular, a precipitation occurred mainly at a pH greater than 5 for Eu(III) and a precipitation and colloid formation significantly occurred at a pH greater than 3 for Th(IV). The precipitation and colloid formation of the multivalent nuclides of Eu(III) and Th(IV) therefore should be considered when $K_d's$ are rightly obtained over the pH range where their precipitation and colloid formation become significant at a given concentration.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.101-109
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• 2008

For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to $1.8\;Mg/m^3$ and the bentonite-sand mixture with a dry density of 1.6 and $1.8\;Mg/m^3$ were measured within the gravimetric water content range of 10wt% to 20wt% and the sand fraction range of 10 to 30wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.211-221
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• 2016

Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP ($SiO_2-Al_2O_3-P_2O_5$). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding $B_2O_3$ and $Fe_2O_3$ to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of $Al_2O_3$ or $B_2O_3$ as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.267-277
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• 2016

For the decommissioning of a nuclear power plant, a site characterization and final status survey require a site-specific suite of radionuclides that could potentially still be present in the site during or after the decontamination processes. The United States Nuclear Regulatory Commission (U.S. NRC) requires a Decommissioning Technical Base Document (DTBD) along with a Site Characterization and Historical Site Assessment (HSA) from the utility for decommissioning to proceed. Both the DTBD and HSA are preliminary components of the Radiological Site Survey investigation process and should be included in the final License Termination Plan (LTP) for site release and reuse consideration from the U.S. NRC and the utility company. This study reviews the United States Nuclear Power Plants (U.S. NPPs) decommissioning cases and is especially focused on the methodologies used for determining a site-specific suite of radionuclides before and during the site characterization and final status surveys. In 2017, Kori-1 will be ready for decommissioning and related preparations are ongoing, this review will help Korea to prepare regulatory guidelines and give technical background for the safe and successful decommissioning of NPPs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.235-243
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• 2016

In the near future, many countries, including the Republic of Korea, will face a significant increase in low level radioactive waste (LLW) from nuclear power plant decommissioning. The purpose of this paper is to look at blending as a method for enhancing disposal options for low-level radioactive waste from the decommissioning of nuclear reactors. The 2007 U.S. Nuclear Regulatory Commission strategic assessment of the status of the U.S. LLW program identified the need to move to a risk-informed and performance-based regulatory approach for managing LLW. The strategic assessment identified blending waste of varying radionuclide concentrations as a potential means of enhancing options for LLW disposal. The NRC's position is that concentration averaging or blending can be performed in a way that does not diminish the overall safety of LLW disposal. The revised regulatory requirements for blending LLW are presented in the revised NRC Branch Technical Position for Concentration Averaging and Encapsulation (CA BTP 2015). The changes to the CA BTP that are the most significant for NPP operation, maintenance and decommissioning are reviewed in this paper and a potential application is identified for decommissioning waste in Korea. By far the largest volume of LLW from NPPs will come from decommissioning rather than operation. The large volumes in decommissioning present an opportunity for significant gains in disposal efficiency from blending and concentration averaging. The application of concentration averaging waste from a reactor bio-shield is also presented.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.297-305
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• 2016

An approach to assess inadvertent human intrusion into radwaste repository was proposed with the assumption that the intrusion occurs after a loss of knowledge of the hazardous nature of the disposal facility. The essential boundary conditions were derived on the basis of international recommendations, followed by an overall approach to deal with inadvertent human intrusion. The interrelation between societal factors, human intrusion scenarios, and protective measures is described to provide a concrete explanation of the approach, including the detailed procedures to set up the human intrusion scenario. The procedure for deriving protective measures is also explained with four steps, including how to derive a safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be used effectively to reduce the potential for and/or the consequences of human intrusion during the entire process of realizing a disposal facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.201-209
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• 2016

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except $5{\times}5{\times}2m\;size$) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a $50{\times}50{\times}2m\;size$ facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.253-266
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• 2016

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.253-257
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• 2013

Geology, hydrogeology, and geochemistry are the main technical siting factors of a geological repository for spent nuclear fuels. This paper evaluated the siting criteria of overseas geological repository with related to hydrogeologic properties, such as hydraulic conductivity, partitioning coefficient, dispersion coefficient, boundary condition, and water age. Each country establishes the siting criteria based on its important geological backgrounds and information, and social environment. For example, Sweden and Finland that have decided a crystalline rock as a host rock of a geological repository show different siting criteria for hydraulic conductivity. In Sweden, it is preferable to avoid area where the hydraulic conductivity on a deposition hole scale (~30m) exceeds $10^{-8}m/s$, whereas Finland does not decide any criterion for the hydraulic conductivity because of limited data for it. In addition, partitioning coefficients should be less than 10-1 of average value in Swedish crystalline bedrock. However, the area where shows 100 times less than average partitioning coefficients of radionuclides in crystalline rock should be avoided in Sweden. In German, the partitioning coefficients for the majority of the long-term-relevant radionuclides should be greater than or equal to $0.001m^3/kg$. Therefore, it is strongly required to collect much and exact information for the hydrogeologic properties in order to set up the siting criteria.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.245-251
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• 2013

The energy band and the G-factor method were compared to determine the exposure rate from the measured spectrum using a NaI(Tl) scintillation detector. First, G-factors of a 3"${\Phi}X3$" NaI(Tl) detector mounted to a EFRD 3300, which means the environmental radiation monitor, in Korea Atomic Energy Research Institute (KAERI) were calculated for several directions of incident photons through the MCNP modeling, and the optimum G-factor applicable to that monitor was then determined by comparing the results both the energy band method and the G-factor method. The results for these spectrometric determinations were also compared with the dose rate from a HPIC radiation monitor around a EFRD 3300. The measured value at the EFRD 3300 based on a 3"${\Phi}X3$" NaI(Tl) detector was $7.7{\mu}R/h$ and its difference was shown about $3{\mu}R/h$, when compared with the results from a HPIC radiation moditor. Since a HPIC is known to be able to measure cosmic rays with the relatively high energy, the difference between them was caused by cosmic rays which were not detected in a 3"${\Phi}X3$" NaI(Tl) detector.