• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.28-41
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• 2004

The study is aimed to assess the usefulness of the mixture design for spent resin immobilization in cement. Although a considerable amount of research has been carried out to determine the limits for the composition of an acceptable resin-cement mixture, no efficient experimental strategy exists that explores the full properties of waste form against composition relationship. In order to gain an overall view, this report introduces the method of mixture design and mixture analysis, and describes the design of experiment of the 5-component mixture with the constraint conditions. The mathematic models of 28-day compressive strength varying with the ingredients are fitted, and the main effect and interaction effect of two ingredients are identified quantitatively along with the graphical interpretation using the response trace plot and contour plots.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3617-3627
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• 2023

This study presents a method to solidify spent ion-exchange resin (IER) in a metakaolin-based geopolymer and shows results of mechanical strength, immersion, leaching, irradiation, and thermal cycling tests for waste acceptance criteria (WAC) to repository. The geopolymer waste form with 20 wt% of simulated spent IER met the WAC in South Korea (ROK), and the leaching tests of various sized-waste forms up to 15.0 × 30.0 cm and waste loadings up to 20 wt% for 1-5 d and 1-90 d achieved a leachability index, L_i > 6. In a leaching test for 5 d, the cumulative fraction leached (CFL) for Cs, which leached the most, was linearly correlated with the square root of leaching time for all waste forms, and L_i increased as the size of the waste form increased. The CFL was also correlated with elapsed time in the 90 d leaching test. The correlations among CFL, time, and volume-to-surface area ratio of waste forms used to estimate the L_i of Cs of a 200-L sized geopolymer with 15 wt% IER showed the L_i values as 14.73 (5 d) and 17.71 (90 d), respectively, indicating that the large-sized geopolymer waste form met the WAC.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.273-281
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• 2021

The radiological safety of the spent resin treatment facility with a¹⁴C treatment capacity of 1 ton/day was evaluated in terms of the external and internal exposure of worker according to operation scenario. In terms of external dose, the annual dose for close work for 1 h/day at a distance of more than 1 m (19.8 mSv) satisfied the annual dose limit. For 8 h of close work per day, the annual dose exceeded the dose limit. For remote work of 2000 h/year, the annual dose was 14.4 mSv. Lead shielding was considered to reduce exposure dose, and the highest annual dose during close work for 1 h/day corresponded to 6.75 mSv. For close work of 2000 h/year and lead thickness exceeding 1.5 cm, the highest value of annual dose was derived as 13.2 mSv. In terms of internal exposure, the initial year dose was estimated to be 1.14E+03 mSv when conservatively 100% of the nuclides were assumed to leak. The allowable outflow rate was derived as 7.77E-02% and 2.00E-01% for the average limit of 20 mSv and the maximum limit of 50 mSv, respectively, where the annual replacement of the worker was required for 50 mSv.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.311-320
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• 2006

Removal characteristics of $H^{14}CO_3$ ion from IRN-150 mixed resin contaminated with $^{14}C$ radionuclide and a gasification behavior of $^{14}C$ radionuclide to $^{14}CO_2$ were investigated. The stripping solutions used for the removal of $^{14}C$ from spent resin were $NaNO_3,\;Na_3PO_4,\;NH_4H_2PO_4,\;H_3PO_4$. The influence of stripping solution concentration on the desorption characteristics of inactive $HCO_3$ ion into stripping solution from IRN-150 mixed resin and the gasification of this ion to $CO_2$ was analyzed. The gasification behavior to $CO_2$ by using NaOH, $HNO_3$, HCl was also compared to that of phosphate solution. Real spent resin stored in Wolsung nuclear power plant was used to evaluate the gasification characteristics of $^{14}C$ radionuclide to $^{14}CO_2$. Gamma radionuclides such as $^{137}Cs,\;^{60}Co$ in residual striping solutions after desorption experiment were analyzed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.253-254
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.130-134
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• 2003

The quantity of solid waste produced during HANARO operation for the years from 1996 to 2002 has been investigated and the interrelation with the reactor power output has been analysed. The waste amount produced during this period was $69, 598{\ell}$ and the processing expense was 1, 878, 000 won. The waste amount and processing expense per reactor power output are $3.01{\ell}/MWD$ and 811 won/MWD, respectively. A dry equipment for the spent resin was developed and the spent resin volume has been reduced by about 40% by using it.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.65-82
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• 2018

With the significant increase in spent ion-exchange resin generation, to meet the requirements of Waste Acceptance Criteria (WAC) of the Wolsong disposal facility in Korea, blending is considered as a method for enhancing disposal options for intermediate level waste from nuclear reactors. A mass balance formula approach was used to enable blending process with an appropriate mixing ratio. As a result, it is estimated around 44.3% of high activity spent resins can be blended with the overall volume of low activity spent resins at a 1:7.18 conservative blending ratio. In contrast, the reduction of high activity spent resins is considered a positive solution in reducing the amount of spent resins stored. In an economic study, the blending process has been proven to lower the disposal cost by 10% compared to current APR1400 treatment. Prior to commencing use of this blending method in Korea, coordinated discussion, and safety and health assessment should be undertaken to investigate the feasibility of fitting this blending method to national policy as a means of waste predisposal processing and management in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.375-375
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.276.2-276
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• 1999