• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2606-2613
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• 2022

Spent resin often exceeds radiation limits for safe disposal, creating a need for commercial-scale treatment techniques to reduce resin radioactivity. In this study, the radiological safety of a commercialized spent resin treatment device with a treatment capacity of 1 ton/day was evaluated. The results confirm that the device is radiologically safe in the event of an accident. This device desorbs ¹⁴C from the spent resin, allowing disposal as low-level waste instead of intermediate-level waste. The device also reduces overall waste by recycling the extracted ¹⁴C. Potential accident scenarios were explored to enable dose assessments for both internal and external exposure while preventing further spillage of the device and processing the spilled resin. The scenarios involved the development of a surface fracture on the resin mixture separator and microwave systems, which were operated under pressure and temperature of 0-6 bar and 0-150 ℃, respectively. In the case of accidents with separator and microwave device, the maximum allowable working time of worker were derived, respectively, considering external and internal exposures. When wearing the respirator corresponding to APF 50, in the case of the microwave device accident scenario, the radiological safety was confirmed when the maximum worker worked within 132.1 h.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.442-448
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• 2022

Intermediate-level radioactive waste (ILW) is not subject to legal approval for cave disposal in Korea. To solve this problem, a spent resin treatment device that separates ¹⁴C-containing resin from zeolite/activated carbon and desorbs ¹⁴C through a microwave device has been developed. In this study, we evaluated the radiological safety of the operators performing repair work in the event of a failure in such a device treating 1 ton of spent resin mixture per day. Based on the safety evaluation results, it is possible to formulate a design plan that can ensure the safety of workers while developing a commercialized device. When each component of the resin treatment device can be repaired from the outside, the maximum and minimum allowable repair times are calculated as 263.2 h and 27.7 h for the ¹⁴C-detached resin storage tank and zeolite/activated carbon storage tank, respectively. For at least 6 h per quarter, the worker's annual dose limit remains within 50 mSv/year; further, over 5 years, it remained within 100 mSv. At least 6 h of repair time per quarter is considered, under conservative conditions, to verify the radiological safety of the worker during repair work within that time.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1935-1945
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• 2023

The radiological safety of workers in a newly developed microwave-based spent resin treatment facility was assessed based on work location and operational scenarios. The results show that the remote-operation room worker was exposed to maximum annual dose of 3.19E+00 mSv, which is 15.9% of the dose limit, thereby confirming radiological safety. Inside the pathway, annual doses in the range of 7.87E-02-2.07E-01 mSv were measured initially at the mock-up tank and later at the point between the spent resin separation and treatment parts. The dose of emergency maintenance workers was below the dose limit (4.08E-03-4.99E+00 mSv); however, before treatment (separation and microwave), the dose of maintenance and repair workers exceeded the dose limit. The doses of the effluent removal workers at the zeolite and activated carbon storage tank and spent resin storage tank were the lowest at 2.79E-01-2.87E-01 mSv and 9.27E-01 mSv in "1 h" and "4-5 h of operation", respectively. The immediately lower and upper layers of the facility room exhibited the highest annual doses of 1.84E+00 and 3.22E+00 mSv, respectively. Through this study, a scenario that can minimize the dose considering the movement of spent resin through the facility can be developed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.214-221
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• 2005

Spent ion-exchanged resin generated from various purification systems in CANDU reactor is causing concern due to a limited storage capacity and safe disposal. As a suggestion for a proper treatment technology for the spent ion-exchanged resin containing a high activity of C­14 radionuclide which would be classified as Class A and C wastes, a fundamental study for the development of C-14 removal technology from a spent resin was performed. The adsorption characteristics of the inactive $HCO_3^-$ ion and other ions in a stripping solution on IRN-150 mixed resin was evaluated and the removal technology of the $HCO_3^-$ ion adsorbed on IRN-150 by an alkaline stripping method was proposed.

• 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.373-384
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• 2006

Spent ion-exchanged resin generated from various purification systems in CANDU reactor was contaminated with high activity of $^{14}C$ radionuclide. This paper describes the results of fundamental study to develop the applicable technology for the treatment of this spent resin. Based on the adsorption capacity of inactive $HCO_3$ ion and other anions on IRN-150 mixed resin, the removal characteristics of $HCO_3$ ion adsorbed on to IRN-150 by various stripping solutions were evaluated. Maximum adsorption amount of the $HCO_3$ ion onto IRN-150 raw resin was about 11 mg-C/g-resin which agrees with the theoretical adsorption amount of this resin. Adsorption affinity of various anions such as $CS,\;CO,\;Na\;NH_4$ was analyzed in single and multi-component systems. From the results of removal characteristics of the $HCO_3$ ion adsorbed on IRN-150 by various stripping solutions, $NH_4H_2PO_4$ stripping solution is more effective than $NaNO_3,\;Na_3PO_3$ solutions for the complete removal of $^{14}C$ radionuclide from the IRN-150 spent resin.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.227-228
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• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.163-164
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.161-162
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.231-232
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• 2017