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

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.02a
• /
• pp.28-41
• /
• 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
• /
• v.54 no.2
• /
• pp.442-448
• /
• 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
• /
• v.55 no.5
• /
• pp.1935-1945
• /
• 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
• /
• 2017.10a
• /
• pp.211-212
• /
• 2017

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

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4543-4553
• /
• 2023

The dose from a possible accident at a microwave-based spent resin mixture treatment facility that was to be installed and operated at the Wolsong nuclear power plant was analyzed to evaluate the radiological safety prior to its installation and operation. The dose to which workers and nearby residents are likely to be exposed was calculated based on the atmospheric dispersion and deposition factors using the XOQDOQ code. The highest atmospheric dispersion factors were 1.349E-05 s/m³ (workers) and 1.534E-06 s/m³ (residents). The highest doses due to emissions from the mock-up tank before operation were 1.91E-06 mSv (workers) and 1.78E-07 mSv (residents). Even after 3 h of operation, emissions from the mock-up tank had the greatest impact ranging from 4.63E-08 to 1.24E-06 mSv (workers) and 2.74E-10 to 1.16E-07 mSv (residents), respectively. The doses were 7.09E-09-4.55E-07 mSv and 4.18E-11-4.25E-08 mSv at 4-5 h of operation, and the maximum doses after operation reached 5.69E-07 mSv and 5.31E-08 mSv for the workers and residents, respectively. Even at the exclusion area boundary (EAB), 4.76E-08-9.51E-07 mSv (annual dose:9.52E-05–1.90E-03 mSv/y) was below the dose limit of the EAB, and the safety of the facility installation inside the NPP was confirmed.

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

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.327-328
• /
• 2018

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3824-3836
• /
• 2021

Workers' safety from radiological exposure in a 1 ton/day capacity spent resin treatment facility was evaluated according to the operating times and outflow rate due to process related leakages. The conservative annual dose based on the operating times of the workers exceeded the dose limit by at least 7.38E+01 mSv for close work. The realistic dose range was derived as 1.62E+01 mSv-6.60E+01 mSv. The conservative and realistic annual doses for remote workers were 1.33E+01 mSv and 3.00E+00 mSv respectively, which were less than the dose limit. The MWR was identified as the major contributor to worker exposure within the 1 h period required for removal of radioactive materials. The dose considering both internal and external exposures without APF was derived to be 1.92E+01 mSv for conservative evaluation and 4.00E+00 mSv for realistic evaluation. Furthermore, the dose with APF was derived as 7.27E-01 mSv for conservative evaluation and 1.51E-01 mSv for realistic evaluation. Considering the APF for leakage from all parts, the dose range was derived as 1.25E+00 mSv-2.03E+00 mSv for conservative evaluation and 2.61E-01 mSv-4.23E-01 mSv for realistic evaluation. Hence, it was confirmed that radiological safety was secured in the event of a leakage accident.