• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Journal of Radiation Industry
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• v.17 no.4
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• pp.501-508
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• 2023

There are various factors that have a negative impact on safety over a long period of time after the closure of a radioactive waste disposal facility. In particular, it is important to limit substances that accelerate radionuclide migration while inhibiting adsorption between radionuclides and the subsurface medium. Through this study, a method for deriving a quantitative criteria evaluation method is proposed for cellulose among materials that accelerate the movement of these radionuclides after closure of the disposal facility. Since Sweden's SKB is representative worldwide for preparing criteria for cellulose in disposal facilities, it analyzed Sweden's acceptance criteria method and presented a method that can be applied domestically. The decomposition characteristics of cellulose and the adsorption and dissolution characteristics of ISA among degradation products were reviewed, and quantitative analysis of cement materials that create a high pH environment favorable for cellulose decomposition was also included. In addition, the total amount of the finally disposable cellulose material can be derived by using the volume information of the waste containing the cellulose material. Through this methodology for calculating the total amount of cellulose, it is expected that subsequent studies will be conducted to secure data reflecting the environmental conditions of radioactive waste disposal facilities in Korea. In addition, it is expected to be utilized as a good method to evaluate the impact of other complexing agents other than cellulose and to suggest the amount of disposal.

• 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.56 no.9
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• pp.3942-3949
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• 2024

This study focuses on investigating the chemical degradation characteristics of cementitious barriers used in low-and intermediate-level radioactive waste repository by reactive transport modeling. The impact of the blending with supplementary cementitious materials (SCMs) in the barriers on the chemical degradation was evaluated to find the optimum barrier design. A number of different barrier designs were examined by replacing ordinary Portland cement (OPC) by SCMs (i.e., fly ash, silica fume, and blast-furnace slag). The simulation results showed that silica fume blended barrier has better durability against chemical degradation by rainwater compared to fly ash or blast-furnace slag blended barriers. In addition, the chemical durability of silica fume-based barrier increased with increasing replacement level up to about 20 %. It seems that the amount of formed calcium silicate hydrate (CSH) in the initial cement-based barrier highly affects the overall chemical durability. The newly developed reactive transport model demonstrated its capability for understanding the barrier performance and investigating the optimal design of the barrier system.

• Environmental Engineering Research
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• v.17 no.2
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• pp.59-63
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• 2012

This paper proposes a modified Ludzack-Ettinger (MLE) type membrane bioreactor (MBR) as a method of treatment for wastewater from food waste disposer. Micro-membrane filtration allows for an extremely low concentration of suspended solids in the effluent. The effluent of the reactor in question is characterized by a relatively high level of non-biodegradable organics, containing a substantial amount of soluble microbial products and biomass. Results obtained in this paper by measurement of membrane fouling are consistent with biomass concentration in the reactor, as opposed to chemical oxygen demand (COD). The MLE process is shown to be effective for the treatment of wastewater with a high COD/N ratio of 20, resulting in are markedly high total nitrogen removal efficiency. Denitrification could be improved at a higher internal recycle ratio. Despite the low concentration of influent phosphorus, the phosphorus concentration of the outflow is seen to be relatively high. This is because outflow phosphorous concentration is related to COD consumption, and the process operates at along solids retention time.

• International Journal of Fluid Machinery and Systems
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• v.3 no.4
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• pp.285-291
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• 2010

To provide required flow rates of cooling or circulating water properly, approach flow conditions of vertical pump systems should be in compliance with state of the art acceptance criteria. The direct inflow should be vortex free, with low pre-rotation and symmetric velocity distribution. Physical model investigations are common practice and the best tool of prediction to evaluate, to optimize and to document flow conditions inside intake structures for vertical pumping systems. Optimization steps should be accomplished with respect to installation costs and complexity on site. The report shows evaluation of various approach flow conditions inside a compact waste water pumping station. The focus is on the occurrence of free surface vortices and the evaluation of air entrainment for various water level and flow rates. The presentation of the results includes the description of the investigated intake structure, occurring flow problems and final recommendations.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.81-94
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• 1991

In order to develop the acceptance criteria for the low and intermediate level radioactive wastes for the land disposal: the following items were reviewed : classifications of radioactive wastes is respect to disposal, basic requirements and criteria that have to be considered during waste management from the origin to disposal. From these studies, the standard test methods to evaluate radioactive waste forms（or packages) were shown.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1495-1506
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• 2023

For the safe disposal of high-level radioactive waste using Engineered Barrier Systems (EBS), bentonite buffer is used by its high swelling capability and low hydraulic conductivity. When the bentonite buffer is contacted to heated pore water containing ions by radioactive decay, chemical alterations of minerals such as illitization reaction occur. Illitization of bentonite indicates the alteration of expandable smectite into non-expandable illite, which threatens the stability and integrity of EBS. This study intends to provide a thorough review on the information underlying in the illitization of bentonite, by covering basic clay mineralogy, smectite expansion, mechanisms and observation of illitization, and illitization in EBS. Since understanding of smectite illitization is crucial for securing the safety and integrity of nuclear waste disposal systems using bentonite buffer, this thorough review study is expected to provide essential and concise information for the preventive EBS design.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.161-172
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• 2017

To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classification logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.369-380
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• 2017

In order to manage the uncertainty about the evaluation and analysis of the human intrusion scenario of the Gyeongju Low and Intermediate Level Radioactive Waste(LILW) disposal facility, the calculation result by the GENII code was assessed using the RESRAD code, which was developed to evaluate the radiation effects of contaminated soil. The post-drilling scenario was selected as a human intrusion scenario into the near-surface disposal facility to analyze the uncertainty of the modeling by identifying any limitations in the simulation of each code and comparing the evaluation results under the same input data conditions. The results revealed a difference in the migration of some nuclides between the codes, but confirmed that the dose trends at the end of the post-closure control period were similar for all exposure pathways. Based on the results of the dose evaluation predicted by RESRAD, sensitivity analysis on the input factors was performed and major input factors were derived. The uncertainty of the modeling results and the input factors were analyzed and the reliability of the safety evaluation results was confirmed. The results of this study can be applied to the implementation 'Safety Case Program' for the Gyeongju LILW disposal facility.