According to acceptance of radioactive waste, homogeneous waste such as concentrated liquid waste and spent resin must be solidified to reduce radiological hazards and protect public health and the ecology. However, when using a High Integrity Containers (HIC), it is stated that homogeneous waste can be disposed of without applying the solidification test requirements. PCHIC, developed in korea, is composed of polyethylene (PE, interior), polymer concrete (PC, filler), and steel (external reinforcement). Currently, PC-HIC will be used as a packaging container for low-level liquid waste and spent resin waste. PE has a lower shielding efficiency compared to PC, but it offers the economic advantage of lower production costs. Therefore, cost savings can be expected if very low-level waste is packaged and disposed of HIC made only of PE materials (PEHIC). Despite the economical advantage of PE-HIC, PE-HIC has not been used domestically since NRC (Nuclear Regulatory Commission) reported that PE-HIC couldn’t meet the mechanical integrity criteria for radiation exsure. However, according to IAEA (International Atomic Energy Agency) research, it has been reported that mechanical integrity of PE-HIC is not affected when the absorbed dose is below 50 kGy. Therefore, in this study, Radiological impact of VLLW packaged into PE-HIC is evaluated to confirm that the absorbed dose is below 50 kGy, which then be used to assess feasibility of PE-HIC to be used as packaging and disposal container for radioactive waste. Radiological impact of VLLW packaged into PE-HIC is evaluated to confirm that the absorbed dose is below 50 kGy, which then be used to assess feasibility of PE-HIC to be used as packaging and disposal container for radioactive waste. The feasibility of using PE-HIC as packaging-disposal containers for radioactive waste will be reviewed. In this study, the radiation effects of only waste packaged in PE-HIC were considered, and additional assumptions are as follows. - Nuclides subject to radioactivity evaluation: Co-60, Cs-137 - Radioactivity concentration: very low-level radioactive wastel concentration limit - Target waste: waste resin - PE-HIC dimensions: outer diameter: 1,194 mm, height: 1,290 mm, and thickness 88 mm (PCHIC internal PE shape) Considering the above assumption, the exposure rate was evaluated using the MicroShield program. Since the density of PE-HIC in the MicroShield program was assumed as the density of air. The absorbed dose was recalculated through density correction of the derived exposure rate. As a result, it was confirmed that absorbed dose was about 2-3 mGy over 300 years. As a result of dose evaluation by MicroShield, it is judged that the mechanical integrity of PEHIC as an packaging of VLLW can be proved by confirming that the absorption dose irradiated to PE-HIC by internal waste is less than 50 kGy.

The treatment process for Spent Filter(SF) of Kori-1 was developed that includes the following : 1) Taking out by robot system 2) Screening by ISOCS 3) Collection of representative samples using a sampling machine 4) Compression 5) Immobilization 6) Packaging and nuclide analysis and 7) Delivery/disposal. Although the robot system, ISOCS, sampling machine and immobilization facility are essentially required for building the above processing but decision to build the compression system and nuclide analysis system must be made after reviewing the need and cost benefit for their construction. In addition, for effcient SF treatment, it is necessary to determine the nuclide concentration range of the SF to which immobilization will be applied. In this study, a cost benefit analysis was performed on existing and alternative methods for processes related to compression treatment, nuclide analysis and immobilization methods, which are greatly affected by economics and efficiency according to the design. First, although the disposal cost is reduced with reducing the number of packaging drums by compressed and packaged but the expected benefits not be equal to or greater than the cost invested in building a compression system. As a result, non-compressed treatment of SF is expected to be economical because the construction cost of compression system is more expensive than the benefits of reducing disposal costs by compression. Second, a cost benefit analysis of direct and indirect nuclide analysis methods was performed. For indirect analysis, scaling factors should be developed and the drum scanner suitable for the analysis for DAW should be improved. As a result, direct analysis applied grouping options is expected to be more economical than indirect analysis requiring the cost for developing scaling factors and improving the scanner. Third, it is timeconsuming and inefficient to distinguish and collect filters that are subject to be immobilized according to the waste acceptance criteria among the disorderly stored SFs in the filter rooms. If the benefits of immobilization of the SFs selectively are not greater than the benefits of immobilization of all SFs, it can be economical to immobilize all SFs regardless of the nuclide concentration of them. As a result, it is more economical to immobilize all SFs with various nuclide concentrations than to selectively immobilize them. The conclusion of this study is that it is not only cost-effective but also disposal-effective to design the treatment process of SF to adopt non-compressed processing, direct analysis and immobilization of all SFs.

Radioactive waste (hereinafter referred to as mixed waste) containing hazardous substances (heavy metals, organic and inorganic waste liquids, asbestos, etc.) has been continuously generated from domestic nuclear power plants, nuclear facilities, and other industrial facilities, and heavy metals were released during the dismantlement of Kori Unit 1 and Wolseong Unit 1. Lead, cadmium, mercury, arsenic), asbestos, decontamination waste liquid (organic/inorganic waste liquid), etc. may be generated. Although hazardous waste related to the nuclear industry continues to be generated, only the regulation direction for hazardous substances is presented in the provisions related to hazardous substances in the delivery regulations for low and intermediate-level radioactive waste and the acceptance criteria for low and intermediate-level radioactive waste disposal facilities. In particular, because there is no clear definition of “hazardousness” and specific standards such as concentration and characteristics for classification of hazardous substances, as well as hazard removal procedures when the hazardousness of radioactive waste is confirmed, no hazardous substances have been delivered in Korea to date and many mixed wastes are stored at each generation facility or at the NPP. As a plan to improve delivery standards related to mixed waste is being prepared recently, it is believed that if the acceptance standards are revised accordingly, it will be possible to confirm the suitability for disposal of drums produced after the establishment of the acceptance standards in 2015. However, it is believed that securing disposal suitability for waste that was packed in 200L drums and compressed under super high pressure in the absence of specific technical standards and regulatory guidelines for the disposal of radioactive waste containing hazardous substances would still remain a difficult problem. In this report overseas acceptance standards related to hazardous waste were reviewed and a plan to secure the disposal suitability of 200 L drums compressed with of super high pressure was proposed.

Every engineering decision in radioactive waste management should be based on both technical and economic considerations. Especially, the management of low-level radioactive waste (LLW) is more critical on economic concerns, due to its long-term and continuous nature, which emphasizes the importance of economic analysis. In this study, economic factors for LLW management were discussed with appropriate engineering applications. Two major factors that should be taken into account when assessing economic expectations are the accuracy of the results and its proper balancing with ALARA philosophy (As Low As Reasonably Achievable). The accuracy of the results depends on the correct application of alternatives within a realistic framework of waste processing. This is because the LLW management process involves variables such as component type, physical dimensions, and the monetary value at the processing date. Two commonly used alternatives are the simplified lump sum present worth and levelized annual cost methods, which are based on annual and capital costs. However, these discussions on alternatives not only pertain to the time series value of operational costs but also to future technical advancements, which are crucial for engineers. As new research results on LLW treatment emerge, proper consideration and adoption should be given to technical cost management. As safety is the core value of the entire nuclear industry, the ALARA philosophy should also be considered in the cost management of LLW. The typical cost of exposure in man-rem has ranged from $1,000 to $20,000 over the past decades. However, with increasing concerns about health and international political threats, the cost of man-rem should be subject to stricter criteria, even the balancing of costs and safety concerns is much controverse issue. Throughout the study, the importance of incorporating proper engineering insights into the assessment of technical value for the financial management of LLW was discussed. However, it’s essential to remember that financial management should not be solely assessed based on the size of expenses but rather by evaluating the current financial status, the value of money at the time, and anticipated future costs, considering the specific context and timeframe.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.