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.

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.

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.

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

With the recent concern regarding cellulose enhancing radionuclide mobility upon its degradation to ISA, disposal of cellulosic wastes is being held off until the disposal safety is vindicated. Thus, a rational assessment should be conducted, applying an appropriate cellulose degradation model considering the disposal environment and cellulose degradation mechanisms. In this paper cellulose degradation mechanisms and the disposal environment are studied to propose the best-suitable cellulose degradation model for the domestic 1st phase repository. For the cellulose to readily degrade, the pH should be greater than 12.5. As in the case of SKB, 1BLA is excluded from the safety assessment because the pH of 1BLA remains below 12.5. Furthermore, despite cellulose degradation occurring, it does not always produce ISA. At low Ca2+ concentration, the ISA yield rate is around 25%, but at high Ca2+ concentration, the ISA yield rate increases up to 90%. Thus, for the cellulose to be a major concern, both pH and Ca2+ concentration conditions must be satisfied. To satisfy both conditions, the cement hydration must be in 2nd phase, when the porewater pH remains around 12.5 and a significant amount of Ca2+ ion is leaching out from the cement. However, according to the safety evaluation and domestic research, 2nd phase of cement hydration for silo concrete would achieve a pH of around 12.4, dissatisfying cellulose degradation condition like in 1BLA. Thus, cellulose degradation would be unlikely to occur in the domestic 1st phase repository. To derive waste acceptance criteria, a quantitative evaluation should be conducted, conservatively assuming cellulose is degraded. To conduct a safety evaluation, an appropriate degradation model should be applied to determine the degradation rate of cellulose. According to overseas research, despite the mid-chain scission being yet to be seen in the experiments, the degradation model considering mid-chain scission is applied, resulting in an almost 100% degradation rate. The model is selected because the repositories are backfilled with cement, achieving a pH greater than 13, so extensive degradation is reasonably conservative. However, under the domestic disposal condition, where cellulose degradation is unlikely to occur, applying such model would be excessively conservative. Thus, the peeling and stopping model derived by Van Loon and Haas, which suggests 10~25% degradation rate, is reasonably conservative. Based on this model, cellulose would not be a major concern in the domestic 1st phase repository. In the future, this study could be used as fundamental data for planning waste acceptance criteria.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

Glass wool, the primary material of insulation, is composed of glass fibers and is used to insulate the temperature of steam generators and pipes in nuclear power plants. Glass fiber is widely adopted as a substitute for asbestos classified as a carcinogen. The insulations used in nuclear power plants are classified as radioactive waste and most of the insulation is Very Low-Level Waste (VLLW). It is packaged in a 200 L drum the same as a Dry Active Waste (DAW). In the case of the insulations, it is packaged in a vinyl bag and then charged into the drum for securing additional safety because of the fine particle size of the fiberglass. A safety assessment of the disposal facility should be considered to dispose of radioactive waste. As a result of analyzing overseas Waste Acceptance Criteria (WAC), there is no case that has a separate limitation for glass fiber. Also, in order to confirm that glass fibers can be treated in the same manner as DAW, research related to the diffusion of glass fibers into the environment was conducted in this paper. It was confirmed that the glass fiber was precipitated due to the low flow velocity of groundwater in the Gyeongju radioactive waste repository and did not spread to the surrounding environment due to the effect of the engineering barrier. Therefore, the glass fiber has no special issue and can be treated in the same way as a DAW. In addition, it can be disposed of in the disposal facility by securing sufficient radiological safety as VLLW.

Colloid migration is an important topic in post-closure safety assessment of radioactive waste repository as radionuclide can be adsorbed onto colloidal particles and migrated along with the colloids. This would reduce retardation of radionuclide migration, thus increasing the released concentration into biosphere. Recently, glass fiber waste has been found to contain small sized crushed glass fiber particles (GFPs), and concerns regarding the colloidal impact of GFP is being discussed. In this study, relevance of assessing GFPs facilitated radionuclide transport in the disposal environment of 1st phase disposal facility. Colloidal impact assessment can be divided into two sections, colloid mobility, and colloid sorption assessments. Considering GFP being denser than water, fluid velocity of 1st phase disposal facility is too slow to initiate movement of such dense particles. GFPs would remain settled, and no colloidal impact is expected. In this study, sorption assessment mainly focused to analyze the possible impact if migration of GFP does occur. The GFP is mainly composed of SiO2 and few other metal oxides. Due to high composition of SiO2 in the GFPs, negative surface charge is induced onto the surface of the GFPs in alkaline environment. This negatively charged surface can attract free positive ions (ex. Ni, Co, Fe, etc.) in the repository, and these ions would be adsorbed onto the surface of the GFPs via coulomb force. Thus, if GFPs migrate, colloid facilitated radionuclide transport can be expected. However, before being released into the biosphere, particles must pass through the engineered and natural barriers, where ion-colloid-rock interactions could result in transfer of radionuclide from one media to another. At Naka Research Center, Japan, ion-colloid-rock interactions are experimented with bentonite colloid, and the result showed that despite colloid’s sorption ability was 10 times higher than the barrier material, the overall released radionuclide concentration has negligible change. To reflect such phenomenon, coulomb attractive force of GFPs and concrete is calculated and compared, which the result showed that glass fiber was 10 times weaker than concrete. Considering the Japan’s experimental result, glass fiber facilitated transport would not enhance the radionuclide release into the biosphere. Nonetheless, assuming GFPs being mobile in 1st phase disposal facility, GFPs’ sorption ability is found to be negligible compared to the concrete of the repository, thus radionuclide transport is not expected to be enhanced. In future, this study could be used as basis for further colloidal impact analysis for the safety assessment of the repository.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Currently, in domestic nuclear power plants (NPP), the spent filters (SFs) used for the purpose of reducing and purifying the radiation of the primary cooling water system are temporarily stored in an untreated state. In order to dispose of SFs, radioactive nuclide analysis (RNA) of SFs is required to be conducted. As segmented gamma scanner (SGS) is already being used in Kori NPP, utilizing SGS for RNA of SFs would be practical and economical. In this paper, factors required to be considered to improve accuracy of SGSs for RNA of SFs are studied. The analysis of the nuclide inventory of the packaging drum for radioactive waste should be performed by the indirect drum nuclide analysis method. The material of the SFs is iron (SS304) on the outside, and paper on the inside. In addition, to meet disposal acceptance criteria, radioactive waste drums are packaged in thick grouting or shielding drums. Therefore, it is necessary to derive an appropriate correction method for high inhomogeneity and thick media. Considering these factors, evaluating radionuclides inventory plans to measure gamma rays in SGS mode. Correct the gamma ray measurement by examining the medium attenuation factor and error factors. In this way, the inventory of gamma nuclides is calculated, and the specific radioactivity of beta ray and alpha particle emitting nuclides other than gamma rays is planned to be calculated by applying scaling factors.

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Thus, evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, safety assessments conducted in Sweden and UK are studied, and the factors required to be considered for appropriate safety assessment of cellulose is analyzed. SKB (Sweden) conducted safety assessment of cellulose degradation as a part of long-term safety assessment of SFR. SKB determined that ISA would impact sorption of trivalent and tetravalent radionuclides (Eu, Am, Th, Np, Pa, Pu, U, Tc, Zr and Nb) at concentration higher than 10−4–10−3 M, and impact sorption of divalent radionuclides (Ni, Co, Fe, Be and Pb) at concentration higher than 10−2 M. Then, SKB conservatively set the upper limit of ISA concentration to be 10−4 M and conducted cellulose degradation evaluation on each waste package type, considering the expected disposal environment of SFR. Based on the calculated results, some of the waste packages showed concentration of ISA to be higher than 10−4 M, so SKB conservatively developed waste acceptance criteria to prevent ISA being produced to an extent of affecting the safety of the repository. SKB conducted safety assessment only for the repositories with pH above 12.5 and excluded 1BLA from the safety assessment as the expected pH of 1BLA is around 12, which is insufficient for cellulose to degrade. However, SKB set disposal limit for 1BLA as well, to minimize potential impact in future. Serco (UK) conducted safety assessment of cellulose degradation for the conceptual repository, which is a concrete vault with cementitious backfill. Serco estimated that the pH of repository would maintain around 12.4. Serco conservatively assumed that the pH would be sufficient for cellulose degradation to occur partially, and suggested application of appropriate degradation ratio for safety assessment of cellulose degradation. To conduct appropriate safety assessment of cellulose degradation, an appropriate ISA concentration limit based on radionuclide inventory list, and an appropriate cellulose degradation ratio based on the pH of disposal environment should be determined. As for guidance, below pH 12.5, cellulose degradation is not expected, and between pH 12.5–13, partial cellulose degradation is expected. In future, this study could be used as fundamental data to evaluate safety of the repository.

Near-surface disposal facility is more susceptible to intrusion than underground repository, resulting in more possible pathways for contaminant release. Alike human intrusion, animals (e.g. Ants, Moles, etc.) could intrude into the disposal site to excavate burrows, which could cause direct release of contaminants to biosphere. In this paper, animal intrusion is demonstrated using GoldSim’s commercial contaminant transport module and impact on the integrity of the near-surface disposal facility is evaluated in terms of fractional release rate of the contaminants. In this study, the near-surface disposal facility is modelled with a single concrete vault to contain radionuclide according to LLW concentration limit stated in NSSC notice No.2020-6. The release of contaminants is modelled to occur directly after the institutional control period, and the contaminants are mostly transported from the concrete vault to cover layers via diffusion. To produce mathematical model of the release of the contaminants due to animal intrusion, firstly, the fraction of burrow volume for each cover layer is calculated separately for each animal species, based on their maximum possible intrusion depth. In this study, fractions of burrow volume for ants and moles are calculated based on their maximum possible intrusion depths, where for ants is 2–3 m, and for moles is 0.1–0.135 m. Then, assuming that the contaminants are distributed homogeneously throughout each cover layers by diffusion, fraction of contaminants transported into the uppermost layer via excavation of the burrow is calculated for each layer based on burrow volume, and fraction of contaminants removed from the uppermost layer to the layers below via collapse of the burrow is also calculated based on the burrow volume. Lastly, the net transportation of contaminants into and out of the burrow via excavation and collapse, respectively, is calculated and demonstrated using direct transfer rate function of the GoldSim. Based on the simulated result, the maximum mass flux is too minor to cause a meaningful impact on the safety. The peak mass flux of the most sensitive radionuclide, I-129, is witnessed at around year 1,470, with a flux value of 5.36×10−6 g·yr−1. This minor release of the contaminants could be due to cover layers being much thicker than the maximum possible intrusion depth of the animals, preventing the animal intrusion into the deeper layers of higher radionuclide concentration. In future, this study can be used to provide a guidance and fundamental data for scenario development and safety evaluation of the near-surface disposal facility.

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, the maximum safe disposal amount cellulose is evaluated considering the disposal environment of silos of 1st phase disposal facility. The key factor governing the impact of cellulose degradation is pH of disposal environment, as cellulose is known to degrade partially at pH above 12.5, and completely at pH above 13. Thus, disposal environment should be analyzed as to determine the extent of degradation. As silos are constructed with large amount of cement, porewater within concrete walls would be of very high pH. However, for high pH porewater to be released into the pores of crushed rock, which is filling up the silos, lower pH groundwater (commonly pH 7) should flow into the silos through the concrete walls. This causes dilution of the high pH concrete porewater, resulting in a lower pH as the silos are filled, reaching to expected pH of 11.8–12.3, which is below cellulose degradation condition. Thus, cellulose degradation is not expected, but to quantitatively evaluate safe disposal amount of cellulose, partial degradation is assumed. Upon literature review, the most conservative ISA concentration, enhancing radionuclide mobility, is determined to be 1.0×10−4 M and to reach this concentration, cellulose mass equivalent to 6wt% of cement of the repository, is required to be degraded. However, this ratio is derived based on complete degradation of cellulose into ISA, so for partial degradation, degradation ratio and yield ratio of ISA should be considered. Commonly, cellulosic material (e.g. cotton, paper, etc.) has degree of polymerization (DP) between 1,000–2,000, and with this DP, degradation ratio is estimated to be about 10%. Furthermore, yield ratio of ISA is known to be 80%. Considering all these aspects, about 1.79×107 kg of cellulose could be disposed, which if converted into number of drums, considering cellulose content of dry active waste, more than 100,000 drums (200 L) could be disposed with negligible impact on safety. Based on the result, negligible impact of cellulose degradation is expected for safety of 1st phase disposal facility. In future, this study could be used as fundamental data for revising waste acceptance criteria.

It has been discovered that the isosaccharinic acid (ISA) formed in a cellulose degradation leachate were capable of forming soluble complexes with thorium, uranium (IV) and plutonium. Since 1993, the ISA has received particular attention in the literature due to its ability to complex a range of radionuclides, potentially affecting the migration of radionuclides. ISA is formed as a result of interactions between cellulosic materials within the waste inventory and the alkalinity resulting from the use of cementitious materials in the construction of the repository. In an alkaline cementitious environment, cellulose degrades mainly via a peeling-off reaction. The main degradation product is ISA, a polyhydroxy type of ligand forming stable complexes with tri- and tetravalent radionuclides. ISA can have an adverse effect on the sorption of radionuclides to an extent which depends on its concentration in the cement pore water and potentially enhance their mobility. The concentration of ISA is governed by several factors such as cellulose loading, cement porosity, extent of cellulose degradation, etc. The sorption of ISA on cement, however, is the process which governs the concentration of ISA in the pore water. According to the experimental result from a literature, the ISA concentration in facilities with a cellulose loading of 5% is calculated to be of the order of 10−4 M. At this level, the effect of cellulose degradation products on radionuclide sorption is negligibly small. Recently in Korea, cellulous limits as waste acceptance criteria is studying and planning to prepare the detailed requirement for near surface radioactive waste disposal facilities. It is desirable to suggest consideration on cellulose disposal limits around the time that the regulatory body and concern organizations establish the cellulose disposal limits as follows. Firstly, identify the cellulose effect on the sorption of the nuclides as cementitious disposal environments such as affected nuclides, threshold value and contribution to radiological risks under domestic disposal environment. Secondly, make sure and consider the difference between lab-scale experimental conditions and probability occurring in real disposal conditions such as probability for generation and persistence of pH in cellulosic material disposal conditions and cellulosic material disposal methods. Finally, consider characterization of cellulosic material such as polymerization, contents of cellulose in law material and time of degradation process. As a result, desirable cellulose limits are to set up for both safety and economic aspect.

Deep geological disposal (DGD) of spent nuclear fuels (SNF) at 500 m–1 km depth has been the mainly researched as SNF disposal method, but with the recent drilling technology development, interest in deep borehole disposal (DBD) at 5 km depth is increasing. In DBD, up to 40SNF canisters are disposed of in a borehole with a diameter of about 50 cm, and SNF is disposed of at a depth of 2–5 km underground. DBD has the advantage of minimizing the disposal area and safely isolating highlevel waste from the ecosystem. Recently, due to an increasing necessity of developing an efficient alternative disposal system compared to DGD domestically, technological development for DBD has begun. In this paper, the research status of canister operation technology and plans for DBD demonstration tests, which subjects are being studied in the project of developing a safety-enhancing high-efficiency disposal system, are introduced. The canister operation technology for DBD can be divided into connection device development and operation technology. The developing connection device, emplacing and retrieving canisters in borehole, adopted the concept of a wedge thus making replacement equipment at the surface unnecessary. The new connection device has the advantage of being well applied with emplacement facilities only by simple mechanical operation. The technology of operating a connection device in DBD can be divided into drill pipe, coiled tubing, free-drop, and wireline. The drill pipe is a proven method in the oil industry, but requiring huge surface equipment. The coiled tubing method uses a flexible tube and shares disadvantages as the drill pipe. The free-drop is a convenient method of dropping canister into a borehole, but has a weakness in irretrievability in an accident. Finally, the wireline method can be operational on a small scale using hydraulic cranes, but the number of operated canisters at once is limited. The test facility through which the connection device is to be tested consists of dummy canister, borehole, lifting part, monitoring part, and connecting device. The canister weight is determined according to the emplacement operation unit. The lifting part will be composed following wireline consisting of a crane, a wire and a winding system. The monitoring part will consist of an external monitoring system for hoists and trolleys, and an internal monitoring system for the connection device’s location, pressure, and speed. In this project, a demonstration test will be conducted in a borehole with 1km depth, 10 cm diameter provided by KAERI to verify operation in the actual drilling environment after design improvement of the connecting device. If a problem is found through the demonstration test, the problem will be improved, and an improved connection device will be tested to an extended borehole with a 2 km depth, 40 cm diameter.