Journal of Nuclear Fuel Cycle and Waste Technology

The accident at the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) resulted in the deposition of large quantities of radionuclides over parts of eastern Japan. Radioactive contaminants have been observed over a large area including forests, cities, rivers and lakes. Due to the strong adsorption of radioactive cesium by soil particles, radioactive cesium migrates with the eroded soil, follows the surface flow paths, and is delivered downstream of population-rich regions and eventually to coastal areas. In this study, we developed a model to simulate the transport of contaminated sediment in a watershed hydrological system and this model was compared with observation data from eroded soil observation instruments located at the Korea Atomic Energy Research Institute. Two methods were applied to analyze the soil particle size distribution of the collected soil samples, including standardized sieve analysis and image analysis methods. Numerical models were developed to simulate the movement of soil along with actual rainfall considering initial saturation, rainfall infiltration, multilayer and rain splash. In the 2019 study, a numerical model will be used to add rainfall shield effect by trees, evaporation effect and shield effects of surface water. An eroded soil observation instrument has been installed near the Wolsong nuclear power plant since 2018 and observation data are being continuously collected. Based on these observations data, we will develop the numerical model to analyze long-term behavior of radionuclides on land as they move from land to rivers, lakes and coastal areas.

Sediment transport,Soil erosion numerical modeling,Rainfall,Image processing analysis,Sieve analysis

In accordance with the decommissioning plan for the Kori Unit 1 NPP, the reactor coolant system will be chemically decontaminated as soon as possible after permanent shutdown. This study developed the chemical decontamination process though the development project of decontamination technology of reactor coolant system and dismantled equipment for NPP decommissioning, which has been carried out since 2014. In this study, Oxidation/reduction process was conducted using system decontamination process development equipment of lab scale and was divided into unit and continuous processes. The optimal process time was derived from the unit process, and decontamination agent and the number of process were derived through the continuous processes. Through the unit process, the oxidation process took 5 hours and the reduction process took 4 hours. As optimum decontamination agent, the oxidizing agent was 200 mg·L^-1 Permanganic acid + 200 mg·L^-1 Nitric acid and the reducing agent was 2000 mg·L^-1 Oxalic acid. In the case of the number of processes, all oxide films were removed during the two-cycle chemical decontamination process of STS304 and SA508. In the case of Alloy600, all oxide films were removed when chemical decontamination was performed for three cycles or more.

System decontamination,Oxidation,Reduction,Permanganic acid,Oxalic acid

High-level radioactive waste (HLW) such as spent fuel is inevitably produced when nuclear power plants are operated. A geological repository has been considered as one of the most adequate options for the disposal of HLW, and it will be constructed in host rock at a depth of 500~1,000 meters below ground level with the concept of an engineered barrier system (EBS) and a natural barrier system. The compacted bentonite buffer is one of the most important components of the EBS. As the compacted bentonite buffer is located between disposal canisters with spent fuel and the host rock, it can restrain the release of radionuclides and protect canisters from the inflow of groundwater. Because of inflow of groundwater into the compacted bentonite buffer, it is essential to investigate soil-water characteristic curves (SWCC) of the compacted bentonite buffer in order to evaluate the entire safety performance of the EBS. Therefore, this paper conducted laboratory experiments to analyze the SWCC for a Korean Ca-type compacted bentonite buffer considering dry density, confined or unconfined condition, and drying or wetting path. There was no significant difference of SWCC considering dry density under unconfined condition. Furthermore, it was found that there was higher water suction in unconfined condition that in confined condition, and higher water suction during drying path than during wetting path.

Engineered barrier system,Compacted bentonite buffer,Soil-water characteristic curve

As candidates for accident-tolerant fuels, ceramic microcell fuels, which are distinguished by their peculiar microstructures, are being developed; these fuels have UO₂ grains surrounded by cell walls. They contribute to nuclear fuel safety by retention of fission products within the UO₂ pellet, reducing rod pressure and incidence of SCC failure. Cesium, a hazardous fission product in terms of amount and radioactivity, can be captured by chemical reactions with ceramic cell materials. The capture-ability of cesium therefore depends on the thermodynamics of the capturing reaction. Conversely, compositional design of cell materials should be based on thermodynamic predictions. This study proposes thermodynamic calculations to evaluate the cesium capture-ability of three ceramic microcell compositions: Si-Ti-O, Si-Cr-O and Si-Al-O. Prior to the calculations, the chemical and physical states of the cesium and the cell materials were defined. Then, the reactivity was evaluated by calculating the cesium potential (ΔG_Cs) and oxygen potential (ΔG_O2) under simulated LWR circumstances of normal operation. Based on the results, cesium capture is expected to be spontaneous in all cell compositions, providing a basis for the compositional design of ceramic microcell fuels as well as a facile way for evaluating cesium capture.

Accident-tolerant fuel,Ceramic microcell UO₂,Cesium capturing,HSC Chemistry,Cesium potential,Oxygen potential

The paper presents the results of a multi-criteria comparative evaluation of 12 feasible Russian nuclear energy deployment scenarios with thermal and fast reactors in a closed nuclear fuel cycle. The comparative evaluation was performed based on 6 performance indicators and 5 different MCDA methods (Simple Scoring Model, MAVT / MAUT, AHP, TOPSIS, PROMETHEE) in accordance with the recommendations elaborated by the IAEA/INPRO section. It is shown that the use of different MCDA methods to compare the nuclear energy deployment scenarios, despite some differences in the rankings, leads to well-coordinated and similar results. Taking into account the uncertainties in the weights within a multi-attribute model, it was possible to rank the scenarios in the absence of information regarding the relative importance of performance indicators and determine the preference probability for a certain nuclear energy deployment scenario. Based on the results of the uncertainty/sensitivity analysis and additional analysis of alternatives as well as the whole set of graphical and attribute data, it was possible to identify the most promising nuclear energy deployment scenario under the assumptions made.

Thermal reactors,Fast reactors,Closed nuclear fuel cycle,MCDA,Uncertainty

A decontamination technology for radioactive liquid wastes was newly developed and hypothetically applied to the liquid waste management system (LWMS) of the nuclear power plant (NPP) to evaluate its decontamination efficacy for the purpose of the fundamental reduction of spent resins. The basic principle of the developed technology is to convert major radionuclide ions in the liquid wastes into inorganic crystal minerals via chemical or biological techniques. In a laboratory batch experiment, the biological method selectively removed more than 80% of cesium within 24 hours, and the chemical method removed more than 95% of cesium. Other major nuclides (Co, Ni, Fe, Cr, Mn, Eu), which are commonly present in nuclear radioactive liquid wastes, were effectively scavenged by more than 99%. We have designed a module including the new technology that could be hypothetically installed between the reverse osmosis (R/O) package and the organic ion-exchange resin in the LWMS of the APR1400 reactor. From a technical evaluation for the virtual installation, we found that more than 90% of major radionuclides in the radioactive liquid wastes were selectively removed, resulting in a large volume reduction of spent resins. This means that if the new technology is commercialized in the future, it could possibly provide drastic cost reduction and significant extension of the life of resins in the management of spent resins, consequently leading to delay the saturation time of the Wolsong repository.

Spent Resins,Radionuclide,Cesium,APR1400,Ion-Exchange Resin

As the present single-layer repository concept requires too large an area for the site of the repository, a multi-layer repository concept has been suggested to improve the disposal density. The effects of the excavation damaged zone around the multi-layer repository constructed in the deep host rock on the temperature distribution in the repository were analyzed. For the thermal analysis of the multi-layer repository, the hydrothermal model was used to consider the resaturation process occurring in the buffer, backfill and rock. The existence of an excavation damaged zone has a significant effect on the temperature distribution in the repository, and the maximum peak temperatures of double-layer and triple-layer repositories can rise to 7℃ and 12℃, respectively depending on the size of the excavation damaged zone and the degree of decrease of the thermal conductivity. The dominant factor affecting the peak temperature in the multi-layer repository is the decrease of thermal conductivity in the excavation damaged zone, and the excavation damaged zone formed around the deposition hole has more significant effects on the peak temperature than does the excavation damaged zone formed around the disposal tunnel.

Multi-layer repository,Peak temperature,Excavation damaged zone,Hydrothermal model,Spent fuel

The decommissioning of nuclear power plants should be prepared by quantitative and qualitative risk assessment. Radiological and non-radiological hazards arising during decommissioning activities must be assessed to ensure the safety of decommissioning workers and the public. Decommissioning experiences by U.S. operators have mainly focused on deterministic risk assessment, which is standardized by the U.S. Nuclear Regulatory commission (NRC) and focuses only on the consequences of risk. However, the International Atomic Energy Agency (IAEA) has suggested an alternative to the deterministic approach, called the risk matrix technique. The risk matrix technique considers both the consequence and likelihood of risk. In this study, decommissioning stages, processes, and activities are organized under a work breakdown structure. Potential accidents in the decommissioning process of NPPs are analyzed using the composite risk matrix to assess both radiological and non-radiological hazards. The levels of risk for all potential accidents considered by U.S. NPP operators who have performed decommissioning were estimated based on their consequences and likelihood of events.

Decommissioning,Risk assessment,Risk matrix,Potential accidents,Hazards

The decommissioning of one nuclear power plant in a multi-unit nuclear power plant (multi-unit NPP) site may pose radiation exposure risk to decommissioning workers. Thus, it is essentially required to evaluate the exposure dose of decommissioning workers of operating multi-unit NPPs nearby. The ENDOS program is a dose evaluation code developed by the Korea Atomic Energy Research Institute (KAERI). As two sub-programs of ENDOS, ENDOS-ATM to anticipate atmospheric transport and ENDOS-G to calculate exposure dose by gaseous radioactive effluents are used in this study. As a result, the annual maximum individual dose for decommissioning workers is estimated to be 2.31×10^-3 mSv·y^-1, which is insignificant compared with the effective dose limit of 1 mSv·y^-1 for the public. Although it is revealed that the exposure dose of operating multi-unit NPPs does not result in a significant impact on decommissioning workers, closer examination of the effect of additional exposure due to actual demolition work is required. The calculation method of this study is expected to be utilized in the future for planned decommissioning projects in Korea. Because domestic NPPs are located in multi-unit sites, similar situations may occur.

Multi-unit NPP,Decommissioning workers,Exposure dose,Impact of multi-unit NPP

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, ³⁶Cl and ⁴¹Ca are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of ³⁶Cl and ⁴¹Ca in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of ³⁶Cl and ⁴¹Ca using the activation calculation code ORIGEN2.

Bioshielding concrete,Radionuclide inventory,³⁶Cl,⁴¹Ca

Zr electrorefining is demonstrated herein using Zirlo tubes in a chloride-fluoride mixed molten salt in the presence of AlF₃. Cyclic voltammetry reveals a monotonic shift in the onset of metal reduction kinetics towards positive potential and an increase in intensity of the additional peaks associated with Zr-Al alloy formation with increasing AlF₃ concentration. Unlike the galvanostatic deposition mode, a radial plate-type Zr growth is evident at the top surface of the salt during Zr electrorefining at a constant potential of −1.2 V. The diameter of the plate-type Zr deposit gradually increases with increasing AlF₃ concentration. Scanning electron microscopy–energy-dispersive X-ray spectroscopy (SEM-EDX) and X-ray photoelectron spectroscopy (XPS) analyses for the plate-type Zr deposit show that trace amount of Al is incorporated as Zr-Al alloys with different chemical compositions between the top and bottom surface of the deposit. Addition of AlF₃ is effective in lowering the residual salt content in the deposit and in improving the current efficiency for Zr recovery.

Zirconium recovery,Electrorefining,Cladding hull,Molten salt,Zr-Al alloys

The current status and prospect of decommissioning technology development at KAERI are reviewed here. Specifically, this review focuses on four key technologies: decontamination, remote dismantling, decommissioning waste treatments, and site remediation. The decontamination technologies described are component decontamination and system decontamination. A cutting method and a remote handling method together with a decommissioning simulation are described as remote dismantling technologies. Although there are various types of radioactive waste generated by decommissioning activities, this review focuses on the major types of waste, such as metal waste, concrete waste, and soil waste together with certain special types, such as high-level and high-salt liquid waste, organic mixed waste, and uranium complex waste, which are known to be difficult to treat. Finally, in a site remediation technology review, a measurement and safety evaluation related to site reuse and a site remediation technique are described.

Nuclear power plant,Decommissioning,Decontamination,Radioactive waste,Site remediation

Currently, the development of evaluation technology for vibration and shock load characteristics and spent nuclear fuel structural integrity under normal conditions of transport is being conducted in the Republic of Korea. This is the first such research conducted in the Republic of Korea and, thus, previous international studies need to be investigated and will be referred to in the ongoing project. Before 2000, several studies related to measurement of vibration and shock loads on spent nuclear fuel were conducted in the US. US national research institutes conducted uniaxial fuel assembly shaker tests, concrete block tests, and multi-axis fuel assembly tests between 2009 and 2016. In 2017, multi-modal transportation tests including road, sea, and rail transport were also performed by research institutes from the US, Spain and the Republic of Korea. Therefore, test preparation procedures, acceleration and strain measurement results, and finite-element and multi-body dynamics analysis were investigated. Based on the measured strain data, the preliminary conclusion was obtained that the measured strain was too small to cause damage to spent nuclear fuel rods. However, this conclusion is a preliminary conclusion that only reviews part of the results; a detailed review is being conducted in the US. The investigation of international studies on spent nuclear fuel structural integrity tests considering vibration and shock loads under normal conditions of transport in the US will be useful data for the project being conducted in the Republic of Korea.

Normal condition of transport,Vibration and shock,Spent nuclear fuel,Structural integrity,Transportability

For design and performance assessment of a high-level radioactive waste (HLW) disposal system, it is necessary to understand the characteristics of coupled thermo-hydro-mechanical (THM) behavior. However, in previous studies for the Korean Reference HLW Disposal System (KRS), thermal analysis was performed to determine the spacing of disposal tunnels and interval of disposition holes without consideration of the coupled THM behavior. Therefore, in this study, TOUGH2-MP/FLAC3D is used to conduct THM modeling for performance assessment of the Korean Reference HLW Disposal System (KRS). The peak temperature remains below the temperature limit of 100℃ for the whole period. A rapid rise of temperature caused by decay heat occurs in the early years, and then temperature begins to decrease as decay heat from the waste decreases. The peak temperature at the bentonite buffer is around 96.2℃ after about 3 years, and peak temperature at the rockmass is 68.2℃ after about 17 years. Saturation of the bentonite block near the canister decreases in the early stage, because water evaporation occurs owing to temperature increase. Then, saturation of the bentonite buffer and backfill increases because of water intake from the rockmass, and bentonite buffer and backfill are fully saturated after about 266 years. The stress is calculated to investigate the effect of thermal stress and swelling pressure on the mechanical behavior of the rockmass. The calculated stress is compared to a spalling criterion and the Mohr-Coulumb criterion for investigation of potential failure. The stress at the rockmass remains below the spalling strength and Mohr-Coulumb criterion for the whole period. The methodology of using the TOUGH2-MP/FLAC3D simulator can be applied to predict the long-term behavior of the KRS under various conditions; these methods will be useful for the design and performance assessment of alternative concepts such as multi-layer and multi-canister concepts for geological spent fuel repositories.

High-level radioactive waste,Korean reference HLW disposal system (KRS),Performance assessment,Coupled thermo-hydro-mechanical behavior,TOUGH2-MP/FLAC3D

Disposal nonconformity of radioactive wastes refers to radioactive wastes that need to be treated, solidified and packaged during operation or decommissioning of NPPs, and are typically exemplified by particulate radioactive wastes with dispersion characteristics. These wastes include the dried powders of concentrated wastes generated in the process of operating NPPs, slurry and sludge, various powdered wastes generated in the decommissioning process (crushed concrete, decontamination sludge, etc.), and fine radioactive soil, which is not easy to decontaminate. As these particulate wastes must be packaged so that they become non-dispersive, they are solidified with solidification agents such as cement and polymer. If they are treated using existing solidification methods, however, the volume of the final wastes will increase. This drawback may increase the disposal cost and reduce the acceptability of disposal sites. Accordingly, to solve these problems, this study investigates the pelletization of particulate radioactive wastes in order to reduce final waste volume.

Disposal nonconformity radioactive waste,Powdered waste,Pelletization,Volume reduction,Solidification,Disposal

Regulatory agency and licensee are preparing for the site restoration of Kori unit 1, the first commercial NPP in Korea, scheduled for 2031. Developing regulatory guidelines and strategies is essential for effective restoration work. Unfortunately, Korea does not have experience of site restoration of commercial NPPs. Therefore, it is important to review cases from experienced countries to establish a strategy and regulatory standards. The U.S. has had numerous soil remediation experiences using RESRAD and MARSSIM. However, formalized evaluation methodologies for subsurface soil have not yet been established in MARSSIM. This survey focused on subsurface soil remediation by reviewing the five decommissioned NPPs under regulation of the US NRC. Overall process of remediating a contaminated subsurface soil and groundwater was reviewed to identify considerations and lessons that could be applicable in Korea. In addition, an applied methodology for evaluation of contaminated subsurface soil and related major issues between regulatory agency and licensees were reviewed in detail to support establishment of remediation strategy for Kori unit 1.

Site restoration,Soil remediation,Subsurface soil,RESRAD,MARSSIM

A user-friendly modeling interface is developed for a process-based total system performance assessment framework (APro) specialized for a generic geological disposal system for high-level radioactive waste. The APro modeling interface is constructed using MATLAB, and the operator splitting scheme is used to combine COMSOL for simulation of multiphysics and PHREEQC for the calculation of geochemical reactions. As APro limits the modeling domain to the generic disposal system, the degree of freedom of the model is low. In contrast, the user-friendliness of the model is improved. Thermal, hydraulic, mechanical and chemical processes considered in the disposal system are modularized, and users can select one of multiple modules: “Default process” and multi “Alternative process”. APro mainly consists of an input data part and calculation execution part. The input data are prepared in a single EXCEL file with a given format, and the calculation part is coded using MATLAB. The final results of the calculation are created as an independent COMSOL file for further analysis.

Radioactive waste,Geological disposal,Total system performance assessment (TSPA),Process-based,Modeling interface

Radioactive materials are released into the air and deposited on the surface soil after a nuclear accident. Radionuclides deposited in soil are transported by precipitation to nearby environments and contaminate the surface water system. Basic data on surface watershed and soil erosion models have been collected and analyzed to evaluate the behavior of radionuclides deposited on surface soil after a nuclear accident. Data acquisition and analysis in aquatic environment were performed to investigate the physical characteristics and variation of biota in rivers and lakes of the Nakdong river area near the Wolsong nuclear power plant. For these purposes, a digital map, and hydrological, water quality and biota data were gathered and a systematic database (DB) was constructed in connection with them. Constructed aquatic DB will be supplied and used in surface watershed and soil erosion models for investigation of long-term movement of radionuclides in adsorptive form in surface soil. Finally, basic data and established models will be utilized for general radiological impact assessment in aquatic environment.

Radionuclides,Surface soil deposition,Aquatic environment data analysis,Construction of DB

Radioactive contaminant from a nuclear facility moves to the ecosystem by run-off or groundwater flow. Among the two mechanisms, contaminant plume through a river can be easily detected through a surface water monitoring system, but radioactive contaminant transport in groundwater is difficult to monitor because of lack of information on flow path. To understand the contaminant flow in groundwater, understanding of the geo-environment is needed. We suggest a method to decide on monitoring location and points around an imaginary nuclear facility by using the results of site characterization in the study area. To decide the location of a monitoring well, groundwater flow modeling around the study area was conducted. The results show that, taking account of groundwater flow direction, the monitoring well should be located at the downstream area. Also, monitoring sections in the monitoring well were selected, points at which groundwater moves fast through the flow path. The method suggested in the study will be widely used to detect potential groundwater contamination in the field of oil storage caverns, pollution by agricultural use, as well as nuclear use facilities including nuclear power plants.

Radioactive contaminant,Nuclear facility,Groundwater flow,Site characterization,Monitoring well,Monitoring methodology

Since the 1970s, aircraft crash accidents have been considered as one of the severest external events that should be evaluated for license application of nuclear reactors. After the 9.11 terrorist attacks, many countries have performed safety assessment against intentional or targeted aircraft crashes into nuclear related facilities. In some countries, assessment against targeted aircraft crash was enforced by regulation and considered an important task for license approval. Safety assessment against aircraft crash is a technically difficult task and many countries manage R&D programs to improve its reliability. In this paper, regulations of many countries regarding safety assessment against aircraft crash are summarized, separating regulations for accident aircraft crash and those for targeted aircraft crash. Research performed in various countries on safety assessment of nuclear facility against aircraft crash are summarized, with a focus on spent nuclear fuel dry storage facilities.

Spent nuclear fuel,Dry storage,Aircraft crash,Safety assessment,High speed impact

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

Spent PWR Fuel,Pyroprocessing,Filter waste,Salt waste,Metal waste

Decommissioning is a critical issue in Korea. Although compared with the operation of nuclear power plants the release of radioactive materials during decommissioning is not expected to be significant, residents should always be protected from radiation exposure. To manage this effectively, Annual Release obj!ectives (ARO) and Annual Release Limits (ARL) were derived from dose standards in the NSSC Notice and dose limit for the public. Based on meteorological data for the three years from 2008 to 2010 in the Shin Kori nuclear power plant site, atmospheric dispersion and ground deposition factors of gaseous effluent were evaluated using the XOQDOQ computer code. The exposure dose was evaluated using the ENDOS-G computer code. Because of differences in radiological sensitivity according to age groups, the results of Annual Release obj!ectives (ARO) and Annual Release Limits (ARL) showed significant differences depending on the radionuclides. The evaluation methodology of this study will provide meaningful information for radioactive effluent management for decommissioning of nuclear power plants.

Gaseous effluents,Decommissioning,Annual Release obj!ectives (ARO),Annual Release Limits (ARL)

The A-KRS (Advanced Korean Reference Disposal System) is the disposal concept for pyroprocessed waste, which has been developed by the Korea Atomic Energy Research Institute. In this disposal concept, the amount of high-level radioactive waste is minimized using pyrochemical process, called pyroprocessing. The produced pyroprocessed waste is then solidified in the form of monazite ceramic. The final product of ceramic wastes will be disposed of in a deep geological repository. By the way, the decay heat is generated due to the radioactive decay of fission products and raises the temperature of buffer materials in the near field of radioactive waste repository. However, the buffer temperature must be kept below 100℃ according to the safety regulation. Usually, the temperature can be controlled by variation of the canister interdistance. However, KAERI has modelled thermal analysis under the boundary condition, where the waste canisters are in direct contact with each other. Therefore, a reliable temperature analysis in the disposal system may fail because of unknown thermal resistence values caused by the spatial gap between waste canisters. In the present work, we have performed thermal analyses considering the gap between heating elements and canisters at the beginning of canister loading into the radioactive waste repository. All thermal analyses were performed using the COMSOL software package.

A-KRS,Heat transfer modeling,Vertical disposal hole,Contact condition,Clearance

The detection performances of the NaI(Tl), LaBr₃(Ce) and CeBr₃ scintillation detectors, which can be used to rapidly evaluate the major artificial radionuclides deposited on the soil surface in a nuclear accident or radiological emergency, were compared. Detection performance was assessed by calculating the minimum detectable activity (MDA). The detection efficiency of each detector for artificial radionuclides was semi-empirically determined using mathematical modelling and point-like sources having certified radioactivity. The background gamma-ray energy spectrum for MDA evaluation was obtained from relatively wide and flat grassland, and the MDA values of each detector for the major artificial radionuclides that could be released in nuclear accidents were calculated. As a result, the relative MDA values of each detector regarding surface deposition distribution at normal environmental radiation level were evaluated as high in the order of the NaI(Tl), LaBr₃(Ce), and CeBr₃ detectors. These results were compared based on each detector’s intrinsic and measurement environment background, detection efficiency, and energy resolution for the gamma-ray energy region of the radionuclide of interest.

In-situ gamma-ray spectrometry, Scintillation detector, NaI(Tl), LaBr₃(Ce), CeBr₃, MDA

Suction of unsaturated bentonite buffers is a very important input parameter for hydro-mechanical performance assessment and design of an engineered barrier system. This study analyzed suction measurement techniques and constitutive models of unsaturated porous media reported in the literature, and suggested suction measurement techniques and constitutive models suitable for bentonite buffer in an HLW repository. The literature review showed the suction of bentonite buffer to be much higher than that of soil, as measured by total suction including matric suction and osmotic suction. The measurement methods (RH-Cell, RH-Cell/Sensor) using a relative humidity sensor were suitable for suction measurement of the bentonite buffer; the RH-Cell /Sensor method was more preferred in consideration of the temperature change due to radioactive decay heat and measurement time. Various water retention models of bentonite buffers have been proposed through experiments, but the van Genuchten model is mainly used as a constitutive model of hydro-mechanical performance assessment of unsaturated buffers. The water characteristic curve of bentonite buffers showed different tendencies according to bentonite type, dry density, temperature, salinity, sample state and hysteresis. Selection of water retention models and determination of model input parameters should consider the effects of these controlling factors so as to improve overall reliability.

Suction,Water characteristic curve,Water retention model,Unsaturated bentonite,Buffer,High-level waste repository

Based on the 8^th Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, ²³⁵U enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial 235U enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with ²³⁵U enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with ²³⁵U enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

Spent fuel,Discharge burnup,Cooling time,Deep geological repository

The composition of the reduced fuel produced in the electrolytic reduction process of pyroprocessing affects the concentration change of UCl₃, an important operating variable of the electrorefining process. In this study, we examined the concentration change of UCl₃ in the electrorefiner according to the content of TRU and RE elements in the reduced fuel and the concentration of Li₂O introduced in the electrorefiner accompanied with the reduced fuel. Considering only the TRU and RE elements, the concentration of UCl₃ decreased with increasing the number of electrorefining operation batch. In order to operate one campaign (20 batches) of electrorefining process, it was found that additional injection of UCl₃ should be conducted more than 3 times. On the other hand, the concentration of UCl3 in the electrorefiner changed significantly depending on the concentration of Li₂O and, accordingly the number of operable electrorefining batches decreased rapidly, showing that the concentration of Li₂O is an important operating variable in electrorefining. Therefore, the results of this study show that to maintain the concentration of UCl₃ in the electrorefiner, the operation mode should be set by taking into account the effect of Li₂O as well as the TRU and RE elements contained in the reduced fuel.

Electrolytic reduction,Reduced fuel,Concentration of Li₂O,Electrorefining,Concentration of Ucl₃

This technical note introduces a newly-proposed concept of deep horizontal drillhole disposal of spent nuclear fuel, and considers how it can be applied in the Korean environment. This disposal concept, in which high-level radioactive waste is disposed in deep horizontal drillholes installed with directional drilling technique, is expected to have great advantages over the existing deep mined repository concept in economics and safety. Since this concept is still at the idea level, however, it is necessary for worldwide expert groups to demonstrate its safety and performance. In addition, the development of guidelines by the regulatory body should be supported. The Korean circumstances, which include a narrow territory and a high population density, as well as the amount of spent nuclear fuel, make the NIMBY (Not In My Back Yard) phenomenon very strong and the siting conditions difficult. Under these conditions, if the disposal section of deep horizontal drillhole concept can be located at the continental shelf, with a stable environment, rather than in a coastal land area, it is expected to alleviate the psychological anxiety of the local community and stakeholders. Moreover, even when constructing a centralized deep mined repository in the future, it is necessary to consider locating the repository in the continental shelf.

Spent nuclear fuel,High-level radioactive waste,Deep horizontal drillhole disposal,Deep mined repository,Deep borehole disposal

Waste characterisation associated with nuclear site decommissioning relies on radiochemical analysis of a diverse range of sample types, requiring extensive validation of analytical techniques using matrix-matched materials. The absence of relevant reference materials has hindered robust method development and validation. The paper discusses how method validation in support of nuclear waste characterisation can be achieved without using reference materials. The key stages in an analytical procedure are evaluated and a multi-stage approach is proposed with the ultimate aim of determining an operational envelope for an analytical procedure.

Decommissioning,Radioactive waste characterisation,Validation,Radioanalytical techniques

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^{-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.}

CANDU Spent nuclear fuel,Transportation cask,Burnup,Minimum cooling period,Radiation Shielding

Large amounts of concrete waste are likely to arise from the decommissioning of a Kori-1 nuclear power plant. Several studies have been conducted on decommissioning concrete waste in recent decades, however, they have been limited to contaminated concrete issues or were small pilot-scale experiments. This study constructed two industrial-scale models of on-site concrete waste management for clean as well as contaminated concrete. To evaluate the performance of both the models, simulations were conducted using the Flexsim software. The concrete particle size distribution of Kori-1 and concrete processor properties based on widely used construction equipment were used as sources of input data for the simulations. It was observed that it may take over two years to complete the on-site concrete management processes owing to the performance of existing processors. In addition, it was demonstrated that it is essential to identify bottlenecks in the system and enhance the performance of the relevant processors to avoid delays of the decommissioning schedule. Our results suggest that this novel approach can contribute to developing schedules or expediting delayed activities in the Kori-1 decommissioning project.

Kori-1,Nuclear power plant decommissioning,Concrete waste management,Logistics,Discrete event,Continuous simulation

Based on spent fuels characteristics from domestic nuclear power plants and a disposal scenario from the current basic plan for high-level radioactive waste management, an improved disposal system has been proposed that enhances disposal efficiency and economic effectiveness compared to the existing disposal system. For this purpose, two disposal canisters concepts were derived from the length of the spent fuel generated from the nuclear power plants. In the disposal scenario, the acceptable amount of decay heat for each disposal container was determined, taking into account the discharge and disposal times of spent fuels in accordance with the current basic plan. Based on the determined decay heat of the two types of disposal canisters and the associated disposal system, thermal stability analyses were performed to confirm their suitability to the proposed disposal system design requirement and disposal efficiency assessment. The results of this study confirm 20% reduction in the disposal area and 20% increase in disposal density for the proposed disposal system compared to the existing system. These results can be used to establish a spent fuel management policy and to design a viable commercial disposal system.

Spent fuels,Disposal Canister,Deep geological disposal system,Disposal scenario,Decay heat,Engineered barrier,Thermal requirement,Disposal efficiency

The application of ion crystallization technology was considered as a way to increase the operating efficiency and improve the operating performance of a liquid waste management system (LWMS) in the Advanced Power Reactor 1400 (APR 1400). Although ion crystallization technology has not been practically applied to Nuclear Power Plants (NPPs) until now, a previous experimental study demonstrated that it is possible to selectively remove at least 95% of various nuclide ions present in the liquid radioactive waste of NPPs. We reviewed the possibility of applying ion crystallization technology to the existing LWMS by applying the nuclide removal rate of ion crystallization technology and prepared a way to improve the existing LWMS in the APR 1400. Furthermore, we determined the optimized application location of ion crystallization technology in the existing LWMS by considering decontamination characteristics of the ion crystallization technology and the existing LWMS design features and operating experiences. The application of ion crystallization technology to the liquid waste collection tank, where liquid radioactive materials are collected, will have the least impact on the existing design while providing the greatest improvement. It is expected that the application of ion crystallization technology to the current APR 1400 or new NPPs would increase the operating efficiency of the LWMS and result in an improvement of system performance.

LWMS,Reverse Osmosis,Demineralizer Module (Cation Bed, Mixed Bed)

Decontamination wastewater generated from the HyBRID decontamination process of the primary system in a nuclear power plant contains impurities such as sulfate ions, metal ions containing radioactive nuclides, and hydrazine (carcinogenic agent). For this reason, it is necessary to develop a technology to remove these impurities from the wastewater to a safe level. In this study, it has been conducted to remove the impurities using a decontamination wastewater surrogate, and a treatment process of the HyBRID decontamination wastewater has been established. The performance and applicability of the treatment process have been verified through 1 L scale of replicates and a pilot scale (300 L/batch) test.

Nuclear power plant,Primary system,HyBRID decontamination process,Decontamination wastewater,Treatment process

After the permanent shutdown of K1 in 2017, decommissioning processes have attracted great attention. According to the current decommissioning roadmap, the dismantling of the activated components of K1 may start in 2026, following the removal of its spent fuel. Since the reactor vessel (RV) and reactor vessel internal (RVI) of K1 contain massive components and are relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. For achieving maximum efficiency nuclear waste management processes for K1, we present activation analysis of the segmentation process and waste classification of the RV and RVI components of K1. For RVI, the active fuel regions and some parts of the upper and lower active regions are classified as intermediate-level waste (ILW), while other components are classified as low-level waste (LLW). Due to the RVI’s complex structure and high activation, we suggest various underwater segmentation techniques which are expected to reduce radiation exposure and generate approximately nine ILW and nineteen very low level waste (VLLW)/LLW packages. For RV, the active fuel region and other components are classified as LLW, VLLW, and clearance waste (CW). In this case, we suggest in-situ remote segmentation in air, which is expected to generate approximately forty-two VLLW/LLW packages.

K1,Waste management,Decommissioning,Segmentation,Reactor vessel,Internal

Japan enacted the “Designated Radioactive Waste Final Disposal Act” for the geological disposal of high-level radioactive waste in 2000 and began the site selection process. However, no local government wanted to participate in the siting process. Therefore, in 2015, the Japanese government developed a new site selection process during the literature survey step, and on June 28, 2017 they published a “Nationwide Map of Scientific Features for Geological Disposal” created with the aim of promoting public participation from local governments. This map illustrated the requirements and criteria to be considered in the early or conceptual stages of securing a geological repository and was useful for improving public understanding and exchanging opinions with local governments by analyzing the suitability of different geological disposal sites.

Geological disposal,High-level waste,Siting,Requirement and criteria,Nationwide map of scientific features

Engineering scale in-situ demonstration of engineered barrier system performance (In-DEBS) based on the advanced Korean reference repository system (A-KRS) was planned from 2012. The whole system was designed and manufactured by the Korea Atomic Energy Research Institute (KAERI) and installed in the KAERI Underground Research Tunnel (KURT) in May, 2016. Operation started in July, 2016 and quality thermal, hydro, and mechanical (THM) data have been produced. This is the first engineering scale in-situ test of a deep geological repository system performance for high level radioactive waste disposal in Korea. The main purpose of this test is the development of verification technology for engineered barrier system (EBS) performance of deep geological repository system by demonstration of manufacturing techniques for large scale disposal canisters and buffer, demonstration of installation of unit module of disposal system, acquisition of long-term THM behavior data around domestic bentonite EBS, validation of the modeling techniques for coupled THM behavior, and development and demonstration of monitoring techniques for major parameters in buffer. With the successful installation and operation of In-DEBS, we have secured demonstration and modeling techniques for coupled THM behavior of EBS using in-situ test data, which form the core of Research, Development, and Demonstration (RD&D) for High-level Radioactive Waste (HLW) disposal technology, as well as securing manufacturing and installation technology of engineering scale EBS. It is expected that the results of the In-DEBS test will provide important information for technology development of a future real scale experiments of Korean repository system.

In-DEBS,Engineered barrier system,Buffer,KURT,Deep geological disposal,Coupled THM behavior,In-situ test

In order to understand behaviors of engineered barriers for geological disposal of high-level radioactive wastes, an In-DEBS (In-situ Demonstration of Engineered Barrier System) test was planned. Geological and hydrogeological data of KURT (KAERI Underground Research Tunnel) site were obtained and analyzed to select an appropriate position for the In-DEBS test. Proper research galleries in the extended section of KURT were decided on for the test using regional data of the KURT site. Borehole investigations were carried out to examine local properties of the research galleries in order to characterize groundwater flow and evaluate groundwater inflow at the research galleries. The data from the borehole investigations provided the basis for selecting the position of a large-diameter borehole for the In-DEBS test as well as other practical requirements for implementing the test in the KURT conditions. From in-situ investigation of a wall of the large-diameter borehole, groundwater inflow and distribution of hydraulic conductivity at the wall were evaluated. The hydrogeological data and analysis results in this study can be applied to understand and simulate the evolution of the behavior of In-DEBS and nearby rocks.

In-DEBS,EBS (Engineered Barrier System),Geological disposal,Hydrogeology,Groundwater inflow

The obj!ective of this study is to propose a design for the In-DEBS (In-situ Demonstration of Engineered Barrier System) and suggest optimized conditions for manufacturing engineered barriers. The In-DEBS design was made from the AKRS (Advanced Korea Reference disposal System of HLW), based on performance experience carried out at KAERI (Korea Atomic Energy Research Institute) and from literature reviews. The optimal manufacturing conditions were derived from the test production with regard to the canister and the bentonite buffer. The performance of the manufactured engineered barriers was also evaluated in advance of in-situ application. In-DEBS was designed at 1/2.3 scale of AKRS. An aluminum mold-type heater of 4.2 kW was used to simulate the heating from nuclear fuel. A new methodology using a floating die press and CIP (Cold Isostatic Press) was introduced to fabricate a reliable engineering-scale buffer block for in-situ experiments, leading to a homogeneous and uniform distribution in the densities of the buffer blocks. From the test results, it was found that a floating die press of 40 MPa and CIP of 50 MPa are necessary to satisfy AKRS design premises for the buffer blocks (dry density > 1.6 kg·cm-3). With regard to the THM sensor installations, CNC (Computer Numerical Control) was applied to shape the sensor holes.

High-level nuclear waste disposal,THM coupled behavior,In-situ test,In-DEBS,Buffer,Canister

The paper describes the process of developing and implementing In-DEBS (In-situ Demonstration of Engineered Barrier System) for analyzing thermal-hydraulic-mechanical (THM) coupled behavior characteristics of an engineered barrier system in a high-level waste disposal system. The engineered barrier system is composed of a disposal canister, buffer, and near field rock. In-DEBS is designed to analyze the THM coupled behavior due to the inflow of underground water and the heat of disposal canisters in the buffer and near field rock. Since it is difficult to assemble the buffer block while installing sensors and heaters in a disposal hole, the buffer blocks were assembled from the outside. OBPA (One-Body Pre-Assembly) assembled with buffer blocks, sensors and heater requires special transport equipment to accurately insert it into the disposal hole because the total weight is about three tones. In this study, the fully assembled buffer block was inserted by developing a crane exclusively for In-DEBS, which can be moved into an accurate position by lifting more than five tones. To analyze the THM coupled behavior of the near field rock, 4 test holes were cored around In-DEBS and a total of 40 temperature sensors and 5 pore pressure sensors were installed in the test hole. A total of 185 sensors for temperature, relative humidity, pressure and displacement sensors were installed in the In-DEBS. All sensors were connected to a multichannel measuring system that stores field data in real time.

In-DEBS,THM coupled behavior,Engineered barrier system,Buffer,Near-field rock

In 2012, an in-situ Demonstration of Engineered Barrier System (In-DEBS) was planned with the aim of investigating the coupled thermo-hydro-mechanical (THM) behavior of both the engineered barrier system (EBS) and the natural barrier system (NBS) at the advanced Korean reference repository system (A-KRS). The relevant in-situ experiment has been in operation in Research Gallery 3 of the KAERI Underground Research Tunnel (KURT) since May 2016. For analysis of the in-situ experiment data and investigation of the THM coupled behavior, it is necessary to understand thermal, hydraulic, and mechanical properties of KURT granite and bentonite buffer material. This paper summarizes the thermal, hydraulic, and mechanical properties of KURT granite and Gyeongju bentonite presented in previous studies. Moreover, it suggests various models capable of providing input parameters for interpretative modeling of In-DEBS experiment or for predictive numerical simulations of In-DEBS or A-KRS.

KAERI Underground Research Tunnel (KURT),In-situ Demonstration of Engineered Barrier System (In-DEBS),Engineered Barrier System (EBS),Natural Barrier System (NBS),THM properties

This paper is intended to describe the field experimental data obtained from the initial operation of In-DEBS (Engineering scale in-situ demonstration of engineered barrier system), which is installed to analyze the THM (thermal, hydro, and mechanical) coupled behavior of engineered barrier system. On the top, which is plugged with concrete, heat is generated during the curing process of concrete. The heat affects the buffer and the near-field rock. The temperature of the buffer and the near field rock becomes less and less farther away from the plugging. Sensors and measurement system were verified using these phenomena. To protect the measurement system, the heater temperature was gradually increased to 100℃ over one week, and the temperature of the heater surface was maintained at the target temperature. After heater operation, the temperature of the buffer reached the steady state in about four days, and the relative humidity increased and gradually decreased after a certain period of time. Also, the total pressure of the buffer was increased by swelling due to groundwater saturation. After the heater was started, it was observed that the temperature and displacement of the near-field rock continued to increase.

In-DEBS,Engineered barrier system,Buffer,Near field rock,Acquisition,Sensor

Coupled thermo-hydraulic (TH) behavior is expected owing to a decay heat from high-level radioactive waste and water intake from rockmass in high-level radioactive waste disposal system. And coupled thermo-hydro-mechanical (THM) behavior is expected in high-level radioactive waste disposal system owing to thermal stress and swelling pressure generated by changes in temperature and saturation. Since 2016, an in-situ demonstration of an engineered barrier system (In-DEBS) at the KAERI Underground Research Tunnel (KURT) has been performed to investigate the coupled THM behavior in the high-level radioactive waste disposal system and nearby rockmass. In this study, a numerical simulation using TOUGH-MP/FLAC3D and an axisymmetric model was carried out to analyze the in-situ experimental data and estimate the coupled THM behavior. In addition, input parameters and THM models used in this study are evaluated and the practical applicability of TOUGH2-MP/FLAC3D to THM simulations is examined by comparison of in-situ data with calculation results. Temperature and relative humidity evolutions are well captured in the numerical simulation. In addition, even though the calculated displacement is relatively small, its evolution trend shows agreement with the in-situ data. The axisymmetric model was useful for analyzing coupled THM behavior in the In-DEBS experiment however, a 3D analysis including a consideration of the effect of KURT geometry is necessary to thoroughly investigate coupled THM behavior in the rockmass. The input parameters and the THM models in this study can be applied to predict the long-term behavior in the In-DEBS experiment and to carry out a performance assessment of the coupled THM behavior in a high-level radioactive waste disposal system.

KAERI Underground Research Tunnel (KURT),In-situ Demonstration of Engineered Barrier System (In-DEBS),Coupled Thermo-Hydro-Mechanical (THM) behavior,Numerical modeling,TOUGH2-MP/FLAC3D

A simplified flowsheet for pyroprocessing commercial spent fuel is proposed in which the only salt treatment step is actinide drawdown from electrorefiner salt. Actinide drawdown can be performed using a simple galvanic reduction process utilizing the reducing potential of gadolinium metal. Recent results of equilibrium reduction potentials for Gd, Ce, Nd, and La are summarized. A description of a recent experiment to demonstrate galvanic reduction with gadolinium is reviewed. Based on these experimental results and material balances of the flowsheet, this new variant of the pyroprocessing scheme is expected to meet the objectives of minimizing cost, maximizing processing rate, minimizing proliferation risk, and optimizing the utilization of geologic repository space.

Salt waste,Pyroprocessing,Drawdown

Tritium and 14C are two most important radionuclides released from nuclear facilities to the environment, and 14C contributes dominant radiation dose to the population around nuclear power plants. This paper presents an overview of the production, pathway, species and levels of tritium and 14C in nuclear facilities, mainly nuclear power plants. The methods for sampling and collection of different species of tritium and 14C in the discharge gas from the stack in the nuclear facilities, atmosphere of the nuclear facilities and environment are presented, and the features of different methods are reviewed. The on-line monitoring methods of gaseous tritium and 14C in air and laboratory measurement methods for sensitive determination of tritium and 14C in collected samples, water and environmental solid samples are also discussed in detailed. Meanwhile, the challenges in the determination and speciation analysis of tritium and 14C are also highlighted.

Tritium,Carbon-14,Source term,Production,Monitoring,Speciation analysis

Radioactive substances, especially 137Cs discharged in the course of Nuclear Power Plant Accident or maintenance of power plants, cause contamination of the soil. For habitation of residents and reuse of industrial land, it is inevitably necessary to decontaminate the soil. This study examines a soil washing process that has actually been used for washing of radioactivecontaminatedoil. The soil washing process uses a washing agent to weaken surface tension of the soil and cesium, separating cesium from the soil. In this study, in order to raise the efficiency of the process, a flocculating agent was added to the washing water to remove fine soil and cesium. The cesium concentrations before and after applying the flocculating agent to cesium solution were measured through ICP-OES. When using 0.1 g of J-AF flocculating agent in the experiment, the maximum Cs removal performance was approximately 88%; the minimum value was 67%. Species combinations between cesium and soil were predicted using Visual MINTEQ Code; the ability to reuse the washing water or not, and the removal rate of the fine soil, determined via measurement of the turbidity after applying the flocculating agent, were determined.

Soil decontamination,Cesium,Soil washing,Visual MINTEQ

For the immobilization of high-radioactive nuclides such as Cs and Sr by high-temperature thermal decomposition, this study was carried out to investigate the phase transformation with calcined temperature by using TGA (thermogravimetric analysis) and XRD (X-ray diffraction) in the Cs-adsorbed CHA (chabazite zeolite of K type)-Cs and CHA-PCFC (potassium cobalt ferrocyanide)-Cs zeolite system, and Sr-adsorbed 4A-Sr and BaA-Sr zeolite system, respectively. In the case of CHA-Cs zeolite system, the structure of CHA-Cs remained at up to 900℃ and recrystallized to pollucite (CsAlSi2O6) at 1,100℃ after undergoing amorphous phase at 1,000℃. However, the CHA-CFC-Cs zeolite system retained the CHA-PCFC-Cs structure up to 700℃, but its structure collapsed in 900~1,000℃, and then transformed to amorphous phase, and recrystallized to pollucite at 1,100℃. In the case of 4A-Sr zeolite system, on the other hand, the structure of 4A-Sr maintained up to 700℃ and its phase transformed to amorphous at 800℃, and recrystallized to Sr-feldspar (SrAl2Si2O8, hexagonal) at 900℃ and to SrAl2Si2O8 (triclinic) at 1,100℃. However, the BaA-Sr zeolite system structure began to break down at below 500℃, and then transformed to amorphous phase in 500~900℃ and recrystallized to Ba/Sr-feldspar (coexistence of Ba0.9Sr0.1Al2Si2O8 and Ba0.5Sr0.5Al2Si2O8) at 1,100℃. All of the above zeolite systems recrystallized to mineral phase through the dehydration/(decomposition)→ amorphous → recrystallization with increasing temperature. Although further study of the volatility and leachability of Cs and Sr in the high-temperature thermal decomposition process is required, Cs and Sr adsorbed in each zeolite system are mineralized as pollucite, Sr-feldspar and Ba/Sr-feldspar. Therefore, Cs and Sr seen to be able to completely immobilize in the calcining wasteform/(solidified wasteform).

Zeolite thermal decomposition,Thermogravimetry analysis,X-ray diffraction,Pollucite,Leucite,Sr-feldspar,Ba/Sr-feldspar,Calcination,Recrystallization,Immobilization

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of UCl3 into an oxide form using K2CO3 and an electrodeposition of NdCl3 into a metal form in LiCl-KCl-UCl3-NdCl3 system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of UCl3 disappeared in the LiCl-KCl salt when the injection of K2CO3 reached theoretical equivalent for the conversion, and then NdCl3 was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.

Spent nuclear fuel,Pyroprocessing,LiCl-KCl,UCl3,NdCl3,Separation

With the significant increase in spent ion-exchange resin generation, to meet the requirements of Waste Acceptance Criteria (WAC) of the Wolsong disposal facility in Korea, blending is considered as a method for enhancing disposal options for intermediate level waste from nuclear reactors. A mass balance formula approach was used to enable blending process with an appropriate mixing ratio. As a result, it is estimated around 44.3% of high activity spent resins can be blended with the overall volume of low activity spent resins at a 1:7.18 conservative blending ratio. In contrast, the reduction of high activity spent resins is considered a positive solution in reducing the amount of spent resins stored. In an economic study, the blending process has been proven to lower the disposal cost by 10% compared to current APR1400 treatment. Prior to commencing use of this blending method in Korea, coordinated discussion, and safety and health assessment should be undertaken to investigate the feasibility of fitting this blending method to national policy as a means of waste predisposal processing and management in the future.

Spent ion-exchange resins,Blending method,Mass balance approach,Mixing ratio,Waste acceptance criteria

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria (34 kgf·cm-2) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

Cement solidification,Radioactive waste,Compressive strength,Volume expansion,Waste acceptance criteria

In Korea, a huge amount of radioactive concrete waste will be generated through decommissioning of nuclear facilities in the near future; therefore, optimum technology for the treatment of concrete waste should be reviewed thoroughly and the future direction of technology development should be discussed. In this paper, many domestic and foreign examples of generation of radioactive concrete waste were pieced together and the characteristics of radioactive concrete waste were examined. Moreover, we reviewed trends in technology development by analyzing the examples of various studies and practical applications of treatment technologies, such as mechanical decontamination, chemical decontamination, volume reduction, recycling and solidification, and also tried to understand the limitations of existing technologies and determine a direction for technical improvement.

Decommissioning,Radioactive concrete,Contamination,Activation,Treatment technology,Volume reduction

Kori unit 1, the oldest commercial nuclear power plant in South Korea, was permanently shut down in June 2017. There are a lot of things to consider in decommissioning nuclear power plants, and one of them is the radiological environmental impact assessment. Performed to promote the health and safety of residents around the nuclear power plant, radiological environmental impact assessment aims to confirm that off-site radiological dose from radioactive material released from the facility does not exceed the regulatory criteria. There are three main parts of environmental impact assessment: predecommissioning environmental monitoring, environmental monitoring during decommissioning, and impact on nearby residents. At present, although the Korea Nuclear Safety Act stipulates that radiological environmental impact assessment resulting from decommissioning should be carried out, the details have not been specified. Therefore, this paper compares and analyzes guidelines for evaluation of radiological environmental impacts of nuclear power plants overseas, and presents a draft on the assessment of radiological dose resulting from decommissioning according to the Korean situation.

Radiological environmental impact assessment,Decommissioning plan,Environmental report

The objective of this study is to propose a new methodology to fabricate a reliable engineering-scale buffer block, which shows homogeneous and uniform distribution in buffer block density, for in-situ experiments. In this study, for the first time in Korea, floating die press and CIP (Cold Isostatic Press) are applied for the manufacture of an engineering-scale bentonite buffer. The optimized condition and field applicability are also evaluated with respect to the method of manufacturing the buffer blocks. It is found that the standard deviation of the densities obtained decreases noticeably and that the average dry density increases slightly. In addition, buffer size is reduced by about 5% at the same time. Through the test production, it is indicated that the stress release phenomenon decreases after the application of the CIP method, which leads to a reduction in crack generation on the surface of the buffer blocks over time. Therefore, it is confirmed that the production of homogeneous buffer blocks on industrial scale is possible using the method suggested in this study, and that the produced blocks also meet the design conditions for dry density of buffer blocks in the AKRS (Advanced Korea Reference Disposal System of HLW).

High-level radwaste disposal,Engineering-scale,Buffer,CIP,Floating die press,In-DEBS

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP';s storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process).

Dual purpose metal cask, Dry storage, Operation process, Vacuum drying, He backfilling

The guarantee of the performance of the engineered barriers in a geological repository is very important for the long-term safety of disposal as well as the efficient design of the repository. Therefore, the performance of the engineered barriers under repository condition should be demonstrated by in-situ experiments conducted in an underground research laboratory. This article provides a review of the major in-situ experiments that have been carried out over the past several decades at underground research laboratories around the world to validate the performance of engineered barriers of a repository, as well as their results. In-situ experiments to study the coupled thermal-hydraulicmechanical behavior of the engineered barrier system used to simulate the post-closure performance of the repository are analyzed as a priority. In addition, in-situ experiments to investigate the performance of the buffer material under a real repository environment have been reviewed. State-of-the art in-situ validations of the buffer-concrete interaction, and the installation of the buffer, backfill and plug, as well as characterization of the near-field rock and the corrosion of the canister materials are, also performed.

In-situ experiment,Engineered barriers,Geological repository,Validation,Demonstration

As an approach for estimation of the droplet size in the molten salt-liquid metal extraction process, a droplet formation experiment at room temperature was conducted to evaluate the applicability of the Scheele-Meister model with water-mercury system as a surrogate that is similar to the molten salt-liquid metal system. In the experiment, droplets were formed through the nozzle and the droplet size was measured using a digital camera and image analysis software. As nozzles, commercially available needles with inner diameters (ID) of 0.018cm and 0.025 cm and self-fabricated nozzles with 3-holes (ID: 0.0135 cm), 4-holes (ID: 0.0135 cm), and 2-holes (ID: 0.0148 cm) were used. The mercury penetration lengths in the nozzles were 1.3 cm for the needles and 0.5 cm for the self-fabricated nozzles. The droplets formed from each nozzle maintained stable spherical shape up to 20 cm below the nozzle. The droplet size measurements were within a 10% error range when compared to the Scheele-Meister model estimates. The experimental results show that the Scheele-Meister model for droplet size estimation can be applied to nozzles that stably form droplets in a water-mercury system.

Molten salt-liquid metal extraction,Water-mercury,Droplet,Scheele-meister model,Room temperature

As part of planning for waste minimization, decontamination foam has been considered as a potential application for the cleaning of radioactive contaminant. In this study, we synthesized silica particles to improve foam stability by varying synthesis parameters. Cetyltrimethylammonium bromide (CTAB) was found to influence the stability of the decontamination foam. The reason was that higher interaction between SiO2 nanoparticles and surfactant at the air-water interface in aqueous solution is beneficial for foam stability. CTAB can also be used as an additive for the aggregation of silica nanoparticles. In the separation of SiO2 nanoparticles, CTAB plays a critical role in the nanoparticles flocculation because of the charge neutralization and hydrophobic effects of its hydrocarbon tails.

Decontamination foam,Foam stability,SiO2,Cetyltrimethylammonium bromide,Silica separation

Currently, the Korea Atomic Energy Research Institute (KAERI) is planning to build the Ki-Jang Research Reactor (KJRR) in Ki-Jang, Busan. It is important to safely dispose of low-level radioactive waste from the operation of the reactor. The most efficient way to treat radioactive waste is cement solidification. For a radioactive waste disposal facility, cement solidification is performed based on specific waste acceptance criteria such as compressive strength, free-standing water, immersion and leaching tests. Above all, the leaching test is important to final disposal. The leakage of radioactive waste such as 137Cs causes not only regional problems but also serious global ones. The cement solidification method is simple, and cheaper than other solidification methods, but has a lower leaching resistance. Thus, this study was focused on the development of cement solidification for an enhancement of cesium leaching resistance. We used Zeolite and Loess to improve the cesium leaching resistance of KJRR cement solidification containing simulated KJRR liquid waste. Based on an SEM-EDS spectrum analysis, we confirmed that Zeolite and Loess successfully isolated KJRR cement solidification. A leaching test was carried out according to the ANS 16.1 test method. The ANS 16.1 test is performed to analyze cesium ion concentration in leachate of KJRR cement for 90 days. Thus, a leaching test was carried out using simulated KJRR liquid waste containing 3000 mg·L-1 of cesium for 90 days. KJRR cement solidification with Zeolite and Loess led to cesium leaching resistance values that were 27.90% and 21.08% higher than the control values. In addition, in several tests such as free-standing water, compressive strength, immersion, and leaching tests, all KJRR cement solidification met the waste acceptance or satisfied the waste acceptance criteria for final disposal.

Ki-Jang Research Reactor,KJRR,Cement solidification,Acceptance criteria. Zeolite,Loess

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like 99Tc and 129I contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of 99Tc and 129I in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.

Spent fuel,Long-lived fission products,99Tc,129I,Nuclear transmutation,Nuclear reactor

To correctly predict the neutron behavior based on diffusion calculations, it is necessary to adopt well-specified boundary conditions using suitable diffusion approximations to transport boundary conditions. Boundary conditions such as the zero net-current, the Marshak, the Mark, the zero scalar flux, and the Albedo condition have been used extensively in diffusion theory to approximate the reflective and vacuum conditions in transport theory. In this paper, we derive and analyze these conditions to prove their mathematical validity and to understand their physical implications, as well as their relationships with one another. To show the validity of these diffusion boundary conditions, we solve a sample problem. The results showthat solutions of the diffusion equation with these well-formulated boundary conditions are very close to the solution of the transport equation with transport boundary conditions.

Neutrons,Transport,Diffusion,Boundary conditions,Nuclear fuel cycle

Nodal transport methods are proposed for solving the simplified even-parity neutron transport (SEP) equation. These new methods are attributed to the success of existing nodal diffusion methods such as the Polynomial Expansion Nodal and the Analytic Function Expansion Nodal Methods, which are known to be very effective for solving the neutron diffusion equation. Numerical results show that the simplified even-parity transport equation is a valid approximation to the transport equation and that the two nodal methods developed in this study also work for the SEP transport equation, without conflict. Since accuracy of methods is easily increased by adding node unknowns, the proposed methods will be effective for coarse mesh calculation and this will also lead to computation efficiency.

Nodal theory,Transport equation,Diffusion equation,Simplified Even-Parity transport equation,Nuclear fuel cycle

Bird screen meshes are installed at the air inlet and outlet ducts of spent fuel storage casks to inhibit the intrusion of debris from the external environment. The presence of these screens introduces an additional resistance to air flow through the ducts. In this study, a porous media model was developed to simplify the bird screen meshes. CFD analyses were used to derive and verify the flow resistance factors for the porous media model. Thermal analyses were carried out for concrete storage cask using the porous media model. Thermal tests were performed for concrete casks with bird screen meshes. The measured temperatures were compared with the analysis results for the porous model. The analysis results agreed well with the test results. The analysis temperatures were slightly higher than the test temperatures. Therefore, the reliability and conservatism of the analysis results for the porous model have been verified.

Spent fuel,Concrete storage cask,Thermal analysis,Thermal test,Bird screen,Natural convection,Heat removal,Porous media

To overcome the low mechanical strength and corrosion behavior of a carbon steel canister at high temperature condition of a deep borehole, SiC ceramics were studied as an alternative material for the disposal canister. In this paper, a design concept for a SiC canister, along with an outer stainless steel container, was proposed, and its manufacturing feasibility was tested by fabricating several 1/3 scale canisters. The proposed canister can contain one PWR assembly. The outer container was also prepared for the string formation of SiC canisters. Thermal conductivity was measured for the SiC canister. The canister had a good thermal conductivity of above 70 W·m-1·K-1 at 100℃. The structural stability was checked under KURT environment, and it was found that the SiC ceramics did not exhibit any change for the 3 year corrosion test at 70℃. Therefore, it was concluded that SiC ceramics could be a good alternative to carbon steel in application to deep borehole disposal canisters.

SiC ceramics,Deep borehole disposal (DBD) system,Disposal canister,Nuclear waste disposal,High level waste

A literature review on the effects of high temperature and radiation on radiation shielding concrete in Spent Fuel Dry Storage is presented in this study with a focus on concrete degradation. The general threshold is 95℃ for preventing long-term degradation from high temperature, and it is suggested that the temperature gradient should be less than 60℃ to avoid crack generation in concrete structures. The amount of damage depends on the characteristics of the concrete mixture, and increases with the temperature and exposure time. The tensile strength of concrete is more susceptible than the compressive strength to degradation due to high temperature. Nuclear heating from radiation can be neglected under an incident energy flux density of 1010 MeV·cm-2·s-1. Neutron radiation of >1019 n·cm-2 or an integrated dose of gamma radiation exceeding 1010 rads can cause a reduction in the compressive and tensile strengths and the elastic moduli. When concrete is highly irradiated, changes in the mechanical properties are primarily caused by variation in water content resulting from high temperature, volume expansion, and crack generation. It is necessary to fully utilize previous research for effective technology development and licensing of a Korean dry storage system. This study can serve as important baseline data for developing domestic technology with regard to concrete casks of an SF (Spent Fuel) dry storage system.

Spent fuel,Dry storage,Concrete,Damage,Radiation degradation,Thermal degradation

In this study, to replace the ‘J-slot joint’, a joint device between a disposal canister and an emplacement jig in Deep Borehole Disposal process, a novel joint device was designed and tested. The novel joint device was composed of a wedge on top of a disposal canister and a hook box at the end of a winch system. The designed joint device had merits in that it can recombine an emplaced canister freely without the replacement of the joint component. Moreover, it can be applied to various emplacement jigs such as drill pipes, wire-lines, and coiled tubing. To demonstrate the designed joint device, the joint device (Φ 110 mm, H 148 mm), a twin canister string (Φ 140 mm, H 1,105 mm), and a water tube (Φ 150 mm, H 1,500 mm) as a borehole model were manufactured at 1/3 scale. As deployment muds, Na-type bentonite (MX-80) and Ca-type (GJ II) bentonite muds were prepared at solid contents of 7wt% and 28wt%, respectively. The manufactured joint device showed good performance in pure water and viscous muds, with an operation speed of 10 m·min-1. It was concluded that the newly developed joint device can be used for the emplacement and retrieval of a deep disposal canister, below 3~5 km, in the future.

Deep borehole disposal,Emplacement method,Joint device,Deployment mud

During the seven years from 2009 to 2016, PWR SNF (spent nuclear fuel) transportation and storage systems suitable for domestic conditions were developed by the government to cope with the saturation of wet storage capacity in NPPs. One of the developed systems is a multipurpose metal cask applicable for transportation/storage; the other is a concrete cask dedicated to storage. Efficient cask technologies were secured utilizing the characteristics and experience of relevant industrial, academic and research institutes. Technological independence was also achieved through several patent registrations of research outcomes. To prepare for a rapid increase of demand in the near future, technology transfer of secured patents and technologies to the domestic industry was carried out twice in the years of 2016 and 2017.

Transportation/storage system,Dual purpose metal cask,Concrete storage cask,Integrity evaluation,Safety test

In order to understand the long-term behavior of radionuclides in granite environments, geochemical behavior characteristics of uranium in granitic host rock of KURT (KAERI Underground Research Tunnel) were investigated by dissolution experiment with different reaction time and solutions. In the dissolution experiment, significantly increased dissolution levels of uranium from granite powder samples were identified during the reaction time of 0~10 days for reaction solutions (UD-CO3 and UD-Bg) containing a large amount of CO3 2-. On the other hand, significantly increased dissolution levels of uranium were also identified for reaction solutions containing Na and Ca after 60 days. Dissolution of uranium continuously increased in reaction solutions of UD-CO3 (44.61 μg·L-1), UD-Bg (41.01 μg·L-1), UD-Na (26.87μg·L-1), UD-Ca (20.26 μg·L-1), UD-CaSi (17.03 μg·L-1), and UD-Si (10.47 μg·L-1) in the experimental period of ~270 days. However, after day 270, dissolution of uranium showed a decreasing tendency. This is thought to have occurred because existing uranium in granite samples reached the limit of dissolution by interaction with reaction solutions. Concentrations of dissolved uranium and points of maximum concentration value were found to differ depending on the CO3 2- presence in the mixed reaction solution and on the geochemical type of the water. It is estimated that differences in the reaction rate between the granite sample and the reaction solution are due to the influence of dissolved ions in the reaction solution.

High-level radioactive waste,KURT,Granite,Uranium,Dissolution,Uranyl-carbonate complex

Under a pyro-processing concept, an electrolytic reduction process has been developed to reduce uranium oxide in molten salt by electrochemical means as a part of spent fuel treatment process development. Accordingly, a model based on electrochemical theory is required to design a reactor for the electrolytic reduction process. In this study, a 1D model based on the phase-field theory, which explains phase separation behaviors was developed to simulate electrolytic reduction of uranium oxide. By adopting parameters for diffusion of oxygen elements in a pellet and electrochemical reaction rate at the surface of the pellet, the model described the behavior of inward reduction well and revealed that the current depends on the internal diffusion of the oxygen element. The model for the electrolytic reduction is expected to be used to determine the optimum conditions for large scale reactor design. It is also expected that the model will be applied to simulate the integration of pyro-processing.

Pyro-processing,Electrolytic reduction,Modeling,Phase-field

Anthropogenic radioactive noble gases formed by nuclear fission are significant indicators used to monitor the nuclear activity of neighboring countries. In particular, radioactive xenon, owing to its abundant generation and short half-life, can be used to detect nuclear testing, and radioactive krypton has been used as a tracer to monitor the reprocessing of nuclear fuels. Released radioactive noble gases are in the atmosphere at infinitesimal amounts due to their dilution in the air and their short half-life decay. Therefore, to obtain reliable and significant data when performing measurement of noble gases in the atmosphere, the minimum detectable activity (MDA) for noble gases should be defined as low as possible. In this study, the MDA values for radioactive xenon and krypton were theoretically obtained based on the BfS-IAR system by collecting both noble gases simultaneously. In addition, various MDA methods, confidence level and analysis conditions were suggested to reduce and optimize MDA with an assessment of the factors affecting MDA. The current investigation indicated that maximizing the pretreatment efficiency and performance maintenance of the counter were the most important aspects for Xe. In the case of Kr, since sample activities are much higher than those of Xe, it is possible to change the target MDA or to simplification of the analysis system.

85Kr,133Xe,Radioactive noble gas,Minimum detectable activity (MDA),Nuclear detection

In this study, we prepared Li-K-Cd alloy, which meets the requirement of eutectic ratio of Li:K, to maintain the operating temperature of the drawdown process at 500℃ and to achieve the reuse of LiCl-KCl molten salt. The prepared Li-K-Cd alloys were added to LiCl-KCl salt bearing U and Nd at 500℃ to investigate the removal of UCl3 in the salt. The reduction of UCl3 in the salt was examined by measuring the OCP value of salt and analyzing the salt composition by ICP-OES. Reduction was also visually confirmed by change of salt color from dark purple to white. The experimental results reveal that the prepared Li-K-Cd alloy has reductive extractability for UCl3 in salt. By improving the preparation method, the Li-K-Cd alloy can be applied to the drawdown process.

Liquid metal extraction,Li-K-Cd alloy,Uranium chloride,LiCl-KCl eutectic salt

A subcritical facility must ensure nuclear criticality safety under all circumstances. For this purpose, it is essential to have a procedure to validate that calculated values do not exceed upper subcritical limit (USL), determined by quantifying the bias and uncertainty. However, there are several validation methodologies of nuclear criticality safety and these can yield different USL. Therefore, it is necessary to analyze the validity of the methodologies to establish one methodology that can provide the most appropriate USL. In this study, two documents, a guide for validation of nuclear criticality safety calculational methodology (NUREG/CR-6698) and a criticality benchmark guide for light water reactor fuel in transport and storage package (NUREG/CR-6361), are compared and analyzed. In particular, the methodology in NUREG/CR-6361 is applied to the USLSTATS code. However, the analysis results show that the methodology in NUREG/CR-6698 is more appropriate, for several reasons. This is applied to decision of USL to design casks using SCALE code version 6.1.

Nuclear criticality safety,Cask,SCALE6.1,Upper subcritical limit (USL)

The decommissioning of nuclear power plants is generally executed in five steps, including preparation, decontamination, cutting/demolition, waste disposal and environmental restoration. So, for efficient decommissioning of nuclear power plants, worker safety, effects compared to cost, minimization of waste, possibility of reuse, etc., shall be considered. Worker safety and measurement technology shall be secured to exert optimal efficiency of nuclear power plant decommissioning work, for which accurate measurement technology for systems and devices is necessary. Typical In-Situ methods for decommissioning of nuclear plants are CZT, Gamma Camera and ISOCS. This study used ISOCS, which can be applied during the decommissioning of a nuclear power plant site without collecting representative samples, to take measurements of the S/G Water Chamber. To validate the measurement values, Microshield and the GEANT4 code was used as the actual method were used for modeling, respectively. The comparison showed a difference of 1.0×101 Bq, which indicates that it will be possible to reduce errors due to the influence of radiation in the natural environment and the precision of modeling. Based on the research results of this paper, accuracy and reliability of measurement values will be analyzed and the applicability of the direct measurement method during the decommissioning of NPPs will be assessed.

Decommissioning,In-situ technique,ISOCS,MonteCarlo code,GEANT4

A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste. Since the heat generated from spent nuclear fuel in a disposal canister is released to the surrounding buffer materials, the thermal properties of the buffer material are very important in determining the entire disposal safety. Especially, since thermal expansion can cause thermal stress to the intact rock mass in the near-field, it is very important to evaluate thermal expansion characteristics of bentonite buffer materials. Therefore, this paper presents a thermal expansion coefficient prediction model of the Gyeongju bentonite buffer materials which is a Ca-bentonite produced in South Korea. The linear thermal expansion coefficient was measured considering heating rate, dry density and temperature variation using dilatometer equipment. Thermal expansion coefficient values of the Gyeongju bentonite buffer materials were 4.0~6.0×10-6/℃. Based on the experimental results, a non-linear regression model to predict the thermal expansion coefficient was suggested and fitted according to the dry density.

Compacted bentonite buffer materials,Linear thermal expansion coefficient,Regression analysis

The applicability of room temperature ionic liquids (RTILs), 1-alkyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([Cnmim] [Tf2N]), was investigated for the extraction of Am(Ⅲ) and Eu(Ⅲ) from nitric acid using n-octyl(phenyl)-N,N-diisobutyl carbamoylmethyl phosphine oxide (CMPO) and tri-n-butylphosphate (TBP) as extractants. The distribution ratios of Am(Ⅲ) and Eu(Ⅲ) in CMPO-TBP/ [Cnmim][Tf2N] were measured as a function of various parameters such as the concentrations of nitric acid, CMPO, and TBP. The results were compared with those obtained in CMPO-TBP/n-dodecane (n-DD). With comparable concentrations of the extractants, the distribution ratios obtained with RTILs were much higher than those obtained with n-DD. It was observed that the extraction efficiency was less for Eu(Ⅲ) than for Am(Ⅲ). The extraction of Am(Ⅲ) and Eu(Ⅲ) decreased with increases in the feed acidity for all three RTILs. The results suggest that the extraction of Am(Ⅲ) and Eu(Ⅲ) by CMPO in RTILs from nitric acid proceeds through the cation-exchange mechanism. The distribution ratios of Am(Ⅲ) and Eu(Ⅲ) increased with increases in the concentration of CMPO for all three RTILs. A linear regression analysis of the extraction data resulted in a straight line with a slope of about 3, suggesting the involvement of 3 molecules of CMPO during the extraction process.

Ionic liquid,Imidazolium,Solvent extraction,Distribution ratio,Molecular organic solvent,Complexing agent

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

Mixed-spectrum,Reactor physics,Neutron transport simulation,Nonproliferation,Long-lived reactor

The ability to recover the nearly limitless supply of uranium contained within the world’s oceans would provide supply security to uranium based fuel cycles. Therefore, in addition to U.S. national laboratories conducting R&D on a system capable of harvesting seawater uranium, a number of collaborative university partners have developed alternative technologies to complement the national laboratory scheme. This works summarizes the systems analysis of such novel uranium recovery technologies along with their potential impacts on seawater uranium recovery. While implementation of some recent developments can reduce the cost of seawater uranium by up to 30%, other researchers have sought to address a weakness while maintaining cost competitiveness.

Seawater uranium,Front end nuclear fuel cycle,Actinide chemistry,Systems analyses

Korea';s first commercial nuclear power plant, Kori Unit 1, was permanently shut down on June 18, 2017, after 40 years of successful operation. Kori Unit 1 plans to construct a waste treatment facility in the turbine building prior to commencement of dismantling in earnest. Various radioactive wastes are decontaminated, disassembled, cut and melted in the waste treatment facility and sent to the radioactive waste repository. The proportion of metal radioactive waste in dismantled waste is about 70%, of which large metal radioactive waste is mainly generated in the primary circuit and has high radioactivity, so radiation exposure must be managed during disassembly. In this study, the steam generators are selected as large metal radioactive waste, the exposure doses of the dismantling workers are calculated using RESRAD-RECYCLE code and the methods for reducing the exposure doses are suggested.

Decontamination,Decommissioning,Steam generator,Exposure dose calculation

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

고리1호기,계통제염,계통제염범위,잔열제거펌프

When considering the long-term safety assessment of spent-nuclear fuel management, americium is one of the most radio-toxic actinides. Although spectroscopic methods are widely used for the study of actinide chemistry, application of those methods to americium chemistry has been limited. Herein, we purified 241Am to obtain a highly pure stock solution required for spectroscopic studies. Quantitative and qualitative analyses of purified 241Am were carried out using liquid scintillation counting, and gamma and alpha radiation spectrometry. Highly sensitive absorption spectrometry coupled with a liquid waveguide capillary cell and time-resolved laser fluorescence spectroscopy were employed for the study of Am(Ⅲ) hydrolysis and oxalate (Ox) complexation. Am3+ ions under acidic conditions exhibit maximum absorbance at 503 nm, with a molar absorption coefficient of 424 ± 8 cm-1·M-1. Am(OH)3(s) colloidal particles formed under near neutral pH conditions were identified by monitoring the absorbance at around 506-507 nm. The formation of Am(Ox)3 3- was detected by red-shifts of the absorption and luminescence spectra of 4 and 5 nm, respectively. In addition, considerable enhancements of the luminescence intensities were observed. The luminescence lifetime of Am(Ox)3 3- increased from 23 to 56 ns, which indicates that approximately six water molecules are replaced by carboxylate ligands in the inner-sphere of the Am(Ⅲ). These results suggest that Am(Ox)3 3- is formed through the bidentate coordination of the oxalate ligands.

아메리슘,방사분석,UV-Vis 흡수 분광학,시간분해 레이저 형광 분광학,화학종 규명

This study presents a revised crackling core model for the description of UO2 sphere oxidation in air atmosphere. For close reproduction of the sigmoid behavior exhibited in UO2 to U3O8 conversion, the fragmentation effect contributing to the increased reactive surface area and the concept of variable grain conversion time were considered in the model development. Under the assumptions of two-step successive reaction of UO2 → U3O7 → U3O8 and final grain conversion time equivalent to ten times the initial grain conversion time, the revised model showed good agreement with the experimental data measured at 599 – 674 K and a lowest deviation when compared with Nucleation and Growth model and AutoCatalytic Reaction model. The evaluated activation energy at 100% conversion to U3O8, 57.6 kJ·mol-1, was found to be closer to the experimentally extrapolated value than to the value determined in AutoCatalytic Reaction model, 48.6 kJ·mol-1.

UO2 산화,파편화,쪼개짐심 모델,결정립

Hydrazine based reductive dissolution applied on magnetite oxide was investigated. Dissolution of Fe(Ⅱ) and Fe(Ⅲ) from magnetite takes place either by protonation, surface complexation, or reduction. Solution containing hydrazine and sulfuric acid provides hydrogen to break bonds between Fe and oxygen by protonation and electrons for the reduction of insoluble Fe(Ⅲ) to soluble Fe(Ⅱ) in acidic solution of pH 3. In terms of dissolution rate, numerous transition metal ions were examined and Cu(Ⅱ) ion was found to be the most effective to speed up the dissolution. During the cycle of Cu(I) ions to Cu(Ⅱ) ions, the released electron promoted the reduction of Fe(Ⅲ) and Cu(Ⅱ) ions returned to Cu(I) ion due to the oxidation of hydrazine. In the experimental results, the addition of a very low amount of cupric ion (about 0.5 mM) to the solution increased the dissolution rate about 40% on average and up to 70% for certain specific conditions. It is confirmed that even though the coordination structure of copper ions with hydrazine is not clear, the Cu(Ⅱ)/H+/N2H4 system is acceptable regarding the dissolution performance as a decontamination reagent.

제염,환원용해,금속산화물,원자로냉각재계통,폐기물 저감,촉매 효과

Less mature nuclear reactor technologies are characterized by a greater uncertainty due to insufficient detailed design information, operational data, cost information, etc., but the expected performance characteristics of less mature options are usually more attractive in comparison with more mature ones. The greater uncertainty is, the higher economic risks associated with the project realization will be. Within a comparative evaluation of less and more mature nuclear reactor technologies, it is necessary to apply economic risk measures to balance judgments regarding the economic performance of less and more mature options. Assessments of any risk metrics involve calculating different characteristics of probability distributions of associated economic performance indicators and applying the Monte-Carlo method. This paper considers the applicability of statistical risk measures for different economic performance indicators within a trial case study on a comparative evaluation of less and more mature unspecified LWRs. The presented case study demonstrates the main trends associated with the incorporation of economic risk metrics into a comparative evaluation of less and more mature nuclear reactor technologies.

Levelized cost,Net present value,Value-At-Risk,Technology readiness

Since the permanent shutdown of Kori Unit 1 in 2017, a full-scale decommissioning project for a commercial nuclear reactor has been approaching. It is estimated that about 160,000 t of low-activity concrete waste will be produced from decommissioning of one unit of this commercial nuclear power reactor. Accordingly, it is necessary to review whether the effectiveness of the current regulatory framework for clearance waste (i.e. waste stream that meets activity concentration guidelines or dose criteria for clearance set forth in NSSC Notice No. 2017-65) can be maintained for the clearance of a bulk amount of concrete waste. In this regard, the IAEA SRS No. 44, which was used as a basis for revision of the Korean clearance regulations, is thoroughly analyzed and the radiological effects from four different clearance scenarios, along with input values and parameters derived from industrial practices in Korea, were evaluated. Though it is shown that the maximum annual dose from most recycling scenarios will be less than the clearance dose criterion for the normal scenario (i.e. an order of magnitude of 0.01 mSv·y-1), the radiation dose, estimated with conservative assumptions for the banking scenario, may exceed the above clearance dose criteria. Therefore, for safe and sustainable clearance of the bulk amount of concrete waste, it is required to diversify the concrete waste processors, perform more detailed site-specific assessment, and apply limiting conditions to the banking scenario.

자체처분,콘크리트 폐기물,해체,건축 폐기물,재활용

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significantstudies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea

건식저장용기,사용후핵연료,열화현상,염화물 응력부식균열,확률론적 모델

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers’ breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of 5 μm, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker’s body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.

해체,방사성 에어로졸,내부피폭,BiDAS,흡입

The granite melting concept, which was suggested by Gibbʼs group for the closing of a deep borehole, was experimentally checked for KURT granite. The granite melting experiments were performed in two pressure conditions of atmospheric melting with certain inorganic additives and high pressure melting formed by water vaporization. The results of atmospheric tests showed that KURT granite started to melt at a lower temperature of 1,000℃ with NaOH addition and that needle shaped crystals were formed around partially melted crystals. In high pressure tests, vapor pressure was increased by adding water with maximum pressure of about 400 bars. KURT granite was partially melted at 1,000℃ when vapor pressure was low. However, it was not melted at vapor pressures higher than 200 bars. Therefore, it was determined that high pressure with a small amount of water vapor more effectively decreased the melting point of granite. Meanwhile, high temperature and high pressure vapor caused severe corrosion of the reactor wall.

심부시추공 처분,시추공 밀봉,화강암 용융,재결정화

As an alternative to deep geological disposal technology, which is considered as a reference concept, the domestic applicability of deep borehole disposal technology for high level radioactive waste, including spent fuel, has been preliminarily evaluated. Usually, the environment of deep borehole disposal, at a depth of 3 to 5 km, has more stable geological and geo-hydrological conditions. For this purpose, the characteristics of rock distribution in the domestic area were analyzed and drilling and investigation technologies for deep boreholes with large diameter were evaluated. Based on the results of these analyses, design criteria and requirements for the deep borehole disposal system were reviewed, and preliminary reference concept for a deep borehole disposal system, including disposal container and sealing system meeting the criteria and requirements, was developed. Subsequently, various performance assessments, including thermal stability analysis of the system and simulation of the disposal process, were performed in a 3D graphic disposal environment. With these analysis results, the preliminary evaluation of the domestic applicability of the deep borehole disposal system was performed from various points of view. In summary, due to disposal depth and simplicity, the deep borehole disposal system should bring many safety and economic benefits. However, to reduce uncertainty and to obtain the assent of the regulatory authority, an in-situ demonstration of this technology should be carried out. The current results can be used as input to establish a national high-level radioactive waste management policy. In addition, they may be provided as basic information necessary for stakeholders interested in deep borehole disposal technology.

고준위폐기물,사용후핵연료,심층 동굴 처분,심부 시추공 처분,국내 적용성,처분용기,성능 평가

Technology-intensive industries can be used as a major growth engine for resource poor country in the territories. For example, in the case of Korea, nuclear power and radiation technology industry was highly developed, and it was possible to obtain national interests such as solving energy problems within the country and exporting nuclear power plants. On the other hand, there are cases where national damage is caused by erroneous governmental policy-making on technology-intensive sectors. In this study, we analyzed cases of misguided governmental policy-making for technology-intensive industry and three factors were identified. And we tried to develop a rational policymaking model using three types of allisonʼs model in combination. The results of this study are expected to be useful for rational governmental policy-making processes for technology-intensive industries.

앨리슨모형, 정부정책결정, 기술집약적 산업, 원자력 및 방사선

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with CoCl2 and K4Fe (CN)6 solutions. When CHA, with average particle size of more than 10 ㎛, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than 104 mL·g-1) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.

Radioactive cesium,Adsorption,Zeolite,Metal ferrocyanide,Impregnation

Cobalt ferrocyanide (CoFC) or nickel ferrocyanide (NiFC) magnetic nanoparticles (MNPs) were fabricated for efficient removal of radioactive cesium, followed by rapid magnetic separation of the absorbent from contaminated water. The Fe3O4 nanoparticles, synthesized using a co-precipitation method, were coated with succinic acid (SA) to immobilize the Co or Ni ions through metal coordination to carboxyl groups in the SA. CoFC or NiFC was subsequently formed on the surfaces of the MNPs as Co or Ni ions coordinated with the hexacyanoferrate ions. The CoFC-MNPs and NiFC-MNPs possess good saturation magnetization values (43.2 emu∙g-1 for the CoFC-MNPs, and 47.7 emu∙g-1 for the NiFC-MNPs). The fabricated CoFC-MNPs and NiFC-MNPs were characterized by XRD, FT-IR, TEM, and DLS. The adsorption capability of the CoFC-MNPs and NiFC-MNPs in removing cesium ions from water was also investigated. Batch experiments revealed that the maximum adsorption capacity values were 15.63 mg∙g-1 (CoFC-MNPs) and 12.11 mg∙g-1 (NiFC-MNPs). Langmuir/Freundlich adsorption isotherm equations were used to fit the experimental data and evaluate the adsorption process. The CoFC-MNPs and NiFC-MNPs exhibited a removal efficiency exceeding 99.09% for radioactive cesium from 137Cs solution (18-21 Bq∙g-1). The adsorbent selectively adsorbed 137Cs, even in the presence of competing cations.

Metal ferrocyanide,Magnetic nanoparticle,Succinic acid,Radioactive cesium

In this study, desorption characteristics of Cs from clay according to the hydrophobic alkyl chain length of the cationic surfactant were investigated. Alkyltrimethylammonium bromide was used as a cationic surfactant, and the length of the hydrophobic alkyl chain of the cationic surfactant was varied from –octyl to –cetyl. The adsorbed amount of the cationic surfactant on montmorillonite increased with the length of the hydrophobic alkyl chain, and intercalation of the cationic surfactant into the clay interlayer increased the interlayer distances. The Cs removal efficiency was also enhanced with increasing alkyl chain length, and the cationic surfactant with the cetyl group showed a maximum Cs removal efficiency of 99±2.9%.

Clay,Cs,Desorption,Surfactant,Alky chain

A gamma-ray peak of 226Ra (186.2 keV) overlaps with one of 235U (185.7 keV) in a gamma-ray spectrometry system. Though reference peaks of 235U can be used to correct the peak interference of 235U in the analysis of 226Ra, this requires a complicated calculation process and a high limit of quantitation. On the other hand, evaluating 226Ra using the correction constant in the overlapped peak can make a rapid measurement of 226Ra without the complicated calculation process as well as overcome the disadvantage in the indirect measurement of 214Bi, which means the confinement of 222Rn gas in a sample container and a time period to recover the secular equilibrium. About 93 samples with 6 species for raw-materials and by-products were prepared to evaluate the activity of 226Ra using the correction constant. The results were compared with the activity of 214Bi, which means the indirect measurement of 226Ra, to validate the method of the direct measurement of 226Ra using the correction constant. The difference between the direct and indirect measurement of 226Ra was generally below about ± 20%. However, in the case of the phospho gypsum, a large error of about 50% was found in the comparison results, which indicates the disequilibrium between 238U and 226Ra in the materials. Application results of the contribution ratio of 226Ra were below about ± 10% . The direct measurement of 226Ra using the correction constant can be an effective method for its rapid measurement of raw materials and by-products because the activity of 226Ra can be produced with a simple calculation without the consideration of the integrity of a sample container and the time period to recover the secular equilibrium.

226Ra,Gamma-ray spectrometry,Correction constant,Raw material,By-Product

The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.

Decommissioning,Metallic waste,Systems Engineering,Dose evaluation,RESRAD-RECYCLE

The result of a preliminary safety assessment that was completed by applying the radionuclide inventory calculated on the basis of available data from radioactive waste generation agencies suggested that many difficulties are to be expected with regard to disposal safety and operation. Based on the results of the preliminary safety assessment of the entire disposal system, in this paper, a unit package exceeding the safety goal is selected that occupies a large proportion of radionuclides in intermediate-level radioactive waste. We introduce restrictions on the amount of radioactivity in a way that excludes the high surface dose rate of the package. The radioactivity limit for disposal will be used as the baseline data for establishing the acceptance criteria and the disposal criteria for each disposal facility to meet the safety standards. It is necessary to draw up a comprehensive safety development plan for the Gyeongju waste disposal facility that will contribute to the construction of a Safety Case for the safety optimization of radioactive waste disposal facilities.

Inventory,Disposal facility,Radioactive waste,Disposal concentration limit,Safety Case

The pyroprocessing of spent nuclear fuel generates LiCl-KCl eutectic waste salt containing radioactive rare earth nuclides. It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste in a hot-cell facility. In this study, capture and solidification of a rare earth nuclide (Nd) in LiCl-KCl eutectic salt using an inorganic composite with a Li2OAl2O3-SiO2-B2O3 system was conducted to simplify the existing separation and solidification process of rare earth nuclides in LiCl-KCl eutectic waste salt from the pyroprocessing of spent nuclear fuel. More than 98wt% of Nd in LiCl-KCl eutectic salt was captured when the mass ratio of the composite was 0.67 over NdCl3 in the eutectic salt. The content of Nd2O3 in the Nd captured-composite reached about 50wt%, and this composite was directly fabricated into a homogeneous and chemical resistant glass waste in a monolithic form. These results will be utilized in designing a process to simplify the existing separation and solidification process.

Spent nuclear fuel,Pyroprocessing,LiCl-KCl eutectic waste salt,Capturing and solidification of rare earth nuclide,Inorganic composite with Li2O-Al2O3-SiO2-B2O3 system

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B&PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

Concrete storage cask,Flaw depth,Stress ratio,Flow stress,Welding

The interaction of U(VI) (hexavalent uranium) species with natural organic matter (NOM) in KURT (KAERI Underground Research Tunnel) groundwater is investigated using a laser spectroscopic technique. The luminescence spectra of the NOM are observed in the ultraviolet and blue wavelength regions by irradiating a laser beam at 266 nm in groundwater. The luminescence spectra of U(VI) species in groundwater containing uranium concentrations of 0.034-0.788 mg·L-1 are measured in the green-colored wavelength region. The luminescence characteristics (peak wavelengths and lifetime) of U(VI) in the groundwater agree well with those of Ca2UO2(CO3)3(aq) in a standard solution prepared in a laboratory. The luminescence intensities of U(VI) in the groundwater are weaker than those of Ca2UO2(CO3)3(aq) in the standard solution at the same uranium concentrations. The luminescence intensities of Ca2UO2(CO3)3(aq) in the standard solution mixed with the groundwater are also weaker than those of Ca2UO2(CO3)3(aq) in the standard solution at the same uranium concentrations. These results can be ascribed to calcium-U(VI)-carbonate species interacting with NOM and forming non-radiative U(VI) complexes in groundwater.

Groundwater,Natural organic matter,Uranium-carbonate species,Laser spectroscopy,Luminescence

Densities of molten salt mixtures of eutectic LiCl-KCl with UCl3, CeCl3, or LaCl3 at various concentrations (up to 13 wt%) were measured using a liquid surface displacement probe. Linear relationships between the mixture density and the concentration of the added salt were observed. For LaCl3 and CeCl3, the measured densities were significantly higher than those previously reported from Archimedes’ method. In the case of LiCl-KCl-UCl3, the data fit the ideal mixture density model very well. For the other salts, the measured densities exceeded the ideal model prediction by about 2%.

Pyroprocessing,Molten salt,Density,Fuel cycle

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidification host matrix for the noble metal fission product waste in anode sludge from the electro-refining process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confirm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

Pyrochemical process,Spent fuel cladding hull,Noble metal,Metallic waste form,Corrosion property

monstration programs for spent nuclear fuel dry storage have been carried out to produce important and confirmatory data to support safety of dry storage systems and integrity of spent nuclear fuel stored in dry condition. The US initiated the dry storage of spent nuclear fuel and has strict and explicit regulatory stipulations on the integrity of spent nuclear fuel in dry storage. The US has carried out several notable demonstration programs for the initiation and license extension of dry storage. At the very early stage of dry storage, the demonstration programs were focused on proof of the safety of dry storage systems and a demonstration project called the dry cask storage characterization project was performed for the license extension of low burn-up fuel dry storage. Currently, a demonstration program for the license extension of high burn-up fuel dry storage is under way and is expected to continue for at least 10 years. Korea has not yet begun the dry storage of PWR fuel and the US programs can be a good reference and can provide lessons to safely begin and operate dry storage in Korea. In this paper, past and current demonstration programs of the US are analyzed and several recommendations are provided for demonstration programs for the dry storage of spent nuclear fuel in Korea.

Spent nuclear fuel,Dry storage,Demonstration,Retrievability,Safety

A deep geological disposal at a depth of 500 m in stable host rock is considered to be the safest method with current technologies for disposal of spent fuels classified as high-level radioactive waste. The most important requirement is that the temperature of the bentonite buffer, which is a component of the engineered barrier, should not exceed 100℃. In Korea, the amount of spent fuel generated by nuclear power generation, which accounts for about 30% of the total electricity, is continuously increasing and accumulating. Accordingly, the area required to dispose of it is also increasing. In this study, various duplex disposal concepts were derived for the purpose of improving the disposal efficiency by reducing the disposal area. Based on these concepts, thermal analyses were carried out to confirm whether the critical disposal system requirements were met, and the thermal stability of the disposal system was evaluated by analyzing the results. The results showed that upward 75 m or downward 75 m apart from the reference disposal system location of 500 m depth would qualify for the double layered disposal concept. The results of this study can be applied to the establishment of spent fuel management policy and the design of practical commercial disposal system. Detailed analyses with data of a real disposal site are necessary.

Deep geological disposal,Reference disposal system,Double layered repository,Decay heat,Engineered barrier,Thermal analyses,Disposal area,Disposal efficiency

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.

Operational accident analysis,Safety function analysis,Hazard analysis,Risk analysis,Operational accident classification criteria

In this study, air sampling locations in the stack of the Advanced Fuel Science Building (AFSB) at the Korea Atomic Energy Research Institute (KAERI) were assessed according to the ANSI/HPS N13.1-1999 specification. The velocity profile, flow angle and 10 μm aerosol particle profile at the cross-section as functions of stack height L and stack diameter D (L/D) were assessed according to the sampling location criteria using COMSOL. The criteria for the velocity profile were found to be met at 5 L/D or more for the height, and the criteria for the average flow angle were met at all locations through this assessment. The criteria for the particle profile were met at 5 L/D and 9 L/D. However, the particle profile at the cross-section of each sampling location was found to be non-uniform. In order to establish uniformity of the particle profile, a static mixer and a perimeter ring were modeled, after which the degrees of effectiveness of these components were compared. Modeling using the static mixer indicated that the sampling locations that met the criteria for the particle profile were 5-10 L/D. When modeling using the perimeter ring, the sampling locations that met the criteria for particle profile were 5 L/D and 7-10 L/D. The criteria for the velocity profile and the average flow angle were also met at the sampling locations that met the criteria for the particle profile. The methodologies used in this study can also be applied during assessments of air sampling locations when monitoring stacks at new nuclear facilities as well as existing nuclear facilities.

Airborne radioactive effluent,Sampling location,Sampling system,CFD

The purpose of this paper is to evaluate several leading options for the management of radioactive metallic waste against a set of general criteria including safety, cost effectiveness, radiological dose to workers and volume reduction. Several options for managing metallic waste generated from decommissioning are evaluated in this paper. These options include free release, controlled reuse, and direct disposal of radioactive metallic waste. Each of these options may involve treatment of the metal waste for volume reduction by physical cutting or melting. A multi-criteria decision analysis was performed using the Analytic Hierarchy Process (AHP) to rank the options. Melting radioactive metallic waste to produce metal ingots with controlled reuse or free release is found to be the most effective option.

Decommissioning,Decontamination,Radioactive scrap metal,Melting technology VLLW,LLW,Volume reduction

The Korea Atomic Energy Research Institute (KAERI) has developed a safeguards technology for pyroprocessing based on the Safeguards-By-Design (SBD) concept. KAERI took part in a Member-State Support Program (MSSP) to establish a pyroprocessing safeguards approach. A Reference Engineering-scale Pyroprocessing Facility (REPF) concept was designed on which KAERI developed its safeguards system. Recently the REPF is being upgraded to the REPF+, a scaled-up facility. For assessment of the nuclear-material accountancy (NMA) system, KAERI has developed a simulation program named Pyroprocessing Material Flow and MUF Uncertainty Simulation (PYMUS). The PYMUS is currently being upgraded to include a Near-Real-Time Accountancy (NRTA) statistical analysis function. The Advanced Spent Fuel Conditioning Process Safeguards Neutron Counter (ASNC) has been updated as Non-Destructive Assay (NDA) equipment for input-material accountancy, and a Hybrid Induced-fission-based Pu-Accounting Instrument (HIPAI) has been developed for the NMA of uranium/transuranic (U/TRU) ingots. Currently, performance testing of Compton-suppressed Gamma-ray measurement, Laser-Induced Breakdown Spectroscopy (LIBS), and homogenization sampling are underway. These efforts will provide an essential basis for the realization of an advanced nuclear-fuel cycle in the ROK.

Pyroprocessing,Safeguards,REPF,PYMUS,ASNC,HIPAI,LIBS

A geological repository for the disposal of high-level radioactive waste is generally constructed in host rock at depths of 500~1,000 meters below the ground surface. A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste, and it can restrain the release of radionuclides and protect the canister from the inflow of groundwater. Since high temperature in a disposal canister is released to the surrounding buffer material, the thermal properties of the buffer material are very important in determining the entire disposal safety. Even though there have been many studies on thermal conductivity, there have been only few studies that have investigates the specific heat capacity of the bentonite buffer. Therefore, this paper presents a specific heat capacity prediction model for compacted Gyeongju bentonite buffer material, which is a Ca-bentonite produced in Korea. Specific heat capacity of the compacted bentonite buffer was measured using a dual probe method according to various degrees of saturation and dry density. A regression model to predict the specific heat capacity of the compacted bentonite buffer was suggested and fitted using 33 sets of data obtained by the dual probe method.

Compacted bentonite buffer,Specific heat capacity,Dual probe method,Regression analysis

In this study, pitch-based activated carbon fibers (ACFs) were modified with pyrolysis fuel oil (PFO). Carbonized ACF samples were activated at 850℃, 880℃ and 900℃. A scanning electron microscope (SEM) and a BET surface area apparatus were employed to evaluate the indoor radon removal of each sample. Among three samples, the BET surface area and micropore area of ACF880 recorded the highest value with 1,420 m2∙g-1 and 1,270 m2∙g-1. Moreover, ACF880 had the lowest external surface area and BJH adsorption cumulative surface area of pores with 151 m2∙g-1 and 35.5 m2∙g-1. This indicates that satisfactory surface area depends on the appropriate temperature. With the above scope, ACF880 also achieved the highest radon absorption rate and speed in comparison to other samples. Therefore, we suggest that the optimum activation temperature for PFO containing ACFs is 880℃ for effective indoor radon adsorption.

Activated carbon fiber,Pitch,Radon,Adsorption,Radioactive waste

This study was conducted to predict and evaluate the uncertainty of safety after closure of the second phase surface disposal facility of the Gyeongju intermediate and low level repository in Korea. In this study, four scenarios are developed considering both intact and degraded states of multi-layered covers and disposal containers; also, the fluid flow by a rainfall into the disposal facility is simulated. The rainfall conditions are implemented based on the monthly average data of the past 30 years (1985~2014); the simulation period is 300 years, the management period regulated by institutional provisions. As a result of the evaluation of the basic scenario, in which the integrity of both of the containers and the covers is maintained, it was confirmed that penetration of rainfall does not completely saturate the inside of the disposal facility. It is revealed that the multiple cover layers and concrete containers effectively play the role of barrier against the permeation of rainfall.

Multi-layers cover,Engineered barrier system,Fluid flow,LILW disposal center

The present study researched and analyzed decontamination technology for decommissioning a nuclear power plant. The decision-making technique (EXPERT-CHOICE) was used to evaluate and select the optimal decontamination technology. In principle, this evaluation method is generally performed by a group of experts in the relevant field. The results of the weights were calculated by multiplying the weights with regard to each criterion and evaluation score. The evaluation scores were categorized into 3 ranges (high, medium, and low), and each range was weighted for differentiation. The level of the technology analysis was improved by additionally quantifying the weights with regard to each criterion and subdividing criteria into subcriteria. The basic assumption of the evaluation was that the weight values would decided on in an expert survey and assigned to each criterion. The evaluation criteria followed high weight for the ';High'; range. Accordingly, H, M, and L were assigned weights of 10:5:1, respectively. This was based on the EXPERT-CHOICE optimal analysis. The minimum and maximum values were excluded, and the average value was used as the evaluation value for each scenario.

Decommissioning,Decontamination,Physical Decontamination,Chemical Decontamination,EXPERTCHOICE

In the process of site selection for a radioactive waste disposal, site characterization must be carried out to obtain input parameters to assess the safety and feasibility of deep geological repository. In this paper, methodologies of site characterization for radioactive waste disposal in Korea were suggested based on foreign cases of site characterization. The IAEA recommends that site characterization for radioactive waste disposal should be performed through stepwise processes, in which the site characterization period is divided into preliminary and detailed stages, in sequence. This methodology was followed by several foreign countries for their geological disposal programs. General properties related to geological environments were obtained at the preliminary site characterization stage; more detailed site characteristics were investigated during the detailed site characterization stage. The results of investigation of geology, hydro-geology, geochemistry, rock mechanics, solute transport and thermal properties at a site have to be combined and constructed in the form of a site descriptive model. Based on this site descriptive model, the site characteristics can be evaluated to assess suitability of site for radioactive waste disposal. According to foreign site characterization cases, 7 or 8 years are expected to be needed for site characterization; however, the time required may increase if the no proper national strategy is provided.

Radioactive waste disposal,Preliminary site characterization,Detailed site characterization,Site descriptive model

For the safe disposal of intermediate level radioactive waste according to the Nuclear Safety and Security Commission’s notice and KORAD’s management plan for low and intermediate level radioactive waste, the disposal concentration limit was derived based on the IAEA methodology. The evaluation of the derived disposal concentration limit revealed that it is not suitable as a practical limit for intermediate level radioactive waste. This is because the disposal concentration limit according to the IAEA methodology is derived using a single value of radioactive waste density and the disposal facility';s volume. The IAEA methodology is suitable for setting the concentration limit for vault type disposal, which consists of a single type of waste, whereas an underground silo type disposal facility is composed of several types of radioactive waste, and thus the IAEA methodology has limitations in determining the disposal concentration limit. It is necessary to develop and apply an improved method to derive the disposal concentration limit for intermediate level radioactive waste by considering the radioactivity of various types of radioactive waste, the corresponding scenario evaluation results, and the regulatory limit.

Intermediate-level radioactive waste,Concentration limit,Underground silo type disposal facility

The Gyeongju underground silo type disposal facility, approved for use in December 2014, is in operation for the disposal of low and very low-level radioactive wastes, excluding intermediate-level waste. That is why the existing low-level radioactive waste level has been subdivided and the concentration limit value for intermediate-level waste has been changed in accordance with Nuclear Safety Commission Notice 2014-003. For the safe disposal of intermediate-level wastes, new optimization methodology for calculating the concentration limit of intermediate radioactive level wastes at an underground silo type disposal facility was developed. According to the developed optimization methodology, concentration limits of intermediate-level wastes were derived and the inventory of radioactive nuclides was evaluated. The operation and post closurescenarios were evaluated for the derived radioactive nuclide inventory and the results of all scenarios were confirmed to meet the regulatory limit. However, in case of 14C, it was confirmed that additional radioactivity limitation through a well scenario was needed in addition to the limit of disposal concentration. It was confirmed that the derived intermediate concentration limit of radioactive waste can be used as the intermediate-level waste concentration limit for the underground disposal facility. For the safe disposal of intermediate-level wastes, KORAD plans to acquire additional data from the radioactive waste generator and manage the cumulative radioactivity of 14C.

Inventory,Disposal facility,Radioactive waste,Concentration limit,Safety Case

The inventory management of radionuclides is essential for the safe management of disposal facilities. In this study, we compared the activity of dry active waste predicted using the generated waste data and that measured for the accepted waste in the disposal facility. For very low level waste, the measured activity was higher than the predicted activity for 137Cs, 90Sr, 99Tc and 129I. In low level waste, the predicted activity was higher than the measured activity for all radionuclides. We also evaluated the variation in the predicted quantity and total activity of each level of dry active waste through a sensitivity analysis on a scaling factor. This result will contribute to the construction of a Safety Case and safe operation of disposal facilities.

Dry Active Waste (DAW),Radioactive Inventory,Scaling factor,Sensitivity analysis

The concept of deep borehole disposal (DBD) for high-level nuclear wastes has been around for about 40 years. Now, the Department of Energy (DOE) in the United States (U.S.) is re-examining this concept through recent studies at Sandia National Laboratory and a field test. With DBD, nuclear waste will be emplaced in boreholes at depths of 3 to 5 km in crystalline basement rocks. Thinking is that these settings will provide nearly intact rock and fluid density stratification, which together should act as a robust geologic barrier, requiring only minimal performance from the engineered components. The Nuclear Waste Technical Review Board (NWTRB) has raised concerns that the deep subsurface is more complicated, leading to science, engineering, and safety issues. However, given time and resources, DBD will evolve substantially in the ability to drill deep holes and make measurements there. A leap forward in technology for drilling could lead to other exciting geological applications. Possible innovations might include deep robotic mining, deep energy production, or crustal sequestration of CO2, and new ideas for nuclear waste disposal. Novel technologies could be explored by Korean geologists through simple proof-of-concept experiments and technology demonstrations.

Deep borehole disposal,Nuclear waste,Radioactive,Repository

A spectroelectrochemical method has been applied to investigate the electrochemical behaviors and identify the kinds of samarium ions dissolved in high temperature molten LiCl-KCl eutectic. An optically transparent electrode (OTE) fabricated with a tungsten gauze as a working electrode has been used to conduct cyclic voltammetry and potential step chronoabsorptometry. Based on the reversibility of the redox reaction of Sm3+/Sm2+ , which was determined from the cyclic voltammograms, the formal potential and the diffusion coefficient were calculated to be -1.99 V vs. Cl2/Cl- and 2.53 × 10-6 cm2·s-1, respectively. From the chronoabsorptometry results at the applied potential of -1.5 V vs. Ag/AgCl (1wt%), the characteristic peaks of absorption for samarium ions were determined to be 408.08 nm for Sm3+ and 545.62 nm for Sm2+. Potential step chronoabsorptometry was conducted using the anodic and the cathodic peak potentials from the voltammograms. Absorbance analysis at 545.63 nm shows that the diffusion coefficient of Sm3+ is 2.15 × 10-6 cm2·s-1, which is comparable to the value determined by cyclic voltammetry at the same temperature.

Samarium,Molten LiCl-KCl eutectic,Spectroelectrochemical,Cyclic voltametry,Potential step chronoabsometry

In order to build a general model of a high-throughput uranium electrorefining process according to the electrode configuration, numerical analysis was conducted using the COMSOL Multiphysics V5.3 electrodeposition module with Ordinary Differential Equation (ODE) interfaces. The generated model was validated by comparing a current density-potential curve according to the distance between the anode and cathode and the electrode array, using a lab-scale (1kg U/day) multi-electrode electrorefiner made by the Korea Atomic Energy Research Institute (KAERI). The operating temperature was 500℃ and LiCl-KCl eutectic with 3.5wt% UCl3 was used for molten salt. The efficiency of the uranium electrorefining apparatus was improved by lowering the cell potential as the distance between the electrodes decreased and the anode/cathode area ratio increased. This approach will be useful for constructing database for safety design of high throughput spent nuclear fuel electrorefiners.

Electrorefining,High-throughput,Uranium,COMSOL Multiphysics,Tertiary current distribution

The Korea Atomic Energy Research Institute (KAERI) constructed the PyRoprocessing Integrated inactive Demonstration facility (PRIDE) to carry out experiments on engineering scale pyroprocessing. PRIDE has a large processing cell that human workers are not allowed to access, and thus the equipment inside is operated from outside using remote handling systems. It is therefore essential to examine the operability and maintainability of the equipment in view of remote handlingsystems, and the equipment is thoroughly examined in a mockup cell before it is installed in the processing cell. If the equipment is tested in a virtual mockup rather than in a mockup cell, the development cost can be significantly reduced. The PRIDE 3D simulator was integrated for virtual verification of equipment that will be installed in the processing cell. All remote handling devices in the actual PRIDE were also virtually installed in the PRIDE 3D simulator. The 3D model of the equipment was loaded and located in the exact position in the virtual processing cell. A scenario to replace an actual electrode was implemented in the PRIDE 3D simulator. The design of the equipment and the working procedures in the scenario were successfully evaluated. The results demonstrated that the PRIDE 3D simulator can be used successfully as an alternative to actual mockup testing.

Remote handling,3D simulator,Virtual Reality,PRIDE,Pyroprocessing

Processing and equipment were tailored for engineering scale fabrication of UO2 porous pellets, a feed material for the electrolytic reduction process in the PRIDE (PyRoprocessing Integrated DEmonstration) facility at KAERI (Korea Atomic Energy Research Institute). The starting materials, UO2 powder and pre-milled surrogate oxide powders, were proportioned to simulate the chemical composition of spent fuel (so-called Simfuel). The Simfuel powders were homogenized by mixing, compacted into a pellet shape, and finally heat treated using a tumbling mixer, rotary press, and sintering furnace. After sintering at 1450℃ for 24 h in 4% H2-Ar, the average bulk density of the UO2 Simfuel pellets was 6.89 g·cm-3, which meets the standard of the following electrolytic reduction process. In addition, the results of a microstructural analysis demonstrated that the sintered Simfuel UO2 porous pellets accurately simulate the properties of spent fuel in terms of the formation of second phases. These results provide essential information for the massive fabrication of UO2 porous pellets for engineering scale pyroprocessing research.

Pyroprocessing,Engineering scale,Head-end process,Feed material,Porous pellet

The decommissioning of Kori unit 1, which was permanently shut down in June of 2017, will be the first instance of the dismantling of a commercial nuclear power plant in Korea. The disposal of waste during the dismantling process accounts for a large part of the total decommissioning cost. Therefore, structures consisting of activated and contaminated concrete must be economically and safely dismantled by establishing a proper dismantling strategy. This study focuses on optimized dismantling and disposal scenarios pertaining to a biological shield. Several dismantling cases, regulations and technologies related to waste treatment as these practices pertain to nuclear power plants are analyzed. To minimize the amount of waste from the biological shield dismantling process, an optimized dismantling scenario is presented and disposal alternatives for dismantled concrete waste are proposed.

commissioning,Concrete waste,Biological shield,Volume reduction,Waste disposal

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.

Near surface disposal facility,Human intrusion,RESRAD,Safety case,Post-drilling scenario

In Korea, commercial nuclear power plants and research reactors have on-site storage systems for the spent nuclear fuel, but it is difficult to expand the facilities used for the storage systems. If decommissioning of nuclear power plants starts, an amount of high level radioactive waste will be generated. In this study, a radiological impact assessment of the railroad transport of high level radioactive waste was carried out considering radiation workers and the public, using the developed transport container as the transport package. The dose rates for workers and the public during the transport period were estimated, considering anticipated transport scenarios, and the results compared with the regulatory limit. A sensitivity analysis was also carried out by considering the different release ratios of the radioactive materials in the high level radioactive waste, and different distances between the transport container and workers during loading and unloading phases and while attaching another freight car. For all the anticipated transport scenarios, the radiological impacts for workers and the public met the regulatory limits.

High level radioactive waste (HLW),RADTRAN,Railroad transport,Spent nuclear fuel

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility’s design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a 2×10 arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the 2×10 arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

PWR spent nuclear fuel,Concrete cask,Dry storage condition,Radiation shielding analysis,Interim storage facility

For dry storage of the spent nuclear fuel (SNF) stored in the storage pool of a nuclear power plant, essentially all moisture must be removed to prevent corrosion of the assembly and canister internals and/or degradation of fuel cladding integrity after SNF canister loading operation. R&D work is now in progress on a forced gas drying system that can be used to remove residual water in canisters. In this work, preliminary design is performed to manufacture the forced gas drying system. This process includes a case study of dry methods for canister moisture removal, relative codes and standards, confirmation of adequate dryness, needs analysis at plant sites, and characteristics of SNF stored in pools. Through this preliminary design work, we obtained a conceptual flow diagram and preliminary P&ID of the forced gas drying system. The results of this study can be used to determine details of the design to manufacture the forced gas drying system.

Forced gas drying system,Spent nuclear fuel,Dry storage,Dry method,Canister

In this study, thermal-hydrological-chemical modeling for the alteration of a bentonite buffer is carried out using a simulation code TOUGHREACT. The modeling results show that the water saturation of bentonite steadily increases and finally the bentonite is fully saturated after 10 years. In addition, the temperature rapidly increases and stabilizes after 0.5 year, exhibiting a constant thermal gradient as a function of distance from the copper tube. The change of thermal-hydrological conditions mainly results in the alteration of anhydrite and calcite. Anhydrite and calcite are dissolved along with the inflow of groundwater. They then tend to precipitate in the vicinity of the copper tube due to its high temperature. This behavior induces a slight decrease in porosity and permeability of bentonite near the copper tube. Furthermore, this study finds that the diffusion coefficient can significantly affect the alteration of anhydrite and calcite, which causes changes in the hydrological properties of bentonite such as porosity and permeability. This study may facilitate the safety assessment of high-level radioactive waste repositories.

Bentonite,TOUGHREACT,Thermal-hydrological-chemical,Porosity,High-level radioactive waste repository

This paper focuses on the estimation of durability and service-life of reinforced concrete structures in Wolsong Low- and intermediate-level wastes Disposal Center (WLDC) in Korea. There are six disposal silos located in the saturated environment. The silo concrete is degraded due to reactions with groundwater and chemical attacks, and finally it will lose its properties as a transport barrier. The infiltration of sulfate and magnesium, leaching of potassium hydroxide, and chlorine induced corrosion are the most significant factors for degradation of reinforced concrete structure in underground environment. From the result of evaluation of the degradation time for each factor, the degradation rate of the reinforced concrete due to sulfate and magnesium is 1.308×10-3 cm/yr, and it is estimated to take 48,000 years for full degradation while potassium hydroxide is leached in depth of less than 1.5 cm at 1,000 years after the initiation of degradation. In case of chlorine induced corrosion, it takes 1,648 years to initiate corrosion in the main reinforced bar and 2,288 years to reach the lifetime limit of the structural integrity, and thus it is evaluated as the most significant factor.

Radioactive waste,Disposal,Engineered barrier,Concrete degradation,Numerical analysis model,Service life,Steel corrosion

Estimation of groundwater hydrologic cycle pattern is one of the most critical issues in sustainable management of groundwater resources in coastal area. This study estimated groundwater percolation by using the water balance methodology and hydrogeological characteristics of land use and soil. Evapotranspiration was computed by using the Thornthwaite method, and surface runoff was determined by using the SCS-CN technique. Groundwater storage change was obtained as 229 mm/a (17.8% of the average annual rainfall, 1286 mm/a), with 693 mm/a (60.1%) of evapotranspiration and 124 mm/a (9.6%) of surface runoff. Rainfall and groundwater storage change was highly correlated, comparing with the relationships between rainfall and evapotranspiration, and between rainfall and surface runoff. This result indicates that groundwater storage change responds more sensitively to precipitation than evapotranspiration and surface runoff.

Groundwater storage change,Evapotranspiration,Surface runoff,Rainfall,Coastal area

Leading national R&D project to design a PWR spent nuclear fuel interim dry storage system that has been under development since mid-2009, which consists of a dual purpose metal cask and concrete storage cask. To ensure the safe operation of dry storage systems in foreign countries, major confinement monitoring techniques currently consist of pressure and temperature measurement. In the case of a dual purpose metal cask, a pressure sensor is installed in the interspace of bolted double lid(primary and secondary lid) in order to measure pressure. A concrete storage cask is a canister based system made of double/redundant welded lid to ensure confinement integrity. For this reason, confinement monitoring method is real time temperature measurement by thermocouple placed in the air flow(air intake and exit) of the concrete structure(over pack and module). The use of various monitoring technologies and operating experiences for the interim dry storage system over the last decades in foreign countries were analyzed. On the basis of the analysis above, development of the confinement monitoring technology that can be used optimally in our system will be available in the near future.

PWR spent nuclear fuel,Confinement monitoring,Dual purpose metal cask,Dry storage,Pressure sensor

Decontamination is one of the most important technologies for the decommissioning of NPP. The purpose of decontamination is to reduce the Risk of exposure of the decommissioning workers, and to recycle parts of the plant components. Currently, there is a lack of data on the efficiency of the decontamination technologies for decommissioning. In most cases, the local radiation level can be lowered below a regulatory limitation by decontamination. Therefore, more efficient decontamination technology must be continuously developed. This work describes the practical experiences in the United States and the European countries for NPP decommissioning using these decontamination technologies. When the decommissioning of domestic nuclear power plant is planned and implemented, this work will be helpful as a reference of previous cases.

Decommissioning,Decontamination,Chemical decontamination,Mechanical decontamination

The Tomographic Gamma Scan (TGS) technique partitions radioactive waste drums into 10x10x16 voxels and assays both the density and concentration of radioactivity for each voxel thus providing for improved accuracy, when compared to the traditional Non-Destructive Assay(NDA) techniques. It could decrease the degree of precision measurement since there is a trade-off between spatial resolution and precision. This latter drawback is compensated by expanding the Region of Interest (ROI) that differentiates the full energy peaks, which, in turn, results in an optimized degree of precision. The enlarged ROI, however, increases the probability of interference among those nuclides that emit energies in the adjacent spectrum. This study has identified the cause of such interference for the reference nuclide of the TGS technique, 137Cs (661.66 keV, halflife 30.5 years), to be 110mAg (657.75 keV, half-life 249.76 days). A new calibration method of determining the optimized ROI was developed, and its effectiveness in accurately characterizing 137Cs and eliminating the interference was further ascertained.

Radioactive waste characterization,NDA,Tomographic gamma scan,Calibration method of TGS

To meet nuclear regulatory requirements, more than 95% individual radionuclides in the low- and intermediate-level radioactive waste inventory have to be identified. In this study, the radionuclide inventory has been estimated by taking the long-term radioactive waste generation, the development plan of disposal facility, and the new radioactive waste classification into account. The state of radioactive waste cumulated from 2014 was analyzed for various radioactive sources and future prospects for predicting the long-term radioactive waste generation. The predicted radionuclide inventory results are expected to contribute to secure the development of waste disposal facility and to deploy the safety case for its long-term safety assessment.

Inventory,Disposal facility,Radioactive waste,New classification standard,Safety Case

The second-stage near surface disposal facility for low and very low level radioactive waste';s permanent disposal is to be built. During the institutional control period, the inadvertent intrusion of the general public is limited. But after the institutional control period, the access to the general public is not restricted. Therefore human who has purpose of residence and resource exploration can intrude the disposal facility. In this case, radioactive effects to the intruder should be limited within regulatory dose limits. This study conducted the safety assessment of human intrusion on the second-stage surface disposal facility through drilling and post drilling scenario. Results of drilling and post drilling scenario were satisfied with regulatory dose limits. The result showed that post-drilling scenario was more significant than drilling scenario. According to the human intrusion time and behavior after the closure of the facility, dominant radionuclide contributing to the intruder was different. Sensitivity analyses on the parameters about the human behavior were also satisfied with regulatory dose limits. Especially, manual redistribution factor was the most sensitive parameter on exposure dose. A loading plan of spent filter waste and dry active waste was more effective than a loading plan of spent filter waste and other wastes for the radiological point of view. These results can be expected to provide both robustness and defense in depth for the development of safety case further.

Near surface disposal,Human intrusion,Waste emplacement,Institutional control period,Safety Case

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

PWR,Activation inventory,Impurity,MCNP,FISPACT,Decommissioning,Radioactive waste

This study investigated the removal of Sr, which was one of the high radioactive nuclides, by adsorption with Barium (Ba) impregnated 4A zeolite (BaA) from high-radioactive seawater waste (HSW). Adsorption of Sr by BaA (BaA-Sr), in the impregnated Ba concentration of above 20.2wt%, was decreased by increasing the impregnated Ba concentration, and the impregnated Ba concentration was suitable at 20.2wt%. The BaA-Sr adsorption was added to the co-precipitation of Sr with BaSO4 precipitation in the adsorption of Sr by 4A (4A-Sr) within BaA. Thus, it was possible to remove Sr more than 99% at m/V (adsorbent weight/solution volume)=5 g/L for BaA and m/V >20 g/L for 4A, respectively, in the Sr concentration of less than 0.2 mg/L (actual concentration level of Sr in HSW). It shows that BaA-Sr adsorption is better than 4A-Sr adsorption in for the removal capacity of Sr per unit gram of adsorbent, and the reduction of the secondary solid waste generation (spent adsorbent etc.). Also, BaA-Sr adsorption was more excellent removal capacity of Sr in the seawater waste than distilled water. Therefore, it seems to be effective for the direct removal of Sr from HSW. On the other hand, the adsorption of Cs by BaA (BaA-Cs) was mainly performed by 4A within BaA. Accordingly, it seems to be little effect of impregnated Ba into BaA. Meanwhile, BaA-Sr adsorption kinetics could be expressed the pseudosecond order rate equation. By increasing the initial Sr concentrations and the ratios of V/m, the adsorption rate constants (k2) were decreased, but the equilibrium adsorption capacities (qe) were increasing. However, with increasing the temperature of solution, k2 was conversely increased, and qe was decreased. The activation energy of BaA-Sr adsorption was 38 kJ/mol. Thus, the chemical adsorption seems to be dominant rather than physical adsorption, although it is not a chemisorption with strong bonding form.

Strontium,Adsorption/Removal,4A zeolite,Co-precipitation of Sr with BaSO4,Adsorption kinetics,Highradioactive seawater waste

In this study, the dispersion process of pollutant substances in a port under wave and current environments was evaluated by a hydraulic experiment. Once the contaminants washed ashore into the port of Wolseong nuclear power plant, transport processes of pollutants were investigated by tracking the tracer according to the variations of experimental condition through a hydraulic experiment. Several hydraulic experiments were performed to analyze the pollutant discharge rate of the surface coming from the different coastal environments. From the hydraulic experiment results, the tracer concentration decreased exponentially. These results suggested that, after the tracer was transported to the open sea, a different gradient was shown under different conditions. For the case of a diluted condition, the half-life of flow rate was 2.70, 10.40, and 26.39 days for 30, 20 and 10 rpm in the left-side, respectively. The decrease of the tracer concentration under conditions of 30 rpm was much faster than that under conditions of 10 rpm. For the wave condition, the half-life of flow rate was 4.59 and 15.35 days for the right and left side of the port in a hydraulic scale prototype, respectively.

Wolseong nuclear power plant,Pollutants discharge,Hydraulic experiment,Discharge characteristics

Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation.

Decommissioning radioactive waste disposal amount,Decommissioning RPV,Kori unit 1

The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.

Domestic bentonite,Buffer,Geological disposal system,Engineered barrier,Ca-bentonite

The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the 3×3 short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

Spent fuel,Drop accident,Impact velocity

Radioactive waste container designs should comply with the requirements for safety (i.e., transportation, storage, disposal) and other criteria such as economics and technology. These criteria are also applicable to the future management of the large amount of LLW and VLLW to arise from decontamination and decommissioning (D&D) of nuclear power plants, which have different features compared to that of wastes from operation and maintenance (O&M). This paper proposes to develop a set of standard containers of multi-purpose usage for transportation, storage and disposal. The concepts of the containers were optimized for management of D&D wastes in consideration of national system for radioactive waste management, in particular the Gyeongju Repository and associated infrastructures. A set of prototype containers were designed and built : a soft bag for VLLW, two metallic containers for VLLW/LLW (a standard IP2 container for sea transport and ISO container for road transport). Safety analyses by simulation and tests of these designs show they are in compliance with the regulatory requirements. A further development of a container with concrete is foreseen for 2016.

Radioactive waste,LLW and VLLW,Standard ISO containers,D&D waste

The properties of wastes for melting need to be considered to minimize the maintenance of refractory and to discharge the molten slags smoothly from a plasma torch melter. When the nonflammable wastes from nuclear facilities such as concrete debris, glass, sand, etc., are melted, they become acid slags with low basicity since the chemical composition has much more acid oxides than basic oxides. A molten slag does not have good characteristics of discharge and is mainly responsible for the refractory erosion due to its low liquidity. In case of a stationary plasma torch melter with a slant tapping port on the wall, a fixed amount of molten slags remains inside of tapping hole as well as the melter inside after tapping out. Nonmetallic slags keep the temperature higher than melting point of metal because metallic slags located on the bottom of melter by specific gravity difference are simultaneously melted when dual mode plasma torch operates in transferred mode. In order to minimize the refractory erosion, the compatible refractories are selected considering the temperature inside the melter and the melting behavior of slags whether to contact or noncontact with molten slags. An acidic refractory shall not be installed in adjacent to a basic refractory for the resistibility against corrosion.

Refractory,Nonflammable waste,Plasma torch,Slag,Basicity

Spent fuels from nuclear power plants, as well as high-level radioactive waste from the recycling of spent fuels, should be safely isolated from human environment for an extremely long time. Recently, meaningful studies on the development of deep borehole radioactive waste disposal system in 3-5 km depth have been carried out in USA and some countries in Europe, due to great advance in deep borehole drilling technology. In this paper, domestic deep geoenvironmental characteristics are preliminarily investigated to analyze the applicability of deep borehole disposal technology in Korea. To do this, state-of-the art technologies in USA and some countries in Europe are reviewed, and geological and geothermal data from the deep boreholes for geothermal usage are analyzed. Based on the results on the crystalline rock depth, the geothermal gradient and the spent fuel types generated in Korea, a preliminary deep borehole concept including disposal canister and sealing system, is suggested.

Geoenvironment,Characteristics,Deep borehole disposal(DBD),Spent fuels,High level waste,Application

The Republic of Tajikistan has ten former uranium mining sites. The total volume of all tailings is approximately 55 million tonnes, and the covered area is more than 200 hectares. The safe management of legacy uranium mining and tailing sites has become an issue of concern. Depending on the performance requirements and site-specific conditions (location in an arid, semiarid or humid region), a cover system for uranium tailings sites could be constructed using several material layers using both natural and man-made materials. The purpose of this study is to find a feasible cost-effective cover system design for the Degmay uranium tailings site which could provide a long period (100 years) of protection. The HELP computer code was used in the evaluation of potential Degmay cover system designs. As a result of this study, a cover system with 70 cm thick percolation layer, 30 cm thick drainage layer, geomembrane liner and 60 cm thick barrier soil layer is recommended because it minimizes cover thickness and would be the most cost-effective design.

Cover system,Uranium tailings sites,Remediation,Percolation,HELP code

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except 5 × 5 × 2 m size) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a 50 × 50 × 2 m size facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

Fukushima NPP accident,Decontamination waste,Temporary storage facility,Ambient dose rate,Separation distance

Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP (SiO2-Al2O3-P2O5). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding B2O3 and Fe2O3 to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of Al2O3 or B2O3 as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.

Dechlorination,Solidification,Waste salt,Synthetic inorganic composite,Wasteform,Leach resistance

This study aimed to the adsorption-removal of high- radioactive iodide (I) contained in the initially generated high-radioactive seawater waste (HSW), with the use of AgX (Ag-impregnated X zeolite). Adsorption of I by AgX (hereafter denoted as AgX-I adsorption) was increased by increasing the Ag-impregnated concentration in AgX, and its concentration was suitable at about 30 wt%. Because of AgCl precipitation by chloride ions contained in seawater waste, the leaching yields of Ag from AgX (Ag-impregnated concentration : about 30~35 wt%) was less than those in distilled water (< 1 mg/L). AgX-I adsorption was above 99% in the initial iodide concentration (Ci) of 0.01~10 mg/L at m/V (ratio of weight of adsorbent to solution volume)=2.5 g/L. This shows that efficient removal of I is possible. AgX-I adsorption was found to be more effective in distilled water than in seawater waste, and the influence of solution temperature was insignificant. Ag-I adsorption was better described by a Freundlich isotherm rather than a Langmuir isotherm. AgX-I adsorption kinetics can be expressed by a pseudo-second order rate equation. The adsorption rate constants (k2) decreased by increasing Ci, and conversely increased by increasing the ratio of m/V and the solution temperature. This time, the activation energy of AgX-I adsorption was about 6.3 kJ/mol. This suggests that AgX-I adsorption is dominated by physical adsorption with weaker bonds. The evaluation of thermodynamic parameters (a negative Gibbs free energy and a positive Enthalpy) indicates that AgX-I adsorption is a spontaneous reaction (forward reaction), and an endothermic reaction indicating that higher temperatures are favored.

Iodide,Adsorption/Removal,Ag-impregnated zeolite,Isotherm,Kinetics,Thermodynamic parameters,Highradioactive seawater waste

In the near future, many countries, including the Republic of Korea, will face a significant increase in low level radioactive waste (LLW) from nuclear power plant decommissioning. The purpose of this paper is to look at blending as a method for enhancing disposal options for low-level radioactive waste from the decommissioning of nuclear reactors. The 2007 U.S. Nuclear Regulatory Commission strategic assessment of the status of the U.S. LLW program identified the need to move to a risk-informed and performance-based regulatory approach for managing LLW. The strategic assessment identified blending waste of varying radionuclide concentrations as a potential means of enhancing options for LLW disposal. The NRC’s position is that concentration averaging or blending can be performed in a way that does not diminish the overall safety of LLW disposal. The revised regulatory requirements for blending LLW are presented in the revised NRC Branch Technical Position for Concentration Averaging and Encapsulation (CA BTP 2015). The changes to the CA BTP that are the most significant for NPP operation, maintenance and decommissioning are reviewed in this paper and a potential application is identified for decommissioning waste in Korea. By far the largest volume of LLW from NPPs will come from decommissioning rather than operation. The large volumes in decommissioning present an opportunity for significant gains in disposal efficiency from blending and concentration averaging. The application of concentration averaging waste from a reactor bio-shield is also presented.

Low-level radioactive waste,Concentration averaging,Blending,Disposal,Decommissioning waste,Bio-shield

Soil can be contaminated by radioactive materials due to nuclide leakage following unexpected situations during the decommissioning of a nuclear power plant. Soil decontamination is necessary if contaminated land is to be reused for housing or industry. The present study classifies various soil remediation technologies into biological, physics/chemical and thermal treatment and analyzes their principles and treatment materials. Among these methods, this study selects technologies and categorizes the economics, applicability and technical characteristics of each technology into three levels of high, medium and low by weighting the various factors. Based on this analysis, the most applicable soil decontamination technology was identified.

Soil remediation,Cesium,Soil washing,Washing-electrokinetic

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

DSRS disposal,Radioisotope (RI) wastes,Classification system,Disposal method

For the decommissioning of a nuclear power plant, a site characterization and final status survey require a site-specific suite of radionuclides that could potentially still be present in the site during or after the decontamination processes. The United States Nuclear Regulatory Commission (U.S. NRC) requires a Decommissioning Technical Base Document (DTBD) along with a Site Characterization and Historical Site Assessment (HSA) from the utility for decommissioning to proceed. Both the DTBD and HSA are preliminary components of the Radiological Site Survey investigation process and should be included in the final License Termination Plan (LTP) for site release and reuse consideration from the U.S. NRC and the utility company. This study reviews the United States Nuclear Power Plants (U.S. NPPs) decommissioning cases and is especially focused on the methodologies used for determining a site-specific suite of radionuclides before and during the site characterization and final status surveys. In 2017, Kori-1 will be ready for decommissioning and related preparations are ongoing, this review will help Korea to prepare regulatory guidelines and give technical background for the safe and successful decommissioning of NPPs.

Nuclear decommissioning,Radiological site survey,Site characterization,Final status survey,DandD

Korea Radioactive Waste Agency (KORAD) began to operate the low and intermediate level radioactive waste disposal facility in Gyeongju and to transport the radioactive waste containing radioactive isotopes from Daejeon to the disposal facility for the first time at 2015. For this radioactive waste transportation, in this study, radiological impact assessment is carried out for workers and public. The dose rate to workers and public during the transportation is estimated with consideration of the transportation scenarios and is compared with the Korean regulatory limit. The sensitivity analysis is carried out by considering both the variation of release ratios of the radioactive isotopes from the waste and the variation of the distances between the radioactive waste drum and worker during loading and unloading of radioactive waste. As for all the transportation scenarios, radiological impacts for workers and public have met the regulatory limits.

Radioactive Isotopes (RI),RADTRAN,Radioactive waste,On-road transport,LILW Disposal facility

Kori Unit 1, the first PWR (Pressurized Water Reactor) plant in Korea, began its commercial operation in 1978 and will permanently shut down on June 18, 2017. After moving the spent fuels to SFP (Spent Fuel Pool) system, Kori Unit 1 will perform a full system chemical decontamination to reduce radiation levels inside the various plant systems. This paper will describe the operation concept of the full system chemical decontamination for Kori Unit 1 based on experiences overseas.

Kori Unit 1,Full system chemical decontamination,Radiation level

An approach to assess inadvertent human intrusion into radwaste repository was proposed with the assumption that the intrusion occurs after a loss of knowledge of the hazardous nature of the disposal facility. The essential boundary conditions were derived on the basis of international recommendations, followed by an overall approach to deal with inadvertent human intrusion. The interrelation between societal factors, human intrusion scenarios, and protective measures is described to provide a concrete explanation of the approach, including the detailed procedures to set up the human intrusion scenario. The procedure for deriving protective measures is also explained with four steps, including how to derive a safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be used effectively to reduce the potential for and/or the consequences of human intrusion during the entire process of realizing a disposal facility.

Radioactive waste,Disposal facility,Human intrusion,Protective measure

The metal waste generated during nuclear power plant decommissioning constitutes a large proportion of the total radioactive waste. This study investigates the current status of domestic and international regulatory requirements for clearance and the clearance experience of domestic institutions. The RESRAD-RECYCLE code was used for analyzing the clearance of the metal wastes generated during actual nuclear power plant decommissioning, and assessment of the exposure dose of twenty-six scenarios was carried out. The evaluation results will be useful in preliminary analysis of clearance and recycling during nuclear power plant decommissioning. As a next step, the effects of reducing disposal costs by clearance can be studied.

Decommissioning,Decontamination,Clearance,Recycle,Metal waste

Based on the results of groundwater flow system modeling for a hypothetical deep geological repository site, quantitative and spatial distributions of groundwater flow rates at the positions of deposition holes, groundwater travel length and time from the positions to the surface environment were analyzed and used to suggest a method for determining locations of deposition holes. The hydraulic head values at the depth of the deposition holes and a particle tracking method were used to calculate the groundwater flow rates and groundwater travel length and time, respectively. From the results, an approach to designing a layout of deposition holes was suggested by selecting relatively favorable positions for maintaining performance of the disposal facility and screening some positions of deposition holes that did not comply with specific constraints for the groundwater flow rates, travel length and time. In addition, a method for determining a geometrical direction for extension of the disposal facility was discussed. Designing the layout of deposition holes with the information of groundwater flow at the disposal depth can contribute to secure performance and safety of the disposal facility.

Layout of deposition holes,Groundwater flow simulation,Groundwater flow rate,Groundwater pathway

To evaluate the physicochemical and radiological properties of raw materials and by-products in domestic distribution, about 220 samples with 16 species were prepared. We measured the energy spectrum and the chemical content, such as U, Th, and K, using a LaBr3 scintillation detector and ED-XRF. In addition, HPGe detector was used to analyze the radioactivity of 234Th, 234mPa, and 214Bi in uranium decay series and 228Ac, 212Pb, and 208Tl in thorium decay series, and 40K. The correlation between characteristic variables, such as the count rate in several ROIs, chemical content, and radioactivity, was assessed to infer the radioactivity of natural radionuclides through a rapid screening method. Based on the results, a characteristic database for raw material and by-product in domestic distribution was established and it will provide useful information in the analysis procedure and improve the accuracy and reproducibility in the analysis of natural radionuclides.

NORM,Raw materials,By-products,Characteristic database

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5·ORIGEN-2 evaluation system and by considering each component’s chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask’s design life, 60Co had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as 28Al and 24Na had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

PWR spent nuclear fuel,Dual-purpose metal cask,Dry storage,Radioactive characterization analysis,Dcontamination·Decommission

We observed the γ-ray shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of 0.371 cm-1 from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a γ-ray of 137Cs, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of 3,175 kg·m-3. Although the unit weight of the concrete with OSS and OSG was 3,052 kg·m-3, which was lower than the maximum unit weight condition by 123 kg·m-3, its attenuation coefficient was improved by 0.012 cm-1. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced γ-ray shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing.

High-weight concrete,Aggregate,γ -ray shielding,Attenuation coefficient,Compressive strength

Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

High-level radioactive waste,Deep geological disposal,Disposal safety,Safety case,Confidence improvement

The disposal facility in Gyeongju is planning to dispose of 800,000 packages of low- and intermediate- level radioactive waste. This facility will be developed as a complex disposal facility that has various types of disposal facilities and accompanying management. In this study, based on the comprehensive development plan of the disposal facility, a preliminary post-closure safety assessment is performed to predict the phase development of the total capacity for the 800,000 packages to be disposed of at the site. The results for each scenario meet the performance target of the disposal facility. The assessment revealed that there is a significant impact of the inventory of intermediate-level radionuclide waste on the safety evaluation. Due to this finding, we introduce a disposal limit value for intermediate-level radioactive waste. With stepwise development of safety case, this development plan will increase the safety of disposal facilities by reducing uncertainties within the future development of the underground silo disposal facilities.

Low-and intermediated radioactive waste,Complex disposal facility,Safety case,Optimization strategy

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

Spent nuclear fuel,Dual purpose metal cask,Initial enrichment,Burnup,Minimum cooling period

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

MACCS2,COSYMA,OSCAAR,Level 3 Probabilistic Safety Assessment,Off-site consequence assessment,Severe nuclear accident

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP';s storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process).

Dual purpose metal cask,Dry storage,Operation process,Vacuum drying,He backfilling

In the safety assessment of the geological disposal system of the radioactive wastes, the abnormal scenarios, in which the system is impacted by the abnormal events, need to be considered in addition to the reference scenario. In this study, characterization and prediction of well intrusion as one of the abnormal events which will impact the disposal system were conducted probabilistically and statistically for the safety assessment. The domestic well development data were analyzed, and the prediction methodologies of the well intrusion were suggested with a computation example. From the results, the annual well development rate per unit area in Korea was about 0.8 well/yr/km2 in the conservative point of view. Considering the area of the overall disposal system which is about 1.5 km2, the annual well development rate within the disposal system could be 1.2 well/yr. That is, it could be expected that more than one well would be installed within the disposal system every year after the institutional management period. From the statistical analysis, the probabilistic distribution of the well depth followed the log-normal distribution with 3.0363 m of mean value and 1.1467 m of standard deviation. This study will be followed by the study about the impacts of the well intrusion on the geological disposal system, and the both studies will contribute to the increased reliability of safety assessment.

Geological disposal system,Safety assessment,Abnormal event,Well intrusion

The production of nuclear fuel cladding tube is expected to increase with the nuclear power plant expansion. Zirconium(Zr) scrap that is generated during manufacturing is also expected to increase. Zr electrorefining experiment was carried out in the fluoride salt of LiF-KF-ZrF4 using multiple electrode for scale up and improving throughput Zr electrorefiner development. The Zr reduction peak observed at -0.8 V(vs.Ni). Polarization behavior showed that the amount of applied current increases because of decreasing cell resistance as the number of cathode increases. Experimental results showed the highest recovery rate about 98% at lowest current density of 25.64 mA/cm2 using 6 electrodes. XRD and TG analysis result show that pure Zr was recovered 99.92% and ICP analysis shows that lower impurity content than conventional impurity content of the Anode(97.8%). Electrorefining consumes energy about 7.15 kWh/kg less than 39.7% compared to the Kroll process using 6 electrode width of 20 mm and height of 65 mm. Because of increasing cell efficiency and recovery rate, using multiple cathode is determined as an efficient technique for scale up electrorefining Zr scrap.

Zirconium,Multiple cathode,Electrorefining,Throughput,LiF-KF-ZrF4

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

Source term analysis,Decommissioning waste classification,MCNP,MCNPX,ORIGEN-2,Wolsong Unit 1

A large amount of radioactively-contaminated gravel can be produced on the demolition/restoration of facilities related the back end of fuel cycle. However, because of the lacking in basic knowledge for decontamination of radioactive-contaminated gravel, this study has performed the basic tests using for soil-washing. To find effective decontamination condition, several experiments were carried out for the selection of optimal decontamination agents. Washing by 0.1 M nitric acid was proved to be more effective than that by distilled water or surfactant for decontamination of uranium-contaminated gravel. In addition, crushing/grinding of uranium-contaminated gravel prior to washing was contributed to increase in of removal efficiency of uranium and reduction of decontamination time. The smaller the sizes of crushed gravel was, the more the removal efficiency increased. Also, small the sized particles improved chances for meeting the clearance requirement of the treated gravel.

Gravel,Washing,Decontamination,Uranium

Since the commercial operation of Kori Unit #1 nuclear power plant(NPP) started in 1978, 23 units at present are operating in Korea. Radioactive wastes will be steadily generated from these units and accumulated. In addition, the life-extension of NPPs, construction of new NPPs and decontamination and decommissioning research facilities will cause radioactive wastes to increase. Recently, Korea has revised the new classification criteria as was proposed by IAEA. According to the revised classification criteria, low-level, very-low-level and exempt waste are estimated to about 98% of total disposal amount. In this paper, current status of overseas cases and disposal method with new classification criteria are analyzed to propose the most reasonable method for estimating the amount of decommissioning waste when applying the new criteria.

Decontamination,Decommissioning,New classification criteria,Radioactive waste,Disposal

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

IRIS collimator,Radiation detector,Dose rate,Radioactivity,MCNP code

The low and intermediate-level radioactive waste generated in Korea is disposed of at Wolsong Disposal Facility. For the safety of a disposal facility, it must be assessed by considering some abnormal scenarios including human intrusion as well as those by natural phenomena. The human intrusion scenario is a scenario that an incognizant man of the disposal facility will be occurred by the drilling. In this paper, the well usage scenario was classified into the human intrusion event as the probability of the well drilling is very low during the man’s lifecycle and then was assessed by using conservative assumptions. This scenario was assessed using the dilution factor of contaminants released from a disposal facility and then it was introduced the applied methodology in this study. The assessed scenario using this methodology is satisfied the regulatory limits.

Scenario,Human intrusion,Well,Contaminant,Dilution factor

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

Safety case,Complementary indicator,Low- and intermediate level waste (LILW),Radioactive waste disposal facility,Radionuclide flux

Safety assessments of geological radioactive waste disposal systems, need to consider the abnormal scenario in which a system is impacted by external events in addition to a reference scenario. In this study, the characterization and prediction of an earthquake as an external event which will impact disposal systems were conducted probabilistically and statistically for the safety assessment. The domestic earthquake data were analyzed, and the prediction methodologies of the earthquake were suggested with a computational example. From the results, the earthquake occurrence rates in Korea ranged from 0.4 /yr to 36.2 /yr depending on the data set and the completeness magnitude. From a conservative point of view, the earthquake occurrence rate in the disposal system was suggested as 5.4×10-4 /yr considering the area of the disposal system. At that time, the completeness magnitude of an earthquake was 2.3. This study will be followed by an appraisal of impacts associated with external events on the geological disposal system, and it will contribute to improvements in reliability of the safety assessment.

Geological disposal system,Safety assessment,External event,Earthquake

In this study, laboratory test equipment was designed and installed to evaluate the thermal behavior of bentonite, which is used as a buffer in high-level waste disposal systems. The thermal analysis was conducted to verify the test results using ABAQUS, a finite element analysis code. In view of the seasonal changes seen during the test, the thermal behavior of bentonite with a temperature of outside air was evaluated. Of the cases examined, the results of the analysis model using stainless steel (Case 3) approximates to the test results, showing an error of about 1℃. The results of the thermal analysis into seasonal temperature distributions are consistent with trends seen in lab-test results. These analyses show that the effects of the thermal conductivity of the material surrounding the buffer and outside air temperature, are very important factors in the thermal behavior of bentonite. In the future, it is expected that a moisture saturation test of a bentonite buffer containing a heat source will be carried out. Therefore, the development of a numerical analysis model is required for the prediction and verification of the laboratory test results.

Laboratory test,Bentonite,ABAQUS,Thermal conductivity,Thermal behavior

55Fe and 63Ni are key factors in deciding the proper handling of the decommissioning of radioactive waste from nuclear facilities. For determining beta emitting radionuclides, the dismantled waste samples should be completely decomposed and separated from the sample matrix. This study reports the comparison results of the recovering efficiencies of Iron and Nickel with wet digestion methods that use various acids and alkali-fusion methods. Various matrices of NIST SRMs (1646a, 1944, 8704, 2709a, and 1633c), the recovering efficiencies of using alkali-fusion methods ranged from 95.3 to 98.3% for Iron, and from 86.6 to 88.1% for Nickel within about 2% of relative standard deviation. On the other hand, those using one of the three wet digestion methods ranged from 77.9 to 105.3% for Iron and from 40.1 to 78.5% for Nickel with over 10% of relative standard deviation. Therefore, one may draw the conclusion that the analytical results derived from Iron and Nickel using alkali-fusion methods are fairly reliable due to the recovering efficiencies observed.

55Fe,63Ni,Decommissioning,Pretreatment,Wet digestion,Alkali-fusion

For the analysis of 90Sr, which is a pure beta emitter, radiochemical separation from the main interfering elements such as Ca, Ba and Ra is required due to their similarity in chemical behavior to strontium. This study describes a radioanalytical procedure using extraction chromatography for separating Sr from interfering elements. The maximum capacity of the resin for Sr was approximately 6 mg per 1.5 mL of bed volume (BV). The recovery of Sr on the resin 1.5 mL (BV) was quantitative for the calcium level of 200 mg at the flow rate of 1 mL min-1. However the chemical yield declined by increasing the flow rate by up to 5 mL min-1 even at the calcium level of 200 mg. When using the same BV of Sr resin, the performance of the resin was enhanced as the cross-sectional area of the Sr resin column is small.

Radiostrontium,Sr resin,Automated radionuclide separator,Optimization

This research studied a scaling analysis for the selection of proper heat generation at tube for 1/4-scale storage vaults. First of all, the temperature field and velocity distribution of an original scale storage vault were analyzed and then numerical analysis of a 1/4-scale storage vault was performed to compare each model. The proper heat generation for a 1/4-scale storage vault, at which the temperature and velocity field of a 1/4-scale storage vault showed the best agreement with that of the original storage vault, was evaluated with proposed dimensionless parameters. The behavior of temperature and velocity of fluid in the 1/4-scale case were most similar to those of the original scale, using a heat flux 1.3 times higher than that seen in the original scale, which was approximately 190 W.

Storage vault,Scaling analysis,Dimensionless parameters,CFD

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

PWR Spent Nuclear Fuel,Dual purpose metal cask,Burnup Credit,MCNP,Criticality,Reactor operating parameter,End effect,Misload

Technetium-99m (99mTc) is an important, short-lived decay product of molybdenum-99 (99Mo), and it is considered the backbone of the modern nuclear diagnostic procedures. Since fission of 235U is the main source of production of 99Mo, either highly-enriched uranium (HEU) targets or low-enriched uranium (LEU) targets are irradiated in the research reactors. The use of LEU targets is being promoted by the international community to avoid the proliferation issues linked with the use of HEU. In order to define the waste management strategy at the planning stage of establishment of an LEU based 99Mo production facility, the impact of the use of LEU targets on the radioactive waste stream of the 99Mo production facility was analyzed. Because the volume of uranium waste is estimated to increase six times, the use of high uranium density targets and the utilization of hot isostatic pressing were recommended to reduce the increased waste volume from the use of LEU based targets.

Molybdenum-99,Low Enriched Uranium,Target,Radioactive Waste

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.

Spent Fuel Pool Islanding,Decommissioning,Regulatory Framework,Spent Fuel Management

This study was performed on the removal of Cs, one of the main high- radioactive nuclides contained in the high-radioactive seawater waste (HSW), by adsorption with IE911 (crystalline silicotitanate type). For the effective removal of Cs and the minimization of secondary solid waste generation, adsorption of Cs by IE911 (hereafter denoted as IE911-Cs) was effective to carry out in the m/V (ratio of absorbent weight to solution volume) ratio of 2.5 g/L, and the adsorption time of 1 hour. In these conditions, Cs and Sr were adsorbed about 99% and less than 5%, respectively. IE911-Cs could be also expressed as a Langmuir isotherm and a pseudo-second order rate equation. The adsorption rate constants (k2) were decreased with increasing initial Cs concentrations and particle sizes, and increased with increasing ratios of m/V, solution temperatures and agitation speeds. The activation energy of IE911-Cs was about 79.9 kJ/mol. It was suggested that IE911-Cs was dominated by a chemical adsorption having a strong bonding form. From the negative values of Gibbs free energy and enthalpy, it was indicated that the reaction of IE911-Cs was a forward, exothermic and relatively active at lower temperatures. Additionally, the negative entropy values were seen that the adsorbed Cs was evenly distributed on the IE911.

Cesium,Adsorption,Removal,Crystalline silicotitanate,Kinetics,Activation energy,High-radioactive seawater waste

It is necessary to develop an effective waste salt treatment technology for the minimization of radioactive waste generation from the pyroprocessing of spent nuclear fuel. For this reason, the separation characteristics of NdCl3 from LiCl-KCl eutectic salt in a reactive distillation process using Li2CO3 or K2CO3 were observed. NdCl3 was converted into oxychloride (NdOCl) or oxide (Nd2O3) in the reaction model between NdCl3 and the carbonates using HSC-Chemistry, and this result was confirmed in the reactive distillation test of the LiCl-KCl-NdCl3 system using the carbonates. Based on these results, the reactive distillation process conditions were determined to separate NdCl3 into an oxide form (Nd2O3) which can be easily fabricated into a final waste form.

Pyroprocessing,Spent nuclear fuel,LiCl-KCl eutectic salt,Reactive distillation,NdCl3

Decontamination is one of the crucial technologies that are applied during the decommissioning process of nuclear facilities to secure the safety of workers and to minimize the quantity of radioactive waste. Decontamination removes radionuclides on the surface of contaminated metal. Compared with decontamination for operational nuclear facilities, decontamination for nuclear power plants that are being decommissioned needs to remove the more and thicker surface using more aggressive agents or specially developed equipment. This paper analyzed the factors to be considered before planning the decontamination, representative decontamination technologies, and their application procedure,etc .

Decontamination,Design,Methodology,Decommissioning,Criteria

Currently, 23 nuclear power plants are in operation at Kori, Uljin, Younggwang and Wolsong site and a reference deep geological disposal system has been developed for the spent fuels generated by them. The reference spent fuel for this disposal system has 4.5wt% of initial enrichment, 55 GWd/MtU of burn-up, and 40 years of cooling time. In this paper, to improve disposal efficiency and economic feasibility, the characteristics of spent fuels from nuclear power plants, such as type and burn-up, were reviewed. A disposal canister concept for shorter length and relatively lower burn-up spent fuels than the reference spent fuels was developed. Based on this canister concept, thermal analyses were carried out and a deep geological disposal concept was proposed. Measures of disposal efficiency such as unit disposal area and disposal density were compared between this disposal system and the reference disposal system. Also, economic feasibility, such as the volume reduction of copper, cast iron, and bentonite, was analyzed and the results of these analyses showed that the disposal system proposed in this paper has an efficiency of at least 20%. These results could be used for establishing spent fuel management policy and designing practical disposal systems for spent fuels.

Spent fuels,Deep geological disposal(DGD) system,Reference spent fuels,Low burn-up,High burn-up,Disposal efficiency

The reaction rates of ZIRLO cladding hulls with cross-section openings were investigated using a thermo-gravimetric analysis system in order to identify whether selective recovery of Zr from oxidized cladding hulls is possible. The experimental results showed that an oxidized ZIRLO cladding hull was not reactive with chlorine gas at 400℃. However, providing fresh cross-sections on one or both ends of the ZIRLO hulls enabled a chlorination reaction. This reaction was completed after 8 hours; a 14% increase on the 7 hours seen with a bare ZIRLO cladding hull. The Sharp-Hancock plot analysis results revealed that the contracting volume model is the best for describing the reaction between the cross-section opened ZIRLO hulls and chlorine gas under the condition of this work. It was concluded that the chlorination process can be employed for oxidized ZIRLO cladding hulls by providing cross-section openings.

ZIRLO,Chlorination,Reaction rate model,Thermo-gravimetric analysis,Cladding hull waste

Radioactive waste processing and disposal is a major problem that needs to be fully addressed by countries that use nuclear power. In recent years, only a few countries have made substantial progress on this issue, and the French stepwise legislative approach on radioactive waste management is evaluated as a successful case. For South Korea, it is still necessary to prepare an adequate national policy for dealing with radioactive waste so the French model has been suggested as a direction of policy making in Korea. Based on comparisons of technical status and sociopolitical indexes in both countries, then this study suggests that the French legislative model is a valid one which may be applicable to the Korean context, especially in relation to resource recycling and social acceptance enhancement strategies.

Radioactive waste,Waste management,Legislative approach,French model,Social acceptance

Developing novel anode materials to replace the Pt anode currently used in electrolytic reduction is an important issue on pyroprocessing. In this study, the electrochemical behavior of TiN was investigated as the conductive ceramic anode which evolves O2 gas during the reaction. The feasibility and stability of the TiN anode was examined during the electrolytic reduction of UO2. The TiN anode could electrochemically convert UO2 to metallic U in a LiCl–Li2O molten salt electrolyte. No oxidation of TiN was observed during the reaction; however, the formation of voids in the bulk section appeared to limit the lifetime of the TiN anode.

Pyroprocessing,Electrolytic reduction,UO2,Anode,TiN

A commercially available copper mesh was investigated as an iodine off-gas capturing medium for pyroprocessing, with an aim to replace costly silver based adsorbents. Theoretical calculation results suggested that the reaction between metallic copper and gaseous iodine will occur spontaneously to produce copper iodide in the temperature range of 100 ~ 500℃. The effect of the reaction temperature on iodine capturing efficiency was investigated by experimentation, and it was found that 5 and 6 wt% of iodine (initial mass 2.0 g) was captured by a single copper mesh (0.26 g) at 300 and 400℃, respectively. The repeated experimental results also suggested that copper utilization can be increased with the help of the spontaneous detachment of the reaction product (CuI) from a copper mesh. The formation of the CuI phase was confirmed using the X-ray diffraction technique, and the surface morphology of the reaction product was observed using scanning electron microscopy.

Iodine capture,Copper iodide,Copper mesh,Off-Gas treatment,Pyroprocessing

Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

Nuclear fuel cycle,External cost,Accident cost,Supply security cost

Radionuclide deposited on the surface of several internal and external systems in a nuclear power plant is created by the activation of corrosion products from nuclear reactor structural materials and fission products. Especially, the constant contact between water and the surface corrodes the inside where primary system makes coolants and corrosion products mixed. Also, these are circulated along the systems. For comparing models, CRUDTRAN, DISER, MIGA-RT and CPAIR codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor that are used at the stage of designing. The corrosion products behavior of domestic PWR primary system was predicted by using CRUDTRAN. This study aims to increase the reliability of corrosion product evaluation model by comparing the actual values and calculated values with the data of a Westing House-type Nuclear Power Plant.

Corrosion product,CRUDTRAN,DISER,MIGA-RT,CPAIR

It is necessary to investigate the electrodeposition behavior of uranium and other elements on the cathode in the electrorefining process to recover the uranium selectively from the reduced metals of the electrolytic reduction process since transuranic elements and rare earth elements is dissolved in the LiCl-KCl eutectic salt. Study on separation factors of U, Ce, Y and Nd based on U and Ce was performed to investigate the deposition behavior of the cathode with respect to the concentration of rare earth elements in LiCl-KCl eutectic salt. After electrorefining with constant current mode by using Ce metal as a sacrifice anode, the contents of U, Ce, Y and Nd in the salt phase and the deposit phase of the cathode were analyzed, and separation factors of the elements were obtained from the analyses. Securing conditions of pure uranium recovery in the elctrorefining process was investigated by considering the separation factors with respect to UCl3 and CeCl3/UCl3 ratio.

Electrorefining,Uranium deposit,Rare earth element,Separation factor,Ce sacrifice anode

External cost is cost imposed on a third party when producing or consuming a good or service. Since the 1990s, the external costs of nuclear powered electricity production have been studied. Costs are a very important factor in policy decision and the external cost is considered for cost comparison on electricity production. As for nuclear fuel cycle, a chosen technology will determine the external cost. However, there has been little research on this issue. For this study, methods for external cost on nuclear power production have been surveyed and analyzed to develop an approach for evaluating external cost on nuclear fuel cycles. Before the Fukushima accident, external cost research had focused on damage costs during normal operation of a fuel cycle. However, accident cost becomes a major concern after the accident. Various considerations for external cost including accident cost have been used to different studies, and different methods have been applied corresponding to the considerations. In this study, the results of the evaluation were compared and analyzed to identify methodological applicability to the external cost estimation with nuclear fuel cycles.

Cost evaluation,External cost,Accident cost,Economics

The buffer is a key component of the engineered barrier system in a high-level radioactive waste (HLW) repository. The present study reviewed the requirements and functional criteria of the buffer reported in literature, and also based on the results, proposed an approach to establish a buffer concept which is applicable to an HLW repository in Korea. The hydraulic conductivity, radionuclide-retarding capacity (equilibrium distribution coefficient and diffusion coefficient), swelling pressure, thermal conductivity, mechanical properties, organic carbon content, and illitization rate were considered as major technical parameters for the functional criteria of the buffer. Domestic bentonite (Ca-bentonite) and, as an alternative, MX-80 (Na-bentonite) were proposed for the buffer of an HLW repository in Korea. The technical specifications for those proposed bentonites were set to parameter values that conservatively satisfy Korea';s functional criteria for the Ca-bentonite and Swedish criteria for the Na-bentonite. The thickness of the buffer was determined by evaluating the means of shear behavior, radionuclide release, and heat conduction, which resulted in the proper buffer thickness of 0.25 to 0.5 m. However, the final thickness of the buffer should be determined by considering coupled thermal-hydraulic-mechanical evaluation and economics and engineering aspects as well.

Buffer,Functional criteria,Buffer concept,Engineered barrier system,High-level waste repository

The precise determination of the activity for each radionuclide in environmental samples requires the self-absorption correction factor. In this research, we derived the self-absorption correction factor for three p-type high purity germanium detectors using the Monte Carlo code MCNPX. These detectors have different characteristics such as crystal diameter, height and size of the core. We compared the calculated full-energy peak efficiency with the experimental value using a standard sample with 1 g/cm3 density and verified the modeling. We simulated the dependency of the full-energy peak efficiency on the 0.3, 0.6, 0.9, 1.0, 1.2 and 1.5 g/cm3 samples and obtained the corresponding self-absorption correction factor. The self-absorption correction factors calculated for the three detectors differ by less than 1% over most of the energy range and sample densities considered. This indicates that the self-absorption correction factors are independent of the crystal characteristics of HPGe detector.

Self-absorption correction factor,High purity germanium detector,Monte Carlo

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

Spent fuel,Interim storage,Dry storage,Rock cavern storage,Feasibility analysis

A high-throughput electrorefiner has been developed for commercialization use by enhancing the uranium recovery from the reduced metal which is produced from the oxide reduction process. It is necessary to scrap and effectively collect uranium dendrites from the surface of the solid cathode for high yield. When a steel electrode is used as the cathode in the electrorefining process, uranium is deposited and regularly stuck to the steel cathode during electrorefining. The sticking coefficient of a steel cathode is very high. In order to decrease the sticking coefficient of the steel cathode effectively, vibration mode was applied to the electrode in this study. Uranium dendrites were scraped and fell apart from the steel cathode by a vibration force. The vibrational scraping of the steel cathode was compared to the self-scraping of the graphite cathode. Effects of the applied current density and the vibration stroke on the scraping of the uranium dendrites were also investigated.

Electrorefining,High-throughput,Scraping,Sticking coefficient,Vibrational scraping

Electrochemical reduction has previously been reported for uranium oxide and mixed oxide nuclear fuel (uranium oxide, plutonium oxide). The laboratory scale electrochemical reduction of plutonium oxide powder is demonstrated in CaCl2-1wt%CaO. The plutonium oxide contained within a permeable steel basket cathode is sacrificed during the process. A graphite anode is also employed during the reduction, leading to a significant contamination of the product.

Plutonium oxide,Molten salt,Electrochemical reduction

The metal product from the electrolytic reduction of uranium oxide in LiCl molten salt retains about 10 ~ 20wt% of the residual salts. Salt vacuum distillation is conducted to separate the residual salt from the metal product and well-performed in a glove box in an argon atmosphere. A dimensionless analysis of the characteristics of a salt vacuum distiller needs to be scaled up for a high capacity process. The vacuum distillation apparatus can be of two different sizes (M-type and P-type). M-type equipment is small in size and exhibits a high recovery rate of more than 95%. A comparison of two salt vacuum distillers was conducted with the dimensionless analysis method. Heat and fluid flows are strongly influenced by the structure of the apparatus and phase transition phenomena of vacuum distillation. The several dimensionless parameters were calculated at the nozzle throat located between the evaporator and the receiver and at different operating temperatures. Both salt vacuum distillers had similar trends of dimensionless parameters. However, the distributions of the parameter values varied with the nozzle geometry and size. The results of the dimensionless analysis will aid the scaling up of the salt distillation process.

Salt distillation,Pyroprocessing,Dimensionless analysis,Scale-up

A new facility, known as Pyrel III, has been installed at ENEA laboratories for pyrochemical process studies under inactive conditions. It is a pilot plant which allows electrorefining and electroreduction experiments to be conducted on simulated fuel. The main component of the plant is a zirconia crucible. The crucible is heated by a furnace which is supported in an externally water-cooled well under the floor of a steel glove-box, where an argon atmosphere is maintained by a continual purge of about 10 L·min-1. The vessel is loaded with LiCl-KCl eutectic salt (59-41 mol%) and is currently operated at 460°C. Several improvements on Pyrel II (the previous operating plant) have been introduced into Pyrel III. They are described in detail, together with the results from the first experimental campaign which used lanthanum metal. Moreover, studies about the treatment of chloride salt wastes from pyroprocesses have been conducted in parallel. They follow two main routes: on one hand, a matrix termed sodalite, a naturally occurring mineral containing chlorine, has been synthesized from a mix of nepheline, simulated exhausted salts and glass frit; on the other hand, a novel method proposed by Korea Atomic Energy Research Institute (KAERI) is under assessment. The final waste forms have been fully characterized with the support of the Politechnique of Milan, by means of density measurements, thermal analysis, and stereomicroscopy observations, FTIR, XRD, and RAMAN spectra, as well as leach tests under static conditions.

Pyroprocesses,Molten Salts,Chloride wastes,Sodalite,SAP matrix,Leach tests

Voltammetry has shown promise as a method to estimate the concentrations of actinides in the molten LiCl-KCl used as an electrolyte in spent nuclear fuel electrorefiners. This salt typically contains several actinides in addition to many active metal fission products (rare earths, Group I & II metals). However, most of the voltammetry studies to date have focused on a single actinide or lanthanide in eutectic LiCl-KCl. This paper examines experimental and analytical techniques that can be used to estimate the concentration of a molten salt mixture containing both lanthanum (III)- and gadolinium(III)-chloride in eutectic LiCl-KCl. The aspects of the experimental procedures and setup that are unique to a multi-lanthanide mixture are briefly discussed. Experimental results from qualitative and quantitative analyses of cyclic voltammetry and open-circuit potentiometry are presented. Due to the close proximity of their standard potentials, extensive analytical work is required to estimate the concentrations. Two approaches are used in this work: peak separation and multivariate analysis. The merits of these two methods will be analyzed and discussed.

Voltammetry,Lanthanide,Electrorefiner,Peak separation,Principle component analysis

A dynamic tube pressure method was proposed for a liquid level measurement. The reliability of our in-house manufactured prototype level measurement system was investigated for water samples in a vial as a preliminary study. The prototype instrument, equipped with a stepper motor and a differential pressure sensor, was used to measure the travel distance of the tube from an initial zero position to the liquid surface. Unlike a conventional bubbler method, our dynamic tube pressure method is based on the abrupt changes in the tube pressure to directly detect the liquid surface. Optimum conditions were determined from the measurements of the travel distance with different-sized tubes at various ambient base pressures and various descending tube speeds. In addition, we proposed a gravimetric calibration method. In the gravimetric calibration method, the travel distances are used instead of the liquid level, which can be obtained from the measurement data of the travel distance. The travel distance versus the weight calibration curve showed a good linear relationship (R2 = 0.9999), and standard deviations of the travel distance over the whole range of experimental conditions were less than 0.1 mm. In a further study, our present system will also be used in the measurement of density and surface tension by minimizing the contact time with high-temperature and highly-corrosive molten salts.

Liquid level,Dynamic tube pressure method,Gravimetric calibration method,Water

The US Department of Energy’s Idaho National Laboratory (INL) has been operating a molten salt electrorefiner at their facility since 1996. The baseline method for disposal of the radioactive salt is the ceramic waste process which generates glass bonded sodalite loaded with chloride salts. This process starts with the high temperature absorption of the salt into zeolite-4A. The salt-loaded zeolite is then blended with glass frit and heated to form a sintered, glass-bonded sodalite. INL is currently assessing alternatives for disposal of the ER salt because of the lengthy processing times, costly equipment and large volume of waste associated with the baseline process. An alternative process was studied, where protonated zeolite was used instead of alkali metal-substituted zeolite. It was found that the metals contained in the salt can replace the protons in the zeolite which are evolved via formation of HCl. From the standpoint of generating a nuclear waste form, the evolution of HCl gas should reduce the weight of the final waste. It has been estimated that the volume of waste produced from immobilizing the INL electrorefiner salt could be reduced by a factor of three using this process followed by sintering the fission product loaded zeolite. Equipment requirements in the hot cell would be significantly simplified, and the time to process all of the waste salt would be reduced by almost a factor of 4. An investigation into the new process has been presented here.

H-Y zeolite,Ion exchange,Salt Occlusion,Nuclear Waste Disposal,Waste Salt Immobilization

A full-scale process has been developed to immobilize fission products that accumulate within the Mark IV electrorefiner (ER) electrolyte at Idaho National Laboratory. ER salt was blended with treatment additives, followed by pressureless consolidation (PC) in a furnace to produce a durable ceramic waste form (CWF). The goal is the development of a process to consolidate actual radioactive ER salt into a form suitable for transportation and disposal. Four batches (300 to 400 kg per batch) of full-scale pre-qualification material preparation runs have been prepared. From these four batches of nonradioactive salt-loaded surrogate material, three full-scale PC trials have been conducted. The first PC test run, established equipment parameters with a basic CWF container design. The second trial included a modified CWF container design, real-time measurement of CWF consolidation, and an audio recording to identify cracking during the CWF cool-down. During the third trial, salt was doped (from the fourth material preparation batch) to create a nonradioactive salt material and to more closely represent actual ER salt. The second and third trials were also used to validate a model developed for the CWF. The CWF model is beneficial for understanding and predicting the physical processes that occur during the heat cycle. This would be particularly useful when the CWF is located in a hot cell, which makes accessing and examining a CWF difficult.

Nuclear Waste Form,Ceramic Waste Form

Waste treatment technology for the separation and solidification of radioactive nuclides generated from the pyrochemical process has been intensively studied to achieve the reduction of radioactive waste volume. The present study reports the separation efficiency of group II fission products in LiCl waste salt generated from a electrolytic reduction process through a layer-melt crystallization method using Sr and Ba nuclides as a surrogate material of group II fission products. The concentrated group II nuclides are converted into stable oxide form in consideration of solidification by a conversion/distillation process, where selective oxidation of group II nuclides proceeds by Li2O oxidant and residual salts are removed by a vacuum distillation process. Finally, to immobilize separated group II nuclides, a preliminary solidification study was conducted using SiO2-B2O3-Al2O3 matrix, and high density glass-based waste form was fabricated under 50wt% waste loading of strontium oxide surrogate material. Through the verification of the crystallization, conversion/distillation, and solidification processes, the treatment flow for the separation and solidification of group II fission products in LiCl waste salt has been established.

Pyrochemical Process,LiCl Waste Salt,Nuclide Separation,Solidification,Waste Form

A large amount of acidic waste solution is generated from the practical electrokinetic decontamination equipments for the remediation of soil contaminated with uranium. After filtration of uranium hydroxides formed by adding CaO into the waste solution, the filtrate was recycled in order to reduce the volume of waste solution. However, when the filtrate was used in an electrokinetic equipment, the low permeability of the filtrate from anode cell to cathode cell due to a high concentration of calcium made several problems such as the weakening of a fabric tamis, the corrosion of electric wire and the adhension of metallic oxides to the surface of cathode electrode. To solve these problems, sulfuric acid was added into the filtrate and calcium in the solution was removed as CaSO4 precipitate. A decontamination test using a small electrokinetic equipment for 20 days indicated that Ca-removed waste solution decreased uranium concentration of the waste soil to 0.35 Bq/g, which is a similar to a decontamination result obtained by distilled water.

Electrokinetic Remediation,Waste solution,Decontamination,Uranium,Reuse,CaSO4

Molten salt electrorefining process achieves uranium deposits at cathode using an electrochemical processing of spent nuclear fuel. In order to recover pure uranium from cathode deposit containing about 30wt% salt, the adhered salt should be removed by cathode process (CP). The CP has been regarded as one of the bottle-neck of the pyroprocess as the large amount of uranium is treated in this step and the operation parameters are crucial to determine the final purity of the product. Currently, related research activities are mainly based on experiments consequently it is hard to observe processing variables such as temperature, pressure and salt gas behavior during the operation of the cathode process. Hence, in this study operation procedure of cathode process is numerically described by using appropriate mathematical model. The key parameters of this research are the amount of evaporation at the distillation part, diffusion coefficient of gas phase salt in cathode processor and phase change rate at condensation part. Each of these conditions were composed by Hertz-Langmuir equation, Chapman-Enskog theory, and interphase mass flow application in ANSYS-CFX. And physical properties of salt were taken from the data base in HSC Chemistry. In this study, calculation results on the salt gas behavior and optimal operating condition are discussed. The numerical analysis results could be used to closely understand the physical phenomenon during CP and for further scale up to commercial level.

Cathode process,Ansys-CFX,Hertz-Langmuir eq.,Pyroprocess,Numerical analysis

A parametric sensitivity to the annual exposure dose rate to the farming exposure group has been probabilistically carried out for three principal elements associated with the nuclide transport behavior in the near-field of the pyroprocessed waste repository system. Credit time for both metal and ceramic containers, annual nuclide release rete, and the degree of loss of bentonite buffer around the container are selected as the elements and investigated for important nuclides. All the elements are shown to be sensitive to the results. Methodology studied through this study and the results are expected to make a good feedback to the repository design. As a follow-up study, separated in Part 2, the A-KRS will be deterministically assessed and then compared among each other with the normal, the worst, and the best case scenarios associated with their extreme values these elements could have.

A-KRS,Pyroprocessed waste,Safety assessment,Probabilistic sensitivity,GoldSim

A parametric sensitivity to the annual exposure dose rate to the farming exposure group has been deterministically carried out for three principal elements identified in the near-field of the pyroprocessed waste repository system as a series study of Part 1 of the coupled paper with the same title. Credit time for both metal and ceramic containers, annual nuclide release rete and the degree of loss of bentonite buffer around the container are selected and investigated deterministically for important nuclides. To this end the A-KRS has been assessed and then compared among each other with the normal, the worst, and the best case scenarios associated with their extreme values these elements could have. All the elements are shown to be sensitive to the results as was in Part 1. Methodology studied through this study and the results are expected to make a good feedback to the repository design.

A-KRS,Pyroprocessed waste,Safety assessment,Deterministic sensitivity,GoldSim

Academic interests in ionic liquid (IL) technologies have been extended to the nuclear industry and the applicability of ionic liquids for processing radioactive materials have been investigated by many researchers. A number of studies have reported interesting results with respect to the spectroscopic and electrochemical behaviors of metal elements included in spent nuclear fuels. The measured and observed properties of metal ions in TBP(tri-butyl phosphate) dissolved ILs have led the development of alternative technologies to traditional aqueous processes. On the other hand, the electrochemical deposition of metal ions in ILs have been investigated for the application of the solvents to aqueous as well as to non-aqueous processes. In this work, a review on the application of ILs in nuclear fuel cycle is presented for the purpose of categorizing and summarizing the notable researches on ILs.

Ionic liquids,Spent nuclear fuel,Extraction,Electrochemical deposition,Radiochemical stability

The public';s access to the disposal facilities should be restricted during the institutional control period. Even after the institutional control period, disposal facilities should be designed to protect radiologically against inadvertent uman intruders. This study is to assess the effective dose equivalent to the inadvertent intruder after the institutional control period thorough the GENII. The disposal unit was allocated with different kind of radioactive waste and the effects of the radiation dose to inadvertent intruder were evaluated in accordance with the institutional control period. As a result, even though there is no institutional control period, all were satisfied with the regulatory guide, except for the disposal unit with only spent filter. However, the disposal unit with only spent filter was satisfied with the regulatory guide after the institutional control period of 300 years. But the disposal unit with spent filter mixed with dry active waste could shorten the institutional control period. So the institutional control period can be reduced through the mixing the other waste with spent filter in disposal unit. Therefore, establishing an appropriate plan for the disposal unit with spent filter and other radioactive waste will be effective for radiological safety and reduction of the institutional control period, rather than increasing the institutional control period and spending costs for the maintenance and conservation for the disposal unit with only spent filter.

Human intrusion assessment,Near surface disposal,GENII,Post-drilling scenario,Waste emplacement,Institutional control period

Unsaturated hydraulic conductivity of the bentonite-buffer was evaluated using the relative humidity data. The method for calculating unsaturated hydraulic conductivity was deduced from the general analytical equation representing the movement of water in unsaturated media, which was applied to the experimental results of water infiltration tests for identifying the behavior of unsaturated hydraulic conductivity according to the water saturation. Unlike the saturated condition, the hydraulic gradient and water flux were irregularly changed, and the unsaturated hydraulic conductivity was increased with increasing the experimental time. Swelling of bentonite grains due to the water absorption increased the volume and size of pore within bentonite, resulting in the increase of water velocity and unsaturated hydraulic conductivity. This result suggested the necessity of further investigation on the correlation between the swelling degree of bentonite-buffer and unsaturated hydraulic conductivity. The method used in this study can be useful technique for evaluating long-term hydraulic performance of bentonite-buffer in the radioactive waste disposal system.

Unsaturated hydraulic conductivity,Relative humidity,Bentonite-buffer,Long-term hydraulic performance

We have developed a smart decontamination device for small-size radioactive scrap metal (SSRSM) necessarily generated from nuclear facilities. This is a multi-modal device such as rotation of magnetic field focusing on the region containing the abrasion pins placed around target and ultrasonic cleaner. Additionally, in order to increase the decontamination efficiency we have modified some configuration of the device so that it could work on them evenly and totally. With the Optimal operating for operation of the new device, we tried to decontaminate some various metal selected as a sample during 15 minutes sequentially using each method, magnetic and ultrasonic device. As a result, the range of decontamination factor has been highly increased to 18~56. After decontamination, all samples were found its activity less than background level.

SSRSM(Small-size Radioactive Scrap Metal),Combined decontamination device,Equipment development,Decontamination factor,Magnetic Abrasive

The repository for the disposal of LILW which is generated from nuclear power plants and industries is expected to be completed in 2014. For the disposal of LILW, it is important to secure a disposal facility itself, but it is also very important to establish a reasonable charging system which all shareholders are satisfied with. Korea’s disposal fee for LILW is higher than other countries’ fee, which is a burden to waste generators as well as the waste management organization. The partial reason for the high disposal fee is put on the high social and construction cost when compared with other countries. However the major reason is put on the excessive borrowing cost that is used for the construction of the LILW disposal facility. In this study, we proposed the way to reduce the excessive borrowing cost for sustainable project managements of LILW disposal by analyzing a cost structure.

Low level radioactive waste disposal,Radioactive waste disposal cost,Borrowing cost of LILW disposal

In the case of the direct measurement of 226Ra using a HPGe gamma-ray spectrometer, the interference between gammarays with 186.21 keV of 226Ra and 185.7 keV of 235U should be corrected to calculate the net peak area in the nergy spectrum. In general, it is very difficult to conduct peaks stripping with difference of about 0.5 keV, although a HPGe with the superior resolution is applied and the maximum channels is applied to the spectrometer. In this study, several interference correction techniques in the direct measurement were surveyed to evaluate the feasibility for the measurement of 226Ra using the gamma-ray spectrometery. Applying the interference corrections to the analysis of raw materials and by-products, the method validation for the direct measurement of 226Ra was conducted by evaluating the measurement uncertainty, linearity, and range. As a result, the optimum method of the interference correction was selected by comparing with the indirect measurement of which progenies of 226Ra, such as 214Pb and 214Bi, were analyzed in the secular equilibrium state.

226Ra,Gamma-ray spectrometry,Interference correction,HPGe detector

The worldwide Status-of-Art survey for the in-situ experiments of the engineered barrier system for HLW underground disposal was performed as a preliminary action for the design of the in-situ demonstration at KURT. Some nations, which have executed or is ongoing the in-situ experiments at their underground research facilities, were summarized in this review. The demonstration projects reviewed were TBT/Sweden-France, LOT/Sweden, HE-E/Switzerland, PRACLAY/Belgium, FEBEX/Spain, HORONOBE/Japan, and BCE/Canada. The investigated items for the projects were mainly their purposes, constitutional structures, test conditions, monitoring parameters and the measuring tools, and test results. In this review, the hardware design and the assembling of the test system were more concentrated rather than their experimental results, because the purpose of this review is to achieve the necessary information for the practical design of the in-situ experiment to be installed at KURT. A mid scale in-situ demonstration of EBS at KURT, that is called IN-DBES, will be launched right after the completion the expanding project of KURT in 2015. It is hoped that the structural design, installing methods, hardware equipments required in the establishment of IN-DEBS will be referred on this review.

Underground disposal,Engineered barrier system,In-situ demonstration,Bentonite buffer

If the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3~5 km and more stable rock formation, it has several advantages. For example, (1)significant fluid flow through basement rock is prevented, in part, by low permeability, poorly connected transport pathways, and (2)overburden self-sealing. (3)Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose-critical radionuclides at the depth. Finally, (4) high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept to the deep geological disposal concept(DGD), very deep borehole disposal(DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, for the preliminary applicability analyses of the DBD system for the spent fuels or high level wastes, the DBD concepts which have been developed by some countries according to the rapid advance in the development of drilling technology were reviewed. To do this, the general concept of DBD system was checked and the study cases of foreign countries were described and analyzed. These results will be used as an input for the analyses of applicability for DBD in Korea.

Spent fuels,High-level radioactive wastes,Deep geological disposal(DGD),Deep borehole disposal(DBD),Safety,Alternative Technology

It is very important to properly understand such “Transport Pathways” elements as “Pipe” and “Cell” pathways in commercial GoldSim Transport Module (GTM) for developing higer quality models and programs for performance assessment of complex radioactive waste repositories. With an illustrative case under an earthquake scenario, by which an increasement in the groundwater flow rate occurs though the geological medium, ways of avoiding possible modeling errors in the nuclide transport modeling in the radioactive waste repository system for its safety assessment by utilizing such pathways are discussed and a proper usage of the pathways is proposed.

GoldSim (GTM),Nuclide transport pathway,Modeling methodology

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating uantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

Activation inventory,Corrosion product,CORA,PACTOLE,CRUDSIM,CREAT

In order to use the nuclear energy as the sustainable energy source, the safe and efficient management of radioactive wastes generated from the nuclear fuel cycle including NPP decommissioning is one of the most important factors. The establishment of acceptance criteria for very low level radioactive wastes generated from decommissioning of nuclear power plant in a large quantity is seemed to play a key role for developing a radioactive wastes disposal strategy as well as NPP decommissioning strategy. In this thesis, we want to review the acceptance criteria of low-and-intermediate-level radioactive wastes in this country through the analysis of other country’s acceptance criteria.

Decommissioning wastes,Very low level radioactive waste disposal,Acceptance criteria of VLLW

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.

Decommissioning,Radiological Inventory,In-Situ Object Counting System,Reactor Head,Direct measurement

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. An electrolytic reduction of the pyroprocessing is a process to reduce oxides into metals using LiCl as an electrolyte and requires a post-treatment process due to the inclusion of residual salt in porous metal products. A vacuum distillation has been adopted for various molten salt systems and could be applied to the post-treatment process of the electrolytic reduction. The residual salt in the metal products includes LiCl, alkali chlorides, and alkaline earth chlorides. In this paper, vapor pressures of chlorides have been estimated and the composition changes on the residual liquid during the vacuum distillation process have been calculated. A model combining a material balance and vapor-liquid equilibrium relations has been proposed under a constant vapor discharging flow rate and liquid composition changes have been calculated using the vapor pressures with respect to a dimensionless time. The behaviors have been compared with temperature and molten salt composition changes to simulate the process condition variation. The distillation of the residual salt has been dominated by LiCl which is the main component of the salt and CsCl of which vapor pressure is higher than that of LiCl would be readily removed. RbCl exhibits similar vapor pressure with LiCl and maintains its composition. However, SrCl2 and BaCl2 of which vapor pressures are much lower than that of LiCl are concentrated with time and expected to be possibly precipitated during the distillation when the initial compositions are increased.

Pyroprocessing,Electrolytic reduction,Vacuum distillation,Chloride vapor pressure,Vapor-liquid equilibrium

Using results of groundwater flow system modeling for a hypothetical deep geological repository site, a distribution of groundwater flow rates at the disposal depth was analyzed and a method of applying this distribution to a safety assessment for a disposal of radioactive wastes was suggested. The distribution of groundwater flow rates was produced by hydraulic heads simulated from regional and local scale groundwater flow models for the hypothetical disposal site. The flow rates at the locations where deposition holes would be located were estimated. These rates were normalized by the maximum of the flow rates in order to probabilistically illustrate a possibility of canister failures at the deposition holes. From the normalized distribution, probabilistic expectations for mass discharges of radionuclides released from the canisters assumed to be failed were calculated and compared with those deterministically estimated under the assumption that the canisters at the same deposition holes were definitely failed. The suggested method can be contributed to constructing a methodology for safety assessment of a geological repository by reflecting natural conditions of a disposal site in more detail.

Groundwater flow at disposal depth,Groundwater flow simulation,Probabilistic analysis,Selection of a deposition hole

Crevice corrosion tests were conducted to examine the corrosion properties of HYBRID (HYdrazine Base Reductive metal Ion Decontamination) which was developed to decontaminate the PWR primary coolant system. To compare the corrosion properties of HYBRID with commonly existing decontamination agents, oxalic acid (OA) and citric oxalic acid (CITROX) were also examined. Type 304 Stainless Steel (304 SS) and Alloy 600 which are major components of the primary coolant system in Pressurized Water Reactor (PWR) were evaluated. Crevice corrosion tests were conducted under very aggressive conditions to confirm quickly the corrosion properties of primary coolant system structure components which have high corrosion resistance. Pitting and IGA were occurred in crevice surface under OA and CITROX conditions. But localized corrosion was not observed under HYBRID condition. Very low corrosion rate of less than 1.3×10-3 μm/h was observed under HYBRID condition for both materials. On the other hand, under OA condition, Alloy 600 indicated comparatively uniform corrosion rate of 4.0×10-2 μm/h but 304 SS indicated rapid accelerated corrosion in lower case than pH 2.0. In case of HYBRID condition, general corrosion and crevice corrosion were scarcely occurred. Therefore, material integrity of HYBRID in decontamination of primary coolant system in pressurized water reactor (PWR) reactor was conformed.

PWR,HYBRID,Crevice corrosion,Reductive decontamination agent,Oxalic acid

Corrosion products are released to the primary coolant in the corrosion process of structural materials. They are deposited on fuel surfaces and activated on exposure to a neutron flux with formation of radionuclides that can become incorporated into out-of-core surface films. To get a clear understanding of activated crud formation process, the specific activity and core residence time of fuel cruds was calculated as a function of exposure time to the core neutron flux on the assumption that parent nuclide is being deposited continuously. Fuel cruds were sampled in the fuel scraping campaign from Hanul Unit 1 Cycle 17 and analyzed for elemental concentration and radioisotope activity.

Corrosion product,Fuel crud,Core residence time,Specific activity

Securing the radiological safety is a prerequisite for the safe management of the naturally occurring radioactive materials (NORM) which cannot be reused. This becomes a crucial focus of our R&D efforts upon the implementation of the Act on Protective Action Guidelines against Radiation in the Natural Environment. To secure the safety, the establishment of technical bases and procedures for securing radiological safety related to the disposal of NORM is required. Thus, it is necessary to analyze the characteristics, to collect the data, to have the radiological safety assessment methodologies and tools, to investigate disposal methods and facilities, and to study the effects of the input data on the safety for the NORM wastes. Here, we assess the environmental impact of the NORM waste disposal with respect to the major domestic and foreign NORM characteristics. The data associated with major industries are collected/analyzed and the status of disposal facilities and methodologies relevant to the NORM wastes is investigated. We also suggest the conceptual design concept of a landfill disposal facility and the management plan with respect to the major NORM wastes characteristics. The radionuclide pathways are identified for the atmospheric transport and leachate release and the environmental impact assessment methodology for the NORM waste disposal is established using a relevant code. The assessment and analysis on the exposure doses and excessive cancer risks for the NORM waste disposal are performed using the characteristics of the representative domestic NORM wastes including flying ash, phosphor gypsum, and redmud. The results show that the exposure dose and the excessive cancer risks are very low to consider any radiation effects. This study will contribute to development in the areas of the regulatory technology for securing radiological safety relevant to NORM waste disposal and to the implementation technology for the Act.

NORM,NORM waste disposal,Radiological safety,Management of NORM waste

Surface oxidation behavior of ZIRLO (ZIRconium Low Oxidation) hulls was investigated using an X-ray photoelectron spectroscopy (XPS) technique. The effects of oxidation time (10-336 h at 500℃) and temperature (400-700℃ for 10 h) were studied. Deconvolution results of the hulls oxidized at 500℃ revealed that a ZrO2 phase appeared after 24 h (11.86%), and an increase in the ZrO2 ratio was observed when the hulls were oxidized for 336 h (17.93%). On the other hand, the ZrO phase which employed 5.68% in the 10 h oxidized sample disappeared when the oxidation time increased to 24 h. The XPS results also showed that an increase in the oxidation temperature resulted in an increase in the ratio of ZrO, which increased from 0 to 5.68, 8.31, and 9.16% when the oxidation temperature increased from 400 to 500, 600, and 700℃, respectively. ZrO2 phase was observed only in the sample that was oxidized at 700℃. The mechanism of ZrO formation was not conclusive, but it was suggested that a formation of hydroxide might have been accelerated at elevated temperatures leading to a formation of a Zr(OH)4 phase. The relationship between the surface oxidation status of the hulls oxidized at 500℃ and their chlorination reaction feasibility was discussed, and it was suggested that the thickness of the oxide layer is an important parameter that determines the chlorination reaction feasibility.

X-ray photoelectron spectroscopy,Cladding hull,ZIRLO,Oxidation,Chlorination

Pyroprocess for treating spent nuclear fuels has been developed based on electrochemical principles. Process simulation is one of the important methods for process development and experimental data analysis and it is also a necessary approach for pyroprocessing. To date, process simulation of pyroprocessing has been focused on electrorefining and there have been not so many investigations on electrolytic reduction. Electrolytic reduction, unlike electrorefining, includes specific features of gas evolution and porous electrode and, thus, different equations should be considered for developing a model for the process. This study summarized required concepts and equations for electrolytic reduction model development from thermodynamic, mass transport, and reaction kinetics theories which are necessitated for analyzing an electrochemical cell. An electrolytic reduction cell was divided and equations for each section were listed and, then, boundary conditions for connecting the sections were indicated. It is expected that those equations would be used as a basis to develop a simulation model for the future and applied to determine parameters associated with experimental data.

Pyroprocessing,Electrolytic reduction,Model,Governing equations,Simulation

Nuclear education & training is an important issue for sustainability of nuclear energy and the safety of the nuclear power plant. The purpose of this paper is to present policy initiatives for establishment of the national nuclear education & training system. It analyzed current status of nuclear manpower and nuclear education & training systems of Korean nuclear organizations and government strategic plans for nuclear manpower education & training. The features of the current nuclear education & training in Korea are institutional diversification and decentralization in Industry-University-Research system. However, linkages and cooperation systematically integrated between institutions are very weak. In addition, duplicated education & training programs and resource allocation, and the resultant inefficiency have been raised as a problem. Therefore, this paper proposed the national nuclear education & training system model as a macro policy initiatives and construction of control tower that manage and adjust overall nuclear education & training.

Nuclear,Education & training,The national nuclear education & training system,Control tower

The first Low- and Intermediate-Level Waste (LILW) disposal facility with 6 silos has been constructed in granite host rock saturated with groundwater in Korea. A two-dimensional numerical modeling on gas migration was carried out using TOUGH2 with EOS5 module in the disposal facility. Laboratory-scale experiments were also performed to measure the important properties of silo concrete related with gas migration. The gas entry pressure and relative gas permeability of the concrete was determined to be 0.97±0.15 bar and 2.44×10-17 m2, respectively. The results of the numerical modeling showed that hydrogen gas generated from radioactive wastes was dissolved in groundwater and migrated to biosphere as an aqueous phase. Only a small portion of hydrogen appeared as a gas phase after 1,000 years of gas generation. The results strongly suggested that hydrogen gas does not accumulate inside the disposal facility as a gas phase. Therefore, it is expected that there would be no harmful effects on the integrity of the silo concrete due to gas generation.

Low- and intermediate-level radioactive waste disposal facility,Gas migration,Silo concrete,Gas entry pressure,Gas permeability

This study evaluated the evolution stage and origin of chemical components of 12 boreholes at crystalline bedrock using multivariate statistical and groundwater quality analyses. Groundwater types are mostly belonged to Na(Ca)-HCO3 and Ca-HCO3 types, indicating that directly reaction of cation exchange (Ca2+ → Na+) prevailed. The degree of groundwater evolution is included the range from low to intermediate stage based on field and laboratory analytical conditions. As a result of multivariate statistical analysis, a typical indicator of groundwater contamination, NO3-, has the positive correlation with Na+ and Cl-. The origin of sea spary (Cl-) has the positive correlation with Na+, SO42-, Mg2+, and K+, while not correlation with Ca2+, Fe2+, HCO3-, F-, and SiO2. The concentration of Cl- and NO3- belongs to general quality of groundwater and not exceeds over the Korean standard for drinking water. And the negative values of saturation index of minerals are calculated with chemical components in groundwater. Therefore, most of chemical components of groundwater in the study area are originated from natural process between rock and groundwater, whereas some of components are derived from sea spary and anthropogenic sources related to agricultural activities.

Groundwater quality,Multivariate statistical analysis,Water-rock interaction,Sea spray,Groundwater type

For several decades, many countries operating nuclear power plants have been studying the various disposal alternatives to dispose of the spent nuclear fuel or high-level radioactive waste safely. In this paper, as a direct disposal of spent nuclear fuels for deep geological disposal concept, the rod consolidation from spent fuel assembly for the disposal efficiency was considered and analyzed. To do this, a concept of spent fuel rod consolidation was described and the related concepts of disposal canister and disposal system were reviewed. With these concepts, several thermal analyses were carried out to determine whether the most important requirement of the temperature limit for a buffer material was satisfiedin designing an engineered barrier of a deep geological disposal system. Based on the results of thermal analyses, the deposition hole distance, disposal tunnel spacing and heat release area of a disposal canister were reviewed. And the unit disposal areas for each case were calculated and the disposal efficiencies were evaluated. This evaluation showed that the rod consolidation of spent nuclear fuel had no advantages in terms of disposal efficiency. In addition, the cooling time of spent nuclear fuels from nuclear power plant were reviewed. It showed that the disposal efficiency for the consolidated spent fuel rods could be improved in the case that cooling time was 70 years or more. But, the integrity of fuels and other conditions due to the longer term storage before disposal should be analyzed.

Spent nuclear fuels,Deep geological disposal(DGD),Disposal efficiency,Rod consolidation,Disposal tunnel,Deposition Hole,Cooling time

A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 §71.45. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation.The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

Spent fuel,Structural integrity,Lifting device,Transport cask

Wolsong Low- and Intermediate-level radioactive waste (LILW) disposal center has two different types of disposal facilities and interacts with the neighboring Wolsong nuclear power plant. These situations impose a high level of complexity which requires in-depth understanding of phenomena in the safety assessment of the disposal facility. In this context, multidimensional radionuclide transport model and hydraulic performance assessment model should be developed to identify more realistic performance of the complex system and reduce unnecessary conservatism in the conventional performance assessment models developed for the 1st stage underground disposal. In addition, the advanced performance assessment model is required to calculate many cases to treat uncertainties or study parameter importance. To fulfill the requirements, this study introduces the development of two-dimensional integrated near-field performance assessment model combining near-field hydraulic performance assessment model and radionuclide transport model for the 2nd stage near-surface disposal. The hydraulic and radionuclide transport behaviors were evaluated by PORFLOW and GoldSim. GoldSim radionuclide transport model was verified through benchmark calculations with PORFLOW radionuclide transport model. GoldSim model was shown to be computationally efficient and provided the better understanding of the radionuclide transport behavior than conventional model.

LILW,Near-surface disposal,Near-field,Hydraulic performance assessment,Two-dimensional radionuclide transport,GoldSim,PORFLOW

The safety case program has been prepared for the development of the disposal facility of low- and intermediate-level radioactive waste in Korea. For the development of the radioactive waste disposal facility, this program can be applied for the safety demonstration of the facility and for the safety judgment of development step based on the international standards and domestic development environment. Systematic safety approach of this program includes the safety strategies such as optimization, robustness, demonstrability and defense-in-depth principle which are based on the safety principle and objectives. From the quality of assessment basis, safety arguments focused on the uncertainty management and the confidence building can assure the disposal safety during the step-wise safety assessment.

Low- and intermediate-level radioactive waste,Disposal facility,Step-wise development,Uncertainty management,Confidence building

The systematic quality assurance activities on documents of the safety assessment are required for the safety case of the low- and intermediate-level radioactive waste disposal facility. In this paper, quality assurance system focused on the input data including the site characterization, groundwater flow, system design and monitoring are prepared and discussed. Rule for the input data selection is suggested and applied for the safety assessment which is based on the in-situ/experiment observations, final facility design and waste pileup plan, engineered barrier, field monitoring, recent biosphere, and radionuclide inventory. The reduction of data uncertainty will be expected to contribute to the safety of disposal facility further.

Quality assurance system,Input data decision,Safety assessment,Low- and Intermediate-Level Radioactive Waste

The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. Since the initial introduction of nuclear power to Korea in 1978, rapid growth in nuclear power has been achieved. This large nuclear power generation program has produced a significant amount of radioactive waste, both low- and intermediate-level waste (LILW) and spent nuclear fuel (SNF); and the amount of waste is steadily growing. For the management of LILW, the Wolsong LILW Disposal Center, which has a final waste disposal capacity of 800,000 drums, is under construction, and is expected to be completed by June 2014. Korean policy about how to manage the SNF has not yet been decided. In 2004, the Atomic Energy Commission decided that a national policy for SNF management should be established considering both technological development and public consensus. Currently, SNF is being stored at reactor sites under the responsibility of plant operator. The at-reactor SNF storage capacity will run out starting in 2024. In this paper, the fundamental principles and steps for implementation of a Korean policy for national radioactive waste management are introduced. Korean practices and prospects regarding radioactive waste management are also summarized, with a focus on strategy for policy-making on SNF management.

Low-and intermediate-level radioactive waste,Spent nuclear fuel management,Management policy and strategy

Nagra was established in 1972 by the Swiss nuclear power plant operators and the Federal Government to implement permanent and safe disposal of all types of radioactive waste generated in Switzerland. The Swiss Nuclear Energy Act specifies that these shall be disposed of in deep geological repositories. A number of different geological formations and sites have been investigated to date and an extended database of geological characteristics as well as data and state-of-the-art methodologies required for the evaluation of the long-term safety of repository systems have been developed. The research, development, and demonstration activities are further supported by the two underground research facilities operating in Switzerland, the Grimsel Test Site and the Mont Terri Project, along with very active collaboration of Nagra with national and international partners. A new site selection process was approved by the Federal Government in 2008 and is ongoing. This process is driven by the long-term safety and feasibility of the geological repositories and is based on a step-wise decisionmaking approach with a strong participatory component from the affected communities and regions. In this paper a brief history and the current status of the Swiss radioactive waste management program are presented and special characteristics that may be useful beyond the Swiss program are highlighted and discussed.

Radioactive waste disposal,Geological repository,High-level waste HLW,Spent nuclear fuel,SF,Site selection,Opalinus Clay

The United States Department of Energy (US DOE) is conducting research and development (R&D) activities under the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R&D activities are ongoing at nine national laboratories, and are divided into storage, transportation and disposal. Storage R&D focuses on closing technical gaps related to extended storage of UNF. Transportation R&D focuses on ensuring transportability of UNF following extended storage, and addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R&D focuses on identifying geologic disposal options and addressing technical challenges for generic disposal concepts in mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal. UFDC R&D goals include increasing confidence in the robustness of generic disposal concepts, reducing generic sources of uncertainty that may impact the viability of disposal concepts, and developing science and engineering tools to support the selection, characterization, and licensing of a repository. The US DOE has also initiated activities in the Nuclear Fuel Storage and Transportation (NFST) Planning Project to facilitate the development of an interim storage facility and to support transportation infrastructure in the near term.

US Nuclear waste management,Used fuel disposition campaign,Storage,Transportation,Disposal

This paper describes basic plans for the development of a radioactive waste disposal facility with the introduction of Nuclear Power Plants (NPPs) for Kenya. The specific objective of this study was to estimate the total projected waste volumes of low- and intermediate-level radioactive waste (LILW) expected to be generated from the Kenyan nuclear power programme. The facility is expected to accommodate LILW to be generated from operation and decommissioning of nuclear power plants for a period of 50 years. An on-site storage capacity of 700 m3 at nuclear power plant sites and a final disposal repository facility of more than 7,000 m3 capacity were derived by considering Korean nuclear power programme radioactive waste generation data, including Kori, Hanbit, and APR 1400 nuclear reactor data. The repository program is best suited to be introduced roughly 10 years after reactor operation. This study is important as an initial implementation of a national LILW disposal program for Kenya and other newcomer countries interested in nuclear power technology.

Nuclear power plants,Radioactive waste,Disposal facility,Newcomer country,Disposal capacity

To characterize quantitatively the transport of 99Tc and the global fallout (137Cs, 90Sr, and 239+240Pu) for soils in Korea, the transport parameters of a convective-dispersion model, apparent migration velocity, and apparent dispersion coefficient were estimated from the vertical depth profiles of the radionuclides in soils. The vertical profiles of 99Tc were measured from a pot experiment for paddy soil that had been sampled from a rice-field around the Gyeongju radioactive waste repository in Korea, and the vertical depth distributions of the global fallout 137Cs, 90Sr, and 239+240Pu were measured from the soil samples that were taken from local areas in Korea. The front edge of the 99Tc profiles reached a depth of about 12 cm in 138 days, indicating a faster movement than the fallout radionuclides. A weak adsorption of 99Tc on the soil particles by the formation of Tc(VII) and a high water infiltration velocity seemed to have controlled the migration of 99Tc. The apparent migration velocity and dispersion coefficient of 99Tc for the disturbed paddy soil were 2.88 cm/y and 6.3 cm2/y, respectively. The majority of the global fallout 137Cs, 90Sr, and 239+240Pu were found in the top 20 cm of the soils even after a transport of about 30 years. The transport parameters for the global fallout radionuclides were 0.01-0.1cm/y (137Cs), 0.09- 0.13cm/y (90Sr), and 0.09-0.18cm/y (239+240Pu) for the apparent migration velocity: 0.21-1.09 cm2/y (137Cs), 0.12- 0.7cm2/y (90Sr), and 0.09-0.36cm2/y (239+240Pu) for the apparent dispersion coefficient.

Transport in soil,Convective-dispersion model,Migration velocity,Dispersion coefficient

Natural green rust (GR) can retard the migration of anions through geological media because it has a Layer Double Hydroxyl (LDH) structure with a positive charge. In this study, the sorption behaviors of anions such as selenite (Se(IV)O32-), selenate (Se(VI)O42-), and iodide (I-) onto green rusts with different structures, i.e., GR(Cl -) and GR(CO32-), were investigated by conducting batch sorption experiments in an anoxic condition. Experimental results showed that selenite was mostly sorbed onto GR(CO32-) and then partly reduced to metal selenium, Se(0). However, little selenate and iodide was sorbed onto GR(CO32-) while some iodide was sorbed onto GR(Cl-). It is presumed from the experimental results that the major sorption mechanism of SeO32- and I- onto green rusts is the anion exchange reaction with the anions existing in the interlayer of the rusts. Green rust, therefore, can play an important role in the retardation of anions migrating through deep geological environments owing to its LDH structure with a high anion exchange capacity.

Green rust,Selenium,Iodide,Anion exchange,Sorption,Reduction

Temperature-dependent hydrolysis behaviors of aqueous U(VI) species were investigated with time-resolved laser fluorescence spectroscopy (TRLFS) in the temperature range from 15 to 75°C. The formation of four different U(VI) hydrolysis species was measured at pHs from 1 to 7. The predominant presence of UO22+, (UO2)2(OH)22+, (UO2)3(OH)5+, and (UO2)3(OH)7- species were identified based on the spectroscopic properties such as fluorescence wavelengths and fluorescence lifetimes. With an increasing temperature, a remarkable decrement in the fluorescence lifetime for all U(VI) hydrolysis species was observed, representing the dynamic quenching behavior. Furthermore, the increase in the fluorescence intensity of the further hydrolyzed U(VI) species was clearly observed at an elevated temperature, showing stronger hydrolysis reactions with increasing temperatures. The formation constants of the U(VI) hydrolysis species were calculated to be log K02,2 = -4.0 ± 0.6 for (UO2)2(OH)22+, log K03,5 = -15.0 ± 0.3 for (UO2)3(OH)5+, and log K03,7 = -27.7 ± 0.7 for (UO2)3(OH)7- at 25°C and I = 0 M. The specific ion interaction theory (SIT) was applied for the extrapolation of the formation constants to infinitely diluted solution. The results of temperature-dependent hydrolysis behavior in terms of the U(VI) fluorescence were compared and validated with those obtained using computational methods (DQUANT and constant enthalpy equation). Both results matched well with each other. The reaction enthalpies and entropies that are vital for the computational methods were determined by a combination of the van’t Hoff equation and the Gibbs free energy equation. The temperature-dependent hydrolysis reaction of the U(VI) species indicates the transition of a major U(VI) species by means of geothermal gradient and decay heat from the radioactive isotopes, representing the necessity of deeper consideration in the safety assessment of geologic repository.

U(VI),Temperature-dependent Hydrolysis,Chemical Speciation,TRLFS

A large quantity of radioactive waste was generated during the decommissioning of the KRR and UCF. The radioactive waste was packed into 200 liter drums and 4m3 containers and these were temporarily stored onsite until their final disposal in the national repository facility. Some of the releasable waste was freely released and utilized for non-nuclear industries. The combustible wastes were treated by the utilization of an incinerator with a capacity of on average 20 kg/hr.

Waste Management,Decommissioning Waste,Volume Reduction,Incineration

A novel methodology to evaluate remote dismantling equipment for a reactor pressure vessel (RPV) in a decommissioning project is presented in this paper. The remote dismantling equipment, mainly composed of cutting tools and positioning equipment, is absolutely required to cut and handle highly radioactive and large components in nuclear power plants (NPPs); this equipment has a great effect on the overall success of the decommissioning project. Conventional evaluation methods have only focused on cutting technologies or positioning equipment, although remote dismantling equipment cannot achieve its goal without organic interaction between the cutting tools and the positioning equipment. In this paper, the cutting tools and the positioning equipment are evaluated by performance parameters according to their original characteristics, the relationship between the two systems, and common factors. Finally, the remote dismantling equipment used in recent decommissioning projects has been evaluated based on the proposed methodology. The results of this paper are expected to be useful for future decommissioning projects.

Evaluation methodology,Remote dismantling system,Cutting tool,Positioning eguipment and Reactor pressure vessel

The rare earth oxide wastes consisting of major 8 nuclides Y, La, Ce, Pr, Nd, Sm, Eu and Gd, are generated during the salt waste treatment of PyroGreen process. The final form of the rare earth is generated as the oxide state. In this study, six candidate glasses were developed to evaluate the feasibility for vitrifying the rare earth oxide wastes within the borosilicate glass system. The solubilities of the mixture of the rare earth oxide waste showed less than 25wt% at 1,200℃, less than 30wt% at 1,300℃, respectively. It means that solubility is increased with the temperature increment. The liquidus temperature of the homogeneous glass with 20wt% waste loading was determined as less than 950℃. In more than solubility of rare earth oxides glass, formation of rare earth-oxide-silicate crystal in glass-ceramic occurred as the secondary phase. As their viscosity at melting temperature 1,200~1,300℃ was less than 100 poise, electrical conductivity was higher than 1 S/cm, 20~25wt% waste loading glasses with good surface homogeneity are judged to have good operability in cold crucible induction melter. Other physicochemical properties of the developed glasses are going to be experimented in the future.

PyroGreen,Vitrification,Rare earth oxide,Glass formulation,Waste solubility

The Geyoungju Ca-bentonite with dry density of 1.6 g/cm3 has been considered as a standard buffer material for the disposal of high level waste in KAERI disposal system design. But it had relatively lower thermal conductivity compared with other surrounding materials, that was one of key parameters to limit the increase of the disposal density in the disposal system. In this study, various additives were selected and mixed with the Ca-bentonite in different mixing methods in order to increase the thermal conductivity from 0.8 W/mK to 1.0 W/mK. As an additive, CNT (Cabon Nano Tube), graphite, alumina, CuO, and Fe2O3 were selected, which are chemically stable and have good thermal conductivity. As mixing methods, dry hand-mixer mixing, wet milling and dry ball mill mixing were applied for the mixing. Above all, the ball mill mixing was proved to be most effective since the produced mixture was most homogeneous and showed higher increase in the thermal conductivity. From this study, it was confirmed that the thermal conductivity for the Geyoungju Ca-bentonite could be improved by adding small amount of highly thermal conductive material to 1.0 W/mk. In conclusion, it was believed that the experimental results will be valuable in the disposal system design if the additive effects on the swelling and permeability on the compact bentonite are also approved in further studies.

HLW,Thermal conductivity,Dry density,Water contents,Additive,Mixture

LCC (Liquid cadmium cathode) is used for electrowinning in pyroprocessing to recover uranium and transuranic elements simultaneously. It is one of the core technologies in pyroprocessing with higher proliferation resistance than a wet reprocessing because LCC-cell does not separate TRU from uranium. The crucible which holds the LCC is technically important because it should be nonconducting material to prevent deposition of metallic elements on the crucible outer surface. The chemical stability is also crucial factor to choose crucible material due to the strong reactivities of TRU and possible incorporation of Li metal during the operation. In this study, the chemical stabilities of four kinds of representative ceramic materials such as Al2O3, MgO, Yl2O3 and BeO were thermodynamically and experimentally evaluated at 500oC with simulated LCC. The contact angle of LCC on ceramic materials was measured as function of time to predict chemical reactivity. All2O3 showed poorest chemical stability and the pores in BeO contributed to a decreases in contact angle. MgO and Y2O3 have superior chemical stability among the materials.

Pyroprocess,LCC (Liquid Cadmium Cathode),Thermodynamic stability,Chemical stability,Rare-earth elements

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

Nuclear fuel,Failed fuel,Spent fuels,Fission products,Sipping technology,Dry storage

Radionuclide sorption data is necessary for the safety assessment of radioactive waste disposal. However the use of sorption database is often limited due to the accessability. A web-based sorption database program named KAERI-SDB has been developed to provide information on the sorption of radionuclides onto geological media as a function of geochemical conditions. The development of KAERI-SDB was achieved by improving the performance of pre-existing sorption database program (SDB-21C) developed in 1998 and considering user’s requirements. KAERI-SDB is designed that users can access it by using a web browser. Main functions of KAERI-SDB include (1) log-in/member join, (2) search and store of sorption data, and (3) chart expression of search results. It is expected that KAERI-SDB could be widely utilized in the safety assessment of radioactive waste disposal by enhancing the accessibility to users who wants to use sorption data. Moreover, KAERI-SDB opened to public would also improve the reliability and public acceptance on the radioactive waste disposal programs.

Database,Radioactive waste disposal,Radionuclide,Safety assessment,Sorption

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into ZrCl4. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min Cl2 + 70 cc/min Ar flow at 380oC. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

Chlorination reaction,Hull waste,Zr recovery,Spent nuclear fuel,Pyro-processing

To see if slaked lime and organic fertilizer applications to soil are useful as countermeasures for reducing 99Tc concentrations in rice seeds after 99Tc contamination of paddy fields, pot experiments were performed for two different paddy soils in a greenhouse. The upper soils for a depth of about 20 cm were treated with the agricultural materials and 99Tc 15 d before transplanting. The effects were compared using the transfer factor (TF) defined as the ratio of the plant concentration to the soil concentration. In the case of control plants, TF values for brown rice in the two soils were 4.1×10-4 and 4.3×10-4. Of various types of the application, only the application of slaked lime at a lower dose (about 0.6 kg m-2), which led to a 60% reduction in the TF value for one soil, seemed to be worth using as a countermeasure. Little effect of the same application was found in the other soil so it is important to determine the effect averaged for a number of soils. Organic fertilizer applications at both of two different doses increased the TF value. It is considered necessary to perform experiments for slake lime applications at doses lower than the above.

99Tc,Paddy soil,Rice seeds,Transfer factor,Slaked lime,Countermeasure

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls was demonstrated to investigate the effect of the oxidation process on the reaction rate of the chlorination reaction. In chlorination reaction experiments performed for 6 h, where reaction products were collected every 2 h, it was observed that a significant decrease in the reaction rate was caused by the oxidation process (500oC, 10 h under an air atmosphere) within the reaction period of 0 – 2 h. The amount of reaction residue increased from 0.95 to 1.65wt% of initial weights in the fresh and Zry-500-10 (Zry-4 hulls oxidized at 500oC for 10 h under an air atmosphere) hulls, respectively. The purity of the recovered Zr was identical at 99.61wt% for the fresh Zry-4 and Zry-500-10 hulls. Quantitative analysis of the chlorination reaction rate was performed by varying the reaction time from 0.5 to 1.0, 2.0, and 4.0 h. The fitting results showed that the relationship between weight loss and reaction time can be interpreted by a linear line with a slope of 23.35wt%/h for the fresh Zry-4 case, while two linear lines were necessary to fit the results of Zry-500-10. In addition, the slope values were 17.12 and 27.16wt%/h for (0 – 20) and (20 – 100)wt% loss regions, respectively.

Hull waste,Chlorination reaction,Zr recovery,Pyroprocess,Reaction rate

The original pilot-scale electrokinetic equipment suitable to soil contamination characteristics of Korean nuclear facility sites was manufactured for the remediation of soil contaminated with uranium. During the experiment with the original electrokinetic equipment, many metal oxides were generated and were stuck on the cathode plate. The uranium removal capability of the original electrokinrtic equipment was almost exhausted because the cathode plate covered with metal oxides did not conduct electricity in the original electrokinetic equipment. Therefore, the original electrokinetic equipment was improved. After the remediation experience for 25 days using the improved electrokinetic remediation equipment, the removal efficiency of uranium from the soil was 96.8% and its residual uranium concentration was 0.81 Bq/g. When the initial uranium concentration of soil was about 50 Bq/g, the electrokinetic remediation time required to remediate the uranium concentration below clearance concentration of 1.0 Bq/g was about 34 days. When the initial uranium concentration of soil was about 75 Bq/g, the electrokinetic remediation time required to remediate below 1.0 Bq/g was about 42 days. When the initial uranium concentration of soil was about 100 Bq/g, the electrokinetic remediation time required to remediate below 1.0 Bq/g was about 49 days.

Pilot-scale Electrokinetic Remediation Equipment,Uranium,Contaminated soil,Remediation,Metal oxide

This study has been carried out to elucidate the characteristics of the dissolution for Thermal Decomposed Solid Waste of uranium-bearing sludge (TDSW), the removal of impurities by an alkalization in a nitric acid dissolving solution of TDSW, and the selective removal (/recovery) of uranium by an acidification in an carbonate alkali solution, respectively. TDSW generated by thermal decomposition of Ubearing sludge which was produced in the uranium conversion plant operation, was stored in KAERI as a solid-powder type. It is found that the dissolution of TDSW is more effective in nitric acid dissolution than oxidative-dissolution with carbonate. At 1 M nitric acid solution, TDSW was undissolved about 30wt% as a solid residue, and uranium contained in TDSW was dissolved more than 99%. In order to the alkalization for the nitric acid dissolving solution of TDSW, carbonate alkalization is more effective with respect to remove the impurities. At the carbonate alkali solution controlled to about 9 of pH, Al, Ca, Fe and Zn co-dissolved with U in dissolution step was removed about 98±1%. On the other hand, U could be recovered more than 99% by an acidification at pH about 3 in a carbonate alkali solution, which was nearly removed the impurities, adding 0.5M H2O2. It was found that uranium could be selectively recovered (/removed) from TDSW.

Uranium Sludge,Solid waste,Nitric acid dissolution,Carbonate dissolution,Akalization,Acidification

There was a study for biological characteristics, except for physico-chemical and mineralogical properties, on the natural bentonite that is considered as a buffer material for the high-level radioactive waste disposal site. A bentonite slurry that was prepared from a local ';Gyeongju bentonite'; in Korea was incubated in a serum bottle with nutrient media over 1 week and its stepwise change was observed with time. From the activated bentonite in the nutrient media, we can find a certain change of both solid and liquid phases. Some dark and fine sulfides began to be generated from dissolved sulfate solution, and 4 species of sulfate-reducing bacteria (SRB) were identified as living cells in samples that were periodically taken and incubated. These results show that sulfate-reducing (or metal-reducing) bacteria are adhering and existing in the powder of bentonite, suggesting that there may be a potential occurrence of longterm biogeochemical effects in and around the bentonite buffer in underground anoxic environmental conditions.

Bentonite,Sulfides,Sulfate-reducing bacteria,Biogeochemical effects

The Korea Atomic Energy Research Institute has developed a performance assessment model implementing the A-KRS concept, which was constructed with the GoldSim. In the A-KRS concept, spent nuclear fuel produced from pressurized-water-reactor operations would be pyroprocessed to reduce waste volume and radioactivity. The wastes to be disposed of in a geologic repository are comprised of metal and ceramic waste forms. In this study, results of simulations conducted to establish credibility and build confidence for the A-KRS model are presented. Specifically, release rates and breakthrough times simulated using the A-KRS model were compared to corresponding results from the U.S. NRC SOAR model. In addition, the A-KRS model results were compared to published release rates from the SKB repository performance assessment. This comparison of the A-KRS model results to other independent performance assessments is expected to form part of a suite of model verification and validation activities to provide confidence that the A-KRS model has been implemented appropriately.

A-KRS,GoldSim,Performance Assessment Model,SOAR,Comparison,verification,Pyroprocess

The Mizunami Underground Laboratory (MIU) Project is a comprehensive research project investigating the deep underground environment within crystalline rock being conducted by Japan Atomic Energy Agency. The MIU Project has three overlapping phases: Surface-based Investigation phase (Phase I), Construction phase (Phase II), and Operation phase (Phase III), with a total duration of 20 years. The overall project goals of the MIU Project from Phase I through to Phase III are: 1) to establish techniques for investigation, analysis and assessment of the deep geological environment, and 2) to develop a range of engineering for deep underground application. For the overall project goals 1), the Phase I goals were set to construct models of the geological environment from all surface-based investigation results that describe the geological environment prior to excavation and predict excavation response. For the overall project goals 2), the Phase I goals were set to formulate detailed design concepts and a construction plan for the underground facilities. This paper introduces geosynthesis procedures for the investigation and assessment of the hydrochemistry of groundwater in crystalline rock.

Mizunami Underground Research Laboratory,Site characterization technology,Groundwater Chemistry

Natural analogue studies play an important role in the safety case which requires multiple lines of evidence including the safety assessment for the geological disposal of radioactive wastes. In this study, foreign status of natural analogue studies was investigated by summarizing natural analogue results according to the research topics related with repository materials and radionuclide migration and retardation. Main results, issues, and applicability of the foreign natural analogue studies were also analyzed. The results of domestic natural analogue studies were classified into studies using uranium ore bodies, rocks, groundwaters, and archeological artifacts, respectively, and their main results were summarized. There are massive materials for natural analogue studies which have been carried out during last several decades but they have not been actively applied to the safety assessment and safety case development for the radioactive waster disposal. Thus, in this study, applicable methods of natural analogues were summarized and a methodology for improving their applicability was examined. Natural analogue study is apparently necessary to improve and illustrate the reliability of safety assessment for a radioactive waste repository. Therefore, it is necessary to develop a methodology and construct a natural analogue information database for the application of the results from natural analogue studies to safety case development.

Natural analogue study,Radioactive waste,Geological disposal,Safety assessment,Safety case

This study gives a conceptual and basic direction to develop a URL (underground research laboratory) program for establishing the performance and safety of a deep geological disposal system in Korea. The concept of deep geological disposal is one of the preferred methodologies for the final disposal of spent nuclear fuel (SNF). Advanced countries with radioactive waste disposal have developed their own disposal concepts reasonable to their social and environmental conditions and applied to their commercial projects. Deep geological disposal system is a multi-barrier system generally consisting of an engineered barrier and natural barrier. A disposal facility and its host environment can be relied on a necessary containment and isolation over timescales envisaged as several to tens of thousands of years. A disposal system is not allowed in the commercial stage of the disposal program without a validation and demonstration of the performance and safety of the system. All issues confirming performance and safety of a disposal system include investigation, analysis, assessment, design, construction, operation and closure from planning to closure of the deep geological repository. Advanced countries perform RD&D (research, development & demonstration) programs to validate the performance and safety of a disposal system using a URL facility located at the preferred rock area within their own territories. The results and processes from the URL program contribute to construct technical criteria and guidelines for site selection as well as suitability and safety assessment of the final disposal site. Furthermore, the URL program also plays a decisive role in promoting scientific understanding of the deep geological disposal system for stakeholders, such as the public, regulator, and experts.

Spent Nuclear Fuel,Deep geological repository,Multi-barrier disposal system,NBS(natural barrier system),EBS(emgineered barrier system),URL(underground research laboratory)

This study is characterized the groundwater flow pattern near crystalline site of Yangbook-Myeon, Gyeongju City, South Korea. From the time series analyses, groundwater level could be classified into 4 types reflecting the hydrogeological characteristics and rainfall pattern. The type Ⅰ (DB1-1, DB1-2) may be directly influenced by rainfall pattern. The type Ⅱ (DB1-3, DB1-7, KB-1, KB-2, KB-3, KB-7, KB-14, KB-15) may be influenced by rainfall event as well as groundwater flow through water-conducting features. The type Ⅲ (DB-5, DB1-6, DB2-2, KB-10, KB-11, KB-13) may be predominantly happens in the crystaline rock mass, groundwater in this type flows through the minor fracture networks rather than direct effect of rainfall event. The type Ⅳ (DB1-8, KB-9) may be influenced by irregular variation of the groundwater level due to anisotropy and heterogeneity of crystalline rock.

Groundwater level,Time series analysis,Rainfall,Fracture,Crystalline rock

In general, boron recovery of 40-90% could be achieved by Reverse Osmosis (RO) membranes in neutral pH condition. As an emerging technology, Forward Osmosis (FO) membrane has attracted growing interest in wastewater treatment and desalination. The objective of this study is to evaluate the possibility of the boron removal in radioactive liquid waste by FO. In this study, the performance of FO was investigated to remove boron in the simulated liquid waste as the factors such as pH, osmotic pressure, ionic strength of solution, etc. The pH of feed solution is a major operating parameter which strongly influences to the permeation of boron and more than 80% of boron content can be separated when conducted at pH values less than 7. The water flux is not influenced but the boron flux and permeation rate tends to decrease in the low salt concentration of 1,000 mg/L. The boron flux increases linearly, but the permeation ratio of reducing boron is nearly constant even with changes in the draw solution concentration.

Forward osmosis,Boron removal,Ionic strength,pH

Electrowinning process in pyroprocessing recovers U (uranium) and TRU (Trans Uranium) elements simultaneously from spent fuels using a liquid cadmium cathode (LCC). When the solubility limit of U deposits over 2.35wt% in Cd, U dendrites were formed on the LCC surface during the electrodeposition at 500℃. Due to the high surface area of dendritic U, the deposits were not submerged into the liquid cadmium pool but grow out of the LCC crucible. Since the U dendrites act as a solid cathode, it prevents the co-deposition of U and TRUs. In this study, the electrodeposition of U onto a LCC was carried out at 440 and 500℃ to compare the morphology and component of U deposits. The U deposits at 440℃ have a specific shape and were stacked regularly at the center of the LCC pool, while the U dendrites (i.e., α-phase) at 500℃ were grow out of the LCC crucible. Through the microscopic observation and XRD analysis, the electrodeposits at 440℃, which have a round shape, were identified as an intermetallic compound such as UCd11. It can be concluded that the LCC electrowinning operation at 440℃ achieves the co-recovery of U and TRU without the formation of U dendrites.

Electrowinning,LCC(Liquid Cadmium Cathode),U dendrite,Intermetallic compound

Since the Fukushima nuclear power plant accident in March 2011, critical views on the increase in operation of nuclear power plants including the safety and the economic feasibility thereof have been expanding across the world. In these circumstances, we are to find out solutions to the controversial questions on whether nuclear power plants are economically more feasible than other energy sources, while the safety thereof is fully maintained. Thereby, nuclear power plants will play a key role as a sustainable energy source in the future as well as at present. To measure the social safety level that Korean people are actually feeling after the Fukushima accident, a method of cost-benefit analysis called the Contingent Valuation Method(CVM) was used, whereby we wanted to estimate the amount of expenses the general public would be willing to pay for the safety based on their acceptance rather than the social safety. As a result of calculating the trade-off value of the economic feasibility versus the safety in nuclear power plants through the survey thereon, it caused the nuclear power generation cost to be increased by 4.75 won/kWh. Reflecting this on the current power generation cost of 39.11 won/kWh would increase the cost to 43.86 won/kWh. It is thought that this potential cost is still more competitive than the coal-fired power generation cost of 67 won/kWh. This result will be available as a basic data for the 2nd Energy Basic Plan to be drawn up this year, presenting policy implications at the same time.

Contingent Valuation Method,Optimum Point,Economic Feasibility,Social Acceptance,Nuclear power generation cost

It is essential to prepare the decommissioning plan for a nuclear power plant (NPP) for the safe decommissioning of the NPP, minimization of the generation of decommissioning wastes, and protection of human beings and environment. Although Kori unit 1 and Wolsong unit 1 will be destined to their decommissioning in Korea in the near future. there is no provisons about preparing the decommissioning plan. In this paper, therefore, the draft guidelines of the decommissioning plan for a NPP were developed by considering the domestic situation, based on the comparative analyses of the regulatory guidelines of the decommissioning plan in U.S., U.K. and France. The draft guidelines are expected to play an important role to modify the domestic laws and regulations on the decommissioning of the NPP, and to give a license holder in charge of decommissioning the detailed instructions for preparing it in advance.

Nuclear power plant,Decommissioning plan,Draft guidelines,Decommissioning wastes

An in situ solute migration system was designed and installed in KAERI Underground Research Tunnel (KURT) constructed in the site of Korea Atomic Energy Research Institute (KAERI) in order to investigate the migration and retardation of non-sorbing and sorbing tracers through a rock fracture. The system is composed of three main parts including injection, extraction, and data treatment. For the selection of a water-conducting fracture, boreholes were drilled. The fractures in the drilled boreholes were investigated using borehole image analysis using borehole image processing system (BIPS). The results of BIPS analysis showed that borehole YH 3-1 and YH 3-2 were connected each other. Moreover, hydraulic tests were carried out to determine the test section with connectivity for the in situ experiments. The in situ solute migration experiments were accomplished to understand the migration of solutes through fractures in KURT using non-sorbing tracers which were fluorescein sodium, eosin-B, bromide and sorbing tracers which were rubidium, nickel, zirconium, and samarium.

Connectivity,Fracture,In situ experiment,KAERI Underground Research Tunnel (KURT),Solute migration

The energy band and the G-factor method were compared to determine the exposure rate from the measured spectrum using a NaI(Tl) scintillation detector. First, G-factors of a 3“ΦX3” NaI(Tl) detector mounted to a EFRD 3300, which means the environmental radiation monitor, in Korea Atomic Energy Research Institute (KAERI) were calculated for several directions of incident photons through the MCNP modeling, and the optimum G-factor applicable to that monitor was then determined by comparing the results both the energy band method and the G-factor method. The results for these spectrometric determinations were also compared with the dose rate from a HPIC radiation monitor around a EFRD 3300. The measured value at the EFRD 3300 based on a 3“ΦX3” NaI(Tl) detector was 7.7 μR/h and its difference was shown about 3 μR/h, when compared with the results from a HPIC radiation moditor. Since a HPIC is known to be able to measure cosmic rays with the relatively high energy, the difference between them was caused by cosmic rays which were not detected in a 3“ΦX3” NaI(Tl) detector.

Energy band,G-factor,Environmental radiation monitor,NaI(Tl),Dose conversion,MCNP

Geology, hydrogeology, and geochemistry are the main technical siting factors of a geological repository for spent nuclear fuels. This paper evaluated the siting criteria of overseas geological repository with related to hydrogeologic properties, such as hydraulic conductivity, partitioning coefficient, dispersion coefficient, boundary condition, and water age. Each country establishes the siting criteria based on its important geological backgrounds and information, and social environment. For example, Sweden and Finland that have decided a crystalline rock as a host rock of a geological repository show different siting criteria for hydraulic conductivity. In Sweden, it is preferable to avoid area where the hydraulic conductivity on a deposition hole scale (~30m) exceeds 10-8 m/s, whereas Finland does not decide any criterion for the hydraulic conductivity because of limited data for it. In addition, partitioning coefficients should be less than 10-1 of average value in Swedish crystalline bedrock. However, the area where shows 100 times less than average partitioning coefficients of radionuclides in crystalline rock should be avoided in Sweden. In German, the partitioning coefficients for the majority of the long-term-relevant radionuclides should be greater than or equal to 0.001 m3/kg . Therefore, it is strongly required to collect much and exact information for the hydrogeologic properties in order to set up the siting criteria.

Geological disposal,Repository,Hydrogeology,Siting factors,Criteria

Applicability of ammonium molybdophosphate/iron oxides-polyacrylonitrile (AMP/IO-PAN) composites on the removal of radionuclides in the radioactive wastewater generated from nuclear power plants was investigated. The composites were characterized using the following analytical techniques: X-ray diffraction (XRD), Fourior transform-infrared (FT-IR) spectroscopy, scanning electron microscopy (SEM), particle size analyzer (PSA), nitrogen adsorption-desorption and magnetic property measurement system (MPMS). 10wt% of AMP/IO-PAN composite has a saturation magnetization of 2.038 emu/g. Single-solute sorptions of Co, Sr and Cs onto 10wt% of AMP/ IO-PAN composite were investigated. The maximum sorption capacities (Q0) predicted by the Langmuir model on 10wt% of AMP/IO-PAN composite were 0.097, 0.086 and 0.66 mmol/g for Co, Sr and Cs, respectively. The maximum sorption capacities (Q0) of Cs predicted by Langmuir model on 0, 10, 20 and 30wt% of AMP/IO-PAN composites were 0.702, 0.655, 0.602 and 0.559 mmol/g, respectively. The maximum sorption capacities (Q0) of Cs decreased with increasing the iron oxide content in the AMP/IO-PAN composites.

Ammonium molybdophosphate,Composites,Ion exchanger,Iron oxides,Radionuclides

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.

Concrete Waste,Cement Waste Form,Decommissioning,Decontamination,Compressive Strength

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

Radionuclide transport simulation,Transport travel time,Multiple source positions,Sorption in a rock matrix,TDRW method

Borehole Disposal Concept (BDC) was initiated by the South African Nuclear Energy Corporation (NECSA) with the view to improve the radioactive waste management practices in Africa. At a time when geological disposal of radioactive waste is being considered, the need to protect ground water from possible radioactive contamination and the investigation of radionuclides migration through soil and rocks of zone of aeration into ground water has becomes very imperative. This is why the Borehole Disposal Concept (BDC) is being suggested to address the problem. The concept involves the conditioning and emplacement of disused sealed radioactive sources in an engineered facility of a relatively narrow diameter borehole (260 mm). Tanzania is operating a Radioactive Waste Management Facility where a number of spent sealed radioactive sources with long and short half lives are stored. The activity of spent sealed radioactive sources range from (1E-6 to 8.8E+3 Ci). However, the long term disposal solution is still a problem. This study therefore proposing the country to adopt the BDC, since the repository requires limited land area and has a low probability of human intrusion due to the small footprint of the borehole.

Borehole Disposal Concept,Spent Sealed Sources,Radioactive Waste Conditioning,Repository

As an alternative of the spent fuel disposal in a geologic repository, a deep borehole disposal concept for disposal at the section of 3 - 5km deep in a borehole has been proposed in several countries. In this paper, the latest reports of Sandia National Laboratories on the borehole disposal researches are analyzed. For implementation of this disposal concept in Korea, a conceptual design of spent fuel disposal canister and a modified deep borehole concept are suggested along with a required disposal area.

Spent fuel,Disposal,Deep borehole,Alternative disposal option

To ensure the reliability of computer programs used for the post-closure safety assessment in the Wolsong LILW Center, the results from MASCOT, SAFE-ROCK and GOLDSIM programs are compared with a problem for degradation. Advantages and disadvantages of each computer programs are individually analyzed. Effects on the individual dose are assessed with each computer programs. MASCOT and SAFE-ROCK showed similar results for 129I and 3H. However, GOLDSIM represented different results for 129I and 3H. It is analyzed further and compared with the fluxes in each barrier of the disposal system. Througout the benchmarking testing of the computer program, the limitation of computer program can be continuously found out for the mature post-closure safety of Korean radwaste disposal system.

Low- and Intermediate-Level Radioactive Waste,Final Disposal,Post-closure safety assessment,MASCOT,SAFE-ROCK,GOLDSIM

Pyroporcessing of spent nuclear fuel generates a considerable amount of LiCl-KCl eutectic waste salt containing radioactive rare earth (RE) chlorides. In this study, a series of processes, which consist of a phosphorylation/distillation process and a solidification process, were performed to minimize volume of the LiCl-KCl eutectic waste salt and solidify a residual waste into a stable form at a relatively low temperature. Over 99wt% of RE chlorides in LiCl-KCl eutectic salt was converted and separated into REPO4 in the phosphorylation/distillation process using a mixture of Li3PO4-K3PO4. The separated REPO4 was solidified into a homogeneous and fine-grained form at 1,050℃ using LIP(Lead Iron Phosphate) as a solidification agent. The final waste volume was reduced below about 10% through the series of the processes.

Pyroprocessing of spent nuclear fuel,LiCl-KCl eutectic salt,Rare earth,Phosphorylation,Distillation,Solidification

As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safetycriteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

Dry Storage,Spent Nuclear Fuel,Degradation Mechanism,Safety Criteria

The equation of the step-drawdown test “sw = BQ + CQp ” written by Rorabaugh (1953) is suitable for drawdown increased non-linearly in the fractured rocks. It was found that value of root mean square error (RMSE) between observed and calculated drawdowns was very low. The calculated C (well head loss coefficient) and P (well head loss exponent) value of well head losses (C Qp) ranged 3.689 × 10-19 ~ 5.825 × 10-7 and 3.459 ~ 8.290, respectively. It appeared that the deeper depth in pumping well the larger drawdowns due to pumping rate increase. The well head loss in the fractured rocks, unlike that in porous media, is affected by properties of fractures (fractures of aperture, spacing, and connection) around pumping well. The C and P value in the well head loss is very important to interpret turbulence interval and properties of high or low permeability of fractured rock. As a result, regression analysis of C and P value in the well head losses identified the relationship of turbulence interval and hydraulic properties. The relationship between C and P value turned out very useful to interpret hydraulic properties of the fractured rocks.

step-drawdown test,fractured rocks,least squares method,well head loss,high or low permeability

The long-lived fission products 79Se and 99Tc have been considered as the major concern nuclides for the disposal of radioactive waste because of their high solubilities and the existence of anionic species in natural water. In this study, the solubilities of FeSe2(s) and TcO2(s), known as respective Solubility Limiting Solid Phase (SLSP) of selenium and technetium, were measured in the KURT (KAERI Underground Research Tunnel) groundwater under various pH and redox conditions. And their solubilities and major species were also calculated using geochemical codes under conditions similar to experimental solutions. Experimental results and calculation for FeSe2 show that the solubility of selenium was found to be below 1x10-6 mol/L under the condition of pH 8 9.5 and Eh=-0.3 ~-0.4 V while the dominant species was identified as HSe-. For TcO2, the solubility of technetium was found to be 5x10-8 1x10-9 mol/L in the solutions of pH 6 ~ 9.5 and Eh<-0.1 V, while the dominant species was TcO(OH)2. However, when the Eh of the solution is -0.35 V, TcO(OH)3 - and TcO4 - are calculated as the dominant species at pH 10.5 ~ 12 and pH>12, respectively.

Selenium,Technetium,Solubility,Species,Groundwater

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at 1170 C after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2 3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.

Sodium-cooled Fast Reactor (SFR),Fuel,Cladding,HT9,Metallurgy

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO2-Al2O3-P2O5) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe2O3 into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for MSAP(Fe=0.1). The experimental results indicated that the addition of Fe2O3 increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B2O3 into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO2-Al2O3-Fe2O3-P2O5-B2O3) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 4 g of wasteform for final disposal. The final volume would be about 3~4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

Immobilization,Volatile radionulide,Sol-gel method,SiO2-Al2O3-P2O5,Metal chloride,pyroprocess

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

IP-2 package,low and intermediate level waste,stacking test,safety evaluation,radiation shielding

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a SiO2-Al2O3-P2O5 (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about 400°C for LiCl and the second was about 700°C for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a SiO2-Al2O3-P2O5 (SAP) could be considered as one of alternatives for the immobilization of waste salt.

Immobilization,Volatile radionulide,Sol-gel method,SiO2-Al2O3-P2O5,Metal chloride,pyroprocess

As an alternative of the high-level radioactive waste disposal in the subsurface repository, a deep borehole disposal is reviewed by several nuclear advanced countries. In this study, the state of the art on the borehole disposal researches was reviewed, and the possibility of borehole disposal in Korean peninsula was discussed. In the deep borehole disposal concept radioactive waste is disposed at the section of 3 - 5km depth in a deep borehole, and it has known that it has advantages in performance and cost due to the layered structure of deep groundwater and small surface disposal facility. The results show that it is necessary to acquisite data on deep geologic conditions of Korean peninsula, and to research the engineering barrier system, numerical modeling tools and disposal techniques for deep borehole disposal.

High-level waste,Borehole disposal,State of the art

This paper gives some basic requirements and preferences of various geological environmental conditions for the final deep geological repository of spent nuclear fuel (SNF). This study also indicates how the requirements and preferences are to be considered prior to the selection of sites for a site investigation as well as the final disposal in Korea. The results of the study are based on the knowledge and experience from the IAEA and NEA/OECD as well as the advanced countries in SNF disposal project. This study discusses and suggests preliminary guideline of the disposal requirements including geological, mechanical, thermal, hydrogeological, chemical and transport properties of host rock with long term geological stabilities which influence the functions of a multi-barrier disposal system. To apply and determine whether requirements and preferences for a given parameter are satisfied at different stages during a site selection and suitability assessment of a final disposal site, the quantitative criteria in each area should be formulated with credibility through relevant research and development efforts for the deep geological environment during the site screening and selection processes as well as specific studies such as productions of safety cases and validation studies using a generic underground research laboratory (URL) in Korea.

Spent Nuclear Fuel,deep geological repository,multi-barrier disposal system,requirement,preference,geoscientific suitability indicator

This study was carried out to remove (/recover) the uranium from the Uranium-bearing Lime Precipitate (ULP). An oxidative dissolution of ULP with carbonate-acidified precipitation and a dissolution of ULP with nitric acidhydrogen peroxide precipitation were discussed, respectively. In point of view the dissolution of uranium in ULP, nitric acid dissolution which could dissolved more than 98% of uranium was more effective than carbonate dissolution. However, in this case, uranium was dissolved together with a large amount of impurities such as Al, Ca, Fe, Mg, Si, etc. and some impurities were also co-precipitated with uranium during a hydrogen peroxide precipitation. On the other hand, in the case of carbonate dissolution-acidified precipitation, U was dissolved less than 90%. Therefore, it was less effective than nitric acid dissolution for the volume reduction of radioactive solid waste. However, it was very effective to recover the pure uranium, because impurities were hardly dissolved and hardly co-precipitated with uranium.

Uranium,Carbonate,Oxidative dissolution,Nitric acid dissolution,Acidified precipitation,Hydrogen peroxide precipitation

With a view to determine a safe speed the limit of a manipulator arm, several experiments was performed with a multi-jointed manipulator in maintenance and decommissioning tasks of nuclear facilities. Under the simulated emergency conditions, which were generated with random combinations of manipulator arm speed, failure probability and failure type, response characteristics of human operators to various malfunctions of a manipulator arm were measured in terms of reaction time, number of false alarm, and number of misses. This paper demonstrated that failure type, manipulator axes and manipulator arm speed has significant effects on human reaction time. As a whole the reaction time was slightly increased with manipulator arm speed, which is showed somewhat different pattern due to failure type. The reaction time to an axis acting on a workpiece directly, which could flex and extend, was fastest and much more its standard deviation was small. Various factors which may affect safe speed were also described.

Characteristics of human operator,Decommissioning,Maintenance,Manipulator,Nuclear facilities,Safe speed

A porous TiO2 pellet was electrochemically converted to the metallic titanium by using a CaCl2 molten salt system at 850 . Ni-TiO2 and graphite electrodes were used as cathode and anode, respectively. The electrochemical behaviour of TiO2 pellet was determined by a constant voltage control electrolysis. Various reaction intermediates such as CaTiO3, Ti2O and Ti6O were observed by XRD analysis during electrolysis of the pellet. Once TiO2 pellet was converted to a porous metallic structure, the porous structure disappeared by sintering and shrinking with increasing the reaction time at high temperature.

titanium sponge,electrochemical reduction,titanium dioxide,molten salt,calcium chloride

Necessity of demonstration test to evaluate the structural integrity of a basket for accident conditions arose during license approval procedure for the WSPP';s dry storage facility named MACSTOR/KN-400. A drop test facility for demonstration was constructed in KAERI site and demonstration tests for basket drop were conducted. As the upper welding region of a loaded basket was collided with a dropped basket during the drop test, the welding in this region was fractured and leakage happened after the drop test. The enhancement of basket design was needed since the existing basket design was not able to satisfy the performance requirement. The directions for design modification were determined and six enhanced designs were derived based on these directions. Structural analyses and specimen tests for each enhanced design were conducted. By evaluating structural analysis results and test results, one among six enhanced designs was decided as a final design for revision. The final design was the one to reduce the height of central post of a basket and to decrease the impact velocity with a dropped basket. Test basket models were fabricated with accordance with the final enhanced design. Additional demonstration test was performed for this test model and all the performance requirements were satisfied.

Spent fuel storage basket,Dry storage,Drop test,Leak test,Penetrant Test,Structural analysis

In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

Radiation,Shielding,MCNP code,Concrete wall,Radwaste drum assay system

The dehydration schemes of rare earth (La, Ce, Nd, Pr, Sm. Eu, Gd, Y) chloride hydrates was investigated by using a dehydration apparatus. To prevent the formation of the rare earth oxychlorides, the operation temperature was changed step by step (80 → 150 → 230 ℃ ) based on the TGA (thermo-gravimetric analysis) results of the rare earth chloride hydrates. A vacuum pump and preheated Ar gas were used to effectively remove the evaporated moisture and maintain an inert condition in the dehydration apparatus. The dehydration temperature of the rare earth chloride hydrate was increased when the atomic number of the rare earth nuclide was increased. The content of the moisture in the rare earth chloride hydrate was decreased below 10% in the dehydration apparatus.

Dehydration,rare-earth chloride hydrates,Anhydrous rare-earth chlorides

The impacts influenced on the performance and safety of a repository are classified as units of Features, Events, and Processes (FEP), for the total system performance assessment (TSPA) related to the permanent disposal of HLW. The importance is evaluated in consideration of the frequency, consequence, regulation, suitability of a specific site, etc. and then these are grouped as a similar FEP. A scenario describing the migration of radionuclide from the repository to the biosphere is derived from understanding the interaction among these groups. KAERI has developed the KAERI FEP lists by review and collation of the foreign studies. The KAERI FEP list has been reviewed by several Korean experts. The five major scenarios describing possible future evolutions of the geological disposal system have been developed by RES and PID methods. Also the CYPRUS which is a KAERI integrated database management system for the total system performance assessment (TSPA) related to the permanent disposal of HLW has been developed and the results of the FEP and scenario development have been uploaded in this system.

HLW,disposal,safety assessment,total system performance assessment,FEP,scenario

Groundwater flow simulations were performed to obtain data of groundwater flow used in a safety assessment for a hypothetical geological disposal facility assumed to be located in the KURT (KAERI Underground Research Tunnel) site. A regional scale modeling of the groundwater flow system was carried out to make boundary conditions for a local scale modeling. And, fracture zones identified at the study site were involved in the local scale groundwater flow model. From the results of the local scale modeling, a hydraulic head distribution was indicated and it was used in a particle tracking simulation for searching pathway of groundwater from the location of the hypothetical disposal facility to the surface where the groundwater reached. The flow distance and discharge rate of the groundwater in the KURT site were calculated. It was thought that the modeling methods used in this study was available to prepare the data of groundwater flow in a safety assessment for a geological disposal facility of radioactive wastes.

Groundwater flow modeling,Hydrogeological model,Particle tracking simulation,Groundwater pathway

The potential repository domains for A-KRS (Advanced Korean Reference Disposal System for High Level Wastes) in geological characteristics of KURT (KAERI Underground Research Tunnel) facility site were proposed to develop a repository system design and to perform the safety assessment. The host rock of KURT facility site is one of major Mesozoic plutonic rocks in Korean peninsula, two-mica granite, which was influenced by hydrothermal alteration. The topographical features control the flow lines of surface and groundwater toward south-easterly and all waters discharge to Geum River. Fracture zones distributed in study site are classified into order 2 magnitude and their dominant orientations are N-S and E-W strike. From the geological features and fracture zones, the potential repository domains for A-KRS were determined spatially based on the following conditions: (1) fracture zone must not cross the repository; and (2) the repository must stay away from the fracture zones greater than 50 m. The western region of the fracture zones in the N-S direction with a depth below 200 m from the surface was sufficient for A-KRS repository. Because most of the fracture zones in N-S direction were inclined toward the east, we expected to find a homogeneous rock mass in the western region rather than in the eastern region. The lower left domain of potential domains has more suitable geological and hydrogeological conditions for A-KRS repository.

A-KRS,KURT,Geological characteristics,Fracture zone,Potential domain for repository

To provide domestic values of input parameters in a safety assessment of radioactive waste disposal under domestic deep underground environments, various kinds of experiments have been carried out under KURT (KAERI Underground Research Tunnel) conditions. The input parameters were classified, and some of them were selected for this study by the criteria of importance. The domestic experimental data under KURT environments were given top priority in the data review process. Foreign data under similar conditions to KURT were also gathered. The collected data were arranged and the statistical calculations were processed. The properties and distribution of the data were explained and compared to foreign values in view of their validity. The following parameters were analysed: failure time and early time failure rate of a container, solubility of nuclides, porosity and density of the buffer, and distribution coefficients of nuclides in the geomedia, hydraulic conductivity, diffusion depth of nuclides, groundwater flow rate, fracture aperture, length of internal fracture, and width of faulted rock mass in the host rock.

input parameter,safety assessment,solubility,sorption,diffusion,migration

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called “A-KRS,” in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

A-KRS pyro-processed waste,repository system,deterministic safety assessment,GoldSim

Geochemical composition of fracture filling minerals and groundwater was investigated to characterize geochemical characteristics of groundwater system at the KURT site. Minerals such as calcite, illite, laumontite, chlorite, epidote, montmorillonite, and kaolinite, as well as I/S mixed layer minerals were detected in the minerals extracted from the fracture surfaces of the core samples. The groundwater from the DB-1, YS-1 and YS-4 boreholes showed alkaline conditions with pH of higher than 8. The electrical conductivity (EC) values of the groundwater samples were around 200 S/cm, except for the YS-1 borehole. Dissolved oxygen was almost zero in the DB-1 borehole indicating highly reduced conditions. The Cl- concentration was estimated around 5 mg/L and showed homogeneous distribution along depths at the KURT site. It might indicate the mixing between shallow groundwater and deep groundwater. The shallow groundwater from boreholes showed Ca-HCO3 type, whereas deep groundwater below 300 m from the surface indicated Na-HCO3 type. The isotopic values observed in the groundwater ranged from -10.4 to -8.2‰ for 18O and from -71.3 to -55.0‰for D. In addition, the isotope-depleted water contained higher fluoride concentration. The oxygen and hydrogen isotopic values of deep groundwater were more depleted compared to the shallow groundwater. The results from age dating analysis using 14C indicated relatively younger (2000 6000yr old) groundwater compared to other european granitic groundwaters such as Stripa (Sweden).

Geochemical characterization,KURT,Groundwater,Stable isotope,Fracture filling mineral

To develop site characterization technologies for a radioactive waste disposal research in KAERI, the geological and hydrogeological investigations have been carried out since 1997. In 2006, the KURT (KAERI Underground Research Tunnel) was constructed to study a solute migration, a microbiology and an engineered barrier system as well as deeply to understand geological environments in in-situ condition. This study is performed as one of the site characterization works around KURT. Several investigations such as a lineament analysis, a borehole/tunnel survey, a geophyscial survey and logging in borehole, were used to construct the geological model. As a result, the geological model is constructed, which includes the lithological model and geo-structural model in this study. Moreover, from the results of the in-situ hydraulic tests, the hydrogeological properties of elements in geological model were evaluated.

Site characterization technologies,KAERI Underground Research Tunnel,Geological model,Hydrogeological properties

A pyroprocess is currently being developed by KAERI to cope with a highly accumulated spent nuclear fuel in Korea. The pyroprocess produces a certain amount of high-level radioactive waste (HLW), which is solidified by a ceramic binder. The produced ceramic waste will be confined in a secure disposal canister and then placed in a deep geologic formation so as not to contaminate human environment. In this paper, the development of a disposal canister was overviewed by discussing mainly its design premises, constitution, manufacturing methods, corrosion resistance in a deep geologic environment, radiation shielding, and structural stability. The disposal canister should be safe from thermal, chemical, mechanical, and biological invasions for a very long time so as not to release any kind of radionuclides.

Disposal canister,Pyroprocess,Ceramic waste,High level waste,Cold spray coating

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

Spent fuel,Pyro-process,Ceramic waste,Metal waste,Deep geological disposal system

A molten salt technology using Li2O-LiCl has been extensively investigated to recover uranium metal from spent fuels in the field of nuclear energy. In the reduction process, it is an important point to maintain the concentration of Li2O. ZrO2 is inevitably contained in the spent fuels because Zr is one of the main components of fuel rod hulls. Therefore, the fate of ZrO2 in Li2O-LiCl molten salt has been investigated. It was found that Li2ZrO3 and Li4ZrO4 were formed chemically and electrochemically and they were not reduced to Zr. The recycling of Li2O is the key mechanism ruling the total reaction in the electrolytic reduction process. However, ZrO2 will have a role as a Li2O sink.

Li2O-LiCl molten salt,Electrolytic reduction,ZrO2,Loss of Li2O

In this study, numerical model for transport of radionuclide and colloid was developed. In order to solve reaction-migration governing equation for colloid and radionuclide, Strang-splitting Sequential Non-Iterative (SNI), which is one of Operator Splitting Method, was used for numerical method and this was coded by MATLAB. From the verification by comparing the simulation results with analytical solution considering only solute transport and rock diffusion, the Pearson’s correlation coefficient was greater than 0.99 which demonstrates the accuracy of the model.

Transport,Colloid,Radionuclide,Numerical model,Advection-dispersion

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of SiO2 and B2O3 for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

Vitrification,Glass formulation,Aluminum metal,Glass fiber

Numerical simulations were performed to evaluate the dispersion characteristics of the pollutant around a Wolsung coastal area at located nuclear power plants. Numerical experiments by using EFDC(Environmental Fluid Dynamics Code) showed good agreements by comparison with the time series and harmonic analysis of the tidal elevations. The released pollutants moved in north direction at flood tide and in south direction at ebb tide. The calculated salinity and temperatures showed good agreements with the observed results by NFRDI(National Fisheries Research & Development Institute). The water circulation due to the variations of the temperature, salinity and tidal components were analyzed to estimate the dispersion characteristics of the pollutant.

Wolsung,EFDC,Water circulation,Harmonic analysis

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.

A-KRS,Pyro-processedradioactive waste,Repository,Probabilistic safety assessment,GoldSim

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in 500oC LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

Zircaloy-4,Hull wastes,Molten salts,Zr oxide,Dissolution,Electrorefining

The current status of the computer codes for the analysis of coupled thermal-hydrological-mechanical behavior occurred in a high-level waste repository was investigated. Based on the reported results on the comparison between the predictions using the computer codes and the experimental data from the in-situ tests, the reliability of the existing computer codes was analyzed. The presented codes simulated considerably well the coupled thermal-hydrological-mechanical behavior in the near-field rock of the repository without buffer, but the predictions for the engineered barrier system of the repository located at saturated hard rock were not satisfactory. To apply the current thermal-hydrological-mechanical models to the assessment of the performance of engineered barrier system, a major improvement on the mathematical models which analyze the distribution of water content and total pressure in the buffer is required.

Thermal-hydrological-mechanical behavior,High-level waste repository,Engineered barrier system,Numerical model,Validation

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advectiondispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed 10-3, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.

Particle tracking simulation,Groundwater pathway,Radionuclide transport,Time domain random walk

Geology, hydrogeology, and geochemistry are the main technical siting factors of a geological repository for spent nuclear fuels. This paper focused on how rock’s different geological conditions, such as topography, soils, rock types, structural geology, and geological events, influence the functions of the geological repository. In the context, the site selection criteria of various countries were analyzed with respect to the geological conditions. Each country established the criteria based on its important geological backgrounds. For example, it was necessary for Sweden to take into account the effect of ice age on the land uplift and sea level change, whereas Japan defined seismic activity and volcanism as the main siting factors of the geological repository. Therefore, the results of the paper seems to be helpful in preparing the siting criteria of geological repository in Korea.

Geological disposal,Repository,Geology,Siting factors,Criteria

Tetraethoxysilane(TEOS) as a silica precursor and hydrochloric acid as an acid catalyst have been used in a surfactant-template synthesis of micrometer-sized mesoporous silica microspheres based on the macroemulsion technique. Increase in the concentration of tetraethoxysilane of the reaction mixture has a serious destructive effect on the particle shape and pore structure. As the acid concentration increases, relatively small microspheres are formed without destroying their spherical morphology of the particles as well as the pore structures. However, due to the attractive interaction between particles in an acidic condition, strong silica agglomerates are formed, and therefore are subject to a post-ultrasonic treatment to separate into an individual single particle.

Mesoporous silica microspheres,emulsion,surfactant,template synthesis,tetraethoxysilane,acid catalyst

To characterize the site specific properties of a study area for high-level radioactive waste disposal research in KAERI, the several geological investigations such as surface geological surveys and borehole drillings were carried out since 1997. Especially, KURT (KAERI Underground Research Tunnel) was constructed to understand the further study of geological environments in 2006. As a result, the first geological model of a study area was constructed by using the results of geological investigation. The objective of this research is to construct a hydrogeological model around KURT area on the basis of the geological model. Hydrogeological data which were obtained from in-situ hydraulic tests in the 9 boreholes were estimated to accomplish the objective. And, the hydrogeological properties of the 4 geological elements in the geological model, which were the subsurface weathering zone, the log angle fracture zone, the fracture zones and the bedrock were suggested. The hydrogeological model suggested in this study will be used as input parameters to carry out the groundwater flow modeling as a next step of the site characterization around KURT area.

High-level radioactive waste disposal,KAERI Underground Research Tunnel,Geological model,Hydrogeological model,Site characterization

An engineering scale test, which is called KENTEX, was carried out to understand and to analyze the coupled thermal, hydrological and mechanical phenomena in the engineered barrier system(EBS) of Korean reference disposal system. Using the experimental data obtained from KENTEX, the water saturation processes in bentonite could be analyzed. From the comparison between the model calculation using ABAQUS and the experimental results, the difference of the water content between them in the unsaturating part was large because the drying phenomena due to moisture redistribution by the temperature gradient could not be included in the model. In the saturating part, the difference of the water content between them was decreased gradually and showed to be small in the full saturation. And the time of about 95% saturation could be estimated about 500 days from the model calculation and experimental results. Also it could be known that the moisture redistribution in the unsaturated part could not be affected on the saturation time of bentonite in the repository. Therefore, it is considered that this model could be used to quantitatively predict the water saturation time in bentonite as EBS for the disposal system.

Saturation Processes,Engineered Barrier,Bentonite Buffer,Coupled THM,Disposal System,KENTEX

In the compacted bentonite containing Ag2O, the transport of iodide ion was investigated by Through-diffusion method. It is confirmed that Iodide ion is transported by diffusion process in the compacted bentonite containing Ag2O as well as in the compacted bentonite without Ag2O. However, the lag-time of iodide ion in the compacted bentonite containing Ag2O is larger than that in the compacted bentonite without Ag2O. The increase of the lag-time was observed in pure iodide ion solution and also in 0.1M NaCl-iodide ion solution. The apparent diffusion coefficient of iodide ion in the compacted bentonite containing Ag2O has lower value than that in the compacted bentonite without Ag2O. The effect of Ag2O on the effective diffusion coefficient was not clearly investigated in the compacted bentonite containing Ag2O while the values of effective diffusion coefficient of iodide ion in the compacted bentonite without Ag2O obtained in this study were similar to those in the compacted bentonite reported in the literature.

Compacted bentonite,Through-diffusion method,Iodide ion,Time-lag,Apparent diffusion coefficient,Effective diffusion coefficient

A mobile laser-induced breakdown detection (LIBD) system was developed for the field measurement of the size and concentration of aquatic colloidal nanoparticles sampled from Korea Atomic Energy Research Institute Underground Research Tunnel (KURT). The established LIBD apparatus is based on the optical detection of a laserinduced plasma by means of a two-dimensional optical imaging method for determining the size of nanoparticle. Calibration curve for determining the size of nanoparticle was obtained from the polystyrene reference particles of a well-defined size. The first direct application was made at KURT for investigating the particle sizes in groundwater. By comparing the size of particles in groundwater with the sizes of reference particles, the mean particle size of approximately 108 26 nm with the concentration lower than 50 ppb was determined.

LIBD,Colloids,Particle size,Particle concentration,Migration of nuclides

Recently, International Atomic Energy Agency and major leading countries in radioactive waste management tend to subdivide the categories of radioactive waste based upon risk-graded approach. In this context, the category of very low level waste has been newly introduced, or optimized management options for this kind of waste have been pursued in many countries. The application of engineered surface landfill type facilities dedicated to dispose of very low level waste has been gradually expanded, and it was analyzed that their design concept of isolation has been much advanced than those of the old fashioned surface trench-type disposal facilities for low and intermediate level waste, which were usually constructed in 1960';s. In addition, the management options for very low level waste in major leading countries are varied depending upon and interfaced with the affecting factors such as: national framework for clearance, legal and practical availability of low and intermediate level waste repository and/or non-nuclear waste landfill, public acceptance toward alternative waste management options, and so forth. In this regard, it was concluded that optimized long-term management options for very low level waste in Korea should be also established in a timely manner through comprehensive review and discussions, in preparation of decommissioning of large nuclear facilities in the future, and be implemented in a systematic manner under the framework of national policy and management plan for radioactive waste management.

Very low-level radioactive waste,Graded Approach,Landfill disposal,International trend,Safe management,Implications

A large volume of uranium electrokinetic leachate has been generated during the electrokinetic decontamination to remove uranium from contaminated soil. The treatment technology for the reuse of the uranium leachate was developed. The concentration of uranium in the generated uranium leachate was 180 ppm and concentrations of Mg(II), K(I), Fe(II), and Al(III) ions ranged from 20 ppm to 1,210 ppm. The treatment process for uranium leachate consisted mainly of mixing and cohesion, precipitation, concentration, and filtration. In order to obtain the pH=11 of a precipitate solution, the calcium hydroxide needs to be 3.0g/100ml and the sodium hydroxide needed to be 2.7g/100ml. The results of several precipitation experiments showed that a mixture of NaOH+0.2g alum+0.15g magnetite was an optimal precipitant for filtration. The average particle size of precipitate with NaOH+alum+0.15g magnetite was 600 . Because the total value of metal concentrations in supernatant at pH=9 was the smallest, sodium hydroxide should be added with 0.2g alum and 0.15g magnetite for pH=9 of leachate.

uranium,electrokinetic,leachate,cohesion,precipitation,filtration

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source (60Co radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

Spent Nuclear Fuel,Transportation/Storage System,Burnup Credit,Radiation Shielding

Polymer-modified cement is the composite material made by partially replacing and strengthening the cement hydrate binders of conventional mortar with polymeric modifiers such as polymer latexes and redispersible polymeric modifiers. It is known that the addition of polymer to cement mortar leads to improved quality, which would be expected to have a high chemical resistance.Therefore, the purpose of this study is to identify the improved chemical resistance, such as low permeability and low ion diffusivity, of the polymermodified cement as a solidification agent for the radwaste. First, polymer-modified cement specimens by latex modification were prepared according to the polymer content from 0 % to 30% to select the optimized polymer content. At those specimens, the water-to-cement (W/C) ratio was maintained to 33% and 50% respectively. After the much curing time, the structural integrity of specimens was evaluated through the compressive strength test and the porosity evaluation by the water immersion method. From the results, 10% of the polymer content at 33% of the W/C ratio was shown to have the most improved quality. Finally, the leaching test referredfrom ANS 16.1 for the specimens having the most improved quality was conducted. Dedicated specimens for the leaching test were then mixed with radioisotopes of 60Co and 137Cs at the specimen preparation.

Polymer-modified cement,Integrity,Compressive strength,Porosity,Immersion,Leaching

This paper introduces the policy and regulations on very low level waste (VLLW) management in China. Given the important decommissioning and site restoration program of the old facility, it is considered necessary to create a new disposal facility dedicated to VLLW. Many general design principles are in common with to the disposal facility for low and intermediate level waste (LILW), namely the isolation of the waste by means of a multibarrier system, but using bentonite and/or high density polyethylene membranes instead of the generalized use of concrete barriers. The design of the facility is consistent with the design of disposal facilities for hazardous waste. The engineering design of two VLLW disposal facilities is introduced.

VLLW,Iandfill Disposal,Multibarrier,disposal cell,Decommissioning

Saline corrosion is one of the major degradation mechanisms for stainless steel type 304 (SS304) dry storage cask during the spent fuel interim storage period. Slow strain rate test (SSRT) and neutral salt spray test (NSS) were performed at 85°C and 200°C with 0.5 wt% sodium chloride mist sprayed on the cold-rolled SS304 specimens of different degrees of reduction in this study. The weight changes of the NSS specimens tested at 85°C for 2000 hours differed greatly from those at 200°C. The weight loss of NSS specimens was not significant at 85°C but the weight gain decreased gradually with increasing the cold-rolled reduction. The yield strength (YS) and ultimate tensile stress (UTS) values obtained from the SSRT tests for lightly cold-rolled specimens in the salt spray environment at 85°C and 200°C are slightly lower than in air. But for those with 20% reductions, the specimen strengths were no longer changed by the saline corrosion. The preliminary results demonstrated that the quality and performance of cold-rolled SS304 is acceptable for fabrication of dry storage casks. However, more work on the corrosion behavior of cold-rolled stainless steel in the saline atmosphere is needed to better understand its long-term performance.

storage cask,spent fuel,saline corrosion,stainless steel,strength

An integrated model for groundwater flow and radionuclide transport analyses is being developed incorporating six underground silos, an excavated damaged zone (EDZ), and fractured host rock. The model considers each silo as an engineered barrier system (EBS) consisting of a waste zone comprising waste packages and disposal container, a buffer zone, and a concrete lining zone. The EDZ is the disturbed zone adjacent to silos and construction & operation tunnels. The heterogeneity of the fractured rock is represented by a heterogeneous flow field, evaluated from discrete fractures in the fractured host rock. Radionuclide migration through the EBS in silos and the fractured host rock is simulated on the established heterogeneous flow field. The current model enables the optimization of silo design and the quantification of the safety margin in terms of radionuclide release.

LILW radioactive waste disposal,Performance assessment,Multiple silos,EDZ,Flow and transport

The release of a site and building from regulatory control is the final stage of the decommissioning process. The MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual) provides overall framework for conducting data collection for a final status survey to demonstrate compliance with site closure requirements. The KAERI carried out establishing a final status survey by using the guidance provided in the MARSSIM for of a site and building of the Korea Research Reactor. The release criteria for a site and building were set up based on these results of the site specific release levels which were calculated by using RESRAD and RESRAD-Build codes. The survey design for a site and building was classified by using the survey dataset and potential contamination. The number of samples in each survey unit was calculated by through a statistical test using the collected data from a scoping and characterization survey. The results of the final status survey were satisfied the release criteria based on an evaluation of the measured data.

Decommissioning,Final Status Survey,MARSSIM,RESRAD,RESRAD-Build

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.

Spent fuels,Storage,Design Basis

The effects of the applied current density, the AgNO3 concentration, the scrubbing liquid flow rate and the NO-air mixed gas flow rate on the NO removal efficiency were investigated by using Ag2+ mediated electrochemical oxidation (MEO). Results showed that the NO removal efficiency increased with increasing the applied current density. The effect of the AgNO3 concentration on the NO removal efficiency was negligibly small in the concentration of AgNO3 above 0.1 M. When the scrubbing liquid flow rate increased, the NO removal efficiency was gradually increased. On the other hands, the NO removal efficiency decreased with increasing the NO-air mixed gas flow rate. As a result of the treatment of NO-air mixed gas by using the MEO process with the optimum operating condition and the chemical absorption process using 3 M HNO3 solution as a scrubbing liquid, the removal efficiency of NO and NOx was achieved as 95% and 63%, respectively.

NO,Mediated electrochemical oxidation (MEO),Current density,Removal efficiency,Gas scrubber

This study was carried out to examine the stability of Na2CO3-H2O2 carbonate solution with aging time in the dissolving solution after oxidative dissolution of U by a carbonate solution, the Cs/Re filtrate after selective precipitation of Cs and Re (as a surrogate for Tc), and the acidification filtrate after precipitation of U by acidification, respectively. The compositions of dissolving solution were not changed with ageing time, and the selective precipitation of Re and Cs was also not affected without regard to the aging time of dissolving solution. The successive removal of Cs and Re from a carbonate dissolving solution was possible. However, the recovery yield of U by acidification was decreased with increasing the aging time of the dissolving solution and the Cs/Re-filtrate, respectively, because U was converted from the uranyl peroxocarbonato complex to the uranyltricarbonate in the solution aged for a long time. Accordingly, it is effective that a dissolving solution and a Cs/Re filtrate are treated within the aging of 7 days, respectively, in order to recover U more than 99%. On the other hand, the recovery of U was carried out in order of the oxidative dissolution of U selective precipitation of Re and Cs precipitation of U by acidification. Almost all of U and a part of FP were co-dissolved in oxidative dissolution step. Over 99% of Re and Cs from the FP co-dissolved with U could be removed by a TPPCl (tetraphenylphosphonium chloride) and a NaTPB (sodium tetraphenylborate), respectively. U was precipitated nearly 100% by acidification to pH 4. Therefore, it was confirmed that the validity of recovery of U by precipitation methods from a SF (spent fuel) in the Na2CO3-H2O2 solution.

U,Cs,Tc,Carbonate solution,Oxidative dissolution,Selective precipitation,Acidification

It is clarified in the radioactive waste transfer regulation that the concentration of radioactive waste for the major radio nuclide has to be examined when radioactive waste is guided to the radioactive waste stores. In case of the low level radioactive waste sample, the analytical results of radioactive waste concentration frequently show a value lower than minimum detectable activity (MDA). Since the MDA value basically depends on the amount of a sample, background value, measurement time, counting efficiency, and etc, it would be necessary to increase a sample amount with a intention of minimizing MDA. In order to measure a concentration of 129I in low and medium level radioactive waste, 129I was collected by using a distillation technique after leaching the simulated radioactive waste sample with a non-volatile acid. The recovery of 129I measured was compared with that measured with column elution technique which is a conventional method using an anion-exchange resin. The recovery of inactive iodide by using the distillation method and column elution were found as 86.5 0.9% and 87.3 2.7%, respectively. The recovery and MDA value calculated for distillation technique when 100 g of extracted solution of 129I was taken, were found to be 84.6 1.6% and 1.2 10-4 Bq/g, respectively. Consequently, the proposed technique with simplified process lowered the MDA value more than 10 times compared to the column elution technique that has a disadvantage of limited sampling amount.

129I,Radioactive waste,MDA,Distillation technique,Recovery

Nuclear power utilities should establish a site-specific groundwater monitoring program for early detection of unplanned radioactive material';s releases which can occur due to degradation of systems, structures and components of the nuclear power plants in order to keep the impact of the unplanned releases on the environment and the residents as low as reasonably achievable. For this end, groundwater flow on site should be evaluated based on characterization of the hydrogeology of a site of concern. This paper aims to provide data necessary for establishing groundwater monitoring program which is currently considered at Kori nuclear power plant 1 by characterizing groundwater flow system on the site based on the existing hydrogeological studies and related documents, and by modeling tritium transport. The results showed that the major groundwater flow direction was south-west and that most of groundwater entered a southern and eastern seas. Although the tritium plume also released into the sea, its rate was delayed by dewatering sump.

Unplanned Release,Groundwater,Tritium,Monitoring,Modeling

In case that the maximum temperature of any surface readily accessible during transport of a spent nuclear fuel (SNF) transport cask exceeds 85°C in the absence of insolation under the ambient temperature of 38°C, personnel barriers or transport hood shall be used to prevent people from casual contact with the transport cask surface. Usually the air temperature within the hood and the hood surface temperature are calculated and further utilized as boundary conditions(free stream temperature and external radiation temperature) for thermal evaluation under normal conditions of transport. In this study, these temperatures are derived using the analytical method based on the heat transfer mechanism around the transport cask under transport hood assuming the thermal equilibrium. By comparing the analytical solutions with the results from the detailed calculations with CFD-computer-code FLUENT 12.1 it is verified that the analytical method is still efficient tool to estimate the temperatures and these temperatures can be further used as boundary conditions for thermal evaluation under normal conditions of transport.

Spent Nuclear Fuel Transport Cask,Transport Hood,Heat Transfer,Natural Convection,Air Flow

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100oC were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

CANDU spent fuel,Geological disposal concept,Thermal requirement,Multi barrier,Unit disposal area,Disposal efficiency

The smooth construction and operation of nuclear facilities requires understanding and support of both the local residents and the national people. It is essential that our country, which should maintain using nuclear energy for national energy security and economic growth, shall improve the social acceptance of nuclear energy. In order to identify the level of social acceptance of nuclear energy, this study investigated the perception of the local residents in Gyeongju and the public in other areas on a nuclear power plant and a low- and intermediate-level radioactive waste disposal facility through an individual interview. The subjects of the investigation were 450 persons. This study identified that perceptions of the respondents were somewhat dependent on the residential area, and derived the implications to be reflected in establishing the customized public-relation strategies.

Social acceptance nuclear power plant,low- and intermediate-level radioactive disposal facility,perception

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

Burnup Credit,Spent Fuel,Transport Cask,Uncertainty,Chemical Assay

The influence of water salinity on the hydraulic conductivities of compacted bentonites with several dry densities were studied. The hydraulic conductivity increases with increasing salinity only when the dry density of bentonite is relatively low. The degree of increase becomes more remarkable at a lower dry density of bentonite. For bentonite with the density of 1.0 Mg/m3 and 1.2 Mg/m3, the hydraulic conductivity of the 0.4 M NaCl solution increases up to about 7 times and 3 times, respectively higher than that of freshwater. However, for the bentonite with a dry density higher than 1.4 Mg/m3, the salinity has an insignificant effect on the hydraulic conductivity, and the hydraulic conductivity is nearly constant within the salinity range of 0.04 to 0.4 M NaCl. The pre-saturation of the bentonite specimen with freshwater has no significant influence on the hydraulic conductivity.

hydraulic conductivity,bentonite,salinity effects,diffuse double layer,electric potential

In this study the complex formation reactions between uranium(VI) and 2,6-dihydroxybenzoate (DHB) as a model ligand of humic acid were investigated by using UV-Vis spectrophotometry and time-resolved laser-induced fluorescence spectroscopy (TRLFS). The analysis of the spectrophotometric data, i.e., absorbance changes at the characteristic charge-transfer bands of the U(VI)-DHB complex, indicates that both 1:1 and 1:2 (U(VI):DHB) complexes occur as a result of dual equilibria and their distribution varies in a pH-dependent manner. The stepwise stability constants determined (log K1 and log K2) are 12.4 0.1 and 11.4 0.1. Further, the TRLFS study shows that DHB plays a role as a fluorescence quencher of U(VI) species. The presence of both a dynamic and static quenching process was identified for all U(VI) species examined, i.e., UO22+, (UO2)2(OH)22+, and (UO2)3(OH)5+. The fluorescence intensity and lifetimes of each species were measured from the time-resolved spectra at various ligand concentrations, and then analyzed based on Stern-Volmer equations. The static quenching constants (log Ks) obtained are 4.2 0.1, 4.3 0.1, and 4.34 0.08 for UO22+, (UO2)2(OH)22+, and (UO2)3(OH)5+, respectively. The results of Stern-Volmer analysis suggest that both mono- and bi-dentate U(VI)-DHB complexes serve as groundstate complexes inducing static quenching.

Uranium(VI),2,6-Dihydroxybenzoic acid,Chelate complex,Complex formation constant,Absorption spectroscopy,TRLFS

The nuclear fuel cycle (NFC) analysis is a study to set a NFC policy and to promote systematic researches by analyzing technologies and deriving requirements at each stage of a fuel cycle. System analysis techniques are utilized for comparative analysis and assessment of options on a considered system. In case that NFC is taken into consideration various methods of the system analysis techniques could be applied depending on the range of an interest. This study presented NFC analysis strategies for the development of a domestic advanced NFC and analysis techniques applicable to different phases of the analysis. Strategically, NFC analysis necessitates the linkage with technology analyses, domestic and international interests, and a national energy program. In this respect, a trade-off study is readily applicable since it includes various aspects on NFC as metrics and then analyzes the considered NFC options according to the derived metrics. In this study, the trade-off study was identified as a method for NFC analysis with the derived strategies and it was expected to be used for development of an advanced NFC. A technology readiness level (TRL) method and NFC simulation codes could be utilized to obtain the required metrics and data for assessment in the trade-off study. The methodologies would guide a direction of technology development by comparing and assessing technological, economical, environmental, and other aspects on the alternatives. Consequently, they would contribute for systematic development and deployment of an appropriate advanced NFC.

Nuclear Fuel Cycle Analysis,Advanced Nuclear Fuel Cycle,Trade-off Study,Technology Readiness Level,Nuclear Fuel Cycle Simulator

When the Fukushima nuclear power plant accident occurred in March of 2011, a hydrogen explosion in the reactor building at the 4th unit of Fukushima plants led to a big surprise because the full core of the unit 4 reactor had been moved and stored underwater at the spent nuclear fuel storage pool for periodic maintenance. It was because the possible criticality in the fuel storage pool by coolant loss may yield more severe situation than the similar accident happened inside the reactor vessel. Fortunately, it was assured to be evitable to an anxious situation by a look of water filled in the storage pool later. In the paper, the safety state of the spent fuel storage pool and rack structures of the domestic nuclear plants would be roughly reviewed and compared with the Fukushima plant case by engineering viewpoint of potential severe accidents.

Severe Accident,Fuel Building,Spent Fuel Storage Rack,Seismic Safety

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

Spent nuclear fuel,Spent nuclear fuel storage facility,Decay Heat,Heat load,Depletion parameter,Storage Scenario

Analysis of foreign and domestic patents for pyroprocessing technology of spent nuclear fuels was carried out in this study. The current status of pyroprocessing technology development in such countries as Korea, USA, Japan and EU was analyzed by classifying the patents for 1975 through 2009 according to registration country, assignee, calendar year and technology area. The major assignees'; activity indices were compared in order to find out whether there is any concentrated area of technical details. Technology competitiveness of the countries was also investigated from the information of patent citation number and family size. Furthermore, some essential unit technologies required for the commercialization of pyroprocessing were derived and examined in the aspect of the state of art as well as the trend of technology development.

Patent analysis,Pyroprocessing,Electrochemical reduction,Electrorefining,Electrowinning,Molten salt waste

In the present work, an electrowinning process in the LiCl-KCVCd system is considered to model and analyze the electrotransport of the actinide and rare-earth elements. A simple dynamic modeling of this process was performed by taking into account the material balances and diffusion-controlled electrochemical reactions in a diffusion boundary layer at an electrode interface between the molten salt electrolyte and liquid cadmium cathode. The proposed modeling approach was based on the half-cell reduction reactions of metal chloride occurring on the cathode. This model demonstrated a capability for the prediction of the concentration behaviors, a faradic current of each element and an electrochemical potential as function of the time up to the corresponding electrotransport satisψing a given applied current based on a galvanostatic electrolysis. The results of selected case studies including five elements (U, Pu, Am, La, Nd) system are shown, and a preliminary simulation is carried out to show how themodel can be used to understand the electrochemical characteristics and provide betler information for developing an advanced electrowinner.

Electrowinner,LiCI-KCl,Liquid cadmium cathode,Diffusion-controlled electrotransport,Actinide elements,Rare-earth elements

The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.

Pyroprocessing,Electrowinning,LCC(Liquid Cadmium Cathode),Uranium dendrite,Mesh type LCC structure,Pyroprocessing

Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing Li2O concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing U3O8 electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between U3O8 and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.

Pyroprocessing,Spent Oxide Fuel,Electrolytic Reduction,Model,Simulation

The electrolytic reduction of a spent oxide fuel involves a liberation of the oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is too corrosive for typical structural materials. Accordingly, it is essential to choose the optimum material for the processing equipment that handles the high molten salt. In this study, hot corrosion studies were performed on bare as well as coated superalloy specimens after exposure to lithium molten salt at 675 for 216 h under an oxidizing atmosphere. The IN713LC superalloy specimens were sprayed with an aluminized NiCrAlY bond coat and then with an Y2O3 top coat. The bare superalloy reveals an obvious weight loss due to spalling of the scale by the rapid scale growth and thermal stress. The chemical and thermal stability of the top coat has been found to be beneficial for increasing to the corrosion resistance of the structural materials for handling high temperature lithium molten salts.

Plasma spray coating,yttria top coat,hot corrosion,structural material,electrolytic reduction

Uranium deposits from an electrorefining process contain about 30% salt. In order to recover pure uranium and transform it into an ingot, the salts have to be removed from the uranium deposits. Major process variables for the salt distillation process of the uranium deposits are hold temperature and vacuum pressure. Effects of the variables on the salt removal efficiency were studied in the previous study[1]. By applying the Hertz-Langmuir relation to the salt evaporation of the uranium deposits, the evaporation coefficients were obtained at the various conditions. The operational conditions for achieving above 99% salt removal were deduced. The salt distilled uranium deposits tend to form the eutectic melt with iron, nickel, chromium for structural material of salt evaporator. In this study, we investigated the hold temperature limitation in order to prevent the formation of the eutetic melt between urnaium and other metals. The reactions between the uranium metal and stainless steel were tested at various conditions. And for enhancing the evaporation rate of the salt and the efficient recovery of the distilled salt, the thermal analysis of the salt distiller was conducted by using commercial CFX software. From the thermal analysis, the effect of Ar gas flow on the evaporation of the salt was studied.

Uranium deposit,salt distillation,eutectic melt,thermal analysis,commercial CFX software,salt removal efficiency

In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare eath oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix(ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilzed with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.

rare earth oxide,immobilization,solid phase sintering,monazite,borosilicate glass

Layer crystallization process was tested for the separation(or concentration) of cesium and strontium fission products in a LiCl waste salt generated from an electrolytic reduction process of a spent oxide fuel. In a crystallization process, impurities (CsCl and SrCl2) are concentrated in a small fraction of the LiCl salt by the solubility difference between the melt phase and the crystal phase. Based on the phase diagram of LiCl-CsCl-SrCl2 system, the separation possibility by using crystallization was determined and the molten salt temperature profile during layer crystallization operation was predicted by using mathematical calculation. In the layer crystallization process, the crystal growth rate strongly affects the crystal structure and therefore the separation efficiency. In the conditions of about 20-25 l/min cooling air flow rate and less than 0.2 g/min/cm2 crystal flux, the separation efficiency of both CsCl and SrCl2 showed about 90 % by the layer crystallization process, assuming a LiCl salt reuse rate of 90 wt%.

pyroprocessing,LiCl salt waste,reuse,Cs/Sr separation,layer crystallization

The distillation rate on LiCl-KCl eutectic salt under different vacuums from 0.5-50 mmHg was first investigated by using both a non-isothermal and a isothermal thermogravimetric (TG) analysis. Based on the non-isothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A distillation rate of 10-4-10-5 mole cm-2 sec-1 is obtainable at temperatures less than 1300K and vacuums of 0.5-50 mmHg. About a 99% salt distillation efficiency was obtained after an hour at a temperature above 1150 K under 50 mmHg in a small scale distillation test system. An increase in the vaporizing surface area is relatively effective for removing residual salt in the remaining particles, when compared to that for the vaporizing time. Over 99.95% of total distillation efficiency was obtained for a 1-h distillation operation by increasing the inner surface area from 4.52 cm2 to 12.56 cm2.

LiCl-KCl eutectic salt,Vacuum distillation,Salt removal,Rare earth precipitate,Thermogravimetric analysis

The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, and the domestic development of electrorefiner have been reviewed. The electrorefiner is composed of an anode basket containing reduced spent fuel such as uranium, transuranic and rare earth elements, and a solid cathode, which are in LiCl-KCl eutectic electrolyte. Oxidation (dissolution) reaction occurs on the anode and a pure uranium is electrochemically reduced (deposited) on the solid cathode. By application of graphite cathode, which has a self-scrapping characteristics for the uranium deposits, and a recovery of the fallen deposits by a screw conveyer, a high-throughput continuous electrorefiner with a capacity of 20 kgU/day has been developed.

Electrorefining,Spent fuel,Uranium,Electrodeposition,Graphite cathode

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-Li2O molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

Electrochemical Reduction,Spent Fuel,Molten Salt,Metallization,Pyroprocessing

The uranium ingot casting process is one of the steps which consolidate uranium deposits produced by electrorefiner as a ingot form in a pryprocessing technique. This paper introduces new design concept of the ingot casting equipment and the performance test results of the lab-scale ingot casting equipment fabricated based on the design concept. Casting equipment produces the uranium ingot by pouring a uranium melt into a mold by tilting a melting crucible. Also it is equipped with a cup which is able to continuously feed uranium deposits into a melting crucible. The productivity could be significantly enhanced by introducing the continuous operation concept.

Pyroprocessing,Uranium casting,Uranium ingot

As nuclear power plants are getting older, interests on a decontaminating process are increasingly attracting more attention. Chemical decontamination is crucial to lower the production of radioactive waste and radiation dose rate. Prior to this, oxidizers and detergents for target material should be chosen so as to decontaminate major systems and components of a nuclear power plant chemically. In order to decontaminate it properly, it is crucial to have information about the chemical composition and crystalline structure of CRUD, analyzing its samples from the target or the decontamination system with components. However, there is no program which enables the extraction of samples directly from the object or the decontamination system with components carrying genuine radioactivity. Therefore, it is limited to samples from corrosion products carrying partial radioactivity as a resource. The composition of CRUD varies considerably depending on refueling cycle because it is closely related to the constituent of basic material. After settling a target, it is crucial to analyze and obtain analytical information about CRUD as a decontamination target. In this paper, various technologies for manufacturing simulated CRUD are introduced as alternatives to unattained samples. A metal oxide or metal hydroxide was used to synthesize simulated cruds having chemical compositions and crystalline stricture similar to the actual one by 12 different methods. CRUD 4(metal oxides in the autoclave vessel) and CRUD 10(metal oxides in a crucible after hydrazing pretreatment)were chosen as the best method for Type 1 and Type 2.respectively. As these CRUD can be synthesized easily without using any specialized equipment or reagents in a short time and in large quantities, they are expected to stimulate the development of decontaminating agents and processes.

Chemical Decon,Decontaminating Agent,CRUD,Pressure Vessel,Decontamination

This paper investigated the characteristics of wasteform containing a spent zeolite used as a separating agent of FPs for recycling LiCl waste which would be generated from pyrochemical process of spent PWR fuel. In this study, a conventional borosilicate and Ca-rich glass were used as a consolidating agent for spent zeolite and it';s mixing ratio was changed in the range, 25~35wt%. The leach rates of Cs and Sr had about 0.1~0.01g/m2day and 0.001~0.0001g/m2day, respectively. The leach resistance of Cs increased with amount of SAP and it showed about 10 times higher in the Ca-rich glass wasteform than in the conventional borosilciate glass wasteform. The compressive strength of wasteform was affected with the amount of glass. Thermal expansion rate of containing 30wt% glass has relatively lower than others. Also, the melting temperature was little changed in given mixing ratio of glass.

spent salt,zeolite,silicate-aluminophosphate(SAP),wasteform,leach rate,compressive strenth

In this study, the adsorption reactions in the binary component system such as kaolinite-humic acid, kaolinite-americium and humic acid-americium were investigated. After performing the basic physico-chemical properties of the kaolinite, the adsorption reactions of the humic acid on the kaolinite were carried out with varying concentration of humic acid and ion strength, and pH. With increasing HA concentration and pH, the sorption of HA onto KA decreased, while the sorption of HA onto KA increased with increasing ionic stre ngth. Also, with varying pH, the adsorption reactions of the americium-kaolinite and americium-humic acid were studied. In the acid and neutral region, Am easily adsorbed on the HA, while the sorption of Am on the HA in the alkali region decreased because of electrostatic repulsion. The results from these studies make it possible to understand the characteristics of adsorption behaviour of the americium by the humic acid in the water environment.

Kaolinite,Humic acid,Americium,Adsorption reaction,pH

In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of (Na2CO3-NaHCO3)-H2O2 was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to 50℃ and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re] >1.

Cs,Re,Tc,Selective precipitation,NaTPB,TPPCl,Carbonate,Solubility

A statistical evaluation methodology was developed to determine the compliance of candidate waste stream with clearance criteria based upon distribution of radionuclide in a waste stream at a certain confidence level. For the cases where any information on the radionuclide distribution is not available, the relation between arithmetic mean of radioactivity concentration and its acceptable maximum standard deviation was demonstrated by applying widely-known Markov Inequality and One-side Chebyshev Inequality. The relations between arithmetic mean and its acceptable maximum standard deviation were newly derived for normally or lognormally distributed radionuclide in a waste stream, using probability density function, cumulative density function, and other statistical relations. The evaluation methodology was tested for a representative case at 95% of confidence level and 100 Bq/g of clearance level of radioactivity concentration, and then the acceptable range of standard deviation at a given arithmetic mean was quantitatively shown and compared, by varying the type of radionuclide distribution. Furthermore, it was statistically demonstrated that the allowable range of clearance can be expanded, even at the same confidence level, if information on the radionuclide distribution is available.

Very low-level radioactive waste,Clearance,Statistics,Distribution,Arithmetic mean,Standard deviation

To obtain the input data for the design and long-term performance assessment of a high-level waste repository, the thermal conductivities of several granite rocks which were taken from the rock cores from the declined borehole were measured. The thermal conductivities of granite were measured under the different conditions of water content to investigate the effects of the water content on the thermal conductivity. A simple empirical correlation was proposed to predict the thermal conductivity of granite as a function of effective porosity and water content which can be measured with relative ease while neglecting the possible effects of mineralogy, structure and anisotropy. The correlation could predict the thermal conductivity of granite with the effective porosity below 2.7% from the KURT site with an estimated error below 10%.

thermal conductivity,granite,estimation model,effective porosity,water content