Journal of Nuclear Fuel Cycle and Waste Technology

This study has been focused on determining the chemical composition of 14C - in terms of both organic and inorganic 14C contents - in reactor coolant from 3 different PWR';s reactor type. The purpose was to evaluate the characteristic of 14C that can serve as a basis for reliable estimation of the environmental release at domestic PWR sites. 14C is the most important nuclide in the inventory, since it contributes one of the main dose contributors in future release scenarios. The reason for this is its high mobility in the environment, biological availability and long half-life(5730yr). More recent studies - where a more detailed investigation of organic 14C species believed to be formed in the coolant under reducing conditions have been made - show that the organic compounds not only are limited to hydrocarbons and CO. Possible organic compounds formed including formaldehyde, formic acid and acetic acid, etc. Under oxidizing conditions shows the oxidized carbon forms, possibly mainly carbon dioxide and bicarbonate forms. Measurements of organic and inorganic 14C in various water systems were also performed. The 14C inventory in the reactor water was found to be 3.1 GBq/kg in PWR of which less than 10% was in inorganic form. Generally, the 14C activity in the water was divided equally between the gas- and water- phase. Even though organic 14C compound shows that dominant species during the reactor operation, But during the releasing of 14C from the plant stack, chemical forms of 14C shows the different composition due to the operation conditions such as temperature, pH, volume control tank venting and shut down chemistry.

PWR,Radiocarbon dioxide,14C chemical form,Reactor coolant,Process sample

Transport package for radioactive material should be structurally safe under puncture drop condition and its safety should be verified by test and numerical analysis. Most finite element analyses for puncture drop have been performed without modeling the impact limiter since failure is occurred in the materials of the impact limiter. This paper presents a new modeling methodology, where an element is eroded in case that the material';s failure criteria are reached at the element';s integration point, to investigate the effect of the impact limiter in the puncture process. The effectiveness of the proposed scheme is shown through the puncture drop analysis of hotcell transport cask, which is under design in KAERI. The results show that about 80 percent of the total impact energy is absorbed due to the deformation of impact limiter. Using the present method the puncture drop can be analyzed more accurately, but it would give conservative results compared to the actual test condition.

Puncture drop,Transport cask,Hypothetical accident condition,Impact limiter,Radioactive material

Dihexyloctanamide(DHOA) was used as an extractant or phase modifier with the diamide extractants in a solvent extraction process for a radioactive liquid waste treatment. The degradation compounds of the DHOA extractant, irradiated with 60Co gamma ray, were octanoic acid and dihexylamine which are identified by a Fourier transform infrared(FT-IR) and gas chromatograph/mass spectrometer(GC/MS) analysis, and determined by the GC/MS with selected ion monitoring(SIM) mode. Retention behavior of octanoic acid, tridecane (internal standard) and dihexylamine in total ion chromatogram (TIC) were 8.65 min., 9.79 min., and 10.27 min., respectively. With increasing the absorbed dose of the γ-ray irradiated DHOA, the concentration of octanoic acid was decreased and that of dihexylamine was increased.

DHOA,Radiolysis,Dihexylamine,Octanoic acid,GC/MS,DHOA,방사선 분해,Dihexylamine,Octanoic acid,GC/MS

An analytical method of 99Tc concentration in soil was set up and discussed considering the 99Tc concentration in Korean soil measured with its analytical method. A selective TEVA resin was used to separate and purify the 99Tc in the soil sample. 99mTc from a commercial 99Mo/99mTc generator was used as a yield tracer for the chemical separation of 99Tc and its problem when using 99mTc as a tracer was discussed. The chemical recovery yield of 99Tc was above 70 %. The optimum conditions of inductively coupled plasma mass spectrometry system(ICP-MS) were set up to determine the 99Tc after the separation process. The minimum detectable activity(MDA) was 15 mBq/kg-dry in this analytical procedure. The 99Tc concentration in soils of Jeju and Kori were measured in the rage of 33.73 - 89.16 mBq/kg-dry. Those values were less than those reported in other countries and seemed to be originated from atmospheric fallout.

99Tc analysis,TEVA resin,Yield tracer,ICP-MS,99Tc concentration

An experiment for uranium sorption onto fresh and weathered biotites was performed. After centrifugation, concentrations of uranium in the supernatants were analyzed using ICP-MS, and biotite samples were investigated using XRD and SEM. With powdered biotites (<3 mm in size), we have conducted uranium sorption experiments about fresh and weathered biotites to obtain uranium sorption amounts in various pH conditions. The uranium sorption was not high at a low pH (e.g., pH 3), but increased with increasing pH. There were lower uranium sorption by the weathered biotites than by the fresh ones, and the difference was much larger at higher pH (e.g., pH 11). The lower sorption values of uranium by the weathered biotites may be caused by a change of mineral surfaces and a chemical behavior of surrounding dissolved elements. It seems that the uranium-mineral interaction has been diminished, especially, in the weathered biotite by a destruction and dissolution of preferential sorption sites on the mineral surfaces and by the colloidal formation from dissolved elements.

Biotite,Weathering,Uranium,Sorption,Colloid

The METRO-K is a model for a radiological dose assessment due to a radioactive contamination in the Korean urban environment. The model has been taken part in the Urban Remediation Working Group within the IAEA';s (International Atomic Energy Agency) EMRAS (Environmental Modeling for RAdiation Safety) program. The Working Group designed for the intercomparison of radioactive contamination to be resulted from the explosion of a radiological dispersal device in a hypothetical city. This paper dealt intensively with a part among a lot of predictive results which had been performed in the EMRAS program. The predictive results of three different models (METRO-K, RESRAD-RDD, CPHR) were submitted to the Working Group. The gap of predictive results was due to the difference of mathemathical modeling approaches, parameter values, understanding of assessors. Even if final results (for example, dose rates from contamintaed surfaces which might affect to a receptor) are similar, the understanding on the contribution of contaminated surfaces showed a great difference. Judging from the authors, it is due to the lack of understanding and information on radioactive terrors as well as the social and cultural gaps which assessors have been experienced. Therefore, it can be known that the experience of assessors and their subjective judgements might be important factors to get reliable results. If the acquisition of a little additional information is possible, it was identified that the METRO-K might be a useful tool for decision support against contamination resulting from radioactive terrors by improving the existing model.

EMRAS program,METRO-K,Radiological dispersal device,Hypothetical city

Social Acceptance is important to decide policy of spent nuclear fuel management. The idea of a public sphere as a receptacle of dynamic process is the core in this discussion. The purpose of this study is to examine the concept, participants, the conflict-structure and agreeable conditions of a public sphere. A public sphere means in this paper, mechanism and systems that various stakeholders'; and public';s participation with spontaneous will can affect decision-making process. For good designing and implementing a public sphere, it is necessary to analysis and cope with political, foreign and security, economic, sociocultural environments, the law and systems around spent nuclear fuel management

a policy of spent nuclear fuel management,Pubic sphere,Social conflict,Social Acceptance

Recently the importance and interest for the microbes has been increased because several important results for the effects of microbes on the radioactive waste disposal have been published continuously. In this study, research status and major results on the various roles and effects of microbes in the radioactive waste disposal have been investigated. We investigated and summarized the roles and major results of microbes in a multi-barrier system consisting of an engineered barrier and a natural barrier which is considered in radioactive waste disposal systems. For the engineered barrier, we discussed about the effects of microbes on the corrosion of a waste container and investigated the survival possibility and roles of microbes in a compacted bentonite buffer. For the natural barrier, the roles of microbes present in groundwaters and rocks were discussed and summarized with major results from natural analogue studies. Furthermore, we investigated and summarized the roles and various interactions processes of microbes and their effects on the radionuclide migration and retardation including recent research status. Therefore, it is expected that the effects and roles of microbes on the radioactive waste disposal can be rigorously evaluated if further researches are carried out for a long-term behavior of the disposal system in the deep geological environments and for the effects of microbes on the radionuclide migration through geological media.

Microbes,Roles and Effects of Microbes,Radioactive Waste,Disposal System,Radionuclide Migration and Retardation

The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt

decay heat,Cs Sr,LiCl,pyroprocess,waste salt

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the 60Co nuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the 60Co nuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. In test of activated concrete waste test, the clean aggregate can be recovery by up to 80%. Also, based on the results of contaminated uranium concrete, contaminated concrete volume reduced by up to 75%.

Radioactive concrete. Contamination,Heating,Paste

Remote detecting system for a radiation contamination using a plastic scintillator and an optical fiber was developed. Using a commercially available silica optical fiber and a plastic scintillator, we tested then for a real possibility as a remote monitoring detector. Also, a plastic scintillator was developed by itself, and evaluated as a radiation sensor. The plastic scintillator was made of epoxy resin, a hardener and an organic scintillation material. The mixture rate of the epoxy resin, hardener and organic scintillator was fixed by using their emission spectrum, transmittance, intensity etc. In this study, in order to decrease the light loss of an incomplete connection between an optical fiber and a scintillator, the optical fiber was inserted into the scintillator during the fabrication process. The senor used a plastic optical fiber and was estimated for its detection efficiency by an optic fiber';s geometric factor.

Remote measurement,Plastic scintillator,Epoxy resin,Optical fiber,Radiation sensor

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a (NH4)2CO3-H2O2 carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that H2O2 is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M (NH4)2CO3-0.5 M H2O2 solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98±2 % for Re and Te, 94±2 % for Cs, and 29±2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10∼20 minutes) due to their high solubility in the (NH4)2CO3 solution regardless of the addition of H2O2, and independent of the concentrations of Na2CO3 and H2O2. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the (NH4)2CO3 concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9∼10 in order to minimize the dissolution of FP.

Oxidative-dissolution,Solubility,Carbonate solution. Oxidant,Fission products

Investigation of resaturation and thermal-hydro-mechanical behavior for the buffer of a repository requires measuring the water content of compacted bentonite. This study investigated the relative humidity of compacted bentonites using a humidity sensor (Vaisala HMT 334) applicable under high temperature and pressure, and then conducted a multi-regression analysis based on the measured results to determine relationships among the water content, relative humidity, and temperature. The relationships for the compacted bentonites with the dry densities of 1,500 kg/m3 and 1,600 kg/m3 were expressed as and , respectively. These were then used to interpret the resaturation of bentonite blocks in the KENTEX test.

water content,compacted bentonite,buffer,HLW repository

Absorption spectra for a quantitative analysis of actinide elements such as U(VI) and Pu(V) were measured by using a liquid waveguide capillary cell (LWCC) which has an optical path length of 1.0 meter. In order to investigate radioactive elements, a LWCC is installed in a glove box and is coupled to a spectrophotometer with optical fibers. Limits of detection (LOD) for the system were determined as 0.74 and 0.35 μM with molar absorption coefficients of 8.14±0.07 (414 nm) and 17.00±0.16 (569 nm) M-1cm-1 for U(VI) and Pu(V) ions, respectively. The measured LOD values are about 30 times more sensitive when compared to those achievable by using a conventional quartz cell with an optical path length of 1.0 cm. As an application with an enhanced sensitivity, a quantitative analysis for micromolar concentrations of Pu(V) has been performed to decrease the uncertainty in the formation constant of the Pu(VI)-OH complex.

Uranium(VI),Plutonium(V),Capillary Cell,Absorption Spectroscopy,Speciation

Cryogenic cutting technology is one of the most suitable technologies for dismantling nuclear facilities due to the fact that a secondary waste is not generated during the cutting process. In this paper, the feasibility of cryogenic cutting technology was investigated by using a computer simulation. In the computer simulation, a hybrid method combined with the SPH (smoothed particle hydrodynamics) method and the FE (finite element) method was used. And also, a penetration depth equation, for the design of the cryogenic cutting system, was used and the design variables and operation conditions to cut a 10 mm thickness for steel were determined. Finally, the main components of the cryogenic cutting system were manufactures on the basis of the obtained design variables and operation conditions.

Cryogenic cutting technology,Smoothed particle hydrodynamics,Cryogenic jet penetration depth equation

An electrolytic system (zinc anode-gallium cathode) was setup to evaluate the performance of several stirrers prepared for this study, where stirrers have been used to prevent uranium from forming dendrite on the cathode in pyrochemical process. In the case of no-stirring condition, zinc dendrites began to grow on the gallium surface in 1 hour and some dendrite grew out of the cathode crucible around 6 hours. When a rectangular stirrer or a tilt stirrer was rotated, at 40 ～ 150 rpm, to mix the liquid gallium cathode, dendritic growth of zinc metal was prevented irrespective of revolution speed, but some of the deposits overflowed out of the cathode crucible owing to the large centrifugal forces at 150 rpm. The harrow stirrer did not nearly retard the dendrite growth at 40 rpm, but the dendrite growth was retarded at higher than 100 rpm and the zinc deposits also did not overflow at 150 rpm. Pounder could also prevent the dendrite growth to some extent but it had some difficulties in operation compared with other types of stirrers.

Pyrochemical process,Dendrite,Stirrer,Gallium cathode,Electrolytic deposition

Study on the chemical speciation of uranium(VI) species, UO22+, UO2(OH)+, (UO2)2(OH)22+, (UO2)3(OH)5+, has been performed by using time-resolved laser-induced fluorescence spectroscopy. Speciation sensitivity which depends on the excitation wavelength was investigated. We obtained the speciation sensitivity in the order of 10-9 M concentration of U(VI) compounds at the excitation wavelength of 266 nm. The fluorescence spectrum and lifetime of UO22+ were carefully measured at pH 1 and ion strength of 0.1 M. The spectrum showed the four characteristic peaks located around 488, 509, 533, 559 nm and the fluorescence lifetime of 1.920.17 . The wavelength shifts of fluorescence peaks and the change of lifetimes for uranium hydrolysis compounds were compared with those of UO22+. We report on the characteristic features, the shifts of peaks to the longer wavelength direction and the prolonged lifetimes, in the fluorescence of the U(VI) hydrolysis compounds.

Uranium(VI),Fluorescence,Chemical Speciation,Laser Spectroscopy

Due to a more favorable climate in the recent relationship between U.S. and North Korea, North Korea nuclear issue is expected to enter the new phase of nuclear verification. From now on, our government should make preparation for taking the appropriate steps against the situation developed after the declaration by North Korea. Therefore, this paper is to identify the problems that may be occurred in the process of verifying and dismantling North Korea';s nuclear program and to suggest the policy considerations that should be incorporated in establishing the action plan for verifying and dismantling her nuclear program, based on the analysis of experiences to verify and dismantle the WMDs in the former Soviet Union and in Iraq, respectively.

North Korea Nuclear Issue,Nuclear Verification,Declaration,Weapons of Mass Destruction,Former Soviet Union,Iraq

Vertical electrokintic-flushing remediation equipment was developed for the remediation of a radioactive soil near nuclear facilities. An optimum reagent was selected to decontaminate the radioactive soil near nuclear facilities with the developed vertical electrokintic-flushing remediation equipment, and the optimum remediation conditions were established to obtain a higher remediation efficiency. Namely, acetic acid was selected as an optimum reagent due to its higher remediation efficiency. When the electrokinetic remediation and the electrokinetic-flushing remediation results were compared, the removal efficiency of 4.6 % and the soil waste solution volume of 1.5 times were increased in the electrokinetic remediation. When the potential gradient within an electrokinetic soil cell was increased by two times (4.0 V/cm), the removal efficiencies of Co2+ and Cs+ were increased by about 4.3%(Co2+ : 98.9%, Cs+: 96.7%). Also, when the reagent concentration was increased from 0.01 M to 0.05M, the removal efficiency of Co2+ was increased but that of Cs+ was decreased. Therefore, the optimum remediation conditions were that the acetic concentration was 0.01M～0.05M, the potential gredient was 4 V/cm, the injection of reagent 2.4 ml/g, and the remediation period was 20days.

Vertical electrokinetic-flushing,Cobalt,Cesium,Radioactive soil,Removal efficiency

To evaluate the efficiency of the decontamination plant for the removal of uranium compounds deposited on the internal surface of UF6 cylinder for its reuse, two demonstration tests of the plant with different ratio of Na2CO3 and H2O2 were carried out, and each test had 5 steps. The main chemical form removed by the tests was to be identified as Na4UO2(CO3)3. More than 50% of uranium was removed by water of the first step, and at the following steps the removal amounts were exponentially decreased. On the other hand, the result shows that the injected amount of Na2CO3, compared with that of the removed uranium, was stoichiometrically excessed. This suggests that the injected amounts of Na2CO3, the generation rate of decontaminated waste, and the decontamination steps could be reduced by a process optimization of the plant.

UF6 cylinder,uranium,decontamination,sodium carbonate

A comprehensive analysis has been conducted on the purity of the uranium product generated from a pyroprocessing of EBR-II spent fuel. The analysis results were compared to the low-level waste criteria for both ROK and USA under a collaborativeprogram between INL and KAERI. It is found that the US LLW definition does not include the activity from any U isotopes, but the Korean one does. The analysis results show that Pu-239 is the only alpha emitting isotope other than U isotopes that exceed the limit in the EBR-II U product. Pu contamination of the product seems to be drastically reduced in a preliminary test of the modified cathode process, and the further development of the proposed technology may be possible to meet the US LLW criteria.

Pyroprocessing,EBR-II,purity,LLW criteria

The spent cationic exchange resins and anionic exchange resins were separated from mixed spent exchange resins by a fluidized bed gravimetric separator. The separated resins were identified by an elemental analysis and thermogravimetric analysis. The each test sample was prepared by diluting the slurry made by wet ball milling the cationic exchange resins and the anionic exchange resins separated as a spherical granular form for 24 hours. The resulting test samples showed a slurry form of less than 75 μm of particle size and 25,000 ppm of CODcr. The decomposition conditions of each test samples from a thermal power plant were obtained with a lab-scale(reactor volume : 220 mL) supercritical water oxidation(SCWO) facility. Then pilot plant(reactor volume : 24 L) tests were performed with the test samples from a thermal power plant and a nuclear power plant successively. Based on the optimal decomposition conditions and the operation experiences by lab-scale facility and the pilot plant, a commercial plant(capacity : 150 kg/L) can be installed in a nuclear power plant was designed.

Supercritical Water Oxidation(SCWO),Ion-exchange resin,Slurry,Pilot test,Gravimetric separator

The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries'; regulations related to safety assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies'; approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

HLW repository,Technical standards,Safety assessment,Assessment period,Peak dose,Uncertainties

To characterize the geological features in the study area for high-level radioactive waste disposal research, KAERI (Korea Atomic Energy Research Institute) has been performing several geological investigations such as geophysical surveys and borehole drillings since 1997. Especially, the KURT (KAERI Underground Research Tunnel) constructed to understand the deep geological environments in 2006. Recently, the deep boreholes, which have 500 m depth inside the left research module of the KURT and 1,000 m depth outside the KURT, were drilled to confirm and validate the results from a geological model. The objective of this research was to investigate hydro-geological conditions using a 3-D geological model around the KURT. The geological analysis from the surface and borehole investigations determined four important geological elements including subsurface weathered zone, low-angled fractures zone, fracture zones and bedrock for the geological model. In addition, the geometries of these elements were also calculated for the three-dimensional model. The results from 3-D geological model in this study will be beneficial to understand hydrogeological environment in the study area as an important part of high-level radioactive waste disposal technology.

high-level radioactive waste disposal,KAERI Underground Research Tunnel,geological model,geological elements,hydrogeological environment

In this study, a thermal behavior of PrCl3 as one of the lanthanide chlorides in LiCl-KCl molten salts was investigated in an oxidizing condition. First, a thermo-gravimetric analysis (TGA) of PrCl3 was carried out by an injection of O2 gas. Based on the results, an oxidation of PrCl3 in the molten salts was performed by sparging O2 gas with changing temperatures. According to the TGA data of PrCl3, a dissociation of PrCl3 occurred rapidly by about 380 oC and a conversion of PrCl3 to PrOCl was completed at about 600 oC. The thermal behavior of PrCl3 in LiCl-KCl molten salts by sparging O2 gas was similar to that of PrCl3 in the TGA test, and PrOCl as a insoluble compound in the molten salts was precipitated into the bottom of the molten salts. A conversion of PrCl3 to PrOCl in the molten salts occurred actively at a higher temperature than 650 oC. And it would be possible to estimate a conversion status of PrCl3 to PrOCl by measuring a Cl2 concentration in a flue gas generated from an oxidation test of PrCl3 in LiCl-KCl molten salts.

TGA,PrCl3,Thermal behavior,LiCl-KCl molten salts,Sparging O2 gas

Electrowinning process recovers uranium with actinide elements from spent fuels and is a key step in the Pyroprocessing because of proliferation resistance. An analysis of heat transfer of the Electrowinning cell was conducted to develop basic tool for designing engineering-scale Electrowinner. For the calculation of the heat transfer, ANSYS CFX commercial code was adapted. As a result of the calculation, the vertical Heating Zone length had great effect upon temperature of LiCl-KCl eutectic salt. To maintain constant temperature in the salt, the Heating Zone length should be three times longer than the height of the salt. However, the argon and salt temperatures were barely affected by the Cooling Zone length. The temperature under the Cell cover was mainly influenced by the number of the cooling plates. When the cooling plates were installed more than the number of 5, temperature under the cover was maintained below 250 ℃. These temperature properties had similar tendency toward the temperature of the Cell which was measured from experiments. Simulated heat transfer information from this study could be used to design engineering-scale Electrowinner.

Pyroprocessing,Electrowinning,Heat transfer,LiCl-KCl eutectic salt,Pyroprocessing

The effect of organic acids on the adsorption of U(VI) onto oxide surfaces (TiO2(anatase), SiO2(amorphous) and Al2O3(amorphous)) has been investigated. Two different organic acids, salicylic and picolinic acids, were used. Changes of adsorption ratio of U(VI), which depend on the existence of organic acids in a sample, were measured as a function of pH. Quantities of adsorbed organic acids, which depend on the existence of U(VI) in a sample, were also measured as a function of pH. It is confirmed that the soluble complex formation of U(VI) with organic acids can deteriorate the adsorption of U(VI) onto TiO2 surface. It is noteworthy that salicylic acid does not affect the adsorption of U(VI) onto SiO2 surface, however, picolinic acid enhances the adsorption of U(VI) onto SiO2 surface. The latter effect can be understood by considering the formation of a ternary surface complex on SiO2 surface, which was confirmed by the co-adsorption of picolinic acid with U(VI) and the change in a fluorescence spectra of U(VI) on surface. In the case of Al2O3, organic acids themselves were largely adsorbed onto a surface without deteriorating the adsorption of U(VI). This would support the possibility of a ternary surface complex formation on the Al2O3 surface, and an additional spectroscopic study is required.

Adsorption,Oxide,Uranium(VI),Salicylic acid,Picolinic acid,Ternary surface complex

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over 100 oC have been proposed. These new disposals have made it possible to introduce the concept of long term storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

CANDU spent fuel,Geological disposal,Thermal requirement,Multi barrier,Buffer,Disposal canister,Long term storage

In the pyroprocessing of spent nuclear fuels, fuel materials are recovered by electrochemical reactions on the surface of electrodes as well as stirring the electrolyte in electrolytic cells such as electrorefiner, electroreducer and electrowinner. The system with this equipment should first be scaled-up in order to commercialize the pyroprocessing. So in this study, the scale-up for those equipment was studied to design large-scale equipment which can be employed in a commercial process in the future. Basically the dimensions of both electrolytic cells and electrodes should be enlarged on the basis of the geometrical similarity. Then the criterion of constant power input per unit volume, characterizing the fluid behavior in the cells, was introduced in this study and a calculation process based on trial-and-error methode was derived, which makes it possible to seek a proper speed of agitation in the electrolytic cells. Consequently examples of scale-up for an arbitrary small scale system were shown when the criterion of constant power input per unit volume and another criterion of constant impeller tip speed were respectively applied.

Scale-up of electrolytic reactors,Fluid behavior of electrolyte,Behavior of Newtonian and non-Newtonian fluids,Apparent viscosity and Effective viscosity

In this study, research status on radionuclide and colloid migration in underground research facilities including KURT (KAERI Underground Research Tunnel) was investigated. Some foreign underground research facilities constructed in crystalline rock formations such as granite were briefly introduced and compared. International joint researches concerned with the radionuclide and colloid migration were investigated particularly for the Grimsel Test Site (GTS) and Äspö Hard Rock Laboratory by analyzing major research items, on-going research projects, and future plans.

Underground research facility,International joint research,Radionuclide,Colloid,Migration,High-level radioactive waste,Deep geological disposal

This study examined the effect of an electrokinetic-flushing remediation for a soil of a high permeability. The soil was sampled from the site around a research atomic reactor which had high hydro-conductivities due to a high content of sand in the soil. The flow rate of the washing reagent was fast at the beginning but it was reduced as time lapsed. In the case of using citric acid as a washing reagent, the flow rate was fastest, 78.7 ml/day. The removal efficiencies of Co2+ and Cs+ from a soil cell with acetic acid were the highest, which were 95.2% and 84.2% respectively. The soil waste-solution volume generated from the electrokinetic remediation was reduced to about 1/20 of that from the soil washing remediation. Meanwhile, the electrokinetic-flushing method enhanced the removal efficiencies of Co2+ and Cs+ from the soil by about 6% and 2% respectively, compared to those by the electrokinetic method. Consequently, it was found that the electrokinetic-flushing method was more effective for the remediation of a soil with a high permeability.

Electrokinetic-flushing,Cobalt,Cesium,High permeability soil,Removal efficiency

In order to transport spent resin in a high integrated container made of high density polyethylene, a method of classifying transportation grade by radioactivity analysis was developed. Radioactivity ratios between radioisotopes in spent resin were derived from radioactivity analysis results about spent resin. The curie-to-dose factors were determined to estimate radioisotope inventory from surface dose rates of spent resin. From the results, boundary value between A type and B type carrier with spent resin in high integrated container made of high density polyethylene was derived to be 1.19 TBq, and the corresponding surface dose rate was 124.2 mSv/h.

Spent Resin,Radioisotope Inventory,Transportation Grade,Dose-To-Curie Factor

In this study, plasma processing of metal surface is experimentally investigated to enhance the surface decontamination efficiency and to find out the reaction mechanism. Cobalt, the major contaminant in the nuclear facilities, and three fluorine-containing gases, CF4/O2, SF6/O2, and NF3 are chosen for the investigation. Thin metallic disk specimens are prepared and their surface etching reactions with the three plasma gases are examined. Results show that the maximum etching rate of 17.2㎛/min. is obtained with NF3 gas at 420℃, while with CF4/O2 and SF6/O2 gas plasmas those of 2.56㎛/min. and 1.14㎛/min. are obtained, respectively. Along with etching experiments, constituent elements of the reaction products are identified to be cobalt, oxygen, and fluorine by AES (Auger Electron Spectroscopy) analysis. It turns out that the oxygen atoms are physically adsorbed ones to the surface from the ambient not participation ones during the analysis after reaction, which supports that the surface reaction of cobalt is mainly to be a fluorination reaction.

Metal surface decontamination,cobalt,reactive plasma etching,Auger electron spectroscopy

A selective separation of Actinide(Ⅲ) by a nPr-BTP/nitrobezene extraction system was studied. The nPr-BTP (2.6-Bis-(5.6-n-propyl-1.2.4-triazin-3-yl)-pyridine) of a environmentally -friendly CHN type was self-synthesized and its compatability with diluent and stability with nitric acid were investigated. At the 0.1M nPr-BTP/nitrobenzene-1M HNO3 and O/A=2, extraction yields of Am used as a representative of Actinide(Ⅲ) and Eu were about 85% and 8%, respectively, and the other RE elements such as Nd, Ce and Y were extracted less than 3% (separation factor of Am and Eu was about 60). Thus, there was no problems in the selective extraction of Actinide(Ⅲ) from RE. The stripping yield of Am with 0.05M HNO3 at O/A=1, however, was about 43% and the maximum stripping yield was 65% at O/A=0.3. It is necessary to develop the stripping system including the stripping agent instead of nitric acid solution.

Actinide(Ⅲ),Rare Earth elements,BTP(Bis triazinyl pyridine),Solvent extraction,Selective separation

Long-term degradation of cement barrier by diffusion was studied with reactive transport modeling. The result of modeling showed that cement barrier was altered about 30cm thickness after 50,000 years. The pH decreased from 13.0 to 11.9 because of depletion of alkali ions, and dissolution/precipitation of portlandite and CSH (Calcium Silicate Hydrate). In addition, porosity increased about 0.3 because of dissolution of portlandite and CSH2.0(Ca2SiO3(OH)2:0.17H2O). The solubility of uranium also increased with the increase of pe value. The results of this study indicate that long-term degradation of cemet can enhance the transport of nuclide by changing pH, pe, porosity in barrier.

Cement,Diffusion,Reactive transport,Porosity,Solubility

A new proposed repository has a final capacity of 800,000 drums radioactive waste. Most of foreign repositories have a general practice of segregating control zones which mainly contributes to classification of degree of control, whether it is called buffer zone or not. Domestic regulatory requirements of establishment of buffer zone in a repository are not much different from those of nuclear power plants for operation period, in which satisfactory design objective or performance objective is the most important factor in determination of the buffer zone. The meaning of buffer zone after closure is a minimum requested area which can prevent inadvertant intruders from leading to non-allowable exposure during institutional control period. Safety assessment with drinking well scenario giving rise to the highest probability of exposure among the intruder';s actions can verify fulfillment of the buffer zone which is determined by operational safety of the repository. At present. for the repository to be constructed in a few years, the same procedure and concept as described in this paper are applied that can satisfy regulatory requirements and radiological safety as well. However, the capacity of the repository will be stepwise extended upto 800,000 drums, consequently its layout will be varied too. Timely considerations will be necessary for current boundary of the buffer zone which has been established on the basis of 100,000 drums disposal.

Radioactive waste repository,Low- and intermediate level waste,Buffer zone,Restricted area

A swelling clay should remain physically and chemically stable for a long time to perform its functions as a buffer material of a high-level waste (HLW) repository. The longevity issues in the swelling clay were reviewed to evaluate their importance in the performance of a repository. The review results suggest that an elevated temperature due to decay heat, groundwater chemistry, high pH environment by concrete, organic matter and microbes, radiation, and mechanical disturbance might significantly affect the long-term performance of a swelling clay as a buffer material. This paper will be used as basic informations to design the swelling clay buffer for a HLW repository.

HLW repository,engineered barrier,buffer,swelling clay,long-term performance

The purpose of the HLW deep geological disposal is to isolate and to delay the radioactive material release to human beings and the environment for a long time so that the toxicity does not affect to the environment. The main requirements for the HLW repository design is to keep the buffer temperature below 100 ˚C in order to maintain its integrity. So the cooling time of spent fuels discharged from the nuclear power plant is the key consideration factors for efficiency and economic feasibility of the repository. The disposal tunnel/disposal hole spacing, the disposal area and thermal capacity required for the deep geological repository layout which satisfies the temperature requirement of the disposal system is analyzed to set the optimized spent fuels cooling time. To do this, based on the reference disposal concept, thermal stability analyses of the disposal system have been performed and the derived results have been compared by setting the spent fuels cooling time and the disposal tunnel/disposal hole spacing in various ways. From these results, desirable spent fuels cooling time in view of disposal area is derived. The results shows that the time reaching the maximum temperature within the design limit of the temperature in the disposal site is likely shortened as the cooling time of spent fuels becomes short. Also it seems that the temperature-rising and-dropping patterns in the disposal site are of smoothly varying form as the cooling atime of spent fuels becomes long. In addition, it is revealed that a desirable cooling time of spent fuels is approximately 40-50 years when spent fuels are supposedly disposed in the deep geological disposal site with its structural scale under consideration in this study.

Spent fuel,Deep geological disposal,Repository layout,Buffer material,Thermal stability,Disposal tunnel/disposal pit spacing,Cooling time

For the purpose of evaluating dose rate to individual due to long-term release of nuclides from the HLW repository, a biosphere assessment model and the implemented code, ACBIO, based on BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To show its practicability and usability as well as to see the sensitivity of compartment scheme or parametric variation to concentration and activity in compartments as well as annual flux between compartments at their peak values, some calculations are made and investigated: For each case when changing the structure of compartments and GBIs as well as varying selected input Kd values, all of which seem very important among others, dose rate per nuclide release rate is separately calculated and analyzed. From the maximum dose rates (Bq/y), flux-to-dose conversion factors (Sv/Bq) for each nuclide were derived, which are to be used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rate (Sv/y) for individual in critical group. It has been also observed that compartment scheme, identification of possible exposure group and GBIs could be all highly sensitive to the final consequences in biosphere modeling

Biosphere modeling,AMBER,Flux to Dose Conversion Factor,HLW repository,Safety Assessment

For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to 1.8 Mg/m3 and the bentonite-sand mixture with a dry density of 1.6 and 1.8 Mg/m3 were measured within the gravimetric water content range of 10 wt% to 20 wt% and the sand fraction range of 10 to 30 wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.

bentonite,bentonite-sand mixture,thermal conductivity,buffer material

The new type phoswich detector consisting of the ZnS(Ag) and plastic scintillator for alpha/beta-ray simultaneous counting was developed for monitoring radiological contamination inside pipes. The detection performance was estimated using the PSD (pulse shape discrimination) method as a function of distance between the scintillator and radioactive source. The attenuation of particles traveling through a thin film for preventing the detector from being contaminated was experimentally estimated. It is concluded from our investigation that the phoswich detector developed can provide a sufficient alpha/beta-ray discrimination. The application of a thin film for preventing the detector from being contaminated was proven to be feasible.

Radiological contamination level inside pipe,Phoswich detector,Pulse shape discrimination,Scintillator

AbstracThermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60 m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are 79.9℃ and 119℃, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as 140.9℃. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.

Radioactive waste management,CANDU Spent nuclear fuel,Long-term dry storage,Retrievability,Natural convection,Disposal Canister

In this study, a sorption experiment of multivalent nuclides such as Eu(III) and Th(IV) relatively stable for redox reactions was carried out for bentonite colloids which had been prepared from the domestic Gyeongju bentonite. The amounts of the nuclides lost by an attachment to bottle walls, by a precipitation, and by a colloid formation were estimated by performing blank tests for the sorption experiments. Sorption coefficients, Kd';s, reflecting the mass losses were obtained and investigated for the sorption of Eu(III) and Th(IV) onto the bentonite colloids. The net sorption coefficients Kd';s considering all the three mass losses were measured as about 106-107 mL/g and 7×106-107 mL/g for Eu(III) and Th(IV), respectively, depending on pH. In particular, a precipitation occurred mainly at a pH greater than 5 for Eu(III) and a precipitation and colloid formation significantly occurred at a pH greater than 3 for Th(IV). The precipitation and colloid formation of the multivalent nuclides of Eu(III) and Th(IV) therefore should be considered when Kd';s are rightly obtained over the pH range where their precipitation and colloid formation become significant at a given concentration.

Europium,Thorium,Bentonite Colloids,Precipitation,Colloid Formation,Sorption,Coefficient Kd

Due to the uncertainties resulting from cost projection, evaluation over long term period, and adequacy of applied discount rate, the economic assessment for back-end fuel cycle is different from each organizations or individuals. In this paper, the features and limitations of some noticeable economic evaluations were investigated and analysed to contribute for the public participation and back-end fuel cycle policy related researches. As a result of analysis, we found that the reprocess and recycling is more economical than direct disposal option, but the result includes high uncertainty that depends on the input parameters.

back-end fuel cycle,economic evaluation,levelised unit cost,uncertainty

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

Spent fuel,Transportation,Logistics analysis,CASK,Repository

Inventories to be disposed of, reference burnup, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intenity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.

spent nuclear fuel,source term,geological disposal system,high-level waste,reprocessing,radioactive waste

The reductive stripping of Np using a n-butyraldehyde (NBA) from loaded organic solution containing Np, which was oxidative-extracted in a system of a 30 % TBP/NDD-2M HNO3 and O/A=2 containing 0.005 M K2Cr2O7 as an oxidant of Np, was studied. The stripping yields of Np was increased with an increasing the NBA concentration, with a decreasing the nitric acid concentration of stripping solution and with a decreasing the reaction temperature. The apparent reductive stripping rate equation was shown by the following equation : -d [Np]Org./dt = 1,524 exp(-2,906/T) [NBA]0.91 [H+]-0.92 [Np]Org. At 1.04 M NBA and 2 M HNO3, the stripping yield of Np and U was 70.1 %, and 7.1 %, respectively, and the separation factor of U over Np (=DU/DNp) was about 30.4. Therefore, it was found that U and Np co-extracted in a system of TBP-HNO3 could be effectively mutual-separated by the NBA.

Neptunium,TBP,NBA,Extraction/stripping,Oxidation/Reduction,Kinetics,Np

Metal filter elements were newly introduced to the high temperature filter(HTF) system in the low- and intermediate-level radioactive waste vitrification plant. In order to evaluate the performance of various metal materials as filter media, elements made of AISI 316L, AISI 904L, and Inconel 600 were included to the test set of filter elements. At the visual inspection to the elements performed after completion of each test, a few dark spots were observed on the surface of some elements. Especially they were found much more at the AISI 316L elements than others. To check the dark spots are the corrosion phenomena or not, two kinds of analyses were performed to the tested filter elements. Firstly, the surfaces or the cross sections of filter specimens cut out from both normal area and dark spot area of elements were analyzed by SEM/EDS. The results showed that the dark spots were not evidences of corrosion but the deposition of sodium, sulfur and silica compounds volatilized from waste or molten glass. Secondly, the ring tensile strength were analyzed for the ring-shape filter specimens cut out from each kind of element. The result obtained from the strength tested showed no evidence of corrosion as well. Conclusionally, depending on the two kinds of analysis, no evidences of corrosion were found at the tested metal filter elements. But the dark spots formed on the surface could reduce the effective filtering area and increase the overall pressure drop of HTF system. Thus, continuous heating inside filter housing up to dew point will be required normally. And a few long-period test should be followed for the exact evaluation of corrosion of the metal filter elements.

Metal Filter Element,High Temperature Filter System,Off-gas,Dust

After Canada has struggled with a radioactive waste problem over for 20 years, the Canadian government finally found out that its approach by far has been lack of social acceptance, and needed a program such as public and stakeholder engagement (PSE) which involves the public in decision-making process. Therefore, the government made a special law, called Nuclear Fuel Waste Act (NFWA), to search for an appropriate nuclear waste management approach. NFWA laid out three possible approaches which were already prepared in advance by a nuclear expert group, and required Nuclear Waste Management Organization (NWMO) to be established to report a recommendation as to which of the proposed approaches should be adopted. However, NFWA allowed NWMO to consider additional management approach if the other three were not acceptable enough. Thus, NWMO studied and created a fourth management approach after it had undertaken an comparison of the benefits, risks and costs of each management approach: Adaptive Phased Management. This approach was intended to enable the implementers to accept any technological advancement or changes even in the middle of the implementation of the plan. The Canadian PSE case well shows that technological R&D are deeply connected with social acceptance. Even though the developments and technological advancement are carried out by the scientists and experts, but it is important to collect the public opinion by involving them to the decision-making process in order to achieve objective validity on the R&D programs. Moreover, in an effort to ensure the principles such as fairness, public health and safety, security, and adoptability, NWMO tried to make those abstract ideas more specific and help the public understand the meaning of each concept more in detail. Also, they utilized a variety of communication methods from face-to-face meeting to e-dialogue to encourage people to participate in the program as much as possible. Given the fact that Korea has been also having a hard time in dealing with spent nuclear fuel management, all of these efforts that Canada has made with a PSE program would give good lessons and implications to the Korean case. In conclusion, as a deliberative participation program, PSE could be a possible breakthrough approach for the Korean spent nuclear fuel management.

Public and Stakeholder Engagement,Spent Nuclear Fuel Management,Adaptive Approach,NWMO

The physical and chemical properties of radioactive waste drums, which have been temporarily stored on site, should be characterized before their shipment to a disposal facility in order to prove that the properties meet the acceptance guideline. The investigation of NDT(Nondestructive Test) method was figured out that the contents in drum, the quantitative analysis of free standing water and void fraction can be examined with X-ray NDT techniques. This paper describes the characteristics of X-ray NDT such as its principles, the considerations for selection of X-ray system, etc. And then, the waste drum characteristics such as drum type and dimension, contents in drum, etc. were examined, which are necessary to estimate the optimal X-ray energy for NDT of a drum. The estimation results were that: ① the proper X-ray energy is under 3 MeV to test the drums of 320 ℓ and less; ② both X-ray systems of 450 keV and/or 3 MeV might be needed considering the economical efficiency and the realization. The number of drums that can be tested with 450 keV and 3 MeV X-ray system was figured out as 42,327 and 18,105 drums (based on storage of 2006. 12), respectively. Four testing scenarios were derived considering equipment procurement method, outsourcing or not, etc. The economical and feasibility assessment for the scenarios was resulted in that an optimal scenario is dependent on the acceptance guide line, the waste generator';s policy on the waste treatment and the delivery to a disposal facility, etc. For example, it might be desirable that a waste generator purchases two 450 keV mobile system to examine the drums containing low density waste, and that outsourcing examination for the high density drums, if all NDT items such as quantitative analysis for ';free standing water'; and ';void fraction';, and confirmation of contents in drum have to be characterized. However, one 450 keV mobile system seems to be required to test only the contents in 13,000 drums per year.

radwaste disposal,nondestructive test,radwaste acceptance guideline,X-ray CT

SITES(Site Information and Total Environmental Data Management System) has been developed for the purpose of systematically managing site characteristics and environmental data produced during the pre-operational, operational, and post-closure phases of a radioactive waste disposal facility. SITES is an integration system, which consists of 4 modules, to be available for maintenance of site characteristics data, for safety assessment, and for site/environment monitoring; site environmental data management module(SECURE), integrated safety assessment module(SAINT), site/environment monitoring module(SUDAL) and geological information module for geological data management(SITES-GIS). Each module has its database with the functions of browsing, storing, and reporting data and information. Data from SECURE and SUDAL are interconnected to be utilized as inputs to SAINT. SAINT has the functions that multi-user can access simultaneously via client-server system, and the safety assessment results can be managed with its embedded Quality Assurance feature. Comparison between assessment results and environmental monitoring data can be made and visualized in SUDAL and SITES-GIS. Also, SUDAL is designed that the periodic monitoring data and information could be opened to the public via internet homepage. SITES has applied to the Wolsong low- and intermediate-level radioactive waste disposal center in Korea, and is expected to enhance the function of site/environment monitoring in other nuclear-related facilities and also in industrial facilities handling hazardous materials.

Radioactive Waste Disposal Facility,Environment Monitoring,Safety Assessment,SITES,SAINT,SUDAL,방사성폐기물 처분시설,환경감시,안전성평가,SITES,SAINT,SUDAL

In deep geological disposal system, the integrity of a disposal canister having spent fuels is very important factor to assure the safety of the repository system. This disposal canister is one element of the engineered barriers to isolate and to delay the radioactivity release from human beings and the environment for a long time so that the toxicity does not affect the environment. The main requirement in designing the deep geological disposal system is to keep the buffer temperature below 100 ˚C by the decay heat from the spent fuels in the canister in order to maintain the integrity of the buffer material. Also, the disposal canister can endure the hydraulic pressure in the depth of 500 m and the swelling pressure of the bentonite as a buffer. In this study, new concept of the disposal canister for the CANDU spent fuels which were considered to be disposed without any treatment was developed and the thermal stability and the structural integrity of the canister were analysed. The result of the thermal analysis showed that the temperature of the buffer was 88.9 ˚C when 37 years have passed after emplacement of the canister and the spacings of the disposal tunnel and the deposition holes were 40 m and 3 m, respectively. In the case of structural analysis, the result showed that the safety factors of the normal and the extreme environment were 2.9 and 1.33, respectively. So, these results reveal that the canister meets the thermal and the structural requirements in the deep geological disposal system.

CANDU spent fuel,Deep geological disposal,Engineered barrier,Disposal canister,Buffer material,Thermal-structural stability

An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight, 235U enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial 235U enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16 16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16 16 Korean Standard Fuel Assembly with a cross section of 21.4 cm 21.4 cm, a length of 453 cm, a mass of 672 kg, the initial 235U enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU

spent nuclear fuel,discharge burnup,pyroprocessing facility,design

A conceptual design study for a pyroprocesing facility, has been carried out in this study, which is available for the recovery of uranium and transuranic elemental group(TRU), that is, reusable as a nuclear fuel especially in a next generation-type reactor. The scale of this facility has been chosen as 20 kg HM/batch, comparatively small engineering size in order to collect scale-up data for the design of a commercial facility as well as to get operational experience. The spent fuel to be handled in this process is as follows : 3.5 % enriched uranium fuel, 35,000MWd/tU and 5-year cooled. The major items considered in the conceptual study are a building lay-out including various hot cells, safety management of the process operation in conjunction with material balance control, radiation safety, inert atmosphere control in shield hot cells, and criticality control of uranium and TRU products.

Pyroprocessing of spent fuels,Nuclear fuel cycle based on dry process,Conceptual design study of pyroprocessing facility

Korea Hydro and Nuclear Power Company (KHNP) conducted site investigations for a low and intermediate-level nuclear waste repository in the Gyeong Ju site. The site characterization work constitutes a description of the site, its regional setting and the current state of the geosphere and biosphere. The main objectives of hydogeological investigation aimed to understand the hydrogeological setting and conditions of the site, and to provide the input parameters for safety evaluation. The hydogeological characterization of the site was performed from the results of surface based investigations, i.e geological mapping and analysis, drilling works and hydraulic testing, and geophysical survey and interpretation. The hydro-structural model based on the hydrogeological characterization consists of one-Hydraulic Soil Domain, three-Hydraulic Rock Domains and five-Hydraulic Conductor Domains. The hydrogeological framework and the hydraulic values provided for each hydraulic unit over a relevant scale were used as the baseline for the conceptualization and interpretation of flow modeling. The current hydrogeological characteristics based on the surface based investigation include some uncertainties resulted from the basic assumption of investigation methods and field data. Therefore, the reassessment of hydrostructure model and hydraulic properties based on the field data obtained during the construction is necessitated for a final hydrogeological characterization.

LILW disposal site,Hydrogeological characteristics,hydro-structural model,hydraulic parameter

We analyzed the statistical properties for the conductive background fractures in the Low and Intermediate Level Waste (LILW) disposal site to conceptualize of its groundwater flow system. The background fractures were classified to fracture sets based on their trends and plunges that were obtained from the borehole logging data, and then the fracture transmissivity distribution was inferred from the fixed interval hydraulic test results. The fracture size distribution of each fracture set was estimated using the fracture density and fracture mapping data. To verify the analyzed results, we compared observed field data to simulated one from the DFN model that was constructed with the analyzed statistical properties of the background fractures, and they showed a good agreement.

LILW disposal site,background fracture,statistical property

Based on the site characterization works in a low and intermediate level waste (LILW) repository site, the numerical simulations for groundwater flow were carried out in order to understand the groundwater flow system of repository site. To accomplish the groundwater flow modeling in the repository site, the discrete fracture network (DFN) model was constructed using the characteristics of fracture zones and background fractures. At result, the total 10 different hydraulic conductivity(K) fields were obtained from DFN model stochastically and K distributions of constructed mesh were inputted into the 10 cases of groundwater flow simulations in FEFLOW. From the total 10 numerical simulation results, the simulated groundwater levels were strongly governed by topography and the groundwater fluxes were governed by locally existed high permeable fracture zones in repository depth. Especially, the groundwater table was predicted to have several tens meters below the groundwater table compared with the undisturbed condition around disposal silo after construction of underground facilities. After closure of disposal facilities, the groundwater level would be almost recovered within 1 year and have a tendency to keep a steady state of groundwater level in 2 year.

LILW repository site,Groundwater flow modeling,Hybrid-EPM approach

The numerical simulations for groundwater flow were carried out to support the input parameters for safety assessment in LILW repository site. As the input parameters for safety assessment, the groundwater flux into the underground facilities during construction, flow rate through the disposal silo after closure of disposal silo and flow pathway from the disposal silo to discharge area were analyzed using the 10 cases groundwater flow simulations. From the total 10 numerical simulation results, the statistics of estimated output were similar to among 10 cases. In some cases, the analyzed input parameters were strongly governed by locally existed high permeable fracture zone at radioactive waste disposed depth. Indeed, numerical simulation for well scenario as a human intrusion scenario was carried out using the hydraulically severe case model. Using the results of well scenario, the input parameters for safety assessment were also obtained through the numerical simulation.

LILW repository site,groundwater flow modeling,safety assessment

This study was carried out to identify the characteristics of hydrochemistry controlling groundwater chemical condition in a repository site of Gyeongju. For this study, 12 bore holes of all monitoring bore holes in the study area were selected and total 46 groundwater samples were collected with depth. In addition, 3 surfacewater samples and 1 seawater sample were collected. For water samples, cations and anions were analyzed. The environmental isotopes(δ18O-δD, Tritium, δ13C, δ34S) were also analyzed to trace the origin of water and solutes. The result of δ18O and δD analysis showed that surface water and groundwater were originated from precipitation. Tritium concentrations of groundwater decreased with depth but high concentrations of tritium indicated that groundwater was recharged recently. The results of ion and correlation analysis showed that groundwater types of the study area were represented by Ca-Na-HCO3 and Na-Cl-SO4, which was caused by sea spray and water-rock interaction. Especially, high ratio of Na content in groundwater resulted from ion exchange. For redox condition of groundwater, the values of DO and Eh decreased with depth, which indicated that reducing condition was formed in deeper groundwater. In addtion, high concentration of Fe and Mn showed that redox condition of groundwater was controlled by the reduction of Fe and Mn oxides.

repository,hydrochemistry,ion exchange,redox condition

Geochemical study on the rocks and minerals of the Gyeongju low and intermediate level waste repository was carried out in order to provide geochemical data for the safety assessment and geochemical modeling. Polarized microscopy, X-ray diffraction method, chemical analysis for the major and trace elements, scanning electron microscopy (SEM), and stable isotope analysis were applied. Fracture zones are locally developed with various degrees of alteration in the study area. The study area is mainly composed of granodiorite and diorite and their relation is gradational in the field. However, they could be easily distinguished by their chemical property. The granodiorite showed higher SiO2 content and lower MgO and Fe2O3 contents than the diorite. Variation trends of the major elements of the granodiorite and diorite were plotted on the same line according to the increase of SiO2 content suggesting that they were differentiated from the same magma. Spatial distribution of the various elements showed that the diorite region had lower SiO2, Al2O3, Na2O and K2O contents, and higher CaO, Fe2O3 contents than the granodiorite region. Especially, because the differences in the CaO and Na2O distribution were most distinct and their trends were reciprocal, the chemical variation of the plagioclase of the granitic rocks was the main parameter of the chemical variation of the host rocks in the study area. Identified fracture-filling minerals from the drill core were montmorillonite, zeolite minerals, chlorite, illite, calcite and pyrite. Especially pyrite and laumontite, which are known as indicating minerals of hydrothermal alteration, were widely distributed in the study area indicating that the study area was affected by mineralization and/or hydrothermal alteration. Sulfur isotope analysis for the pyrite and oxygen-hydrogen stable isotope analysis for the clay minerals indicated that they were originated from the magma. Therefore, it is considered that the fracture-filling minerals from the study area were affected by the hydrothermal solution as well as the simply water-rock interaction.

LILW repository,Gyeongju,Lithology,Fracture-filling mineral,Geochemistry

Post-closure safety assessment for the Wolsong Low- and Intermediate-level radioactive waste Disposal Center is described. Based on assessment context, closure concept and ground water flow characteristics of the disposal site, brief descriptions are included on the assessment scenarios, models, input parameters and tools. Radionuclide transport modeling in the near-field and far-field, gas generation and transport modeling, human intrusion and biosphere transport are also described briefly. Assessment results for each scenarios are shown to meet the performance criteria of regulatory body. Further and continuous efforts to improve the safety of disposal facility will be made during the construction and operational period.

Low- and Intermediate-level Radioactive Waste (LILW),Wolsong LILW Disposal Center,Post-closure Safety Assessment

The purpose of Radioactive Waste Acceptance Criteria(WAC) is to verify a radioactive waste compliance with radioactive disposal facility requirements in order to maintain a disposal facility';s performance objectives and to ensure its safety. To develop WAC which is conformable with domestic disposal site conditions, we furthermore analysed the WAC of foreign disposal sites similar to the Kyung-Ju disposal site and the characteristics of various wastes which are being generated from Korea nuclear facilities. Radioactive WAC was developed in the technical cooperation with the Korea Atomic Energy Research Institute in consideration of characteristics of the wastes which are being generated from various facilities, waste generators'; opinions and other conditions. The established criteria was also discussed and verified at an advisory committee which was comprised of some experts from universities, institutes and the industry. So radioactive WAC was developed to accept all wastes which are being generated from various nuclear facilities as much as possible, ensuring the safety of a disposal facility. But this developed waste acceptance criteria is not a criteria to accept all the present wastes generated from various nuclear facilities, so waste generators must seek an alternative treatment method for wastes which were not worth disposing of, and then they must treat the wastes more to be acceptable at a disposal site. The radioactive disposal facility WAC will continuously complement certain criteria related to a disposal concentration limit for individual radionuclide in order to ensure a long-term safety.

radioactive waste,disposal,waste acceptance criteria,radwaste characteristics,waste form characterization

On July 31, 2008, the Government issued the construction and operation permit for the first low and intermediate level radioactive waste disposal facility in the Republic of Korea. In this paper, the fundamental regulatory framework, regulatory requirements and technical standards of the disposal facility are introduced, and the phased review process adopted for evaluation of the safety of the facility is briefly described. The Atomic Energy Act sets forth a stepwise regulatory framework for the whole life-cycle of the disposal facility such as siting, design, construction, operation, closure and institutional control. More detailed regulatory requirements and technical standards are stipulated in the subsequent regulations of the Atomic Energy Act and a series of Notices issued by the Ministry of Eduction, Science and Technology. The Korea Institute of Nuclear Safety, as entrusted by the Ministry under the Atomic Energy Act, conducted safety review on the disposal facility, and evaluated the compliance with relevant criteria in all technical elements (i.e. siting and structural safety, radiological environmental impact, operational safety, systems and components, quality assurance, and total systematic performance assessment, etc.). The overall safety review process can be phased into inception phase, initial review phase, main review phase and completion phase. The review results were reported to and deliberated by the five Sub-committees of the Special Committee on Nuclear Safety, and then reported to the Ministry. The Ministry issued the construction and operation permit of the disposal facility through the deliberation of the review results by the Nuclear Safety Commission. Hereafter, the safety of the repository will be reassured by a series of subsequent regulatory inspections and reviews under the Atomic Energy Act. In addition, the licensee';s continuous implementation of the "Safety Promotion Plan" may also enhance the long-term safety of the repository and contribute to build-up the confidence of the safety case.

Radioactive Waste,Disposal,Safety Review,Technical Standards

The purpose of site monitoring and investigation is to offer the basic data for performance assessment and design of low- and intermediate-level radioactive waste(LILW) disposal facility by monitoring variations of main site properties continually in the stage of pre-operation, operation and post-closure. Main contents of site monitoring are as follows. In the stage of pre-operation, suitability evaluation for disposal facility and monitoring for constructing and operating disposal facility are performed. In the operation period, monitoring is performed including surroundings to research the influence to environment with operating disposal facility and operate safely and efficiently. In the post-closure period, monitoring about major site properties is performed to prevent the effect of radioactive waste from disposal facility and to secure long-term safety.

Site mornitoring,Disposal facility,Long-term safety

To obtain base-line data in the low- and intermediate-level radioactive waste disposal facility, radiological environment investigation is required at least two years prior to the facility operation near the disposal facility. The investigation has been performed since January 2007 for the Wolsong LILW disposal center. In this paper, investigation procedure and detailed information obtained for six months in 2008 are described. Based on the current results, future investigation planning is also discussed for the radiological environment management of the disposal facility.

Radioactive waste disposal facility,Radiological environment investigation,Environmental radiation monitor,Portable radiation survey meter,Thermo luminance dosimeter,Radiation,radioactivity

Electrorefining experiments were successfully carried out in LiCl-KCl eutectic molten salt with a graphite cathode. It was found that the formation of Uranium-Graphite intercalation compound(UGIC) helped the self-scraping mechanism of the uranium dendrite and the efficiency of the electrorefiner increased due to an elimination of the stripping step. The contaminations of the uranium deposit by rare earth elements was negligible while about 300 ppm of carbon was observed. The carbon contamination is believed to be eliminated by further purification by yttrium reaction. The morphology characteristics of the recovered U deposit was compared to that of steel cathode. These are only qualitative preliminary experimental results, but we believe that further research on this type of activity change the direction of the electrorefining research on spent nuclear fuel.

electrorefining,graphite cathode,intercalation,deposit

In order to compare the correlation of radioactivity ratio between the radwaste streams and the primary coolant of PWR NPPs, A RCS sampling kit was installed to primary coolant system for the collection of the radionuclides during the normal operation of NPPs. RCS samples were collected from PWR type of domestic NPPs through 2004 to 2005, and pretreated with acid microwave digestion or leaching method to assay quantitatively of several interesting radionuclides. The radioactivity ratios of 137Cs to 60Co in a filter cartridge and a resin cartridge were 2.3E-2 and 7.3E-1, respectively. At a same period of the reactor operating cycle, the radioactivity ratios of 137Cs to 60Co were 6.3E-1 for a evaporator bottom, 6.7E-1 for a spent resin, and 5.6E-2 for a dry active waste, so that these radwaste streams were identified as having similar characteristics with the corresponding RCS samples.

RCS Coolant,60Co,137Cs,Radionuclide Assay

Combined method of extraction chromatography and liquid scintillation analysis was established for determinating 55Fe and 63Ni radioactivity in solid samples. Activated concrete wastes generated from dismantling KRR-2 were analysed. The sequential separation including precipitation and extraction chromatography resulted in the above 90% chemical recoveries of Fe. Above 60% recoveries of Ni were obtained by this procedure exception to 43.6 and 46.5% recoveries. The seperation and counting procedure was also confirmed with spiked samples of known quantity. The measured and spiked quantity were agreed with the 3.7% and 0.7% variations in the 55Fe and 63Ni experiments, respectively. The radioactivities of 55Fe in the dismantled concretes are shown from below MDA to maximum 361 Bq/g. The radioactivities of 63Ni in all concrete samples are below MDA. The 63Ni doesn';t exist in dismantled concretes from KRR-2. The radioactivity of 55Fe is decreased rapidly as the sampling depth is increased from the concrete surface.

55Fe,63Ni,dismantled concrete,extraction chromatography,LSC

A conventional single compartment model cannot simulate reasonably the migration phenomenon of contaminants through unsaturated zone, due to the intrinsic unrealistic assumption of the compartment model that contaminants entering a compartment are immediately and uniformly mixed. Although, a multi-compartments model, in which even physically identical layer is divided into multiple compartments, may be used for explaining the retardation of contaminant mass flux along with increasing number of compartments, its numerical modeling is usually time-consuming and appropriate analytical solutions have not been reported yet. In order to improve the conventional compartment models on contaminant migration through unsaturated zone, a series of analytical solutions for multi-compartments model were derived and a generalized constraint under which the results from multi-compartments model can be simply approximated by single compartment model was proposed. The simplified approximation method was verified by a simple numerical analysis on the constraint under hypothetical conditions. It was also proved that the influent contaminant transfer rate from the bulk unsaturated zone can be generally represented into a time-dependent nominal transfer rate rather than a constant. In addition, the nominal transfer rate turned out to be very sensitive to the contaminant transfer rate between compartments in unsaturated zone, but to be almost insensitive to the transfer rate from contaminated zone. It is expected that the simplified approximation method developed in this study can be used for rapid and reasonable estimation of the migration phenomenon of contaminant through unsaturated zone, instead of time-consuming multi-compartments modeling.

Multi‐Compartments Model,Unsaturated Zone,Simplification,Approximation

Quantitative analysis on release behavior of the 85Kr and 14C fission gases from the spent fuel material during the voloxidation and OREOX process has been performed. This thermal treatment step in a remote fabrication process to fabricate the dry-processed fuel from spent fuel has been used to obtain a fine powder The fractional release percent of fission gases from spent fuel materials with burn-up ranges from 27,000 MWd/tU to 65,000 MWd/tU have been evaluated by comparing the measured data with these initial inventories calculated by ORIGEN code. The release characteristics of 85Kr and 14C fission gases during the voloxidation process at 500 ℃ seem to be closely linked to the degree of conversion efficiency of UO2 to U3O8 powder, and it is thus interpreted that the release from grain-boundary would be dominated during this step. The high release fraction of the fission gas from an oxidized powder during the OREOX process would be due to increase both in the gas diffusion at a temperature of 700 ℃ in a reduction step and in U atom mobility by the reduction. Therefore, it is believed that the fission gases release inventories in the OREOX step come from the inter-grain and intra-grain on UO2 matrix. It is shown that the release fraction of 85Kr and 14C fission gases during the voloxidation step would be increased as fuel burn-up increases, ranging from 6 to 12%, and a residual fission gas would completely be removed during the OREOX step. It seems that more effective treatment conditions for a removal of volatile fission gas are of powder formation by the oxidation in advance than the reduction of spent fuel at the higher temperature.

DUPIC,Spent fuel. Voloxidation,OREOX,Oxidized powder,Fractional release percent,85Kr,14C

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to 1,500 ℃ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% H2/Ar environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about 800℃ ~ 1,200℃, but that of green pellet under the reduction atmosphere was 1,100 ℃ ~ 1,400 ℃. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.

DUPIC,Spent fuel. Green pellet,Voloxidation,Sintering,Semi-volatile fission product,Cesium,Fractional release percent

The pollutants (Rn, CH, CO, HS, radioactive gas from radiolysis) were generated from the process of construction and operation of underground repository, and after disposal of low-intermediate radioactive waste inside there must be controlled by a ventilation system to distribute them in area where enough air is supported. Therefore, a suitable technical approach is needed especially at an underground repository that is equipped with many entry tunnels, storage tunnels, exhaust-blowing tunnels, and vertical shafts in complicated network form. For the technical approach of such a ventilation system, WIPP (Waste Isolation Pilot Plant) in U. S and SFR (Slutforvar for Reaktoravfall) low-intermediate radioactive waste repository in Sweden were selected as the models, for calculating the required air quantity, organizing a ventilation network considering cross section, length, surface roughness of the air passage, and describing a calculation of resistance of each circuit. Based on these procedures, a best suited ventilation system was completed with designing proper capacity of fans and operating plan of vertical shafts. As a result of comparing the two repositories based on the geometry dimensions and ventilation facility equipment operation, more parallel circuit as in WIPP, brought decrease in resistance for entire system leading to reduce of operating costs, and the larger cross-sectional area of the SFR, the greater the percentage of disposal capacity. Accordingly, the mixture of parallel circuit of WIPP repository for reducing resistance and SFR repository formation for enlargement of disposal capacity would be the most rational and efficient ventilation system.

Low- Intermediate radioactive waste repository,ventilation system,WIPP,SFR,ventilation network

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by its FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

Soluble Nuclides,Used fuel,Gap,Grain boundary,Inventory

Fenton‘s Reagent is applied to directly dissolve the cation-exchange resin, IRN-77. The characteristics of the experimental procedure is to dry the resin first and FeSO4 solution is completely absorbed into the resin, and then H2O2 is introduced later for an effective reaction between the reagents within the resin. As a characteristic of the dissolution, the lag time is needed for about 1 hour until the main reaction is occurred, which was more affected with the less concentration of FeSO4 and the less initial dose of H2O2. The dose of H2O2 was equally divided into the early stage and after the initial reaction to provide an effective and safe reaction condition. The optimum conditions is appeared that the concentration of FeSO4 is 0.0009 M and the dose of 15 % H2O2 solution is 6-7 volume for the dissolution of unit weight of IRN-77. The effect of the heating on the lag time was checked and the time could be reduced within 5 minutes at 50 oC, which is a relatively low temperature. The large amount of 5 g and 10 g of the resin was also completely decomposed by increasing the dose of H2O2 to 9-10 volume ratio.

Ion-exchange resin,IRN-77,Dissolution,Fenton's reagent,FeSO4,H2O2

It was intended in this study to investigate the effects of various parameters on the chemical reaction or mass transfer yield in a tubular-type nuclear waste treatment equipment. Since such equipments, as a tubular reactor, multistage solvent extractor, and adsorption column, accompany chemical reaction or mass transfer along the fluid-flowing direction, mathematical modeling for each equipment was carried out first. Then their behaviors of the chemical reaction or mass transfer were predicted through computer simulations. The inherent major parameters for each equipment were chosen and their sensitivities affecting the reaction or mass transfer yield were analyzed. For the tubular reactor, the effects of axial diffusion coefficient and reaction rate constant on the reaction yield were investigated. As for the multistage solvent extractor, the backmixing of continuous phase and the distribution coefficient between fluid and solvent were considered as the major parameters affecting the extraction yield as well as concentration profiles throughout the axial direction of the extractor. For the adsorption column, the equilibrium constant between fluid and adsorbent surface, and the overall mass transfer coefficient between the two phases were taken as the major factors that affect the adsorption rate.

tubular reactor,stagewise solvent extractor,adsorption column

In this study, domestic regulatory requirement was investigated for self-disposal of concrete waste from nuclear fuel processing facility. And after self-disposal as landfill or recycling/reuse, the exposure dose was evaluated by RESRAD Ver. 6.3 and RESRAD BUILD Ver.3.3 computing code for radiological assessments of the general public. Derived clearance level by the result of assessments for the exposure dose of the general public is 0.1071Bq/g (3.5% enriched uranium) for landfill and 0.05515 Bq/ (5% enriched uranium) for recycling/reuse respectively. Also, residual radioactivity of concrete waste after decontamination was investigated in this study. The result of surface activity is 0.01Bq/ for -emitter and the result of radionuclide analysis for taken concrete samples from surface of concrete waste is 0.0297Bq/g for concentration of 238U, below 2w/o for enrichment of 235U and 0.0089Bq/g for artificial contamination of 238U respectively. Therefore, radiological hazard of concrete waste by self-disposal as landfill and recycling/reuse is below clearance level to comply with clearance criterion provided for Notice No. 2001-30 of the MOST and Korea Atomic Energy Act.

Clearance level,Concrete waste,Uranium,Residual activity,Exposure dose

The absorption spectra of Pu in stock solution were measured using a UV-Vis-NIR spectrophotometer after dissolving PuO2 with HNO3, HF and HClO4. The spectroscopic characteristics of Pu (III, IV, VI) in acidic, neutral and alkali media were investigated. Also, the intensities and position of major peaks for Pu(VI) were observed with increasing acidic and alkali concentration. The variation of oxidation states of Pu(VI) with an adding reducing reactant was investigated in HCl and NaOH medium.

Pu oxidation state,spectroscopic characteristics,acid/alkali medium

The possibility and limitation of one dimensional nano scale electron shielder is briefly discussed. A Nano scale electron shielder will reduce the weight and size of shielding materials. However, practical application still requires further research. In this work, we discuss general problems related to nano scale electron shielder, by taking an arbitrary one dimensional potential barrier as an example

nano scale electron shielder,quantum transmission,quantum reflection

In this study, uranium migration experiments have been performed using a natural groundwater and a granite core with natural fractures in a glove-box constructed to simulate an appropriate subsurface environment. Groundwater flow experiments using the non-sorbing anionic tracer Br- were carried out to analyze the flow properties of groundwater through the fracture of the granite core. The result of the uranium migration experiment showed a breakthrough curve similar to that of the non-sorbing Br-. This result may imply that uranium migrates as anionic complexes through the rock fracture since uranium can form carbonate complexes at a given groundwater condition. The distribution coefficient of the uranium between the groundwater and the fracture filling material was obtained as low as 2.7 mL/g from a batch sorption experiment. This result agrees well with the result from the migration experiment, showing a faster elution of the uranium through the rock fracture. In order to analyze retardation properties of the uranium through the rock fracture, the retardation factor (~16.2) was obtained by using the obtained from the batch sorption experiment and it was compared with the (~14.3) obtained by using the result from the uranium migration experiment. The values obtained from the both experiments were very similar to each other. This reveals that the retardation of the uranium is mainly occurred by the fracture filling material when the uranium migrates through the fracture of a granite core.

Uranium migration,Glove-box,Rock fracture,Hydraulic properties,Retardation factor,Distribution coefficient

This study was performed to evaluate the co- and sequential separation of Tc, Np and U from the simulated multi-component HLW solution by a TBP (tributyl phosphate)-TOA (tri- octyl amine)/NDD (n-dodecane)-HNO3 extraction system. An optimal condition of (30 % TBP- 0.5 % TOA)/NDD-1 M HNO3 was selected by taking account of a prevention of the 3rd phase and effects of concentration of TBP, TOA and nitric acid on the co-extraction of Tc, Np and U. In that condition, the extraction yields were 81 % (Tc), 85 % (Np), less than 9 % (Am and RE elements), about 8 % (Pd), and less than 5 % (other elements) so that the system developed for the co-extraction of Tc, Np and U was proved to be available. For that, however, more than 99 % of Zr was found to be pre-removed. The co-extracted Tc, Np and U were sequentially separated in order of Tc (stripping agent : 5 M HNO3) → Np by reductive stripping (reductive-stripping agent : 0.1 M AHA) → U (stripping agent : 0.01 M HNO3), and then their separation factors were evaluated. At these conditions, 95 % of Tc, 98 %of Np and 99 % of U could be recovered in each step.

Tc,Np,U,TBP,TOA,Solvent extraction,REDOX,Selective stripping

This study was performed to evaluate the co- and mutual separation for Am, Cm and RE elements from the simulated multi-component solution equivalent to real HLW level by a Zr-DEHPA(di-(2-ethylhexyl) phosphoric acid containing Zirconium)/NDD(n-dodecane)-HNO3 extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M HNO3 was selected taking into consideration of prevention of the third phase, and effects of concentration of DEHPA, nitric acid and impregnant amount of Zr on the co-extraction of Am, Cm and RE. In that condition, the extraction yields were 81% (Am), 85% (Cm), more than 80% (RE elements), 98% (Mo), 85% (Fe), 98% (U), 73% (Np), and less than 5% (other elements) so that the system developed for the co-extraction of Am-Cm/RE was proved to be available. For that, however, U, Np, Mo and Fe was elucidated to have to be removed in advance, and Zr inducing the third phase formation was found to be practically excluded. The co-extracted Am-Cm/RE were sequentially separated in an order of Am-Cm (stripping agent : 0.05 M DTPA-1M Lactic acid of pH 3.6) → RE (stripping agent : 5M HNO3), and then their separation factors were evaluated. At above conditions, Am of 65.4%, Cm of 63.9%, RE (except for Y) of more than 85% were stripped.

Am,Cm,Rare Earth elements,Zr-DEHPA,Co-extraction,Mutual separation

In hydrothermal reaction tests, smectite-to-illite conversion was identified using a domestic bentonite which is favorably considered as a buffer material, and its dependency on various hydrothermal conditions was investigated. The analysis results of the XRD and Si concentration indicated that the smectite-to-illite conversion was a major process of bentonite alteration under the hydrothermal conditions. The temperature, potassium concentration in solution, and pH were observed to significantly affect the smectite-to illite conversion. A model of conversion reaction rate was suggested to evaluate the long-term stability of smectite composing a major constituent of bentonitic buffer. It was expected from the evaluation results that the smectite would keep its integrity for very long disposal time under a normal condition, whileas it might be converted to illite by 50 percent after over 5x104 year of disposal time under a conservative condition and consequently lose its swelling capacity as a buffer material of a repository.

bentonite,smectite,hydrothermal behavior,long-term stability

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analysis were performed for the container in a view of Type B. As a result of structural analyses, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at 134.8 ℃ at room temperature. In 800 ℃ fire test, The thermal barrier of container sustained the inner vessel at 405 ℃ after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was 550 ℃. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used

tritium,container,transport,safety evaluation

This study suggested a new method to stabilize molten salt wastes generated from the pyro-process for the spent fuel treatment. Using conventional sol-gel process, SiO2-Al2O3-P2O5 (SAP) inorganic material that is reactive to metal chlorides were prepared. In this paper, the reactivity of SAP with the metal chlorides at 650~850, the thermal stability of reaction products and their leach-resistance under the PCT-A test method were investigated. Alkali metal chlorides were converted into metal aluminosilicate(LixAlxSi1-xO2-x) and metal phosphate(Li3PO4 and Cs2AlP3O10) while alkali earth and rare earth chlorides were changed into only metal phosphates (Sr5(PO4)3Cl and CePO4). The conversion rate was about 96~99% at a salt waste/SAP weight ratio of 0.5 and a weight loss up to 1100℃ measured by thermogravimetric analysis were below 1wt%. The leach rates of Cs and Sr under the PCT-A test condition were about 10-2g/m2・day and 10-4g/m2・day. From these results, it could be concluded that SAP can be considered as an effective stabilizer for metal chlorides and the method using SAP will give a chance to reduce the volume of salt wasteform for the final disposal through further researches.

Stabilization,Volatile radionulide,Sol-gel method,SiO2-Al2O3-P2O5,Metal chloride,Wasteform

In this work 14C and 3H distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated . It was found that the recovery percent of 14C by the wet oxidation-acid stripping was 81%～100% for the added activity range of 14C, 0.72 Bq～460 Bq, and it was not affected by the kinds of stripping acids, 3 N-HCl, 3 N-HNO3 and 3 N-H2SO4. And the recovery percent of 3H by distillation using the same apparatus was 81%～101% for the added activity range of 3H, 0.60 Bq～435 Bq. Among the tested stripping acids, 3 N-HCl, 3 N-HNO3 and 3 N-H2SO4, only the trapped 3H solution by distillation in 3 N-H2SO4 was compatible with the 3H scintillator, Ultimagold XR. Neither of the 14C and 3H trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 N-H2SO4 stripping contained gamma nuclides. However, some gamma nuclides, 60Co, 134Cs, 137Cs and 54Mn, were found in the trapped 3H solutions of the spent resins by the wet oxidation-3 N-HCl stripping. It was the same for the 3H trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). Meanwhile only two nuclides, 134Cs, and 137Cs, were found in the 14C trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the 14C in the spent resins existed as inorganic carbon form, more than about 70% of the total 14C content. Among the analyzed 30 spent ion exchange resin samples, the average concentration of 14C and 3H for the high radioactive samples, 8 samples, was 19000 Bq/g±41000 Bq/g, 670 Bq/g ±460 Bq/g and that for the low radioactive samples, 22 samples, was 4.2 Bq/g±4.3 Bq/g, 6.0 Bq/g±5.3 Bq/g, respectively. And the average 14C/3H ratio for the high radioactive samples, was higher, 28, than that of low radioactive samples, 0.70. Some linear relationship trend was found between the activity concentrations of 14C and 3H.

Wet oxidation-acid stripping,14C,3H,spent resin

The solubility of U(VI) hydrolysis products was determined by using a laser-induced breakdown detection (LIBD) technique. The experiments were carried out at uranium concentrations in range from 2×10-4 M to 4×10-6 M, pH values between 3.8 and 7.0, the constant ionic strength of 0.1 M NaClO4 and the temperature of 25.0±0.1℃. The solubility product of U(VI) hydrolysis products was calculated from LIBD results by using the hydrolysis constants selected in NEA-TDB. The solubility product extrapolated to zero ionic strength, logK°sp=-22.85±0.23 was calculated by using a specific ion interaction theory (SIT). The spectral features of ionic species in uranium solutions were investigated by using a conventional UV-visible absorption spectrophotometer and a fluorophotometer, respectively. (UO2)2(OH)22+ and (UO2)3(OH)5+ were dominant species at uranium concentration of 2×10-4 M.

Uranium(VI),Solubility,LIBD,Hydrolysis,Speciation,Spectroscopy

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or dispersal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions. For the tests we examined the failure of bolts and the deformation of flange to evaluate a loss or dispersal of radioactive material and measured the shielding thickness using a ultrasonic thickness gauge to assess a loss of shielding integrity. The strains and accelerations acquired from tests were compared with those by analyses to verify the impact dynamic analysis. The analytic results were larger than the those of test so that the analysis showed the conservative results. Finally, we evaluated the safety of the type IP-2 transport package under the stacking test condition using a finite element analysis. Under the stacking test condition, the maximum Tresca stress of the shielding material was 1/3 of the yielding stress. Two kinds of a type IP-2 transport package were safe for the free drop test condition and the stacking test condition.

Type IP-2 transport package,Safety evaluation,Free drop test,Radioactive waste drum

Appropriate numerical models for the simulation of off-gas flow in hot area of the vitrification plant have been developed in this study. The models have been applied to analyze the effect of design parameters of real plant and numerical analyses have been performed for CCM(Cold Crucible Melter), pipe cooler and HTF(High Temperature Filter). At first, the effect of excess oxygen and the ratio of oxygen distribution on combustion characteristics in the CCM has been studied. Next, solidification behavior of radio nuclide in the pipe cooler has been numerically modeled and scrutinized. Finally, flow pattern in accordance with the location of off-gas entrance of the HTF has been compared.

Vitrification Plant,Radio nuclide,CCM,Pipe Cooler,HTF,CFD

Fenton‘s reagent is applied to directly decompose the ion-exchange resins, IRN-78 and the mixed resin with IRN-77. The newly applied procedures is to dry the resin first and the catalyst solution is completely absorbed into the resin, then a limited dose of H2O2 is introduced for an effective reaction between the reagents within the resin. As a characteristic on the decomposition of IRN-78, the resin mixture should be heated to 40oC to induce the initial reaction and lag time is also needed for about 20 minutes until the main reaction occurs. The effectiveness of the decomposition is investigated using CuSO4, Cu(NO3)2 and FeSO4 as a catalyst and the decomposition rate is compared depending on the concentration of each catalyst and the amount of H2O2. The most effective catalyst was found to be FeSO4 for IRN-78 alone and the mixed resin with IRN-77, and FeSO4 showed a special effect that the reaction was initiated without heating and a lag time. Furthermore, the optimum concentration of the catalyst for each resin and the mixed one is suggested in the view point of the amount of H2O2. needed and the stability of the decomposition reaction.

Resins,IRN-77,IRN-78,Fenton';s reagent,Decomposition,FeSO4,CuSO4

In this paper, structural design requirements and safety evaluation criteria of the spent nuclear fuel disposal canister are studied for deep geological deposition. Since the spent nuclear fuel disposal canister emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal canister should be secured. Usually this repository is expected to locate at a depth of 500m underground. The canister which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock is a solid structure with cast iron insert, corrosion resistant overpack and lid and bottom, and entails an evenly distributed load of hydrostatic pressure from underground water and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. If the canister is not designed for all possible external loads combinations, structural defects such as plastic deformations, cracks, and buckling etc. may occur in the canister during depositing it in the deep repository. Therefore, various structural analyses must be performed to predict these structural problems like plastic deformations, cracks, and buckling. Structural safety evaluation criteria of the canister are studied and defined for the validity of the canister design prior to the structural analysis of the canister. And structural design requirements(variables) which affect the structural safety evaluation criteria should be discussed and defined clearly. Hence this paper presents the structural design requirements(variables) and safety evaluation criteria of the spent nuclear fuel disposal canister.

spent nuclear fuel disposal canister,hydrostatic pressure,underground water,bentonite buffer,structural design requirements,structural safety evaluation criteria,structural analysis

An underground research tunnel is essential to validate the integrity of a high-level waste disposal system, and the safety of geological disposal. In this study, KAERI underground research tunnel (KURT) was constructed in the site of Korea Atomic Energy Research Institute(KAERI). The results of the site investigation and the design of underground tunnel were presented. The procedure for the construction permits and the construction of KURT were described briefly. The in-situ experiments being carried out at KURT were also introduced.

underground research tunnel,high-level waste,site investigation

Spent filter wastes of about 1,000 units (200 L) have been stored in the waste storage facility of the Korea Atomic Energy Research Institute since its operation. At the moment, to secure space in a waste storage facility as well as to efficiently manage spent filter wastes, it is necessary to conduct a compaction treatment of these spent filters, and finally, to repack the compacted spent filters into a 200 liter drum. To do that, the spent filter wastes were first classified according to their generation facilities, their generation date and their surface dose rate by investigating the inventory of the spent filters. In order to repack a compacted spent filter in a 200 liter drum, it is first necessary to conduct a radionuclide assessment of a spent filter before compacting it. Therefore, after taking a representative sample from a spent filter without a dismantlement, the nuclide analysis for it will be conducted. And then, after putting a spent filter into a regular drum by conducting the columnar shaping of the hexahedral form of a spent filter, the compaction treatment of the shaped spent filter will be conducted by vertically compacting it.

Spent filter,HEPA filter,nuclide analysis,representative sample,columnar shaping,compaction treatment

Various kinds of RI wastes are discharged from licensed organizations of radioisotopes such as hospitals and clinic organizations, educational organizations, research institutions, and public organizations. Radioiodines such as 125I and 131I are radioisotopes mainly used in nuclear medicine and industry. A method for the determination of radioiodines in RI wastes has been applied to measure low level activity using acid decomposition method and HPGe gamma ray spectrometer. Prior to analysis of real samples, 131I reference solution and 10 g of yellow tissue paper was added to flask in mantle and was heated in 100 mL of 0.4 N K2Cr2O7 and 100 mL of 9 M H2SO4, and then distilled after adding 10 mL of 30 % H2PO3 and 1 mL of 30 % H2O2. The condensed iodine by circulator was extracted into CCl4, then back-extracted into the aqueous phase with 10 mL of 5 % K2SO2 solution. Finally, 131I was measured at 364.48 keV using HPGe gamma ray spectrometer after precipitation and filtration. Chemical yield of three steps such as acid decomposition process, chemical separation process, and precipitation and filtration process was more than 94 % respectively. MDA(Minimum Detectable Activity) of 131I at this analytical condition was 0.16 ßq/g.

Acid decomposition method,Radioiodine,RI waste

Hyperalkaline groundwater formed by groundwater-cement components and its reaction with bedrock in a nuclear waste repository were simulated by geochemical modeling. The result of groundwater-cement components reaction showed that the pH of water was 13.3 and the precipitated minerals were Brucite, Katoite, Calcium Silicate Hydrate(CSH1.1), Ettringite, Hematite, and Portlandite. The result of interaction between such minerals and groundwater sampled in Gyeongju area also showed that the pH of groundwater reached 12.4. Interaction between such hyperalkaline groundwater and granite was simulated by kinetic model during 103 years. This result showed that the final pH of groundwater reached 11.2 and the variation of pH was controlled by dissolution/precipitation of silicate and CSH minerals. Groundwater quality was also determined by dissolution/precipitation of silicate, CSH, oxide minerals. Our results show that geochemical modeling of long-term hyperalkaline groundwater and rock interaction can contribute to the safety assessment of engineered barrier by predicting geochemical condition in repository site.

Nuclear waste depository,Cement,Granite,Geochemical modeling,Kinetic model

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.4 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ϒ-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

Pyroprocessing of spent fuel,Nuclear fuel cycle,Separation of uranium and TRU,Non-proliferation

The granulation equipment of concentrated wastes is manufactured for the polymer solidification of concentrated wastes. It uses liquid sodium silicate as a granulating agent for the granulation of dried powder containing boric acid. The granulating agent is sprayed in the form of droplet and mean size of dried granules is 2～4 mm. The new technology which has been used for the polymer solidification of spent resin in U.S. and certified by Nuclear Regulatory Commission (NRC) is successfully applied to concentrated wastes. This uses in-situ solidification process within drum without mechanical mixing. Maximum loading of waste can be achieved without increasing of waste volume. Polymer waste forms were evaluated with several test such as fire test, compressive strength test, leaching test, immersion test, irradiation test, and thermal cycling test according to standard test procedures.

concentrated waste,boric acid,granulation,polymer solidification

This study deals with the irradiation effects on some selected components which are being used in an Advanced Spent Fuel Conditioning Process (ACP). Irradiation test components have a higher priority from the aspect of their reliability because their degradation or failure is able to critically affect the performance of an ACP equipment. Components that we chose for the irradiation tests were the AC servo motor, potentiometer, thermocouples, accelerometer and CCD camera. ACP facility has a number of AC servo motors to move the joints of a manipulator and to operate process equipment. Potentiometers are used for a measurement of several joint angles in a manipulator. Thermocouples are used for a temperature measurement in an electrolytic reduction reactor, a vol-oxidation reactor and a molten salt transfer line. An accelerometer is installed in a slitting machine to forecast an incipient failure during a slitting process. A small CCD camera is used for an in-situ vision monitoring between ACP　campaigns. We made use of a gamma-irradiation facility with cobalt-60 source for an irradiation test on the above components because gamma rays from among various radioactive rays are the most significant for electric, electronic and robotic components. Irradiation tests were carried out for enough long time for total doses to be over expected threshold values. Other components except the CCD camera showed a very high radiation hardening characteristic. Characteristic changes at different total doses were investigated and threshold values to warrant at least their performance without a deterioration were evaluated as a result of the irradiation tests.

Irradiation test,AC servo motor,Potentiometer,Thermocouple,Accelerometer,CCD camera

For the assessment of the performance and safety of a deep underground radioactive repository system, the thermal, hydraulic, mechanical, and chemical behaviors and their coupling should be studied. In order to analyze the THMC coupling behavior more effectively, which requires complex mathematical models and modelling techniques, DECOVALEX international cooperation project was launched in 1992. Since its beginning, four major stages of the project were successfully completed and THMC modelling techniques for various conditions could be developed. In this study, the current status and major achievements from the project was reviewed and possible benefits of the participation to the project was discussed.

DECOVALEX,THMC coupling,underground repository,geological disposal,code validation,international project

Status on the spent nuclear fuel management policy and R&D plan of the major countries is surveyed. Also the prospect of the future R&D plan is suggested. Recently so-called fuel cycle nations, which have the reprocess policy of the spent fuel, announced new spent fuel management policy based on the advanced fuel cycle technology. The policy is focused to transmute highly radioactive material and material having a very long half-life, and to recycle the Pu and U contained in the spent fuel. In this way the radio-toxity of the spent fuel as well as the amount of the high level waste to be eventually disposed can greatly be reduced. Most of countries selected the wet process as a primary option for the treatment of the spent fuel since the advanced wet process, which is based on the existing PUREX process, looks more feasible as compared with the dry process. The wet process, however, is much more sensitive in terms of proliferation-resistance compared with the dry process. The pure Pu can easily be obtained by simply modifying the process. On the other hand the pure Pu can not be extracted in the dry process based on the high temperature molten salt process such as a pyroprocess. Even though the pyroprocess technology is very premature, it has a great merit. Thus it is necessary for Korea to have a long term strategy for pursuing a spent fuel treatment technology with a proliferation resistance and a great merit for the GEN-IV fuel cycles. Pyroprocess is one of the best candidates to satisfy these purposes.

spent fuel,treatment,wet process,dry process,fuel cycle

The Japanese Government selected Mutsu, Aomori Prefecture as a provisional spent-fuel repository site. This comes as a result of the prefecture's five-year campaign to host the site since 2000. Korea stores spent nuclear fuel within the compound of nuclear power plants, and expects the storage capacity to reach its limit by the year 2016. This compels Korea to learn the cases of Japan. Having successfully hosted Gyeongju as a site for low-to-intermediate-level nuclear waste repository, Korea has already learned the potential process of hosting spent fuel storage site. The striking difference between the two countries in the process of hosting the site is that the Korean government had to offer the local city a large amount of subsidy for hosting through competitive citizens' referendum among candidate cities while it was the leadership of the local municipality that enabled the controversial decision in Japan. It is also a distinguishable characteristics of Japan that not a huge subsidy is provided to the local host city. I hope this study offers an idea to Korea's future effort to select a spent-fuel host site.

the mid-term storage facility,nuclear spent fuel in Japan,City of Mutsu in Japan

The spectroscopic characteristics of Pu (III, IV, V, VI) in the HCl media were investigated by measuring Pu oxidation states using a UV-Vis-NIR spectrophotometer (400-1200 ) after adjusting Pu oxidation states with oxidation/reduction reagents. Pu in stock solution was reduced to Pu(III) with NH2OH HCl, and oxidized to Pu(IV) and Pu(VI) with NaNO2 and HClO4, respectively. Also, Pu(V) was adjusted in the Pu(VI) solution with NH2OH HCl. The major absorption peaks of Pu(IV) and Pu(III) were measured in the 470 and 600 , respectively. The major absorption peaks of Pu (VI) and Pu(V) were measured in the 830 and 1135 , respectively. There was not foundto be significant changes of UV-Vis absorption spectra for Pu(III), Pu(IV) and Pu(VI) with aging time, except that an unstable Pu(V) immediately reduced to Pu(III).

Pu oxidation state,absorption spectrum,spectroscopic characteristics

For the accurate safety assessment of potential radioactive waste disposal site which is located in the crystalline rock it is important to simulate the mass transportation through engineered and natural barrier system precisely, characterized by porous and fractured media respectively. In this work the methods to construct discrete fracture network for the analysis of flow and mass transport through fractured-porous medium are described. The probability density function is adopted in generating fracture properties for the realistic representation of real fractured rock. In order to investigate the intersection between a porous and a fractured medium described by a 2 dimensional rectangular and a cuboid grid respectively, an additional imaginary fracture is adopted at the face of a porous medium intersected by a fracture. In order to construct large scale flow paths an effective method to find interconnected fractures and algorithms of swift detecting connectivities between fractures or porous medium and fractures are proposed. These methods are expected to contribute to the development of numerical program for the simulation of radioactive nuclide transport through fractured-porous medium from radioactive waste disposal site.

Discrete Fracture Network,Fractured-Porous Medium,Fracture Intersection,Connectivity

A demonstration facility for advanced spent fuel conditioning process (ACP) is under construction in KAERI. In this hot cell facility, all process equipments and materials are taken in and out only through the rear door. The working table in front of the process rear door is specially designed to be openable for the efficient use of the space. This paper presents the structural safety analysis of the openable working table, for the normal operational load condition and accidential drop condition of heavy object. Both cases are investigated through static and dynamic finite element analyses. The analysis results show that structural safety of the working table is sufficiently assured and the working table is not collapsed even when an object of 500 is dropped from the height of 50 .

ACP,hot cell,working table,structural safety,FEM,ACP

This study examined the possibility of using dark-striped field mice as a biological indicator for the environmental radio-surveillance. For this study, dark-striped field mice were caught from five areas of Kyonggi, Kyongsang, Chungchong and Cholla provinces. The external morphological characteristics and isoenzymic types of dark-striped field mice were studied after they were captured. Among the external morphological characteristics, the dark-brown coat, dark back stripe, head-to-tail length, tail length, and ear length matched the taxonomical characteristics of darkstriped field mice. The analyses on L-lactate dehydrogenase, aspartate aminotransferase, and malate dehydrogenese revealed that one species of dark-striped field mice, called Apodemus agrarius, was inhabitated throughout a wide range of Korea. On the other hand, the frequency of micronuclei in peripheral polychromatic erythrocytes to survived mice after irradiation also analyzed. The LD50/30 of A. agrarius and ICR mice were approximately 5 Gy and 7.9Gy, respectively. The results of the study reveal that wild A. agrarius have a high potential as a biological monitoring system to determine the impact of radiation in areas such as those within the vicinity of nuclear power plants.

Apodemus agrarius,radiosensitivity,ecological and biological dosimeter,and ICR.

The characteristics of the aluminum waste melting and the distribution of the radioactive nuclides have been investigated for the estimation on the volume reduction and the decontamination of the aluminum wastes from the decommissioning of the TRIGA MARK and research reactors at the Korea Atomic Energy Research Institute(KAERI). The aluminum wastes were melted with the use of the fluxes such as flux A : NaCl-KCl-Na3AlF6, flux B : NaCl-NaF-KF, flux C : CaF2, and flux D : LiF-KCl-BaCl2 in the DC graphite arc furnace. For the assessment of the distribution of the radioactive nuclides during the melting of the aluminum, the aluminum materials were contaminated by the surrogate nuclides such as cobalt(Co), cesium(Cs) and strontium(Sr). The fluidity of aluminum melt was increased with the addition of the fluxes, which has slight difference according to the type of fluxes. The formation of the slag during the aluminum melting added the flux type C and D was larger than that with the flux A and B. The rate of the slag formation linearly increased with increasing the flux concentration. The results of the XRD analysis showed that the surrogate nuclide was transferred to the slag, which can be easily separated from the melt and then they combined with aluminum oxide to form a more stable compound. The distribution ratio of cobalt in ingot to that in slag was more than 40% at all types of fluxes. Since vapor pressures of cesium and strontium were higher than those that of the host metals at the melting temperature, their removal efficiency from the ingot phase to the slag J. Kor. Rad. Waste Soc.

Melting,Arc Furnace,Aluminum Waste,TRIGA,Slag,Distribution

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 - container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by 20 tons.

disposal container,Von Mises stress,spent PWR fuel,safety factor,deep geological repository

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA';s transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

Spent nuclear fuel,Transport system,Transport cask,On-site transport and storage,Transport regulations

Ion exchange resin particles should not be found in steam generator(S/G) sludge. The suspicious spherical resin particles observed in S/G sludge sample were characterized for particle size distribution under optical microscope using the micro-technique, for element analysis by the electron probe micro analysis (EPMA), and for molecular identification by the IR spectroscopy. The particle sizes are distributed from 1 to 200 for the sludge, while 40 to 500 for the spherical resin particles. The results of the elemental analysis showed different major impurities: Si, Al, Mn, Cr, Ni, Zn and Ti for the sludge particles, while Si, Cu, Zn for the spherical resin particles. However, both particles contain Fe as a matrix of magnetite (Fe3O4). IR spectrum of the spherical particles was not quite similar to the IR spectrum of ion exchange resins used in S/G system. These results indicate that the spherical particles are not related to ion exchange resin particles and may be formed by the process of the sludge formation.

Spherical particles,S/G sludge,characterization,micro-technique

Transparency on the Total System Performance Assessment (TSPA) is the key issue to enhance the public acceptance for a radioactive repository. To approve it, all performances on TSPA through Quality Assurance is necessary. The integrated Cyber R&D Platform is developed by KAERI using the T2R3 principles applicable for five major steps ; planning, research work, documentation, and internal & external audits in R&D';s. The proposed system is implemented in the web-based system so that all participants in TSPA are able to access the system. It is composed of three sub-systems; FEAS (FEp to Assessment through Scenario development) showing systematic approach from the FEPs to Assessment methods flow chart, PAID (Performance Assessment Input Databases) being designed to easily search and review field data for TSPA and QA system containing the administrative system for QA on five key steps in R&D’s in addition to approval and disapproval processes, corrective actions, and permanent record keeping. All information being recorded in QA system through T2R3 principles is integrated into Cyber R&D Platform so that every data in the system can be checked whenever necessary. Throughout the next phase R&D, Cyber R&D Platform will be connected with the assessment tool for TSPA so that it will be expected to search the whole information in one unified system.

Total System Performance Assessment,Cyber R&D Platform,Quality Assurance,FEAS,PAID

A demonstration facility for an advanced spent fuel conditioning process (ACP) is under construction at KAERI. In this hotcell facility, the rear door is frequently used since all process equipment and materials are taken in and out only through the rear door. Therefore, both the structural safety and stability of the door are essentially required for the safety of ACP facility. In this paper, the finite element analysis has been performed to investigate the structural safety under the impact condition between the rear door and the door frame. Also the possibility of the rear door being tumbled over by the impact force or the inertia force under a sudden stop conditon has been evaluated. The analysis results demonstrate that the structural safety and stability of the rear door are sufficiently assured for both the impact and the accidential stop conditions.

ACP,hotcell,rear door,structural safety,FEM

Inventories, and characteristics such as dimension, fuel rod array, weight, 235U enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial 235U enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed 36 GWD/MTU and 40 GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be 45 GWD/MTU at the end of 2000';s. From the comprehensive study, it was concluded that the imaginary SNF with 16 16 Korean Standard Fuel Assembly, cross section of 21.4 21.4 , length of 453 , mass of 672 , initial 235U enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

spent nuclear fuel,geological disposal system,high-level waste,reprocessing,radioactive waste

Alpha spectrometry was studied for the determination of 237Np in spent nuclear fuel samples. The optimum condition for the electrodeposition of 237Np was obtained as follows : for 1 1.5 hour of deposition time, at the current intensity of 1.2 1.5 A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on 237Np was decreased as the amount of 237Np decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of 237Np determined by alpha spectrometry after separation in synthetic solution was 98.8±5.1%(n=4). The contents of 237Np in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.

237Np,alpha spectrometry,electrodeposition,anion exchange chromatography,actinides separation

By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

Inconel 600/690,Corrosion Experiment,Corrosion Speed

A coupled T-H-M(Thermo-Hydro-Mechanical) analysis was carried out for KENTEX (KAERI Engineering-scale T-H-M Experiment for Engineered Barrier System), which is a facility for validating the coupled T-H-M behavior in the engineered barrier system of the Korean reference HLW(high-level waste) disposal system. The changes of temperature, water saturation, and stress were estimated based on the coupled T-H-M analysis, and the influence of the types of mechanical constitutive material laws was investigated by using elastic model, poroelastic model, and poroelastic-plastic model. The analysis was done using ABAQUS, which is a commercial finite element code for general purposes. From the analysis, it was observed that the temperature in the bentonite increased sharply for a couple of days after heating the heater and then slowly increased to a constant value. The temperatures at all locations were nearly at a steady state after about 37.5 days. In the steady state, the temperature was maintained at 90 at the interface between the heater and the bentonite and at about 70 at the interface between the bentonite and the confining cylinder. The variation of the water saturation with time in bentonite was almost same independent of the material laws used in the coupled T-H-M processes. By comparing the saturation change of T-H-M and that of HM(Hydro-Mechanical) processes using elastic and poroelastic material model respectively, it was found that the degree of saturation near the heater from T-H-M calculation was higher than that from the coupled H-M calculation mainly because of the thermal flux, which seemed to speed up the saturation. The stresses in three cases with different material laws were increased with time. By J. Kor. Rad. Waste Soc. Vol. 4(2), p.117-131, June 2006. comparing the stress change in H-M calculation using poroelasetic and poroelasetic-plastic model, it was possible to conclude that the influence of saturation on the stress change is higher than the influence of temperature. It is, therefore, recommended to use a material law, which can model the elastic-plastic behavior of buffer, since the coupled T-H-M processes in buffer is affected by the variation of void ratio, thermal expansion, as well as swelling pressure.

Coupled THM process,Bentonite saturation,Material law,HLW repository

Chemical wastes containing small amounts of uranium can not be disposed of them after treatment as an industrial waste, because the uranium concentration in the final dry cake exceeds the exemption level. Especially for the removal of uranium in this study, the method for immobilizing Diphosil powder within alginate beads is adopted to make a bead form from a powdered resin. Sodium alginate bead itself showed a capability to uptake uranium to above 60 %, but the value was decreased to below 30 % after equilibrium. The adsorption rate of uranium increased with the increasing content of Diphosil in the sodium alginate bead. Diphosil resin itself showed very fast uptake of uranium from early stages, and then the rates were leveled off. Diphosil bead showed an improved capability to uptake uranium considering the pure Diphosil content in the composite bead, and provide a considerable potential for further applications of a continuous process by using Diphosil as a bead form.

radioactive chemical wastes,uranium removal,uranium adsorption,Diphosil

A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a free drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

Spent fuel,Dry storage cask,Drop accident condition,Impact limiter,Structural integrity

It is a serious task to the decommissioning of the uranium conversion plant that the demolition of the lagoon sludge. The main component of the sludge is ammonium nitrate and that is the very explosive material. Therefore, the bio-denitrification is a attractive process to remove the nitrate. In this work, some process variables was tested such as incubation temperature, nitrate concentration, electron donor, C/N ratio, seeding ratio, and pH with an anaerobic bacteria as Pseudomonas halodenitrificans. The results would be used as basic data to the continuous bio-denitrification process.

bio-denitrification,Pseudomonas halodenitrificans,uranium conversion,nitrate,sludge

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered on the inner surface of hot cell and surface of equipment in hot cell. It was necessary to develop a filtration equipment to reuse the PFC waste solution generated on PFC decontamination due to the high cost of PFC solution and for minimization of the volume of second waste solution. The filtration equipment was developed to remove hot particulate in PFC waste solution. It was made suitable size and weight in consideration of hot cell gate and crane. And it has wheels for easy movement. Flux of the filtration equipment decreased with particulate concentration increase. It consists of pre-filter(1.4 m) and final-filter(0.2 m) for protection of the flux decrease along filtration time. Its treatment capacity of waste solution is 0.2 L/min.

PFC,filtration equipment,hot particulate,removal efficiency,flux

Decommissioning cost estimation is a very important technique in designing and planning of nuclear facilities'; decommissioning. Decommissioning cost estimation should be made according to the phases of decommissioning activities and installed components of nuclear facilities. In this paper, the basic framework necessary for decommissioning cost estimation is completed so that it could be used as a technique for decommissioning costs estimation by specifying cost items and group components and unit cost factors on which work time is calculated. Also, factors to be considered for decommissioning cost estimation of major activities and tasks are reviewed. Afterwards, these techniques will be utilized as a basic technology to establish methodology of decommissioning cost estimation and evaluation.

Decommissioning Cost Estimation,Cost Items,Unit Cost Factors

Chemical wastes are generated from nuclear facilities and R&D laboratories, but the uranium concentration in the final dried cake is evaluated into 11.2 Bq/g, which exceeds the exemption level of 10 Bq/g for each U isotopes, so the cake is categorized into a radioactive waste. Acid dissolution was applied to extract uranium from the waste sludge, and uranium adsorption on the dissolved solution was experimented by using IRN-77 and Diphosil bead. A large amount of resin was required to get above 80 % of uranium removal, which was found to be due to a large amount of metal ions simultaneously dissolved from the precipitates with uranium. As an alternative method, acid dissolution is applied to the dewatered wet cake of the sludge, and the natural evaporation method is adopted for the dissolved solution. The uranium concentration of the dissolved solution was estimated to be 6.97E-01 Bq/ml, and the specific activity of the final waste sheets is evaluated to be 4.3 Bq/g. These results lead to the suggestion that the application of acid dissolution to the wet cake and the natural evaporation for the dissolved solution is an effective treatment method for chemical wastes containing uranium.

radioactive chemical wastes,uranium removal or adsorption,acid dissolution,natural evaporation

The regulation for the low and intermediate level radioactive waste to be transferred to the disposal facility, recently revised, require that radioactive waste generators should set up waste certification program to verify the radioactive waste conform to the waste acceptance criteria(WAC) before disposal. The radioactive waste disposal facility, scheduled to be constructed in Korea, will institute WAC for the wastes to be transferred to the facility. This WAC is expected to compose of the requirements for the radiological characterization, physical and chemical characterization , physical/chemical restriction, prohibited item, packaging, identification, labeling, and documentation. For the compliance with this regulation, The radioactive waste generators should verify that the waste meet WAC through performance of the waste certification program and are responsible for handing in all the certification documents to the disposal facility. This waste certification program plan was set up as a preliminary program for the certification of radioactive waste generated in Korea Atomic Energy Research Institute (KAERI) and should be further revised until preparation of WAC by disposal agent.

low and intermediate level,Radioactive waste,Certification program

The engineering performance of a high level waste repository is significantly dependent upon the T-H-M behavior in the engineered barrier system. An engineering-scale test facility (KENTEX) was set up to validate the T-H-M behaviors in the buffer of a reference disposal system developed in the 2002. The validation tests started on May 31, 2005 and is now in progress. The KENTEX facility and validation test programme are introduced, and pre-operation calculations are also presented to give information on the sensitive location of sensors and operational conditions . This test will provide information (e.g., large-scale apparatus, sensors, monitoring system etc.) needed for "in-situ" tests, make the validation of a T-H-M model for the T-H-M performance assessment of the reference disposal system, and demonstrate the engineering feasibility of fabricating and emplacing the buffer of a repository.

HLW disposal system,buffer,T-H-M behavior,validation test

Characteristics of heat release process, while the Wigner energy was drawn off the graphite during DSC(Differential Scanning Calorimenter) measurement as an example of annealing process which is one of release methods of Wigner energy that is contained in the irradiated graphite, was studied. Linear temperature rise method in DSC operation was selected to estimate the total Wigner energy content and the heat release rate of each graphite samples, which were located in several positions in the thermal column in KRR-2 research reactor. As an annealing process in DSC operation Wigner energy of the irradiated graphite samples were totally released by heat supplying to the graphite from room temperature to 500 , in DSC. Characteristics of Wigner energy release from the graphite sample was well correlated with the various activation energy model of the kinetic equation.

Irradiated Graphite,Wigner Energy,Annealing,DSC Measurement,Linear Temperature Rise Method

The structural and contamination characteristics of hot cells at KAERI were investigated. The SEM results showed that the size of the hot particulate on the inner surface of the hot cell ranged from 0.2 to 10 . It was found that an inlet flow rate of 15 m/sec was suitable for this developed cyclone with a 49 mm optimum vortex finder length. The results showed that the collection efficiency was about 85% for 3 particles. The collection efficiency didn';t show a sharp increase when the inlet flow rate was faster than 15m/sec. When the temperature of the inlet flow gas was increased, the collection efficiency of the cyclone was slightly decreased. The larger the vortex finder length was, the higher the pressure drop in the cyclone was. The cut size diameter decreased with an increment of the Reynolds number. It was established that the flow in the cyclone was a turbulent flow on the basis of the Reynolds number and this turbulent flow caused a pressure drop in the cyclone. decreased with increasing values of the Reynolds number and it gradually approached a constant value at a higher value of the Reynolds number. Namely, approached approximately 0.045 between 6000 and 8000 of the Reynolds number.

cyclone,hot particulate,collection efficiency,cut size,Reynolds number

Two candidate glasses, AG8W1 and DG-2, have been developed for vitrifying the mixture of low activity resin, zeolite and Dry Active Waste(DAW), and DAW solely, respectively. In order to evaluate the chemical durability of the glasses, two different leaching tests, Product Consistency Test(PCT) and Vapor Hydration Test(VHT), have been performed. As the results of PCT performed from 7 to 120 days, the leach rates of B, Na, Si and Li in the glasses were much lower than those of the benchmark glass(SRL-EA). As the results of VHT performed for 7 days, the leach rates were 2 and 10 g/ /day for AG8W1 and DG-2, respectively. The results of VHT met the regulatory guideline(<50 g/ /day) for the low activity glasses of Hanford in the USA. Consequently, two candidate glasses to be used at a commercial operation in the future showed that their chemical durability is satisfactory according to the results of two leaching tests.

Vitrification,Vitrified Form,Leaching Test,PCT,VHT

In this study, an experimental study on the sorption properties of uranium(VI) onto a bentonite colloid generated from Gyeongju bentonite which is a potential buffer material in a high-level radioactive waste repository was performed as a function of the pH and the ionic strength. The bentonite colloid prepared by separating a colloidal fraction was mainly composed of montmorillonite. The concentration and the size fraction of the prepared bentonite colloid measured using a gravitational filtration method was about 5100 ppm and 200-450 nm in diameter, respectively. The amount of uranium removed by the sorption reaction bottle walls, by precipitation, and by ultrafiltration was analyzed by carrying out some blank tests. The removed amount of uranium was found not to be significant except the case of ultrafiltration at 0.001 M NaClO4. The ultrafiltration was significant in the lower ionic strength of 0.001 M NaClO4 due to the cationic sorption onto the ultrafilter by a surface charge reversion. The distribution coefficient Kd (or pseudo-colloid formation constant) of uranium(VI) for the bentonite colloid was about 104 107 mL/g depending upon pH and ionic strength of NaClO4 and the Kd was highest in the neutral pH around 6.5. It is noted that the sorption of uranium(VI) onto the bentonite colloid is closely related with aqueous species of uranium depending upon geochemical parameters such as pH, ionic strength, and carbonate concentration. As a consequence, the bentonite colloids generated from a bentonite buffer can mobilize the uranium(VI) as a colloidal form through geological media due to their high sorption capacity.

Uranium(VI),Buffer Material,Sorption,Precipitation,Ultrafiltration,Bentonite Colloid,Distribution Coefficient

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the nondetective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tipover test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

Spent nuclear fuel,Dry storage cask,Tip-over analysis,Initial angular velocity,Test for verifying

Vol-oxidizer is a device to convert UO2 pellets into U3O8 powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in UO2 pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the U3O8 measuring the temperature distribution, and the volume of UO2 and U3O8. We manufactured the new vol-oxidizer for a 20 kg HM/batch in UO2 pellets, and then analyzed the characteristics of the U3O8 powder for the verification. The experimental results show that the permeability of the U3O8 throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is 40 . The capacities of new vol-oxidizer for a 20 kg HM/batch in UO2 pellets were agree well with the predictions of design model.

UO2 pellet,Oxidation,U3O8,Design model,Demonstration,Vol-oxidizer,Mesh

In the decommissioning and decontamination(D&D) planning stage, it is important that the scenarios are evaluated from an engineering point of views because the decommissioning work has to be executed economically and safely by following the best scenarios. Therefore, we need to develope several modules to evaluate the decommissioning scenarios. In this paper, the digital mock-up system is constructed in the virtual space to simulate the whole decommissioning process. The schedule evaluating equation and cost evaluation equation are derived to calculate the working time and the expected cost. And in order to easily identify the radiation level about the activated objects, the radiation visualization module is developed. Finally, on the basis of the obtained results from the Digital Mock-up and other important factors, the evaluating method of the scenarios that can indicate the best scenario is described.

evaluation modules,digital mockup system,decommissioning and decontamination,radiation visualization

In the measurement of gamma rays emitted from the nuclide in the radioactive waste drum, to analyze the nuclide concentration accurately, it is necessary to use the proper calibration standards and to correct for the attenuation of the gamma rays. Two drums having a different density were used to analyze the nuclide concentration inside the drum in this study. After carrying out the system calibration, we measured the gamma rays emitted from the standard source inside the model drum with changing the distance between the drum and the detector. The measured values were corrected with the three kinds of gamma attenuation correction methode, as a results, the error was less than 10 % in the low density drum and less than 25 % in the high density drum. The measured activity in the short distance was more accruable than in the long distance. The transmission correction for the mass attenuation showed good results(very low error) compared to the mean density and the differential peak correction method.

NDA,Model drum,Gamma-ray,Attenuation correction,Mass attenuation coefficient,Transmission Source

At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.

spent nuclear fuel,direct disposal,reprocessing,long-term storage,global safety regime

KDC-1 canister for PWR spent fuels which will be used for the Korean Reference Disposal System was developed. The structural analysis of the canister was carried out as a part of the safety analysis. Two conditions, disposal condition and handling condition, were considered for the structural analysis. Three kinds of load cases, normal, abnormal and rock movement, were considered for the disposal condition. The results of the calculation showed that the safety factors from the structural analysis were greater than the design requirements. Two accident scenarios, gripper failure accident and canister drop accident, were analyzed for the handling condition. According to the gripper failure scenario analysis, the handling machine with grippers could be used even in the cases that one or two grippers failed. The maximum von Mises stress from the canister drop accident scenario was 0.762 MPa, which was negligible compared with the yield stress of nodular cast iron. The proposed KDC-1 canister for PWR spent fuels proves to be safe under the repository condition that is based upon the Korean reference disposal system according to the structural analysis for disposal condition and handling condition.

structural analysis,canister,spent fuel,disposal,safety

Removal characteristics of H14CO3- ion from IRN-150 mixed resin contaminated with 14C radionuclide and a gasification behavior of 14C radionuclide to 14CO2 were investigated. The stripping solutions used for the removal of 14C from spent resin were NaNO3, Na3PO4, NH4H2PO4, H3PO4. The influence of stripping solution concentration on the desorption characteristics of inactive HCO3 ion into stripping solution from IRN-150 mixed resin and the gasification of this ion to CO2 was analyzed. The gasification behavior to CO2 by using NaOH, HNO3, HCl was also compared to that of phosphate solution. Real spent resin stored in Wolsung nuclear power plant was used to evaluate the gasification characteristics of 14C radionuclide to 14CO2. Gamma radionuclides such as 137Cs, 60Co in residual striping solutions after desorption experiment were analyzed.

14C radionuclide,Spent resin,Stripping solution,14CO2,Gasification

basic research was conducted on the mineral weathering and geochemical characteristics in the KURT (KAERI Underground Research Tunnel), which was recently constructed at a site in KAERI. Some rock samples exposed during the KURT construction were examined using a microscope and chemical analysis for some micro-changes of the rocks caused by the chemical weathering. The weathered granite has some small and fine cracks around the rock-forming minerals. In particular, there are a characteristic weathering of feldspar mineral and a preferential leaching of Ca component from the mineral dissolution. In addition, by the dissolution of biotite containing Fe2+ component there were iron-oxides precipitates as secondary products into the micro-cracks of around minerals. The results also show that the micro-cracks initiated from the mineral interior are extended and connected into the larger cracks along the grain boundary with the progress of the weathering. Thus, it is considered that some chemicals dissolved from the fresh rock would be involved in the formation of secondary minerals and migrate interacting with them.

KURT,weathering,granite,cracks

The depth distributions of total 3H and 14C activities were characterized for the activated shielding concrete from a decommissioning of KRR-2 using the commercially available tube furnace and a liquid scintillation counter. The correlation of measurement results between 3H, 14C and gammer emitter was evaluated to apply for estimating radionuclide inventory of the concrete waste generated from decommissioning KRR-2. The detection limits for 3H and 14C are 0.048 and 0.028 Bq/g respectively. The specific activities of the 3H and 14C tend to decrease exponentially as the depth of the concrete becomes deeper from the surface. In addition, the 3H and 14C activities were in good correlation with the 60Co activities analysed for the shielding concrete of KRR-2.

3H,14C,Concrete,Activation,Decommissioning,Liquid scintillation counter

In this paper a computer program was developed, which simulates the Li reduction process of PWR spent fuel, and the amount of a produced metal or chloride compound was calculated at the various amount of Li with the program. It establishes a database, which is composed of some characteristics related to a chemical reaction equation and thermodynamic data, and it calculates the transformed rate of PWR spent fuel oxide at the certain amount of Li by using the database as input data. As the results of the performance test of the program, it was validated that the transformed values of oxides, except for Eu2O3 and Sm2O3, were almost the same to within about a 6 % error with those calculated by the previous code and that the calculated amount of Li was also exactly consistent with the theoretical one, which is used for a complete reaction of each oxide in a single chemical reaction. A relationship between Li and the transformed metal of each oxide was analyzed on the basis of the quantities calculated with the verified development program. Of the results, when the amount of Li was given to be 250 mole, the 83.73 percentage of UO2 was transformed into U while the remainder was still to be UO2. In addition, it was appeared that the 297 mole of Li was needed to completely convert UO2 into U.

Li reduction,PWR Spent Fuel,chemical equilibrium,metalization

The cemented and paraffin wastes form which are incorporated the concentrated wastes, the cemented waste form which is incorporated the spent ion-exchange resins, and the miscellaneous waste(decontamination paper) were irradiated up to 10+8 rads at 5.43 x 10+5 rads/hr with Co-60(72,023.9 Ci) as an external irradiation source. As a result, the radiolysis gases such as H2, CH4, N2, C2H6, O2, CO and CO2 were measured in all the wastes. The major gas which was generated in all the wastes was hydrogen(H2). The volume of the generated gases showed a difference from 0.029 cm3.atm/1.1g to 0.788 cm3.atm/1.1g according to the type of wastes, and more was generated in the cemented waste form incorporated a spent ion-exchange resin than in the other wastes. More hydrogen(H2) gas was generated in the decontamination paper waste than in the other wastes, and the G(H2) value was 0.12.

radioactive waste,waste form,characterization,irradiation,radiolysis,G-value

In order to evaluate the potential radiological risks for the reuse of the site after decommissioning of nuclear facilities, a mathematical model was developed and materialized into the Microsoft Excel® spreadsheets frame. A set of input parameter values was proposed, which is useful in the preliminary risk screening step before the detailed evaluation with the site-specific data. It appeared that the screening levels calculated by the present model was agreed with the derived concentration guideline limits resulted from RESRAD Ver.6.2 and the German dose criteria for releasing a nuclear site from regulatory control.

Decommissioning,Site Reuse,Screening,Radiological risk

A dynamic compartment model is presented to predict the contamination level of agricultural plant by 137Cs as a result of a soil deposition. The model considered the processes of a percolation, soil mixing by a plowing before transplanting, plant uptake, leaching to a deep soil, and fixation to a clay mineral. The effects of the soil properties (pH, clay mineral, organic matter content, and exchangeable K), which are spatially varied, on a plant uptake and the leaching rates of 137Cs in a root zone soil were modeled by the Absalom model. To test the validity of the model, the 137Cs aggregated transfer factors(TFa) for rice plants were compared with those observed from some simulated 137Cs soil deposition experiments, which were carried out with respect to rice plants cultivated in seventeen paddy soils of different properties for two consecutive years. Observed 137Cs TFa values of the rice plants did not show an evident trend for the pH and clay content of the soil properties, while they increased with an increasing organic matter content or a decreasing exchangeable K concentration. Predicted 137Cs TFa values of the rice plants were found to be comparable with those observed.

137Cs,Absalom model,soil properties,dynamic compartment model

Spent ion-exchanged resin generated from various purification systems in CANDU reactor was contaminated with high activity of 14C radionuclide. This paper describes the results of fundamental study to develop the applicable technology for the treatment of this spent resin. Based on the adsorption capacity of inactive HCO3- ion and other anions on IRN-150 mixed resin, the removal characteristics of HCO3- ion adsorbed on to IRN-150 by various stripping solutions were evaluated. Maximum adsorption amount of the HCO3- ion onto IRN-150 raw resin was about 11 mg-C/g-resin which agrees with the theoretical adsorption amount of this resin. Adsorption affinity of various anions such as Cs, Co, Na, NH4 was analyzed in single and multi-component systems. From the results of removal characteristics of the HCO3- ion adsorbed on IRN-150 by various stripping solutions, NH4H2PO4 stripping solution is more effective than NaNO3, Na3PO4 solutions for the complete removal of 14C radionuclide from the IRN-150 spent resin.

CANDU reactor,Spent resin,14C radionuclide,Anion,Adsortpion,Stripping solution

The estimated decommissioning cost of nuclear research reactor is calculated by applying a unit cost factor-based engineering cost calculation method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning cost of nuclear research reactor is composed of labor cost, equipment and materials cost. Labor cost of decommissioning costs in decommissioning works are calculated on the basis of working time consumed in decommissioning objects. In this paper, the unit cost factors and work difficulty factors which are needed to calculate the labor cost in estimating decommissioning cost of nuclear research reactor are derived and figured out.

Research Reactor,Decommissioning Cost Estimation,Unit Cost Factor

In design of a deep geological repository for the high level wastes, it is very important that the temperature of the bentonite block should not be over 100 oC to maintain the integrity of the bentonite buffer block from the decay heat. In this study, for the layout of the repository to meet the requirement, the analysis of the disposal tunnel and disposal pit spacing was carried out. To do this, based on the reference repository concept, several cases of cooling times and disposal tunnel and disposal pit spacing were compared. The thermal stabilities of the disposal systems were analyzed in terms of the cooling time and spacing. The results showed that it was more desirable to determine the layout of the repository in terms of disposal pit spacing than the disposal tunnel spacing. The results of these analyses can be used in the deep geological repository design. The detailed analyses with the exact site characteristics data will reduce the uncertainty of the results.

Spent Fuel,HLW,Deep Geological Disposal,Repository Layout,Buffer Block,Thermal Stability,Disposal Tunnel/Disposal Pit Spacing

Regulatory systems for decommissioning of nuclear power plants in several countries, such as Japan, United States of America, Germany, United Kingdom, France, and Republic of Korea, are surveyed. In the survey, regulatory policies, legislations, licensing process, inspection and public involvements for decommissioning are identified and compared. Afterwards, the survey results will be utilized as a reference to establish the improvement directions of domestic regulatory system.

nuclear power plants,decommissioning,regulatory system,licensing process

Anion exchange chromatography and HDEHP extraction chromatography using DTPA-lactic acid as an eluent were applied in series for the separation of 241Am and 244Cm in radwaste samples. The separated elements were determined by electrodeposition at the sodium hydrogen sulfatesodium sulfate buffer solution followed by alpha-spectrometry. The recovery yields of 241Am and 244Cm were 85.2±17 % and 86.3±15.3 %, respectively, from the synthetic solution of spent nuclear fuel sample. The amounts of 241Am and 244Cm determined in radwaste sample solutions of condensate bottoms were at the range of 1.5 1.9 Bq/g and 1.7 Bq/g, respectively.

241Am,244Cm,trans uranium,HDEHP extraction chromatography,electrodeposition,radwaste samples,alpha spectrometry

To evaluate parameters influencing on the dust removal of the High Temperature Filter(HTF) system, a computer simulation of fluid dynamics inside the system had been performed. The results showed that the optimum pulse jet periods were 50ms and 90ms for the 1000mm and 1500mm long filter elements respectively. Dust removal effect was very excellent under the pulse jet pressure of 3 bar. But the distance between the pulse jet nozzle and the venturi of a filter element had no meaningful effect on the performance with the variation from 5mm to 10mm. Compared to the dispersion mode of pulse jet, the collective mode of pulse jet flow was preferable in maintaining the pressure inside the system stable.

Pulse duration,Pulse pressure,High Temperature Filter

Shielded laser ablation system composed of laser system, image analyser, XYZ translator with motion controller, ablation chamber, manipulator and various optics was designed. Nd:YAG laser which can be tunable from 1064 ㎚ to 266 ㎚ was selected as light source. CCD camera(< x 200) was chosen to analyze a crater less than 50 ㎛ in diameter. XYZ translator was composed of three linear stage which can travel 50 ㎚ with a minimum movement of 1 ㎛ and motion controller. Before the performance test, each part of system was optically aligned. To perform the ablation test, the specimen was ablated by 50 ㎛ interval and observed by image analyser. The shape of crater was almost round, indicating laser beam has homogeneous energy distribution. The resolution and magnification of image system were compatible with the design.

micro sampling,laser ablation,highly radioactive specimen,nuclear fuel

From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutorn radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

neutron flux,radiation weighting factor,equivalent dose,calibration,containment building

High temperature - high pressure apparatus was developed to simulate nickel ferrite corrosion productswhich were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 ℃ in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needlelike or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.

radioactive crud,crudsimulation,corrosion product,nickel ferrite,high temperature corrosion

Preparation of ferrocyanide-anion exchange resin and adsorption test of the prepared resin on the Cs+ ion were performed. Adsorption capability of the prepared resin on the Cs+ ion in the simulated citric acid based soil decontamination waste solution was 4 times greater than that of the commercial cation exchange resin. Adsorption equilibrium of the prepared resin on the Cs+ ion reached within 360 minutes. Adsorption capability on the Cs+ ion became to decrease above the necessary Co2+ ion concentration in the experimental range. Recycling test of the spent ion exchange resin by the successive application of hydrogen peroxide and hydrazine was also performed. It was found that desorption of Cs+ ion from the resin occurred to satisfy the electroneutrality condition without any degradation of the resin.

Cs+,ion exchange resin,recycling,soil decontamination waste

Many different kinds of radwastes are discharged from the nuclear power plants, and 129I is included in these radwastes. Recovery test of 129I was evaluated for different radwastes(dry active waste, sludge, spent resin and simulated evaporator bottom). Recovery of 129I for dry active waste by acid leaching with 1.8% NaClO was 74.3%(RSD,2.2%) and 129I for spent rein by alkali fusion method with KOH as a flux agent was 87.7%(RSD,0.9%), respectively. Iodide in simulated evaporator bottom containing a high concentration of borate was adsorbed with anion exchange resin at pH 7 phosphate buffer solution. Recovery of 129I for anion exchange resin was 92.5% and not affected up to 1,200 ㎍/mL H3BO3(as a Boron). Recovery of 129I for the spent resin from nuclear power plant was 87.2% (RSD,1.2%).

sample pre-treatment,129I,radwastes,recovery

This study suggested a new method for the solidification of molten salt waste generated from the electro-metallurgical process in the spent fuel treatment. Using binary material system, sodium silicate and phosphoric acid, metal chlorides were converted into metal phosphate in the microreaction module formed by SiO2 particles. The volatile element in the reaction module would little vaporized below 1100 ℃. After the gel product was mixed with borosilicate glass powder and thermally treated at 1000 ℃, Li exists as Li3PO4 separated from glass phase and, Cs and Sr would be incorporated into an amorphous phase from XRD analysis. In case of the addition of ZrCl4 to the binary system, the gel products were transformed into NZP structure considered as an prospective ceramic waste form after heat-treatment above 700 ℃. From these results, the gel-route pretreatment can be considered as an effective approach to the solidification of molten salt waste by the confirmed process or waste form and this also would be an alternative method on the ANL method using zeolites in USA by the confirmation of its chemical durability as an future work. confirmed process or waste form and this also would be an alternative method on the ANL method using zeolites in USA by the confirmation of its chemical durability as an future work.

Solidification,Volatile radionuclide,Sol-gel method,Pre-treatment,Metal Chloride,Waste form

The size and main ingredient of hot particulate generated during the nuclide experiment in hot cells of nuclear facilities were 0.5300 ㎛ and UO2. A cyclone filter equipment which consists of a cyclone and Bag/HEPA filter was devised to remove hot particulate generated during the nuclide experiment in hot cells of nuclear facilities. The experimental conditions to maximize the collection efficiency of hot particulate were suggested through experiments done with the cyclone filter equipment. With the large size of simulated particulate, the collection efficiency of the particulate was high. When the size of simulated particulate was more than 5 ㎛, the collection efficiency of the particulate was more than 80% and when the size of simulated particulate was less than 1.0 ㎛, the collection efficiency decreased by less than 70%. If the inflow velocity of simulated particulate was increased, the collection efficiency of the particulate was also increased. When the inflow velocity of simulated particulate was more than 12m/sec, the collection efficiency was higher than 70%, but after 17 m/sec inflow velocity, no change observed. The collection efficiency of the simulated particulate can be enhanced with the length of vortex finder inside the chamber. With the length of vortex finder, 7.2cm, the observed collection efficiency of the particulate was the maximum. Moreover, when the sub-cone was attached under the cyclone, the collection efficiency of cyclone increased 2%. It was found that effect by attachment of sub-cone was not serious.

hot particulate,cyclone,collection efficiency,velocity,size,vortex finder,sub-cone

A voloxidizer for a hot cell demonstration, that handles spent fuels of a high radiation level in a limited space should be small and spent fuel powders should not be dispersed out of the equipment involved. In this study a density rate equation as well as the Stokes'; equation has been proposed in order to obtain the theoretical terminal velocity of powders. The terminal velocity of U3O8 has been predicted by using the terminal velocity of SiO2, and then etermination has been the optimum air flow rate which is able to prevent powders from scattering. An equation which has shown a relationship between theoretical terminal velocities of U3O8 and SiO2 has been derived with the help of the Stokes'; equation, and then an experimental verification made for the theoretical Stokes'; equation of SiO2 by means of an experimental device made of acryl. The theoretical terminal velocity based on the proposed density rate equation has been verified by detecting U3O8 powders in a filter installed in the mock-up voloxidizer. As the results, the optimum air flow rates seem to be 20 LPM by the Stokes'; equation while they are 14.5 L/min by the density rate equation. At the experiments with the mock-up voloxidizer, a trace amount of U3O8 seems to be detectable at the air flow rate of 14.5 L/min by the density rate equation, but U3O8 powders of 7 ㎛ diameter seem detectable at the air flow rate of 20 L/min by the Stokes'; equation. It is revealed that 14.5 L/min is the optimum air flowe rate which is capable of preventing U3O8 powders from scattering in the UO2 voloxidizer and the proposeed density rate equation is proper to calculate the terminal velocity of U3O8 powders.

SiO2,UO2 pellet,Terminal velocity,U3O8,Demonstration,Voloxidizer

Isotopes of alkali and alkaline earth metals (AM and AEM) are the main contributors to the heat load and the radiotoxicity of spent fuel (SF). These components are separated from the SF and dissolved in a molten LiCl in an electrolytic reduction process. A mass transfer model is developed to describe the diffusion behavior of Cs, Sr, and Ba in the SF into the molten salt. The model is an analytical solution of Fick';s second law of diffusion for a cylinder which is the shape of a cathode in the electrolytic reduction process. And the model is also applied to depict the concentration profile of the oxygen ion which is produced by the electrolysis of Li2O. The regressed diffusion coefficients of the model correlating the experimentally measured data are evaluated to be greater in the order of Ba, Cs, and Sr for the metal ions and the diffusion of the oxygen ion is slower than the metal ions which implies that different mechanisms govern the diffusion of the metal ions and the oxygen ions in a molten LiCl.

Diffusion,Alkali metal,Alkaline Earth Metal,Spent Fuel,Molten Salt

Effects of the aluminum melting temperature, melting time and a kind of flux agents on the distribution of surrogate nuclide were investigated in the electric furnace at the aluminum melting including surrogate radionuclides(Co, Cs, Sr) in order to establish the fundamental research of the melting technology for the metallic wastes from the decommissioning of the TRIGA research reactor. It was verified that the fluidity of aluminum melt was increased by adding flux agent but it was slightly varied according to the sort of flux agents. The results of the XRD analysis showed that the surrogate nuclides move into the slag phase and then they were combined with aluminum oxide to form more stable compound. The weight of the slag generated from aluminum melting test increased with increasing melting temperature and melting time and the increase rate of the slag depended on the kind of flux agents added in the aluminum waste. The concentration of the cobalt in the ingot phase decreased with increasing reaction temperature but it increased in the slag phase up to 90% according to the experimental conditions. The volatile nuclides such as Cs and Sr considerably transferred from the ingot phase to the slag and dust phase.

Aluminum waste,Melting,Slag,Flux,Nuclide Distribution,Decontamination Factor

The electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt system has been studied in a 10 g U3O8/batch-scale experimental apparatus with an integrated cathode assembly at 650 ℃. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt% Li2O system and the U3O8-LiCl-3 wt% Li2O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

integrated cathode assembly,electrolytic reduction,cyclic voltammogram,uranium oxide,molten salt

The removal of radioactive corrosion products from the reactor coolant through a magnetic filter system is one of the many approaches being investigated as a means to reduce radiation sources and exposures to the operational and maintenance personnel in a nuclear power plant. Many research activities in water chemistry, therefore, have been performed to provide a filtration system with high reliability and feasibility and are still in process. In this study, it was devised the magnetic filter system with permanent and electric magnets to remove the corrosion products in the coolant stream taking an advantage of the magnetic properties of corrosion particles. Permanent magnets were used for separation of corrosion products and electric magnets were utilized for flocculation of colloidal particles to increase in their size. Experiments using only permanent magnets, in the previous study, displayed the satisfactory outcome of filtering corrosion products and indicated that the removal efficiency was more than 90 % for above 5 ㎛ particles. Experiments using electric magnets also showed the good performance of flocculation without chemical agents and exhibited that most corrosion particles were flocculated into lager aggregates about 5 ㎛ and over in diameter. It is, thus, expected that the magnetic filter system with the arrangement of permanent and electric magnets will be an effective way for the removal of radioactive corrosion products with considerably high removal efficiency.

radioactive corrosion products,magnetic filter system,permanent magnets,electric magnets,particle size

Methodologies of NEWTRIT model, NRC model and AIRDOS-EPA model, which are off-site dose assessment models for regulatory compliance from routine releases of tritium into the environment, were investigated. Using the domestic data, if available, the predictive results of the models were compared. Among them, recently developed NEWTRIT model considers only doses from organically bounded tritium (OBT) due to environmental releases of tritiated water (HTO). A total dose from all exposure pathways predicted from AIRDOS-EPA model was 1.03 and 2.46 times higher than that from NEWTRIT model and NRC model, respectively. From above result, readers should not have an understanding that a predictive dose from NRC model may be underestimated compared with a realistic dose. It is because of that both mathematical models and corresponding parameter values for regulatory compliance are based on the conservative assumptions. For a dose by food consumption predicted from NEWTRIT model, the contribution of OBT was nearly equivalent to that of HTO due to relatively high consumption of grains in Korean. Although a total dose predicted from NEWTRIT model is similar to that from AIRDOS-EPA model, NEWTRIT model may be have a meaning in the understanding of phenomena for the behavior of HTO released into the environment.

tritium,NEWTRIT model,NRC model,AIRDOS-EPA model,environmental release

In the stage of conceptual design for the construction and operation of the geologic repository for radioactive wastes, it is important to consider a repository ventilaton system which serves the repository working environment, hygiene & safety of the public at large, and will allow safe maintenance like moisture content elimination in repository for the duration of the repositories life, consturction/operation/closure, also allowing safe waste transportation and emplacement. This paper describes the possible ventilation system design criteria and requirements for the prospective Korean radioactive waste repositories with emphasis on the underground rock cavity disposal method in the both cases of low & medium-level and high-level wastes. It was found that the most important concept is separate ventilation systems for the construction (development) and waste emplacement (storage) activities. In addition, ventilation network system modeling, natural ventilation, ventilation monitoring systems & real time ventilation simulation, and fire simulation & emergency system in the repository are briefly discussed.

radioactive waste repository,ventilation system,repository environment & saftey,ventilation design criteria

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered at inside surface of hot cell and surface of equipment in hot cell. It was necessary to develop a particulate filtration equipment to reuse PFC solution used on PFC decontamination due to its high cost and to minimize the volume of second wastewater. Contamination characteristics of hot particulate were investigated and then a filtration process was presented to remove hot particulate in PFC solution generated through PFC decontamination process. The removal efficiency of PVDF(Poly vinylidene fluoride), PP(Polypropylene), Ceramic(Al2O3) filter showed more than 95%. The removal efficiency of PVDF filter was a little lower than those of other filters at same pressure(3 psi). A ceramic filter showed a higher removal efficiency with other filters, while a little lower flux rate than other filters. Due to inorganic composition, a ceramic filter was highly stable against radionuclides in comparison with PVDF and PP membrane, which generate H2 gas in -radioactivity atmosphere. Therefore, the adoption of ceramic filter is estimated to be suitable for the real filtration process.

PFC,hot particulate,removal efficiency,flux,ceramic(Al203) membrane

The characterization of radiological contamination inside pipes generated during the decommission of a nuclear facility is necessary before pipes can be recycled or disposed. But, existing direct measurements of radioactive contamination level using the survey-meter can not estimate the characteristic of contamination on a local area such as the pipe inside. Moreover, the measurement of surface contamination level using the indirect methods has many problems of an application because of the difficulty of collecting sample and contamination possibility of a worker when collecting sample. In this work, plastic scintillator was simulated by using Monte Carlo simulation method for detection of beta radiation emitted from internal surfaces of small diameter pipe. Simulation results predicted the optimum thickness and geometry of plastic scintillator at which energy absorption for beta radiation was maximized. In addition, the problem of scintillator processing and transferring the detector into the pipe inside was considered when fabricating the plastic detector on the basis of simulation results. The characteristic of detector fabricated was also estimated. As a result, it was confirmed that detector capability was suitable for the measurement of contamination level. Also, the development of a detector for estimating the radiological characteristic of contamination on a local area such as the pipe inside was proven to be feasible.

contamination level inside pipe,Monte Carlo simulation,beta radiation,plastic scintillator

Demand on the nuclear energy in East Asian countries has been grown rapidly to support economic development. After 9.11, nuclear security has become the world wide issue. In addition, unlike to other region, some countries are considering the introduction of nuclear power plants. To meet the challenges a new regional multilateral nuclear approach is proposed aiming at assurance of supply and non -proliferation. The new proposal is based on the principles of confidence building, volunteering, and incentives. The step wise approach is recommended to implement the multilateral system in East Asia.

MNA,Non-proliferation,nuclear cooperation,Spent Nuclear Fuel,Enrichment

Yellowish uranium compounds were enriched at the interface between a Ca-bentonite block and a waste glass, containing about 20% uranium oxide, in contact with the block due to the dissolution of uranium by a synthetic granitic groundwater in Ar atmosphere. The uranium compound formed for 6 years leach time was identified as a beta-uranophane [Ca(UO2)2(SiO3OH)25H2O] using XRD, IR and mass spectrometer. The solubility of the beta-uranophane was measured to be about 10-6mole/L in de-mineralized water at 80℃.

Uranium alteration,Waste glass,Bentonite,Uranophane,Leaching

Natural ventilation in radioactive waste repositories is considered to be less efficient than mechanically forced ventilation for the repository working environment and hygiene & safety of the public at large, for example, controlling the exposure of airborne radioactive particulate matter. It is, however, considered to play an important role and may be fairly efficient for maintaining environmental conditions of the repository over the duration of its lifetime, for example, moisture content and radon (Rn) gas elimination in repository. This paper describes the feasibility of using natural ventilation which can be generated in the repository itself, depending on the conditions of the natural environment during the periods of repository construction and operation. Evidences from natural cave analogues, actual measurements of natural ventilation pressures in mountain traffic tunnels with vertical shafts, and calculations of airflow rates with given natural ventilation pressures indicate possible benefits from passive ventilation for the prospective Korean radioactive waste repository. Natural ventilation may provide engineers with a cost-efficient method for heat and moisture transfer, and radon (Rn) gas elimination in a radioactive waste repository. The overall thermal performance of the repository may be improved. The dry-out period may be extended, and the seepage flux likely would be decreased.

radioactive waste repository,natural ventilation,rock cavern disposal method,thermodynamic method

14C and 3H in the evaporator bottom (EB) discharged from the Nuclear power plant (NPP) were extracted simultaneously into a gaseous 14CO2 and liquefied HTO by using the chemical oxidation, which is the method to oxidize samples completely using potassium persulfate and sulfuric acid. The extracted 14C and 3H were counted by the liquid scintillation counter (LSC) after the quench correction. To examine the recovery of 14C using the radioactive standards, Na214CO3, 14C-alcohol, and 14C-toluene as 14C, and HTO as 3H were used. Also, the most suitable method for oxidizing 14Ctoluene, which is difficult to be oxidized, was investigated through FT-IR spectra according to the concentration of sulfuric acid. With the identical method, 14C and 3H in the EB generated in the NPP were assayed in the range of 8.35 1.38 103 Bq/g and 2.46 × 102 ~ 1.40 × 104 Bq/g, respectively.

14C,3H,radwaste,chemical oxidation,LSC

In this study, conventional head-end processes of spent TRISO fuel have been reviewed to develope more effective treatment methods. The main concerns in the TRISO treatment are to effectively separate the carbon and SiC contained in the TRISO particles. The crush-burn scheme which was considered in the early stages of the development has been replaced by the crush-leach process because of 14C problems as a second waste being generated during the process. However, there are still many obstacles to overcome in the reported processes. Hence, innovative thermomechanical concepts and a molten salt electrochemical approach to breach the coating layers of the TRISO particle with a minimized amount of second waste are proposed in this paper and their principles are described in detail.

TRISO,HTGR,head-end processing,thermal shock,breaching

Nuclides in radioactive wastes are assumed to be transported in the geosphere by groundwater and probably discharged into the biosphere. Quantitative evaluation of doses to human beings due to nuclide transport in the geosphere and through the various pathways in the biosphere is the final step of safety assessment of the radioactive waste repository. To calculate the flux to dose conversion factors (DCFs) for nuclides appearing at GBIs with their decay chains, a template ACBIO which is an AMBER case file based on mathematical model for the mass transfer coefficients between the compartments has been developed considering material balance among the compartments in biosphere and then implementing to AMBER, a general and flexible software tool that allows to build dynamic compartment models. An illustrative calculation with ACBIO is shown.

radioactive waste repository,biosphere,ACBIO,AMBER,dose conversion factor

HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

Type B packaging,Dry spent resin,Accidental conditions,Structural integrity,Free drop,Puncture

It is necessary to forecast the various dismantling activities prior to dismantling nuclear facilities by using various software instead of a physical mock-up system because the dismantling in a contaminated with radioactivity cause the results of an unexpected situation. The component that needs to develop a dismantling mock-up system was examined. There are many component systems such as a decommissioning database system, 3D dosimetric mapping that represents a distribution of a radionuclide contamination, a component of modeling for nuclear facility and devices include the decontamination and decommissioning. The research of software architecture about these components was carried out because these component systems that have been independently doesn';t describe not only to visual an activities of Decontamination and Decommissioning(D&D) but also to evaluate it. The result was established an architecture that consist of an visualization module which could be visualized an D&D activities and a simulation module which can be evaluated a dismantling schedule and decommissioning cost.

Dismantling digital mock-up system,Dismantling database system,3D dosimetric mapping,software architecture

Thermal decomposition and stabilization characteristics of the solid cake after the dissolution of nitrate of the lagoon sludge was investigated. Most of the nitrates were dissolved in the water and removed to the filtrate, but small amount of nitrates, calcium carbonate and uranium were remained in the solid cake. The solid cake was thermally decomposed in the muffle furnace at 900 ℃ for 5 hours. Uranium, which is in the lagoon 1, was stabilized with NaNO3 decomposition to Na2O · 2UO3 form. For the lagoon 2, it is confirmed that CaO, which was created by thermal decomposition of the Ca(NO3)2 and CaCO3, was transferred to Ca(OH)2 in the air with water. Because it is known that Ca(OH)2 is stable material, further additives did not need to the stabilization of the thermal decomposition of the lagoons.

uranium conversion,thermal decomposition nitrate,sludge,stabilization

In this paper, a prototype of the Bridge-Transported Servo Manipulator (BTSM) system is introduced, which has been developed to do operation and maintenance jobs remotely in a hot cell. The system consists of a telescopic transporter, a slave arm, a master arm, and a control system. Several tests such as a positional tracking, a weight handling, reliability, and operability have been performed and test results are presented. Based on the test results, an upgraded system which will be used during demonstrations of the advanced spent fuel conditioning process (ACP) has been designed.

Teleoperation,Servo manipulator,Remote maintenance,Performances tests,ACP

In order to produce only a pH-controlled solution without discharging any unwanted solution, this work has developed a continuous electrolytic system with a pH-adjustment reservoir being placed before an ion exchange membrane-equipped electrolyzer, where as a target solution was fed into the pH-adjustment reservoir, some portion of the solution in the pH-adjustment reservoir was circulated through the cathodic or anodic chamber of the electrolyzer depending on the type of the ion exchange membrane used, and some other portion of the solution in the pH-adjustment reservoir was discharged from the electrolytic system through the other counter chamber with its pH being controlled. The internal circulation of the pH-adjustment reservoir solution through the anodic chamber in the case of using a cation exchange membrane and that through the cathodic chamber in the case of using an anion exchange membrane could make the solution discharged from the other counter chamber effectively acidic and basic, respectively. The phenomena of the pH being controlled in the system could be explained by the electro-migration of the ion species in the solution through the ion exchange membrane under a cell potential difference between anode and cathode and its consequently-occurring non-charge equilibriums and electrolytic water- split reactions in the anodic and cathodic chambers.

pH-control electrolyzer,ion exchange membrane,electro-migration,electrolytic watersplit reaction

To support the design concept and the performance assessment of the cover system for low- and intermediate-level radioactive waste(LILW) disposal facility, a pioneering study is conducted for the tomb of historical age. Research status of the art are investigated and the characteristics of tomb cover are summarized based on the preservation status of historical remains. On-site soil samples are prepared and their unsaturated hydraulic conductivities are measured by an one-step outflow method. Visiting the excavation site of historical tomb and communication with Korean archeological society are required for the further understanding and for the extension to the radioactive waste disposal research.

near-surface disposal facility,tomb,unsaturated soil

Ultrasonic decontamination of the type 304 stainless steel specimen loosely contaminated with Eu2O3 powders was investigated. Decontamination factors (DFs) by the three kinds of ultrasonic media such as water, pure PFC (Perfluorocarbon, C7F16) and a mixed solution of 99.9 vol% PFC and 0.1 vol% anionic surfactant were determined. The determined DF values were 20, 50 and 200, respectively. This significant difference in the decontamination factors for the different decontamination solution was well explained by the surface tension of the media as well as the interaction between the positively charged surface of Eu2O3 powders and the anionic surfactant. Ultrasonic decontamination behavior of the loosely contaminated metal specimens such as plate, pipe, welding specimen and crevice specimen in the mixed solution of PFC and anionic surfactant was also investigated. The contaminants were completely removed for the tested specimens except for the longest specimen. For 6-cm long pipe specimen, however, 98.5 % of the contaminants were removed.

PFC,Ultrasonic,Decontamination,metal,distillation,PFC

The morphology of the high burnup UO2 spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI';s PIEF. The test specimen was used with about 200 of the spent UO2 fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at 1400 for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to 500 and held for 3hrs in the mixed gas (60 He and 100 STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the UO2 matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to 10.

High burn-up fuel,oxidation and annealing test,post-irradiation,fission product,micro-structure

A study on the separation of 99Tc, 94Nb, 55Fe, 90Sr and 59/63Ni in various radioactive wastes discharged from nuclear power plants has been performed for a use in their quantification which is indispensible for the evaluation of the radionuclide inventory. Ni was recovered along with Ca, Mg, Al, Cr, Ti, Mn, Ce, Na, K, and Cu through the sequential separation procedure of Re(as a surrogate of 99Tc), Nb, Fe and Sr by anion exchange and Sr-Spec extraction chromatography. In this research, chemical separation of Ni from the co-existing elements was investigated by cation exchange and Ni-Spec extraction chromatography. Precipitation behaviour of Ni and the co-existing elements with dimethylglyoxime(DMG) was investigated in ammonium citrate/ethanol-H2O and tartaric acid/acetone-H2O in order to purify separated Ni fractions and to prepare 59/63Ni source for the radioactivity measurement using a gas proportional counter. Recovery of Ni separated through ion exchange chromatographic separation procedure was 92.1% with relative standard deviation of 0.9%. In addition, recovery of Ni with DMG in the tartaric acid/acetone-H2O was 85.6% with relative standard deviation of 1.9%.

Radioactive wastes,59/63Ni,Ion exchange chromatography,Extraction chromatography,Precipitation

This paper describes a compartment dynamic model for evaluating the contamination level of tritium in agricultural plants exposed by accidentally released tritium. The present model uses a timedependent growth equation of plant so that it can predict the effect of growth stage of plant during the exposure time. The model including atmosphere, soil and plant compartments is described by a set of nonlinear ordinary differential equations, and is able to predict time-dependent concentrations of tritium in the compartments. To validate the model, a series of exposure experiments of HTO vapor on Chinese cabbage and radish was carried out at the different growth stage of each plant. At the end of exposure, the tissue free water(TFWT) and the organically bound tritium(OBT) were measured. The measured concentrations were agreed well with model predictions.

tritium,compartment dynamic model,agricultural plants

To identify the effect of engineered barriers on the leach rate of cesium from spent PWR fuel under a synthetic granitic groundwater, the related leach tests with and without bentonite or metals have been performed up to about 6 years. The leach rates were decreased as a function of leaching time and then became a constant after a certain period. The period in a bare spent fuel was much longer than that with bentonite or metal sheets. The cumulative fraction of cesium released from the spent fuel with bentonite or with copper and stainless steel sheets was steadily increased, but the fraction from bare fuel was rapidly and then sluggishly increased. However, the values deducted its gap inventory from the cumulative fraction of cesium released from the bare fuel was almost very close to the others. These suggest that the initial release of cesium from bare fuel might be dependant on its gap inventory and the effect of engineered barriers on the long-term leach rate of cesium would be insignificant but the rate with engineered barriers could be reduced in the initial transient period due to their retardation effect. And the long-term leach rate of cesium from spent fuel in a repository would be approached to a constant rate of 2 10-2 g/ -day.

leach rate,spent fuel,cesium,bentonite,metal sheets,repository

The forward dissolution rate of a borosilicate waste glass was determined as an interlaboratory study(ILS) testing program for the evaluation of precision in the measurement of the dissolution rate of a waste glass using a single-pass flow-through(SPFT) test, whose conducting practice has been written for standardization through American Society for Testing and Materials (ASTM). A simulated low-activity waste glass powder with a size of 100/200 mesh was dissolved by lithium buffer solution (pH=10) at 70? under Ar atmosphere. By plotting the dissolution rates as a function of silicon and boron concentration in eluate, the forward dissolution rate of the glass was obtained as about 2.7 10-5 g m-2 s-1 in our laboratory.

SPFT test,forward dissolution rate,waste glass

Bentonite has been considered as a candidate buffer material in the underground repository for the disposal of high-level radioactive waste because of its low permeability, high sorption capacity, self sealing characteristics, and durability in nature. In this study, the potential for separation of bentonite particles caused by the groundwater erosion was studied experimentally for a Korean Cabentonite under the relevant repository conditions. Results showed that bentonite particles can be generated at the bentonite/granite interface and mobilized by the water flow although the intrusion of bentonite into fracture by swelling pressure was observed to be small. Different processes of mobilization of theses colloids from the compacted bentonite block have been identified in this study. The concentration of particles eluted in water was increased as the flow rate increased. Thus the result reveals that the erosion of the bentonite surface due to the groundwater flow together with intrusion processes is the main mechanism that can mobilize bentonite colloids in the fracture of the granite.

Calcium Bentonite,Separation of Bentonite Particles,Erosion of Bentonite,Stability of Bentonite Colloids,Rock/Buffer Interface

In this study, analysis of the disposal tunnel spacing and disposal pit pitch was carried out, as a factor of the design to estimate the scale and layout of the repository. To do this, based on the reference repository concept and the engineered barrier concept, several cross sections of the disposal tunnel and disposal pit were established. After then, the mechanical and thermal stabilities of the established tunnels were analyzed. Also, an optimized disposal tunnel spacing and the disposal pit pitch reducing the excavation volume was proposed. The results of these analyses can be used in the deep geological repository design. The detailed analyses by the exact site characteristics data to reduce the uncertainty of the site and the modification for the optimization are required.

Deep Geological Disposal,HLW,Repository Layout,Excavation Volume,Tunnel Spacing,Pit Pitch

The strong protection method of radiation emergency preparedness is the evacuation when a great deal of radionuclide material is released to environment. Required factors for evacuation time estimate of Ulchin nuclear power plant site were investigated. The traffic capacity and the traffic volume by season were investigated for the traffic analysis and simulation within EPZ of Ulchin site. As a result, the background traffic volume by season were established. The NETSIM code was applied to simulate for 12 scenarios in the event of normal traffic/summer peak traffic/winter peak traffic, daytime/night, and normal weather/adverse weather conditions. The results showed that the evacuation time required for total vehicles to move out from EPZ took generally 210 315 minutes. The evacuation time took longer about 45 minutes at night than in the daytime, and 45 minutes in adverse weather than normal condition.

Public Evacuation,Nuclear Emergency,Emergency Preparedness,Traffic Analysis,NETSIM

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be 2.10×10-3, 2.97×10-3 and 1.01×10-1 mSv/h, respectively. And those calculated by MCNP-4C are 1.60×10-3, 2.99×10-3 and 7.88×10-2 mSv/h, respectively. The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01mSv/h for the operation area and 0.15mSv/h for the service (maintenance) area.

Shielding,Shielding Design,Hot Cell,QAD-CGGP,MCNP,Effective Dose

This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 under the normal condition. The off-normal condition has an environmental temperature of 38 . An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.

thermal analysis,spent fuel,storage cask,heat transfer,natural convection,thermal radiation,fluid flow

Engineered barrier test facility is specially designed to demonstrate the performance of engineered barrier system for the near-surface disposal facility under the domestic environmental conditions. Comprehensive measurement systems are installed within each test cell. Long- and short-term monitoring of the multi-layered cover system can be implemented according to different rainfall scenarios with artificial rainfall system. Monitoring data on the watercontent, temperature, matric potential, lateral drainage and percolation of cover-layer system can besystematically managed by automatic data acquisition system. The periodic measurement data are collected and will beanalyzed by a dedicated database management system, and provide a basis for performance verification of the disposal cover design.

Engineered barrier,multi-layered cover system,water content,matric potential,lateral drainage,percolation

Final disposal of radioactive waste generated from Nuclear Power Plant (NPP) requires the detailed infonnation about the characteristics and the quantities of radionuclides in waste package. Most of these radionuclides are difficult to measure and expensive to assay. Thus it is suggested to the indirect method by which the concentration of the Difficult-to-Measure (DTM) nuclide is estimated using the correlations of concentration - it is called the scaling factor - between Easy-to-Measure (Key) nuclides and DTM nuclides with the measured concentration of the Key nuclide. In general, the scaling factor is detennined by the log mean average (LMA) method and the regression method. However, these methods are inadequate to apply to fission product nuclides and some activation product nuclides such as 14C 90Sr. In this study, the artificial neural network (ANN) method is suggested to improve the conventional SF detemúnation methods - the LMA method and the regression method. The root mean squared eπors (RMSE) of the ANN models are compared with those of the conventional SF detenrúnation models for 14C and 90Sr in two parts divided by a training part and avalidation part. The SF detemúnation models are arranged in the order of RMSEs as the following order: ANN model

scaling factor,artificial neural network,inventory,LMA,regression

The cost-effective reduction of occupationa1 radiation dose (ORD) at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORD data of existing plants. Through the data ana1ysis, it is required to identify what are the jobs of repetitive high ORD at the nuclear power plant. In this study, Percentile Rank Sum Method (PRSM) is proposed to identify repetitive high ORD jobs, which is based on non-parametric statistical theory. As a case study, the method is applied to ORD data of maintenance and repair jobs at Kori units 3 and 4 that are pressurized water reactors with 950 MWe capacity and have been operated since 1986 and 1987, respectively in Korea. The results was verified and va1idated, and PRSM has been demonstrated to be an efficient method of analyzing the data.

Occupational radiation dose,ALARA,Non-parametric statistica1 method

A variety of factors such as the pattern of intake (acute or chronic), monitoring interval and the characteristics of the radionuclides could have a significant influence on the estimates for the intake and internal dose. The relative differences of the assessed intakes based on the assumption of an acute intake to that of a chronic intake were evaluated by using the predicted bioassay quantity in the whole body or organs for an acute and chronic intake through the inhalation of 125I, 137C, 235U with the AMAD of 1 ㎛ and 5 ㎛ for the monitoring intervals of 7, 14, 30, 60, 90, 120, 180, 360 days, respectively. The relative difference of the assessed intakes based on the intake pattern is affected by the monitoring interval, radionuclide and absorption type, but the particle size has little influence on the difference of the assessed intakes based on the intake pattern. The maximum monitoring interval, which is defined as the monitoring interval that the relative difference of the assessed intakes based on the assumption of an acute intake to that of a chronic intake is less than 10%, is 60 d for 125I with Type F, 180 d for 137C with Type F, 90 d for 235U with Type M, and 360 d for 235U with Type S. It was concluded that an intake pattern has little influence on the estimates of the assessed intake in the case where the monitoring interval is shorter than the maximum monitoring interval for each radionuclide.

Acute Intake,Chronic Intake,Monitoring Interval,and Relative Difference

The exposure dose fonn recycling of a 1arge amount of the stee1 scrap from the KRR-1&2 decommissioning activities was evaluated, and also the clearance 1eve1(draft) was derived. The maximum individual dose and collective dose were evaluated by modifying intemal dose conversion factor which was based on the concept of effective dose in ICRP 60, app1ied to the RESRAD-RECYCLE ver 3.06 computing code, lAEA Safety Series 111-P-1.1 and NUREG-1640 as the assessment too1. The result of assessment for individua1 dose and collective dose is 23.9 μSv per year and 0.11 man · Sv per year respectively. The clearance levels were ultimately detennined by extracting the most conservative value form the results of the generic assessment and specific assessment methodo1ogies. The result of clearance level for radionuclides(Co60, CS137) is 1ess than 1.14x10-1 Bq/g to comp1y with the clearance criterion(maximum individua1 dose : 10 IlSv per year, collective dose : 1 man · Sv per year) provided for Korea Atomic Energy Act and re1evant regulations.

C1earance Leve1,Stee1 Scrap,KRR-1&2,Individual dose,Col1ective dose

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-l. To calculate neuσon flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95 % of the total activity in RPV came from the nuclides of 56Fe, 60CO, 59Ni and 63Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25x106 GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was performed. The comparison results showed a good agreement.

Decommissioning,DORT,ORIGEN2,Activation Products Inventory

Regulations and guidelines for radioactive waste disposal require detailed infonnation about the characteristics of radioactive waste drums prior to transport to the disposal sites. However, estimation of radionuclide concentrations in the drummed radioactive waste is difficult and unreliable. In order to overcome this difficulty, scaling factor (SF) method has been used to assess the activities of radionuclides, which could not be directly analyzed. A radioactive waste assay system has been operated at Korean nuc1ear power plant (KORI site) since 1996 and consolidated SF concept has played a dominant role in the detenIlÍnation of radionuclide concentrations. However, SFs are somewhat dispersive and limited in KORI site. Therefore establishment of the assay system using more improved SFs is planned and progressed. In this paper, the scope of research is briefly introduced. For the selection of more reliable activity determination method, the accuracy of predicted SF values for each activity determination method is compared. From the comparison of each activity determination method, it is recommended that SF detenuination method should be changed from the arithmetic mean to the geometrical mean for more reliable estimation of radionuclide activity. Arithmetic mean method and geometric mean method are compared based on the data set in KORI system. And, this change of SF determination method will prevent an inordinate over-estimation of radionuclide inventory in radwaste drum.

scaling factor,radioactive waste,activity,inventory,arithmetic mean,geometric means

Studies were conducted to select the candidate buffer material for a high-level waste (HLW) repository in Korea. This paper presents the hydraulic properties, the swelling properties, the thermal properties, and the mechanical properties as well as the radionuclide release-retarding capacity of Kyungju bentonite as part of those studies. Experimental results showed that the hydraulic conductivities of the compacted bentonite were very low and less than 10-11m/s. The values decreased with increasing the dry density of the compacted bentonite. the swelling pressures were in the range of 0.66 MPa to 14.4 MPa and they increased with increasing the dry density. The themal conductivities were in the range of 0.80 kcal/mh℃ to 1.52 kcal/mh℃ . The unconfined compressive strenth, Young’s modulus and Poison’s ratio showed the range of 0.55 MPa to 8.83 MPa, 59 MPa to 1275 MPa, and 0.05 to 0.20, respectively, when the dry densities of the compacted bentonite were 1.4 Mg/㎥ to 1.8 Mg/㎥ . The diffusion coeffcients in the compacted bentonite were measured under and oxidizing condition. The values were 1.7 x 10-10 ㎡/s to 3.4 x 10-10 ㎡/s for electrically neutral tritium (H-3), 8.6 x 10-14 ㎡/s to 1.3 x 10-12 ㎡/s for cations (Cs, Sr, Ni), 1.2 x 10-11 ㎡/s to 9.5 x 10-11/s for anions (I, Tc), and 3.0 x 10-14 ㎡/s to 1.8 10-13 ㎡/s for actinides (U, Am), when the dry densities were in the range of 1.2 Mg/㎡ to 1.8 Mg/㎡ . The obtained results will be used in assessing the barrier properties of Kyungju bentonite as a buffer material of a repository in Korea.

Barrier properties,Bentonite,Buffer,HLW Repository

Neutron induced prompt gamma-ray spectroscopy(NIPS) system measures the prompt gammaray emitting by the interaction of a neutron with various materials. This system will be of great benefit to scientists worldwide, since it provides the non-destructive measurement of many element in either solid or liquid wastes. In this study, the full-energy-peak (FEP) efficiency calibration for a HPGe detector was constructed in the -ray energy range from 80 keV to 8 MeV, using 133Ba and 152Eu RI sources and 35Cl(n, )36Cl thermal neutron captured reaction. The FEP efficiency curve for the higher energies using the 35Cl(n, )36Cl reaction was normalized with the curve obtained from the RI sources, since the accurate activity of its prompt γ-ray is unknown. The average thermal neutron flux was theoretically calculated using the FEP efficiency curve for the KCl standard solutions. The NIPS system equipped with a γ-γ coincidence setup with two n-type coaxial HPGe detectors was considered in order to reduce the interfering γ-ray background. The FEP efficiency curve for the γ-γ coincidence system was also obtained for full energy range. The performance of the normal and coincidence NIPS system was tested by comparing signal-to-noise ratio in each mode using the reference sample.

Efficiency calibration,Neutron,Prompt gamma-ray,Coincidence,HPGe detector

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP';s) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

HLW disposal,Scenario,FEP,Total system performance assessment,Initial container failure,Dose,MASCOT-K

A computer software program, called BIDAS (BIoassay Data Analysis Software) is developed to interpret the bioassay measurement data in terms of intakes and the committed effective dose using models currently recommended by the International Commission on Radiological Protection (ICRP) to describe the behavior of the radioactive materials within the body. The program consists of three modules; first, a database module to manage the bioassay data, second, another databasee module to store the predicted bioassay quantities of each radionuclide and finally, a computational module to estimate the intake and committed effective dose calculated with the bioassay quantity measurement values from either an acute or chronic exposure of the radionuclies within the body. This paper describes the features of the program as well as the quality assurance check results of the BIDAS software program.

Bioassay data,Biokinetic models,Intake,Committed effective dose,Acute exposure,Chronic exposure

Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 kW/m of maximum linear power and 1,770 MWd/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 ㎛ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2 2.5 ㎛ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.

EPMA,nuclear fuel,post irradiation examination,DUPIC

An experimental study on the sorption of U(VI) onto a Korean granite was performed as a function of the geochemical parameters such as contact time, pH, ionic strength, and carbonate concentration using a batch procedure. The distribution coefficient,, was about 1-200 mL/g depending on the experimental conditions. The sorption of U(VI) onto granite particles was greatly dependent upon the contact time, pH, and carbonate concentration, but insignificantly dependent on the ionic strength. It was noticed that the sorption of U(VI) onto granite particles was highly correlated with the uranium speciation in the solution, which was dependent on the pH and carbonate concentrations. It was deduced from the kinetic sorption experiment that a two-step first-order kinetic behavior could dominate the kinetic sorption of U(VI) onto granite particles. In the alkaline range of a pH above 7, U(VI) sorption was greatly decreased and this might be due to the formation of anionic U(VI)-carbonate aqueous complexes as predicted by the speciation calculations.

Granite,Sorption,Uranium(VI),Geochemical parameters,Distribution coefficient

H2SO4 and citric acid were used as additives for the electrokinetic remediation experiment to increase removal efficiency of Cs and Co from the radioactive soil waste stored for more than 10 years. The average effluent velocity discharged from the elctrokinectic remediation experimental column was 2.010 cm/min and the discharged soil wastewater volume for 10 days is 3.6 pore volume of the column. 97% of Co in the column was decontaminated for 10 days of operation, while only 54% of Cs was decontaminated. These results are considered that the absorption equilibrium coefficient of Cs is higher than that of Co. The predicted values of the residual concentration by the proposed mathematical model were well coincided with the experimental results within the experimental error range

Electrokinetic Soil Remediation,Model,H2SO4,Citric acid,137Cs,60Co

Advanced spent fuel Conditioning Process(ACP) project which is under development for efficient spent fuel management has finished process feasibility study and is preparing type hot cell construction for process experimentaton. Radiation dose evaluation for the radioactive nuclides were preliminarily performed for normal operation and accident case with the basic concept design report, the meteorological data and the recent site specific data. According to the production and release rate of nuclides, dose evaluations for residents around facility were performed. The evaluation result shows a safe margin for regulation limits and SAR(Safety Analysis Report) limit of IMEF(Irradiated Material Examination Facility) where this facility will be constructed.

ACP,hotcell,dose evaluation,IMEF

In order to simultaneously vitrify the ion exchange resin(IER) and combustible dry active waste(DAW) generated from Korean nuclear power plants, a vitrification pilot test was conducted using an induction cold crucible melter(CCM). The energy necessary for startup of the glass using a Ti-ring was evaluated as about 290 kWh. The power supplied from a high frequency generator to melt the glass properly was ranged from 160 to 190 kW without any interruption. When the mixture of the IER and DAW was fed into the CCM, the concentration of CO was lowered up to 1/40 compared to feeding the IER solely. It may be caused by the DAW which can produce about 1.8 times higher heat compared to the IER. When the swelling phenomenon occurred in the glass melt, the concentration of NO2, oxidizing gas, was higher than NO, reducing gas. Total feed amounts of the IER and DAW were 368 and 751 kg, respectively. And then, about 74 of volume reduction factor was achieved.

Vitrification,DAW,IER,Ignition,Swelling

In order to vitrify the combustible dry active waste (DAW) generated from Korean Nuclear Power Plants, a glass formulation development based on waste composition was performed. A borosilicate glass, DG-2, was formulated to vitrify the DAW in an induction cold crucible melter (CCM). The processability, product performance, and volume reduction effect of the candidate glass were evaluated using a computer code and were measured experimentally in the laboratory and CCM. The glass viscosity and electrical conductivity as the process parameters were in the desired ranges. Start-up and maintaining glass melt of the candidate glass were favorable in the CCM. The product of the glass product such as chemical durability, phase stability, and density was satisfactory. The vitrification process using the candidate glass was also evaluated assuming that it was operated as economically as possible.

dry active waste (DAW),glass formulation,CCM,viscosity and electrical conductivity

The contaminative influence of beef due to the inhalation of air and the ingestion of soil of livestock, both of which are dealt with as minor contaminative pathways in most radioecological models but may not be neglected, was comprehensively investigated with the improvement of the Korean food chain model DYNACON. As the results, it was found that both pathways can not be neglected at all in the contamination of beef in the case of an accidental release during the non-grazing period of livestock. The ingestion of soil was more influential in the contamination of beef than the inhalation of air over most time following an release. If precipitation is encountered during an accidental release, contaminative influence due to the ingestion of soil was far greater compared with the cases of no precipitation. This fact was more distinct for a long-lived radionuclide 137Cs than a short-lived radionuclide 131I (elemental iodine). Compared with the results for milk performed prior to this study, the contaminative pathways due to the inhalation of air and the ingestion of soil were more important in beef because of longer biological half-lives. On the other hand, in the case of an accidental release during the grazing period of livestock, radioactive contamination due to the ingestion of pasture was dominant irrespective of the existence of precipitation during an accidental release. It means that contaminative influence due to the inhalation of air and the ingestion of soil is negligible, like the cases of milk.

dynamic food chain model,livestock,beef,inhalation of air,ingestion of soil

A bridge transported servo manipulator (BTSM) system for the advanced spent fuel conditioning process (ACP) has been developed to overcome the limitation of access, which is a drawback of mechanical master-slave manipulators (MSM) for the equipment maintenance. The servo manipulator is composed of a slave manipulator attached to the telescoping tubesets equipped with the overhead bridge installed at a hot cell and a master manipulator installed at an out-of-hot cell. Each manipulator has 7 degrees-of-freedom (DOF): a body rotation, an upper-arm tilt, a lower-arm tilt, a lower-arm rotation, a wrist pan & tilt, and a grasp motion. A wire-driven mechanism for a lower-arm rotation, a wrist pan and tilt, and a grasp motion of the manipulator has been adopted to increase the handling capacity compared to the manipulator weight and decrease the friction. The main disadvantage of the wire-driven mechanism is that if one link is in motion, other links can be affected. In this paper, the transmission characteristics among the wire-driven links have been formulated to overcome this drawback. The unexpected behaviors are confirmed by analyses of transmission characteristics as well as experiments. Also, the experimental results show that the unexpected behaviors are greatly decreased by the proposed compensation equations.

ACP,MSM,Bridge transported,Servo manipulator,Wire-driven,Equipment maintenance

Important operation parameters and performance of a high temperature ceramic candle filter system were evaluated through a series of demonstration tests at a pilot-scale vitrification plant. At the initial period of each test, due to the growth of dust cake on the surface of ceramic candles, the pressure drop across the filter media increased sharply. After that it became stable to a certain range and varied continuously proportion to the face velocity of off-gas. On the contrary, at the initial period of each test, the permeability of filter element decreased rapidly and then it became stable. Back flushing of the filter system was effective under the back flushing air pressure range of 3 5 bar. Based on the dust concentrations measured by iso-kinetic dust sampling at the inlet and outlet point of HTF, the dust collection efficiency of HTF evaluated. The result met the designed performance value of 99.9%. During the demonstration tests including a hundred hour long test, no specific failure or problem affecting the performance of HTF system were observed.

Radioactive Waste Vitrification,High Temperature Filter,Ceramic Candle,Face Velocity,Permeability,Pressure Drop

A Li recovery technology has been developed and related experimental verification efforts were carried out to improve the economical viability and environmental friendliness of the ';Advanced Spent Fuel Conditioning Process'; being developed at KAERI. This technology is characterized by the combination of 1) the electrolysis of Li2O in a molten salt by using a porous non-conducting magnesia container at the cathode, 2) the separation of the Li in the container from the molten salt by elevating the container above the level of a molten salt, 3) the transport of the Li in the container by using a vacuum siphon to a separated reservoir. Li was semi-continuously recovered from a LiCl-Li2O molten salt with a more than 95% yield by using the developed technology.

Li Recovery,Electrowinning,LiCl,Li2O,Molten Salt

Most of the low-level liquid radioactive wastes generated from PWR plants are classified into high or low total suspended solid(HTDS or LTDS), and into radiochemical and radioactive laundry waste. Although the evaporation process has a high decontami- nation ability, it has several problems such as corrosion, foam, and congestion. A new liquid waste disposal process using the ion-exchange demineralizer(IED), instead of the current evaporation process, has been introduced into the Yonggwang NPP #5 and 6 . These two methods have been compared to understand the differences in this study. Aspects compared here were the released radioactivity amount of the liquid radioactive wastes, the dose of off-site residents, the decontamination factor, and the amount of the solid radioactive wastes. The IED system is designed to discharge higher radioactivity about 20% than the evaporating system, and the actual radioactivity released from the evaporating and IED system were 0.473mCi and 1.098mCi, respectively. The radioactivity released from the IED was 2.32 times higher than that of the evaporating system. The dose of off-site residents was 2.97×m㏜ for the evaporating system, and 6.47×m㏜ for IED. The decontamination factor(DF) of the evaporator is, in most cases, far lower than the lower limits of detection(LLD) with the Ge-Li detector. Due to the low concentration of the liquid wastes collected from the liquid waste system, the decontamination factor of IED is very low. Since there is not enough data on the amount of solid radioactive wastes generated by the evaporation system, the comparison on these two systems has been conducted on the basis of the design, and the comparison result was that the evaporating system generated more wastes about 40% than IED.

on-exchange demineralizer(IED),evaporator,liquid radioactive waste,dose of off-site residents,decontamination factor,Treatment

The conditions are obtained for the decomposition of solid radioactive wastes, including ion exchange resin, zeolite, charcoal, and sludge from nuclear power plant. In the process of decomposing the radioactive wastes was used the microwave acid digestion method with mixed acid. The solution after acid digestion by the following method was colorless and transparent. Each solution was analyzed with ICP-AES and AAS and the recovery yield for 5 different elements added into the simulated radioactive wastes were over 94 %. The elemental analysis of destructive low and medium level radioactive wastes by the proposed microwave acid digestion conditions concerning the chemical characteristics of each radioactive waste are expected to be useful basic data for development of optimal glass formulation.

Solid radioactive wastes,Microwave acid digestion,Inductively coupled plasma atomic emission spectrometry,Pre-treatment,Vitrification technology

In order to enhance an oxidation resistance of the pure uranium metal under air condition, a small quantity of niobium(Nb) which is known to mitigate metal oxidation is added into uranium metal as an alloying element. A simulated metallic uranium alloy, U-Nb has been fabricated and then oxidized in the range of 200 to 300 ℃ under the environment of the pure oxygen gas. The oxidized quantity in terms of the weight gain(wt%) has been measured with the help of a thermogravimetric analyzer. The results show that the oxidation resistance of the U-Nb alloy is considerably enhanced in comparison with that of the pure uranium metal. It is revealed that the oxidation resistance of the former with the niobium content of 1, 2, 3 and 4 wt% is : 1) 1.61, 7.78, 11.76 and 20.14 times at the temperature of 200 ℃ ; 2) 1.45, 5.98, 10.08 and 11.15 times at 250 ℃ ; and 3) 1.33, 4.82, 8.87 and 6.84 times at 300 ℃ higher than that of the latter, espectively. Besides, it is shown that the activation energy attributable to the oxidation is 17.13 21.92 kcal/mol.

Oxidation Kinetics,Uranium,Uranium Alloy,Spent Nuclear Fuel,Oxidation Resistance

Soil washing equipment was developed for decontamination of radioactively contaminated soil generated during normal operation or decommissioning and verification experiments were performed. Decontamination efficiency above 80% was achieved. In case of low radiation level and large particle size, decontamination efficiency was higher. According to the ratio of volume of water to soil quantity, decontamination efficiency was higher in case of high radiation level. Redecontamination using decontaminated soil was effective in case of small particles. Using soil washing equipment, radioactivity of contaminated soil generated in nuclear power plant can be decreased and volume of soil for disposal can be decreased. And this equipment can be used in decommissioning.

contaminated soil,Cs,particle separation,soil washing,decontamination equipment

Construction of tunnels in a deep crystalline host rock for a potential High-Level Radioactive Waste(HLW) repository inevitably generates an excavation disturbed zone (EDZ). There have been a series of debates on whether a rmeability in an EDZ increases or not and what would be the maximum depth of an EDZ. Recent studies show mixed opinions on permeability. However, there has been an international consensus on the thickness of an EDZ; 30 cm for TBM and 1 meter for controlled blast. One of the impacts of an EDZ is on determining the distance between adjacent deposition holes. The void gap by the excavation hinders relaxation of temperature profiles so that the current Korean reference designing distance between holes should be stretched out more to keep the maximum temperature in a buffer region below 100 degrees Celsius. The other impact of an EDZ is on the long-term post closure radiological safety. To estimate the impact, the reference scenario, the well scenario, is chosen. Released nuclides diffuse through a bentonite buffer region experiencing strong sorption and reach a fracture surrounded by a porous medium. Inside a fractured porous region, radionuclides migrate by advection and dispersion with matrix diffusion into a porous medium. Finally, they reach a well assumed to be a source of potable water for local residents. The annual individual dose is assessed on this well scenario to find out the significance of an EDZ. A profound sensitivity study was performed, but all results show that the impact is negligible. Even though the role of an EDZ turns out to be limited on overall safety assessment, still it is worthwhile to study the chemical role of an EDZ, such as a potential source for natural colloids, potential sealing of an open fracture by fine clay particles generated by the process of an EDZ, and alteration of a sorption mechanism by an EDZ in the future.

EDZ (Excavation Disturbed Zone),High-Level Radioactive Waste Disposal,Total System Performance Assessment,Well Scenario

During the last two years, Site Information and Total Environmental database management System (SITES) ver. 1.0 has been developed for the systematic SITES Database Module (SDM), which includes site information, facility information and environmental information. The SITES includes the module for site environmental monitoring system and safety assessment (M&A) system for the nuclear facility. SITES is expected to be an effective system for the radioactive waste disposal management facility. Currently, SITES ver. 2.0 is under development after the SITES ver. 1.0 that is focused on the M&A system. The main purpose of this paper is to introduce and try to account for the major development in the concept of SEMS sub-module of the M&A module. The SEMS is purposed of development of the program for real time environmental monitoring, prediction, and automatic alarm system using SITES Database and related information.

SITES,SAS,SEMS,M&A,Information Processing Module,GIS Module

This paper shows our experiment is performed to understand the exposure tendency of 137Cs according to the height of area and also, to supplement it by considering chemical characters of 137Cs exposed to the soil. The samples we use for this experiment are from the general flat area of Yeonggwang county where it has NPPs, the high places of Keumjung & Bulgap mountains, and Naejan mountain where it is quite far from the NPPs. The data from this experiment show that the exposure of 137Cs is not harmful since its range is around 252 Bq/kg-dry in most of sampled soils such as from the general flat area, the high place of Keumjung mountain where is 2 km away from the NPPs, the other high place of Bulgap mountain where is about 20 km away from the NPPs, and Naejan mountain where it is far from the NPPs. Not like the general flat area, however, the data show that the higher the area is the more 137Cs is exposed. That is, at the top of mountains, the more 137Cs is exposed compared to at the bottom area. It is almost 2 6 times more than the general flat area of Yeonggwang county where it has NPPs. The data also show that the spread of 137Cs is deeply related to the geographical(the height of area, rainfall, etc..) factors and chemical factors of soils. As the geographical factors, there are far more chances to be exposed of 137Cs at the high area of mountains through the air compared to at lower area and therefore, we can get more high-leveled readings of 137Cs at the high area while it is low-leveled ones at the general flat area even if both of them have the same soil conditions. Regarding the chemical factors of soil, it is clarified that the CEC is the key factor. The CEC means the capability of sticking 137Cs accumulated into the soil. Hence, the more CEC it has the more high-leveled readings of 137Cs we get under the same geographical condition.

137Cs,soil,CEC,Activity,geographical factors

The underground research tunnel is essential to validate the integrity of a reference high-level waste disposal system, and the safety of geological disposal. In this study, a basic design of an underground research tunnel (URT) was tried to be developed. The candidate site for URT was described briefly, and it was intended to suggest the basic concept of the underground research tunnel. In order to develop the design of URT based on the basic concept, design equirements were established. Based on the basic concept and the design requirements, the basic design of URT was performed. Research items to be studied in the URT were also derived in this study.

underground research tunnel,high-level waste,disposal system,basic design,design requirements

As a follow up to the Agenda 21';s policy statement for safe management of radioactive waste adopted at Rio Conference held in 1992, the UN invited the LAEA to develop and implement indicators of sustainable development for the management of radioactive waste. The IAEA finalized the indicators in 2002, and is planning to calculate the member states'; values of indicators in connection with operation of its Net-Enabled Waste Management Database system. In this paper, the basis for introducing the indicators into the radioactive waste management was analyzed, and calculation methodology and standard assessment procedure were simply depicted. In addition, a series of innate limitations in calculation and comparison of the indicators was analyzed. According to the proposed standard procedure, the indicators for a few major countries including Korea were calculated and compared, by use of each country';s radioactive waste management framework and its practices. In addition, a series of measures increasing the values of the indicators was derived so as to enhance the sustainability of domestic radioactive waste management program.

radioactive waste,indicator,sustainable development,preliminary calculation

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2SO4 electrolyte, a NaNO3 neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na2SO4 solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO3 on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO2, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO3 solution with the current density of 100 mA/cm2. It was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO3 neutral salt electrolyte by reducing and radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

Electrochemical Decontamination,Neutral Salt Electrolyte,Uranium Compound,Current Density,Metallic Waste,Radioactivity

This study proposed a new electrolytic reduction technology that is based on the integration of simultaneous uranium oxide metallization and Li2O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li2O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt, increase of metallization yield, and simplification of process.

Electrolytic Reduction Technology,Uranium Oxide,Uranium Metal,Lithium Chloride,Molten Salt

To understand the long-term leach behavior of a borosilicate waste glass in a repository, the leaching experiment with three kinds of simulated borosilicate waste glasses has been carried out since the middle of 1997. The five years results indicate that a boron would be applied as an indicator of a long-term leaching of their borosilicate waste glasses and that their long-term leach rates have a tendency to be close to about 0.03g/m2-day even though their compositions and their ratios of the surface area to the volume of leachate are different.

leach,borosilicate glass,repository condition,long-term,indicator

In this study, the maintenance process by the servo manipulator has been developed for the parts of the equipment, which are unable to reach out by the Master-Slave Manipulator(MSM) in a hot cell. To do this, a virtual mock-up is implemented using the virtual prototyping technology. Using this mock-up, the workspace of the manipulators in the hot cell and the operator';s view through the wall-mounted lead glass have been analyzed. In addition, the path planning of the servo manipulator using the collision detection function of the virtual mock-up has been established. From these, the maintenance process for the parts of the equipment, which are located at the outside of the MSM';s workspace using the servo manipulator has been proposed and verified through the graphic simulation. It is revealed that the proposed remote maintenance process of the equipment can effectively be used in the real hot cell operation. It is also believed that the implemented virtual mock-up of the hot cell can effectively be applied in analyzing the various hot cell operation and enhancing the reliability and safety in a hot cell remote handling for the spent fuel management.

Hot cell,Remote Handling,Master-Slave Manipulator,Servo manipulator,Virtual Mock-up,Graphic Simulation

Adsorption and desorption characteristics of methyl iodide at high temperature conditions up to 250℃ by TEDA-impregnated activated carbon and silver-ion exchanged zeolite(AgX-10), which are used for radioiodine retention in nuclear facility, were experimentally evaluated. In the range of temperature from 30℃ to 250℃, the adsorption capacity of base activated carbon decreased sharply with increasing temperature but that of TEDA-impregnated activated carbon showed higher value even at high temperature ranges. Especially, the residual amount of methyl iodide after desorption on TEDA-AC represented 30% lower value than that on AgX-10. However, it can be used as an adsorbent for the removal of methyl iodide up to 150℃ if it is preventing explosion by ignition. The breakthrough curves of methyl iodide in the fixed bed packed with AgX-10 up to 400℃ were compared upon the effects of bed temperatures, bed depth and input concentration of methyl iodide. Removal mechanism of methyl iodide on AgX-10 was proposed, based on the analysis of by-product gas generated from adsorption reaction.

methyl iodide,zeolite,activated carbon,AgX,TEDA-AC,adsorption,desorption

In a pressurized water reactor, radioactive corrosion products (CRUD) in primary coolant system are one of the major sources for the occupational radiation exposure of the personnel in a nuclear power plant. Through the recent trend of long term fuel cycle in a nuclear power plant, radioactive corrosion products deposited in reactor core for a long time are also the cause of Axial Offset Anomaly (AOA) having an effect on reactor power by the hideout of boron. CRUD consist primarily of magnetite, nickel ferrite, cobalt ferrite, and so on. They have the characteristic of strong magnetism. Therefore it is performed the conceptual design to develop the filter which removes the CRUD by magnetic field that is generated by an arrangement of permanent and electric magnets. Contrary to the conventional filter, the proposed filter does not interrupt the fluid flow, so there is little pressure drop and it can be used continuously. It is expected to be applied as one of the technologies for removal of the CRUD.

Radioactive Corrosion Product,Magnetic Separation,Magnetic Filler,Permanent Magnets

In order to predict the dynamic behaviors of uranium and cobalt in a fixed bed at various influent pH values of liquid waste, the adsorption system is regarded as a multi-component adsorption between each ionic species in the solution. Langmuir isotherm parameters of each species were extracted by incorporating equilibrium data with the solution chemistry of the uranium and cobalt using IAST. Prediction results were in good agreement with the experimental data, except for a high concentration and pH. Although there was some limitations in predicting the cobalt adsorption, this method may be useful in analyzing a complex adsorption system where various kinds of ionic species exist in a solution.

Surface-modified activated carbon,Uranium,Cobalt,Adsorption,Dynamic behavior

IAEA, FT-04-020, and ANS 16.1, standard leaching test methods, were evaluated comparatively with their test results. Leaching index of 60Co and 137Cs by ANS 16.1 method for waste forms of paraffin and cement were above 6.0. Their leaching behavior were depending on the type of matrix and leachant. Leachability of 60Co for cement waste form was higher in simulated seawater than de-mineralized water, and was higher in de-mineralized water for paraffin waste form. Leachability of 60Co was contrary to 137Cs. Cumulative fraction leached of 60Co was higher in order of IAEA > ANS > FT in a cement waste form.

radioactive waste ion cement,paraffin,leaching test,IAEA Hepse′ method,ANS 16.1. FT-04-020