• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.35-44
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• 2017

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.87-96
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• 2004

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 $10^{-11}$m/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 ㎫ and they increased with increasing the dry density. The thermal conductivities were in the range of 0.80 ㎉/m $h^{\circ}C$ to 1.52 ㎉/m $h^{\circ}C$. The unconfined compressive strength, Young's modulus and Poison's ratio showed the range of 0.55 ㎫ to 8.83 ㎫, 59 ㎫ to 1275 ㎫, and 0.05 to 0.20, respectively, when the dry densities of the compacted bentonite were 1.4 Ms/㎥ to 1.8 Mg/㎥. The diffusion coefficients in the compacted bentonite were measured under an oxidizing condition. The values were $1.7{\times}10^{-10}$m^2$/s to 3.4{\times}10^{-10}$m^2$/s for electrically neutral tritium (H-3), 8.6{\times}10^{-14}$m^2$/s to 1.3{\times}10^{-12}$m^2$/s for cations (Cs, Sr, Ni), 1.2{\times}10^{-11}$m^2$/s to 9.5{\times}10^{-11}$m^2$/s for anions (I, Tc), and 3.0{\times}10^{-14} $m^2$/s to 1.8{\times}10^{-13}$m^2$/s $for actinides (U, Am), when tile 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.n Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.1-7
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• 2009

This study has been focused on determining the chemical composition of $^{14}C$ - in terms of both organic and inorganic $^{14}C$ contents - in reactor coolant from 3 different PWR's reactor type. The purpose was to evaluate the characteristic of $^{14}C$ that can serve as a basis for reliable estimation of the environmental release at domestic PWR sites. $^{14}C$ 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 $^{14}C$ 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 $^{14}C$ in various water systems were also performed. The $^{14}C$ 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 $^{14}C$ activity in the water was divided equally between the gas- and water- phase. Even though organic $^{14}C$ compound shows that dominant species during the reactor operation, But during the releasing of $^{14}C$ from the plant stack, chemical forms of $^{14}C$ shows the different composition due to the operation conditions such as temperature, pH, volume control tank venting and shut down chemistry.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.189-197
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• 2012

Geochemical composition of fracture filling minerals and groundwater was investigated to characterize geochemical characteristics of groundwater system at the KURT site. Minerals such as calcite, illite, laumontite, chlorite, epidote, montmorillonite, and kaolinite, as well as I/S mixed layer minerals were detected in the minerals extracted from the fracture surfaces of the core samples. The groundwater from the DB-1, YS-1 and YS-4 boreholes showed alkaline conditions with pH of higher than 8. The electrical conductivity (EC) values of the groundwater samples were around $200{\mu}S/cm$, except for the YS-1 borehole. Dissolved oxygen was almost zero in the DB-1 borehole indicating highly reduced conditions. The Cl- concentration was estimated around 5 mg/L and showed homogeneous distribution along depths at the KURT site. It might indicate the mixing between shallow groundwater and deep groundwater. The shallow groundwater from boreholes showed $Ca-HCO_3$ type, whereas deep groundwater below 300 m from the surface indicated $Na-HCO_3$ type. The isotopic values observed in the groundwater ranged from -10.4 to -8.2‰ for ${\delta}^{18}O$ and from -71.3 to -55.0‰for ${\delta}D$. In addition, the isotope-depleted water contained higher fluoride concentration. The oxygen and hydrogen isotopic values of deep groundwater were more depleted compared to the shallow groundwater. The results from age dating analysis using $^{14}C$ indicated relatively younger (2000~6000yr old) groundwater compared to other european granitic groundwaters such as Stripa (Sweden).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.45-53
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• 2012

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a $SiO_2-Al_2O_3-P_2O_5$ (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about $400^{\circ}C$ for LiCl and the second was about $700^{\circ}C$ for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a $SiO_2-Al_2O_3-P_2O_5$ (SAP) could be considered as one of alternatives for the immobilization of waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.27-36
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• 2012

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP ($SiO_2-Al_2O_3-P_2O_5$) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of $Fe_2O_3$ into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP( Fe=0.1). The experimental results indicated that the addition of $Fe_2O_3$ increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing $B_2O_3$ into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP ($SiO_2-Al_2O_3-Fe_2O_3-P_2O_5-B_2O_3$) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3~4 g of wasteform for final disposal. The final volume would be about 3~4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.281-290
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• 2018

In order to understand the long-term behavior of radionuclides in granite environments, geochemical behavior characteristics of uranium in granitic host rock of KURT (KAERI Underground Research Tunnel) were investigated by dissolution experiment with different reaction time and solutions. In the dissolution experiment, significantly increased dissolution levels of uranium from granite powder samples were identified during the reaction time of 0~10 days for reaction solutions ($UD-CO_3$ and UD-Bg) containing a large amount of $CO_3{^{2-}}$. On the other hand, significantly increased dissolution levels of uranium were also identified for reaction solutions containing Na and Ca after 60 days. Dissolution of uranium continuously increased in reaction solutions of $UD-CO_3$ ($44.61{\mu}g{\cdot}L^{-1}$), UD-Bg ($41.01{\mu}g{\cdot}L^{-1}$), UD-Na ($26.87{\mu}g{\cdot}L^{-1}$), UD-Ca ($20.26{\mu}g{\cdot}L^{-1}$), UD-CaSi ($17.03{\mu}g{\cdot}L^{-1}$), and UD-Si ($10.47{\mu}g{\cdot}L^{-1}$) in the experimental period of ~270 days. However, after day 270, dissolution of uranium showed a decreasing tendency. This is thought to have occurred because existing uranium in granite samples reached the limit of dissolution by interaction with reaction solutions. Concentrations of dissolved uranium and points of maximum concentration value were found to differ depending on the $CO_3{^{2-}}$ presence in the mixed reaction solution and on the geochemical type of the water. It is estimated that differences in the reaction rate between the granite sample and the reaction solution are due to the influence of dissolved ions in the reaction solution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.117-131
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• 2006

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 $90^{\circ}C$ at the interface between the heater and the bentonite and at about $70^{\circ}C$ 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 H-M(Hydro-Mechanical) processes using elastic and poroelastic material mod31 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 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.311-318
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• 2010

U. S. NRC Regulation Guide 1.52 regulating ESF ACS in nuclear power plants has been revised to revision 3. To apply reduction of operability test time, allowance of alternative challenge agents for in-place leak test of HEPA filters, and upgrade of Methyl Iodide penetration acceptance criterion in activated carbon performance test suggested in Reg. Guide 1.52(Rev.3) on Yonggwang units 5 and 6 ESF ACSes, technical feasibility study was carried out with on-site experiments as well as experiments with a lab-scale model. It was confirmed that the moisture in the system returned to the level before the test in 1 or 4 days even though the moisture was removed during the operability test lasting more than 10 hours. Therefore, it is appropriate to perform monthly operability test in 15 minutes just long enough to check the operability of equipment. To change challenge material for in-place HEPA filter leak test, size of aerosol, production rate, and leak detection capability were compared for DOP and PAO. It was concluded that PAO can be substituted for DOP in nuclear power plants. The upgrade of Methyl Iodide penetration acceptance criterion from 0.175 % to 0.5 % in active carbon filter bed deeper than 4 inches was to conform to the change of activated carbon performance test method to ASTM D3803(1989). It was confirmed that Methyl Iodide penetration acceptance criterion of 0.5 % under $30^{\circ}C$, relative humidity 95 % condition was conservatively good enough for testing performance of active carbon insitu. The licence change of Yonggwang units 5 and 6 has been completed based on this study.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.171-178
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• 2008

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.