• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.37-43
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• 2012

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.161-170
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• 2012

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.151-159
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• 2012

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.143-149
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• 2012

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.13-19
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• 2012

The long-lived fission products $^{79}Se$ and $^{99}Tc$ have been considered as the major concern nuclides for the disposal of radioactive waste because of their high solubilities and the existence of anionic species in natural water. In this study, the solubilities of $FeSe_2(s)$ and $TcO_2(s)$, known as respective Solubility Limiting Solid Phase (SLSP) of selenium and technetium, were measured in the KURT (KAERI Underground Research Tunnel) groundwater under various pH and redox conditions. And their solubilities and major species were also calculated using geochemical codes under conditions similar to experimental solutions. Experimental results and calculation for $FeSe_2$ show that the solubility of selenium was found to be below $1{\times}10^{-6}mol/L$ under the condition of pH 8~9.5 and Eh=-0.3~-0.4 V while the dominant species was identified as $HSe^-$. For $TcO_2$, the solubility of technetium was found to be $5{\times}10^{-8}{\sim}1{\times}10^{-9}mol/L$ in the solutions of pH 6~9.5 and Eh<-0.1 V, while the dominant species was $TcO(OH)_2$. However, when the Eh of the solution is -0.35 V, $TcO(OH)_3^-$ and $TcO_4^-$ are calculated as the dominant species at pH 10.5~12 and pH>12, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.1-11
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• 2012

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.113-122
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• 2007

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-sorting 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 $K_d$ 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 $R_d({\sim}16.2)$ was obtained by using the $K_d$ obtained from the batch sorption experiment and it was compared with the $R_d({\sim}14.3)$ 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.123-132
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• 2007

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)/$NDD(n-dodecane)-HNO_3$ extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M $HNO_3$ 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)${\rightarrow}RE$ (stripping agent : 5M $HNO_3$), 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.133-143
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• 2007

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 $(n-dodecane)-HNO_3$ extraction system. An optimal condition of (30% TBP-0.5% TOA)/NDD-1 M $HNO_3$ 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 $HNO_3$)${\rightarrow}Np$ by reductive stripping (reductive-stripping agent : 0.1 M AHA)${\rightarrow}U$ (stripping agent : 0.01 M $HNO_3$), 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

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 analyses were performed for the container in a view of Type B. As a result of structural analysis, 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^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. 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.