• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.10 no.1
• /
• pp.1-11
• /
• 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)
• /
• v.5 no.2
• /
• pp.113-122
• /
• 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)
• /
• v.5 no.2
• /
• pp.123-132
• /
• 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)
• /
• v.5 no.2
• /
• pp.133-143
• /
• 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)
• /
• v.5 no.2
• /
• pp.155-169
• /
• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.4
• /
• pp.275-286
• /
• 2014

This study evaluated the evolution stage and origin of chemical components of 12 boreholes at crystalline bedrock using multivariate statistical and groundwater quality analyses. Groundwater types are mostly belonged to Na(Ca)-$HCO_3$ and Ca-$HCO_3$ types, indicating that directly reaction of cation exchange ($Ca^{2+}{\rightarrow}Na^+$) prevailed. The degree of groundwater evolution is included the range from low to intermediate stage based on field and laboratory analytical conditions. As a result of multivariate statistical analysis, a typical indicator of groundwater contamination, $NO_3$-, has the positive correlation with $Na^+$ and $Cl^-$. The origin of sea spary ($Cl^-$) has the positive correlation with $Na^+$, $SO{_4}^{2-}$, $Mg^{2+}$, and $K^+$, while not correlation with $Ca^{2+}$, $Fe^{2+}$, $HCO_3{^-}$, $F^-$, and $SiO_2$. The concentration of $Cl^-$ and $NO_3{^-}$ belongs to general quality of groundwater and not exceeds over the Korean standard for drinking water. And the negative values of saturation index of minerals are calculated with chemical components in groundwater. Therefore, most of chemical components of groundwater in the study area are originated from natural process between rock and groundwater, whereas some of components are derived from sea spary and anthropogenic sources related to agricultural activities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.4
• /
• pp.267-274
• /
• 2014

The first Low- and Intermediate-Level Waste (LILW) disposal facility with 6 silos has been constructed in granite host rock saturated with groundwater in Korea. A two-dimensional numerical modeling on gas migration was carried out using TOUGH2 with EOS5 module in the disposal facility. Laboratory-scale experiments were also performed to measure the important properties of silo concrete related with gas migration. The gas entry pressure and relative gas permeability of the concrete was determined to be $0.97{\pm}0.15bar$ and $2.44{\times}10^{-17}m^2$, respectively. The results of the numerical modeling showed that hydrogen gas generated from radioactive wastes was dissolved in groundwater and migrated to biosphere as an aqueous phase. Only a small portion of hydrogen appeared as a gas phase after 1,000 years of gas generation. The results strongly suggested that hydrogen gas does not accumulate inside the disposal facility as a gas phase. Therefore, it is expected that there would be no harmful effects on the integrity of the silo concrete due to gas generation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.4
• /
• pp.315-334
• /
• 2014

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.4
• /
• pp.271-280
• /
• 2013

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.4
• /
• pp.311-324
• /
• 2013

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