• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.4
• /
• pp.429-435
• /
• 2019

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

• Journal of the Mineralogical Society of Korea
• /
• v.27 no.4
• /
• pp.293-299
• /
• 2014

The bentonite, a buffer material, is essential for the deep geological disposal of HLW (high-level radioactive waste), and it is important to know its characteristic long-term evolution in the underground environment. With an assumption that the concentration of aqueous copper ions will increase if copper-coated materials on a metal canister are corroded, we examined some characteristic ion-exchanges and cation release phenomena occurring in the bentonite clay (montmorillonite) interacted with aqueous Cu cations. During the interaction between dissolved copper and bentonite, Na rather than Ca cations in the expandable clay were preferentially replaced by Cu ions in the experiment. In addition, the Cu-exchanged montmorillonite was characterized by an asymmetric X-ray diffracted pattern with relatively collapsed interlayers compared to the raw sample. These results indicate that the gradual change of the original bentonite property may occur in a underground disposal condition. We are going to further study the characteristic chemical and mineralogical changes of the bentonite buffer to be used for the disposal site by conducting additional experiments.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.10 no.4
• /
• pp.263-272
• /
• 2012

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

• Nuclear Engineering and Technology
• /
• v.46 no.6
• /
• pp.847-856
• /
• 2014

Spent resin waste containing a high concentration of $^{14}C$ radionuclide cannot be disposed of directly. A fundamental study on selective $^{14}C$ stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive $HCO_3{^-}$ ion, as the chemical form of $^{14}C$, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of $NO_3{^-}$ > $HCO_3{^-}{\geq}H_2PO_4{^-}$. Thus the competitive adsorption affinity of $NO_3{^-}$ ion in binary systems appeared far higher than that of $HCO_3{^-}$ or $H_2PO_4{^-}$, and the selective desorption of $HCO_3{^-}$ from the resin was very effective. On one hand, the affinity of $Co^{2+}$ and $Cs^+$ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of $Cs^+$ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, $NH_4H_2PO_4$ solution was preferable for the stripping of $^{14}C$ from the spent resin.

• Nuclear Engineering and Technology
• /
• v.45 no.7
• /
• pp.969-978
• /
• 2013

A full-sized model for the horizontally oriented metal cask containing 21 spent fuel assemblies has been considered to evaluate the internal natural convection behavior within a dry shield canister (DSC) filled with helium as a working fluid. A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values ($3{\times}10^6{\sim}3{\times}10^7$) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results. The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to $1.5{\times}10^6$ < Ra < $1.0{\times}10^7$ are proportional to 0.5 power of the Rayleigh number, while the Nusselt numbers for $1.0{\times}10^7$ < Ra < $8.0{\times}10^7$ are proportional to 0.27 power of the Rayleigh number. These results agreed well with the trends of the experimental data for Ra > $1.0{\times}10^7$.

• Economic and Environmental Geology
• /
• v.42 no.5
• /
• pp.471-477
• /
• 2009

An experimental removal of dissolved uranium (U) exsiting as uranyl ion (${UO_2}^{2+}$) was carried out using Shewanella p., iron-reducing bacterium. By the microbial reductive reaction, initial U concentration ($50{\mu}M$) was constantly decreased, and most U were removed from solution after 2 weeks. Major mechanism that U was removed from the solution was adsorption, precipitation and mineralization on the microbe surface. Under the transmission electron microscopy, the U adsorbed on the microbe was observed as being crystallized and eventually enlarged to several ${\mu}m$ sizes of minerals by combining with individual microbes and organic exudates. It seems that such U growth and mineralization on the microbial surface could affect the U behavior in a radioactive waste disposal site. Thus, the biogechemical reaction of metal-reducing bacteria observed in this experiment could give an affirmative measure that the microbial activity may retard U movement in subsurface environment.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.83-83
• /
• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. Distillation process was employed for the cathode processing. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. In this study, a mesh-covered crucible was investigated for the sat distillation of electrorefiner uranium deposits. A liquid salt separation step and a vacuum distillation step were combined for salt separation. The adhered salt in uranium deposits was efficiently removed in the mesh-covered crucible. The salt distiller was operated simply since repeated cooling - heating step was not necessary for the change of the crucible. The operation time could be reduced by the use of the mesh-covered crucible and the combined operation of the two steps. A method to preserve a vacuum level was proposed by double O-rings during the operation of the distiller with the mesh-covered crucible. After the salt distillation, the salt content was measured and was below 0.1wt% after the salt distillation. The residual salt after the salt distillation can be removed further during melting of uranium metal.

• Economic and Environmental Geology
• /
• v.56 no.4
• /
• pp.421-433
• /
• 2023

The final disposal of spent nuclear fuel(SNF) from nuclear power plants takes place in a deep geological repository. The metal canister encasing the SNF is made of cast iron and copper, and is engineered to effectively isolate radioactive isotopes for a long period of time. The SNF is further shielded by a multi-barrier disposal system comprising both engineering and natural barriers. The deep disposal environment gradually changes to an anaerobic reducing environment. In this environment, sulfide is one of the most probable substances to induce corrosion of copper canister. Stress-corrosion cracking(SCC) triggered by sulfide can carry substantial implications for the integrity of the copper canister, potentially posing a significant threat to the long-term safety of the deep disposal repository. Sulfate can exist in various forms within the deep disposal environment or be introduced from the geosphere. Sulfate has the potential to be transformed into sulfide by sulfate-reducing bacteria(SRB), and this converted sulfide can contribute to the corrosion of the copper canister. Bentonite, which is considered as a potential material for buffering and backfilling, contains oxidized sulfate minerals such as gypsum(CaSO4). If there is sufficient space for microorganisms to thrive in the deep disposal environment and if electron donors such as organic carbon are adequately supplied, sulfate can be converted to sulfide through microbial activity. However, the majority of the sulfides generated in the deep disposal system or introduced from the geosphere will be intercepted by the buffer, with only a small amount reaching the metal canister. Pyrite, one of the potential sulfide minerals present in the deep disposal environment, can generate sulfates during the dissolution process, thereby contributing to the corrosion of the copper canister. However, the quantity of oxidation byproducts from pyrite is anticipated to be minimal due to its extremely low solubility. Moreover, the migration of these oxidized byproducts to the metal canister will be restricted by the low hydraulic conductivity of saturated bentonite. We have comprehensively analyzed and summarized key research cases related to the presence of sulfates, reduction processes, and the formation and behavior characteristics of sulfides and pyrite in the deep disposal environment. Our objective was to gain an understanding of the impact of sulfates and sulfides on the long-term safety of high-level radioactive waste disposal repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.2
• /
• pp.179-186
• /
• 2006

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.

• Nuclear Engineering and Technology
• /
• v.20 no.2
• /
• pp.105-111
• /
• 1988

The chemical effects resulting from the capture of the thermal neutron by manganese in various crystalline permanganates, that is, potassium permanganate ammonium permangante and barium permanganate, have been investigated. The effect of pH of solvent on the distribution of radioactive manganese chemical species, that is, cationic $^{56}$ Mn, $^{56}$ MnO$_2$ and $^{56}$ MnO$_4$$^{[-10]}$ produced in the permanganates by $^{55}$ Mn(n, ｒ) $^{56}$ Mn reaction was studied by using various adsorbents and ion-exchanger, that is, zeolite A-3, kaolinite, alumina, manganese dioxide and Dowex-50 The distribution of radioactive MnO$_4$$^{[-10]}$ in kaolinite and alumina has higher than that in other adsorbents and ion-exchanger at a representative pH value of 4, 7 and 9, respectively. The yield of radioactive MnO$_4$$^{[-10]}$ is higher at pH 4 End pH 9 than at pH 7. The thermal annealing behavior of recoil manganese atoms produced in the permanganates by $^{55}$ Mn(n, ｒ) $^{56}$ Mn reaction was also studied. The retention of MnO$_4$$^{[-10]}$ in the thermal annealing is increased as annealing temperature increases when it was treated at 10$0^{\circ}C$ and 13$0^{\circ}C$. The recoil effect of permanganates was explained by the hot zone model.