• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.1-7
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• 2007

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(U-GIC) 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.

• Journal of Korean Society on Water Environment
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• v.23 no.2
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• pp.201-205
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• 2007

This paper reports the amount of $^{222}Rn$ and $^{238}U$ in 18 sites of ground water and 30 sites of surface water. The instrument used to count $^{222}Rn$ activity was the liquid scintillation counter (LSC) which could resolute ${\alpha}$ and ${\beta}$ radiations. And $^{238}U$ was analyzed by the inductively coupled plasma (ICP). Radon and Uranium were not detected in raw and treated water which were sampled in a water treatment plant. However, radon ($^{222}Rn$) was high concentration in ground water from Jeon-la, Gang-won. So was uranium ($^{238}U$) in case of ground water from Gang-won, Choong-chung. Radon ($^{222}Rn$) activities were detected less than 15 pCi/L at 5 sampling points, 15~300 pCi/L at 7 sampling points, 300~4000 pCi/L at 6 sampling points. However, Radon ($^{222}Rn$) activities of all ground water samples were less than 4,000 pCi/L, which was bellow American Alternative Maximum Contamination Level (AMCL). Uranium ($^{238}U$) concentrations were less than $0.1{\mu}g/L$ at 5 sampling points, from $0.1{\mu}g/L$ to $20{\mu}g/L$ at 13 sampling points. Uranium was not detected in about 30% of the whole samples, but the concentration ranged from relatively low to high concentrations depending on the sampling point. The minimum detectable activity (MDA) of radon was 15 pCi/L. and the detection limit of uranium was $0.1{\mu}g/L$.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.27-39
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• 1990

Appreciable radiation exposures certainly were occurred in the reactor burn-up, the nuelear fall-out and the surroundings of nuclear installations with radioactive effluents. Therefore, radioactive nuclides is not only potentially hazardous to workers of nuclear power plants and related industrials, but also the wokers who handle radioactive nuclides in biochemical research and nuclear medicine diagnostics. And in the case of occurring the nuclear accidents, the early medical treatment of radiation injury should be necessary but little is established medical procedures to decontaminate the victims of internal contamination of radioactive nuclides in korea. Accordingly, to achieve the basic data for protective roles and medical treatment of radiation injury, the present studies were carrid out to evaluate the decontamination of uranium by the chemical drugs. The results observed were summarized as follows: 1. The combined treatmet group of sodium bicarbonate and saline with uranyl nitrate injection simultaneously and the dithiothreitol group that was administered 30 minutes after uranyl nitrate injection were increased significantly in the change of body weight than uranyl nitrate-only group (P<0.005). 2. All the experimental groups were increased the fluid intake and urine volume on the uranyl nitrate-induced acute renal failure. but the combined treatment group of sodium bicarbonate and saline with uranyl nitrate injection simultaneously and the dithiothreitol group that was administered 30 minutes after uranyl nitrate injection have the higher increment of fluid intake and urine volume (P<0.05). 3. When sodium bicarbonate and saline was treated with uranyl nitrate injection simultaneously. and dithiothreitol was administered 30 minutes after uranyl nitrate injection. there was significantly reduced in BUN concentration (P<0.01). 4. When dithiothreitol was administered 30 minutes after uranyl nitrate injection. there was reduced more significantly on the increment of serum creatinine concentration than that observed in uranyl nitrate-only group(P<0.01). but when the combined treatment of sodium bicarbonate and saline with uranyl nitrate simultaneously, there was still. albeit much less marked. decrease in serum creatinine concentration. 5. The sodium bicarbonate and saline was treated with uranyl nitrate simultaneously and dithiothreitol was administered at 30 minutes after uranyl nitrate were excreted markedly higher urine creatinine concentration than the uranyl nitrate-only group. 6. Uranyl nitrate has been used in experimental animals to produce hydropic degeneration and swelling of proximal tubules, disappearance of microvilli and brush border or necrosis in the kidney and centrilobular necrosis, congestion, and telangiectasia of the liver. When the sodium bicarbonate and saline was treated with uranyl nitrate simultaneously, and dithiothreitol was administered. 30 minutes after uranyl nitrate, there was more marked the protective effect than uranyl nitrate-only group. Finally, if the sodium bicarbonate and saline may administered as quickly as possible each time that some risk for internal contamination, with uranium, and dithiothreitol is administered 30 minutes after uranium contamination, there ameliorates the course of uranyl nitrate-induced acute renal failure.and this effect is assocciated with prevention of uranium (heavy metal)-induced alterations in BUN, serum creatinine, urine creatinine, fluid intake, urine volume and body weight.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.72-80
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• 2004

Electrolytic dissolution study was carried out to evaluate the applicability of electrochemical decontamination process using a neutral salt electrolyte as a decontamination technology for the recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant using SUS-304 and Inconel-600 specimen as the main materials of internal system components of the plant. The effects of type of neutral salt as an electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as $UO_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion plant were peformed in $Na_2SO_4$ and $NaNO_3$ solution. It was verified that the electrochemical decontamination of the dismantled metallic wastes was quite successful in $Na_2SO_4$ and $NaNO_3$ neutral salt electrolyte by reducing $\beta$ radioactivities below the level of self disposal with authorization within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

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 Na2s04 electrolyte, a NaNO$_3$ 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 Na$_2$SO$_4$ 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 NaNO$_3$ 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 UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3785-3795
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• 2024

The present study concerned with the activity concentrations of natural radionuclides (²³⁸U, ²³⁴U, ²³⁰Th, ²²⁶Ra, ²³²Th, ⁴⁰K and, ²³⁵U) in ten sedimentary rock samples collected from fault zone, Gabal Um Hamd, southwestern Sinai, Egypt. These samples were investigated to study their behavior during a part of geologic time. The activity concentrations were measured using γ-ray spectrometry (HPGe detector). The investigated samples were analyzed for major oxides using the XRF technique. The results demonstrated high average activity concentrations of ²³⁸U, ²³⁴U, ²³⁰Th, ²²⁶Ra, ²³²Th, ⁴⁰K and, ²³⁵U than the worldwide average values as reported by UNSCEAR 2008. Theil diagram showed that there are accumulation and leaching of uranium in some samples in the two sides of the fault zone. It is noticed that the ages of uranium depositions for the samples collected from the downthrown of the fault zone vary from 121.5 to 440.1 ky, while for the sample collected from the upthrown of the fault is 210.9 ky. The ²³⁰Th/²³²Th activity ratios range between 4.55 and 91.04 for downthrown samples and between 4.75 and 6.05 for upthrown samples which are smaller than 20 except for two samples, indicating a contamination of the samples by detrital ²³⁰Th. After subtraction of the detrital ²³⁰Th, the corrected ages for downthrown samples vary from 119.1 to 231.7 ky while for upthrown samples vary from 164.4 to 390 ky.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.91-101
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• 2007

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 $0.05515Bq/cm^2$ (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 $0.01Bq/cm^2\;for\;{\alpha}-emitter$ and the result of radionuclide analysis for taken concrete samples from surface of concrete waste is 0.0297Bq/g for concentration of $^{238}U$, below 2w/o for enrichment of $^{235}U$ and 0.0089Bq/g for artificial contamination of $^{238}U$ 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.

• Economic and Environmental Geology
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• v.51 no.6
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• pp.531-542
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• 2018

Batch sorption experiments were performed to remove the uranium (U) in groundwater by using the bamboo charcoal. For 2 kinds of commercialized bamboo charcoals in Korea, the U removal efficiency at various initial U concentrations in water were investigated and the optimal sorption conditions to apply the bamboo charcoal were determined by the batch experiments with replicate at different pH, temperature, and reaction time conditions. From results of adsorption batch experiments, the U removal efficiency of the bamboo charcoal ranged from 70 % to 97 % and the U removal efficiency for the genuine groundwater of which U concentration was 0.14 mg/L was 84 %. The high U removal efficiency of the bamboo charcoal maintained in a relatively wide range of temperatures ($10{\sim}20^{\circ}C$) and pHs (5 ~ 9), supporting that the usage of the bamboo charcoal is available for U contaminated groundwater without additional treatment process in field. Two typical sorption isotherms were plotted by using the experimental results and the bamboo charcoal for U complied with the Langmuir adsorption property. The maximum adsorption concentration ($q_m:mg/g$) of A type and C type bamboo charcoal in the Langmuir isotherm model were 200.0 mg/g and 16.9 mg/g, respectively. When 2 g of bamboo charcoal was added into 100 mL of U contaminated groundwater (0.04 ~ 10.8 mg/L of initial U concentration), the separation factor ($R_L$) and the surface coverage (${\theta}$) maintained lower than 1, suggesting that the U contaminated groundwater can be cleaned up with a small amount of the bamboo charcoal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

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 radionuclide 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 radionuclide 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. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.