• Nuclear Engineering and Technology
• /
• v.47 no.3
• /
• pp.293-305
• /
• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.110-120
• /
• 2005

For utilizing vitrification to treat low and intermediate level waste, industrial pilot plant was designed and constructed in October 1999 at Daejon, Korea through the joint research program among NETEC, MOBIS and SGN. More than 70 tests were performed on simulated IER, DAW etc. including key nuclide surrogate(Cs, Co); this plant has been shown to vitrify the target waste effectively and safely, however, some dust are generated from the HTF(High Temperature Filter) as a secondary waste. In case of long term operation, it is also concerned that pipe plugging can be occurred due to deposited dust in cooling pipe namely, connecting pipe between CCM(Cold Crucible Melter) and HTF. In this regard, we have developed the special complementary system of the off-gas treatment system to recycle the dust from HTF to CCM and to remove the interior dust of cooling pipe. Main concept of the dust recycling is to feed the dust to the CCM as a slurry state; this system is regarded as of an important position in the viewpoint of volume reduction, waste disposal cost and glass melt control in CCM. The role of DRS(Dust Recycling System) is to recycle the major glass components and key nuclides; this system is served to lower glass viscosity and increase waste solubility by recycling B, Na, Li components into glass melt and also to re-entrain and incorporate into glass melt like Cs, Co. Therefore dust recycling is helpful to control the molten glass; it is unnecessary to consider a separate dust treatment system like a cementation equipment. The effects of Dust Cleaner are to prevent the pipe plugging due to dust and to treat the deposited dust by raking the dust into CCM. During the pilot vitrification test, overall performance assessment was successfully performed; DRS and Dust Cleaner are found to be useful and effective for recycling the dust from HTF and also removing the dust in cooling pipe. The obtained operational data and operational experiences will be used as a basis of the commercial facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.1
• /
• pp.39-52
• /
• 2007

Quantitative analysis on release behavior of the $^{85}Kr\;and\;^{14}C$ fission gases from the spent fuel material during the voloxidation and OREOX process has been performed. This thermal treatment step in a remote fabrication process to fabricate the dry-processed fuel from spent fuel has been used to obtain a fine powder The fractional release percent of fission gases from spent fuel materials with burn-up ranges from 27,000 MWd/tU to 65,000 MWd/tU have been evaluated by comparing the measured data with these initial inventories calculated by ORIGEN code. The release characteristics of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation process at $500^{\circ}C$ seem to be closely linked to the degree of conversion efficiency of $UO_2\;to\;U_3O_8$ powder, and it is thus interpreted that the release from grain-boundary would be dominated during this step. The high release fraction of the fission gas from an oxidized powder during the OREOX process would be due to increase both in the gas diffusion at a temperature of $500^{\circ}C$ in a reduction step and in U atom mobility by the reduction. Therefore, it is believed that the fission gases release inventories in the OREOX step come from the inter-grain and inter-grain on $UO_2$ matrix. It is shown that the release fraction of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation step would be increased as fuel burn-up increases, ranging from 6 to 12%, and a residual fission gas would completely be removed during the OREOX step. It seems that more effective treatment conditions for a removal of volatile fission gas are of powder formation by the oxidation in advance than the reduction of spent fuel at the higher temperature.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.15 no.2
• /
• pp.26-29
• /
• 2011

Purpose: The general test method of the Korean Pharmacopeia specifies the test method on the clauses of quality control after manufacturing. According to KFDA Guidance for Medicines, standards of residual solvents regulates the maximum permissible dose of acetonitrile as 400 ppm, ethanol as 5,000 ppm, and acetic acid as 5,000 ppm. This study aims at identifying the type of resiual solvents in the final [F-18]FDG vial of an automatic synthesizer and measure its residual quantity. Materials and Methods: The center carried out residual solvents test of [F-18]FDG injection using Agilent Technologies 7890A with a Flame Ionization Detector. The column of Agilent Technologies 7890A used in measuring of residual solvents was CP WAX column ($30m{\times}0.53mm{\times}1.0{\mu}m$) and analysis condition was split mode 1:1 at the initial temperature $70^{\circ}C$ which was increased $20^{\circ}C/minute$ after two minutes and maintained at the final $140^{\circ}C$ for two minutes. The analysis method was as following: Firstly, ethanol-acetonitrile-acetic acid mixture was classified into four types of concentration (250-25-250 ppm, 1,000-100-1,000 ppm, 3,000-300-3,000 ppm, and 6,000-600-6,000 ppm), and $1.0{\mu}L$ of each type of concentration was injected into gas chromatography followed by an analysis of its peak domain. Then, a calibration-curve by the external standard method was drawn based on the analysis result. Results: While ethanol and acetonitrile were detected in TRACERlab MX, FASTlab had additional acetic acid. The residual quantity of the ethanol-acetonitrile-acetic acid mixture evaluated using the calibration-curve was average 72 ppm ethanol, 54 ppm acetonitrile, and 1030 ppm acetic acid for FASTlab, whereas average 439 ppm ethanol and 79 ppm acetonitrile for TRACERlab MX. This indicated that both of them were within the maximum permissible dose. Conclusion: Solvent residues in the [F-18]FDG injection were all within maximum permissible doses and proper to be used to examine a patient. The result indicated that types and quantities of solvent resides of radioactive pharmaceuticals vary depending on the automatic synthesizer.

• Korean Journal of Mineralogy and Petrology
• /
• v.35 no.3
• /
• pp.313-331
• /
• 2022

Natural or native abiotic molecular hydrogen (H₂) is a major component in natural gas, however yet its importance in the global energy sector's usage as clean and renewable energy is underestimated. Here we review the occurrence and geological settings of native hydrogen to demonstrate the much widesprease H₂ occurrence in nature by comparison with previous estimations. Three main types of source rocks have been identified: (1) ultramafic rocks; (2) cratons comprising iron (Fe²⁺)-rich rocks; and (3) uranium-rich rocks. The rocks are closely associated with Precambrian crystalline basement and serpentinized ultramafic rocks from ophiolite and peridotite either at mid-ocean ridges or within continental margin(Zgonnik, 2020). Inorganic geological processes producing H₂ in the source rocks include (a) the reduction of water during the oxidation of Fe²⁺ in minerals (e.g., olivine), (b) water splitting due to radioactive decay, (c) degassing of magma at low pressure, and (d) the reaction of water with surface radicals during mechanical breaking (e.g., fault) of silicate rocks. Native hydrogen are found as a free gas (51%), fluid inclusions in various rock types (29%), and dissolved gas in underground water (20%) (Zgonnik, 2020). Although research on H₂ has not yet been carried out in Korea, the potential H₂ reservoirs in the Gyeongsang Basin are highly probable based on geological and geochemical characteristics including occurrence of ultramafic rocks, inter-bedded basaltic layers and iron-copper deposits within thick sedimentary basin and igneous activities at an active continental margin during the Permian-Paleogene. The native hydrogen is expected to be clean and renewable energy source in the near future. Therefore it is clear that the origin and exploration of the native hydrogen, not yet been revealed by an integrated studies of rock-fluid interaction studies, are a field of special interest, regardless of the presence of economic native hydrogen reservoirs in Korea.

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

• Journal of radiological science and technology
• /
• v.37 no.4
• /
• pp.261-265
• /
• 2014

The ${\gamma}$-ray concentration and gross-${\beta}$ activity by age group were measured in the teeth of males and females of the domestic residents. They were divided into 7 age groups from 10s to the age of 70s. The gross-${\beta}$ activity concentration was measured by using the Tennelec XLB measuring instrument filled with P10 gas (argon 90%, methane 10%). The ${\gamma}$-ray was measured through the ${\gamma}$-ray spectroscopic analytical method by using the high purity germanium (HPGe) radiation detector. The range of gross-${\beta}$ activity concentration was measured 0.089 to 0.32 Bq/kg in females and 0.13 to 0.26 Bq/kg in males. From the ${\gamma}$-ray spectroscopic analysis of the teeth, the natural radioactive isotopes of $^{40}K$, $^{208}Tl$, $^{228}Ac$ and $^{234}Th$ were detected and their measured ${\gamma}$-ray activity concentrations were found to be 20.7, 21.9, 3.88 and 5.24 Bq/kg, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.2
• /
• pp.149-157
• /
• 2005

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered at inside surface of hot cell and surface of equipment in hot cell. It was necessary to develop a particulate filtration equipment to reuse PFC solution used on PFC decontamination due to its high cost and to minimize the volume of second wastewater. Contamination characteristics of hot particulate were investigated and then a filtration process was presented to remove hot particulate in PFC solution generated through PFC decontamination process. The removal efficiency of PVDF(Poly vinylidene fluoride), PP(Polypropylene), Ceramic(Al$_{2}$O$_{3}$ filter showed more than 95$\%$. The removal efficiency of PVDF filter was a little lower than those of other kiters at same pressure(3psi). A ceramic filter showed a higher removal efficiency with other filters, while a little lower flux rate than other filters. Due to inorganic composition, a ceramic filter was highly stable against radio nuclides in comparison with PVDF and PP membrane, which generate H$_{2}$ gas in e-radioactivity atmosphere. Therefore, the adoption of ceramic filter is estimated to be suitable for the real nitration process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.1
• /
• pp.79-84
• /
• 2007

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.217-226
• /
• 2006

The structural and contamination characteristics of hot cells at KAERI were investigated. The SEM results showed that the size of the hot particulate on the inner surface of the hot cell ranged from 0.2 to $10{\mu}m$. It was found that an inlet flow rate of 15 m/sec was suitable for this developed cyclone with a 49 mm optimum vortex finder length. The results showed that the collection efficiency was about 85% for $3{\mu}m$ particles. The collection efficiency didn't show a sharp increase when the inlet flow rate was faster than 15m/sec. When the temperature of the inlet flow gas was increased, the collection efficiency of the cyclone was slightly decreased. The larger the vortex finder length was, the higher the pressure drop in the cyclone was. The cut size diameter decreased with an increment of the Reynolds number. It was established that the flow in the cyclone was a turbulent flow on the basis of the Reynolds number and this turbulent flow caused a pressure drop in the cyclone. $Stk^{1/2}_{50}$ decreased with increasing values of the Reynolds number and it gradually approached a constant value at a higher value of the Reynolds number Namely, $Stk^{1/2}_{50}$ approached approximately 0.045 between 6000 and 8000 of the Reynolds number.