• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.367-368
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.131-132
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• 2012

• The Journal of Engineering Geology
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• v.23 no.2
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• pp.171-186
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• 2013

A drilling project was undertaken to characterize the geochemical relationship and the occurrence of radioactive materials at a test site among public-use groundwaters previously known to have high occurrence of uranium and radon-222 in the Daejeon area. A borehole (121 m deep) was drilled and core rocks mainly consist of two-mica granite, and associated with pegmatite and dykes of intermediate composition. The groundwater samples collected at six different depths in the borehole by a double-packed system showed the pH values ranging from neutral to alkaline (7.10-9.3), and electrical conductivity ranging from 263 to 443 ${\mu}S/cm$. The chemical composition of the borehole groundwaters was of the $Ca-HCO_3(SO_4+Cl)$ type. The uranium and Rn-222 contents in the groundwater were 109-1,020 ppb and 9,190-32,800 pCi/L, respectively. These levels exceed the regulation guidelines of US EPA. The zone of the highest groundwater uranium content occurred at depths of 45 to 55m. The groundwater chemistry in this zone (alkaline, oxidated, and high in bicarbonate) is favorable for the dissolution of uranium into groundwater. The dominant uranium complex in groundwater is likely to be $(UO_2CO_3)^0$ or $(UO_2HCO_3)^+$. Radon-222 content in groundwater shows an increasing trend with depth. The uranium and thorium contents in the core were 0.372-47.42 ppm and 0.388-11.22 ppm, respectively. These levels are higher values than those previously been reported in Korea. Microscopic observations and electron microprobe analysis(EPMA) revealed that the minerals containing U and Th are monazite, apatite, epidote, and feldspar. U and Th in these minerals are likely to substitute for major elements in crystal lattice.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

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

• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.179-182
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• 2001

The chloride volatilization method for the recovery of zirconium and removal of uranium from zirconium containing metallic wastes formed in spent fuel reprocessing was studied using the simulated alloy waste, i.e. the mixture of Zr foil and UO$_2$/U$_3$O$_{8}$ powder. When the simulated waste was heated to react with chlorine gas at 350- l00$0^{\circ}C$, the zirconium metal changed to volatile ZrCl$_4$showing high volatility ratio (Vzr) of 99%. The amount of volatilized uranium increases at higher temperatures causing lowering of decontamination factor (DF) of uranium. This is thought to be caused by the chlorination of UO$_2$ with ZrCl$_4$vapor. The highest DF value of 12.5 was obtained when the reaction temperature was 35$0^{\circ}C$. Addition of 10 vol.% oxygen gas into chlorine gas was effective for suppressing the volatilization of uranium, while the volatilization ratio of zirconium was decreased to 68% with the addition of 20 vol.% oxygen. In the case of the mixture of Zr foil and U$_3$O$_{8}$, the V value of uranium showed minimum (44%) at 40$0^{\circ}C$ with chlorine gas giving the highest DF value 24.3. When the 10 vol.% oxygen was added to chlorine gas, the V value of zirconium decreased to 82% at $600^{\circ}C$, but almost all the uranium volatilized (Vu=99%), which may be caused by the formation of volatile uranium chlorides under oxidative atmosphere.ere.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Journal of the Korean Society for Precision Engineering
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• v.30 no.1
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• pp.113-119
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• 2013

To develop a new nuclear fuel, it needs to make a test fuel rod and carry out burn-up test in the test loop of a research reactor to check the irradiation characteristics of the nuclear fuel. At that time, several sensors such as thermocouples, LVDTs and SPNDs are needed to be attached in and out of the fuel rod and connect them with instrumentation cables. Then, the instrumentation cables deliver the signals measured by the sensors to the measuring device located outside of the reactor pool. In particular, to install a thermocouple in a fuel rod, it needs to drill off holes on the alumina blocks and sintered $UO_2$ pellets. However, because the hardness of a sintered $UO_2$ pellet is 700 Hv (or HRC 61) and that of an alumina block is 1480 Hv, a special drilling machine which adapts a diamond coated drill bit had developed. In this study, several case experiments have been carried out to find an optimal drilling condition of the drilling machine. And, using the optimal drilling condition, minimum numbers of the holes that a drill bit can drill off are verified.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.181-185
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• 1998

모의 혼합산화물인 (U,Ce)O$_2$ 에 dopant 인 Li$_2$O 와 SiO$_2$ 를 첨가한 소결체의 압축크립변형거동을 수소분위기, 온도 1773K 에서 응력(10-120MPa)을 변화시켜 조사하였다. Dopant 를 첨가할 경우 정상상태 크립변형속도는 첨가하지 않은 경우보다 크게 증가하는 것으로 나타났다. 증가한 원인으로는 Li$_2$O 를 첨가한 경우 우라늄 확산계수의 증가에 기인되며, SiO$_2$ 를 첨가한 경우에는 SiO$_2$ 가 glassy phase 로 입계에 위치하여 입계이동이 용이하게 되어 정상상태 크립변형속도가 증가한 것으로 사료된다. 또한 저응력구간에서 (U,Ce)O$_2$ 의 크립활성화에너지는 109.6 kcal/mol 로 $UO_2$ 의 크립활성화에너지(94.2kca1/mol)보다 더 크게 나타났다.

• Journal of Integrative Natural Science
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• v.6 no.2
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• pp.104-110
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• 2013

1-Aza-12-crown-4 macrocyclic ligand was combined with styrene (2th petroleum in 4th class hazardous materials) divinylbenzene copolymer having 1%, 2%, 3%, and 6% crosslinks by a substitution reaction, in order to synthesize resin. These synthetic resins were confirmed by chlorine content, elementary analysis and IR-spectrum. As the results of the effects of pH, equilibrium arrival time, crosslink of synthetic resin, and dielectric constant of a solvent on uranium ion adsorption for resin adsorbent, the uranium ion showed high adsorption at pH 3 or over and adsorption equilibrium of uranium ion was about 2 hours. In addition, adsorption selectivity for the resin in methanol solvent was the order of uranium ($UO_2{^{2+}}$) > iron ($Fe^{3+}$) > lutetium ($Lu^{3+}$) ions, adsorbability of the uranium ion was in the crosslinks order of 1%, 2%, 3%, and 6% was increased with the lower dielectric constant.