• Applied Chemistry for Engineering
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• v.28 no.3
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• pp.345-350
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• 2017

The sintered density of uranium oxide pellets for pressurized water reactors is generally analyzed with pellet's samples completed with the sintering process. In this paper, the sintered density was analyzed by the newly developed method measuring the chromatography of ammonium diuranate, a precursor of uranium oxide, by a spectrophotometer (CM-5, Konica Minolta) before completing the sintering process. As a result of the sintered density analysis based on the brightness, color coordinate values (L, a, b) obtained from five ammonium diuranate samples by a spectrophotometer and the trend line of sintered density analyzed by a previous method, the sintered density with respect to the L value was observed with 0.9967 of the decision factor $R^2$. In case of a value, $R^2$ value was 0.9534 indicating lower reliability than that of the L value. However, b value with $R^2$ value of 0.4349 showed a very low correlation.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.191-198
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• 2008

The influence of fission products' contents on the DUPIC fuel powder and pellet properties was experimentally evaluated using SIMFUEL as a surrogate for actual spent PWR fuel due to the high radioactivity of spent fuel. Pure $UO_2$ and SIMFUEL pellets with fission products equivalent to a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used as impurities in this study. The specific surface area of the powder milled after the OREOX treatment increased and resulted in sintered pellets with a theoretical density (TD) higher than 95%, regardless of the impurity contents. However, the grain size of the sintered pellets decreased with the increasing impurity contents. As a result of the dissolved oxides in $UO_2$ from the impurity groups, the specific surface area of the OREOX powder increased with an increase of the impurities. The grain size of the sintered pellets was significantly decreased by the metallic and oxide precipitates.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.564-569
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• 2003

During fuel irradiation tests, all parts of cylindrical structure with multiple holes act as heat sources due to fussion heal and ${\gamma}-flux$. The high temperature is especially generated over $2500^{\circ}C$ in the center of pellet. Due to the high temperature, many problems occur, such as melting of pellet and declining of heat transfer between cladding and coolant. [n this study, it is attempted 10 minimize the temperature of pellet using optimization method about geometric variables. For thermal and optimization analysis or structure. the finite element method code. ANSYS 5.7 is used. In this procedure. subproblem approximation method is used to the optimization methods. Through the optimum design process, the temperature of sealed basket type is reduced from $2537^{\circ}C$ to $2181^{\circ}C$ and the temperature of open basket type is reduced from $2560^{\circ}C$ to $2106^{\circ}C$.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.10a
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• pp.1055-1058
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• 2005

In the ACPF(Advanced spent nuclear fuel Conditioning Process Facility), the spent fuel pellets which are highly radioactive materials are separated with its clad and are fed into the next conditioning process. For this, at the other facility called PIEF(Post Irradiation Examination Facility) a spent fuel rod, 3.5 m long, is cut by 25 cm long which is suitable length fur the decladding process. These rod-cuts are packed into the capsule and are moved to the ACPF. Once the capsule is unloaded in the ACPF, the rod-cut is taken out one-by-one from the capsule and installed on the decladding device. In these processes, the crushed spent fuel pellet can be scattered inside the facilities and thus it contaminate the hot cell. In this paper, we developed the specially designed capsule which prevents the pellets scattering and remarkably reduces the leading and unloading time of the rod-cuts.

• 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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• 2005.05a
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• pp.385-386
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• 2005

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.323-330
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• 2002

Various types of compounds were tested with the aspects of decomposition and formation of residue in a $CO_2$ or 7H$_2$+93$N_2$ atmosphere. The evaporation temperature range of each compound was determined from thermogravimetric curve. Decomposition of dicarbon amide, stearic acid, acrowax and zinc stearate was studied by thermogravimetry in $CO_2$ or in 7H$_2$+93$N_2$ atmosphere. All compounds were decomposed in $CO_2$ atmosphere at lower than 40$0^{\circ}C$, but the residue, ZnO remained for zinc stearate. ZnO did not decompose in $CO_2$ atmosphere up to 130$0^{\circ}C$, but reduced into Zn metal and disappeared in the temperature range of $600^{\circ}C$ to 120$0^{\circ}C$ in 7H$_2$+93$N_2$ atmosphere. The effect of residue, which trapped in closed pores of sintered pellet, on the thermal stability was studied using the resintering test at 1$700^{\circ}C$ in 7H$_2$+93$N_2$ atmosphere. In the case of oxidative sintered pellet with admixing zinc stearate, the cavity formation accompanied with a density drop after resintering is due to the pressure of the Zn gases trapped in the isolated pores.

• Journal of the Korean Crystal Growth and Crystal Technology
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• v.17 no.2
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• pp.75-81
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• 2007

Densification and grain growth have been investigated in 5 wt% $U_3O_8$ seeded $UO_2$ and compared with those of the common $UO_2$ pellet. $UO_2$ compacts and $U_3O_8$ seeded $UO_2$ compacts were sintered at $1300{\sim}1700^{\circ}C$ for 0 h to 4 h. Density and grain size of the sintered pellets were measured by the water immersion method and the image analyzer, respectively. The seeded pellet has a slightly lower density during the intermediate sintering stage. However, the difference of density between two pellets decreases to about 0.5%TD with increasing the sintering temperature. The grain size of the two pellets is similar until $1600^{\circ}C$ but that of the seeded pellet rapidly increases with increasing the sintering temperature.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3516-3525
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• 2022

The high capture cross-section (𝜎_c) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high 𝜎_c as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd₂O₃ and Er₂O₃ respectively. Models comprise nuclear fuel with a homogeneous mixture of Er₂O₃, AmO₂, SmO₂, and HfO₂ with Gd₂O₃ as well as the coating of PaO₂ and ZrB₂ on the Gd₂O₃ pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er₂O₃ and SmO₂ showed expected results namely higher values of these parameters than the reference design of Gd₂O₃ at the early burnup period. The highest value was found for the model of PaO₂ and Gd₂O₃ followed by ZrB₂ and HfO₂. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd₂O₃ and absorber coating thickness.

• Resources Recycling
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• v.11 no.4
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• pp.3-10
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• 2002

The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.