• Nuclear Engineering and Technology
• /
• 제36권1호
• /
• pp.104-111
• /
• 2004

The melting temperatures of $UO_2,\;UO_2-6wt\%Gd_{2}O_3,\;UO_2-12wt\%Gd_{2}O_3,\;UO_2-2wt\%Er_{2}O_3,\;and\;UO_2-4wt\%Er_{2}O_3$ fuels were measured. Fuel materials were loaded in a tungsten capsule of which shape met the black body condition. The melting temperature was measured by the thermal arrest method during heating of the capsule in an induction furnace. The measured melting temperature of $UO_2$ fuel was $2815{\pm}20^{\circ}C$. The solidus and liquidus temperatures of $UO_2-Gd_{2}O_3\;and\;UO_2-Er_{2}O_3$ had also been measured, and it was observed that the solidus temperatures of them were lower than the liquidus temperature by $15{\sim}25^{\circ}C$. Measured melting temperatures of $UO_2,\;UO_2-Gd_{2}O_3\;and\;UO_2-Er_{2}O_3$ fuels were as follows:

• Nuclear Engineering and Technology
• /
• 제40권1호
• /
• pp.21-36
• /
• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• 한국결정성장학회지
• /
• 제17권2호
• /
• pp.75-81
• /
• 2007

[ $UO_2$ ] 소결체와 $U_3O_8$종자를 5wt% 첨가한 $UO_2$ 소결체의 치밀화 과정 및 입자성장 양상을 소결 온도 및 시간을 변수로 하여 분석하였다. $UO_2$ 성형체와 5wt% $U_3O_8$ 종자 첨가 성형체를 수소분위기에서 $1300^{\circ}C$에서 $1700^{\circ}C$로 온도를 올려가며 0시간에서 4시간 소결하여 밀도와 입자크기를 측정하였다. $1300^{\circ}C$까지는 종자 첨가에 상관없이 거의 같은 밀도를 가졌지만 온도가 올라가면서 종자 첨가 소결체의 치밀화가 저하되었다가 $1700^{\circ}C$ 근처에서 거의 비슷한 밀도를 가지게 된다. 입자성장의 경우, $1600^{\circ}C$에서는 종자 입자를 제외하면 기지상의 입자 크기는 거의 비슷하지만 $1700^{\circ}C$ 이후에서는 종자첨가 $UO_2$ 소결체의 입자성장이 종자가 첨가되지 않을 경우에 비하여 2배 이상 빠르게 진행되었다.

• 한국재료학회지
• /
• 제30권10호
• /
• pp.533-541
• /
• 2020

UO₂ kernels, a key component of fuel elements for high temperature gas cooled reactors, have usually been prepared by sol-gel methods. Sol-gel processes have a number of advantages, such as simple processes and facilities, and higher sphericity and density. In this study, to produce 900 ㎛-sized UO₂ particles using an external gelation process, contact length extension of the NH₃ gas of the broth droplets pass and the improvement of the gelation device capable of spraying 14 M-NH₄OH solution are used to form 3,000 ㎛-sized liquid droplets. To produce high-sphericity and high-density UO₂ particles, HMTA, which promotes the gelation reaction in the uranium broth solution, is added to diffuse ammonium ions from the outside of the gelation solution during the aging process and generate ammonium ions from the inside of the ADU gel particles. Sufficient gelation inside of ADU gel particles is achieved, and the density of the UO₂ spheres that undergo the subsequent treatment is 10.78 g/㎤; the sphericity is analyzed and found to be 0.948, indicating good experimental results.

• 한국분말재료학회지
• /
• 제16권2호
• /
• pp.115-121
• /
• 2009

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for $UO_2$ kernel preparation. In this study, ADU compound particles were calcined to $UO_3$ particles in air and Ar atmospheres, and these $UO_3$ particles were reduced and sintered in 4%-$H_2$/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to $UO_3$ phases at $500^{\circ}C$. At $600^{\circ}C$, the $U_3O_8$ phase appeared together with $UO_3$. After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric $UO_2$. As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of $U_3O_7$ and $U_4O_9$.

• Nuclear Engineering and Technology
• /
• 제43권2호
• /
• pp.141-148
• /
• 2011

This paper presents the quantitative analysis results of research on the burnup effect on the nuclear fuel cycle cost of BeO-$UO_2$ fuel. As a result of this analysis, if the burnup is 60 MWD/kg, which is the limit under South Korean regulations, the nuclear fuel cycle cost is 4.47 mills/kWh at 4.8wt% of Be content for the BeO-$UO_2$ fuel. It is, however, reduced to 3.70 mills/kWh at 5.4wt% of Be content if the burnup is 75MWD/kg. Therefore, it seems very advantageous, in terms of the economic aspect, to develop BeO-$UO_2$ fuel, which does not have any technical problem with its safety and is a high burnup & long life cycle nuclear fuel.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.391-392
• /
• 2005

• 한국세라믹학회지
• /
• 제33권6호
• /
• pp.631-636
• /
• 1996

The powder characteristics of UO2 via AUC prepared by precipitation from a UN with AC soiution produced from nuclear fuel powder conversion plant and that of the existing facility were compared. Mean particle size of AUC powder was decreased and agglomerates were much occured in case of using the AC solution that that of the gases but other properties such as particle size distribution and shape of particle are thought to be similarly. In compaction of UO2 powder the breaking pressur of agglomerated UO2 powder and the sintered density of final UO2 pellet from AC solution were measured 1.45$\times$108 N/m2 and 10.52 g/cc, These values could be used in nuclear fuel powder fabrication process.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.393-394
• /
• 2005

• 한국재료학회지
• /
• 제10권1호
• /
• pp.55-61
• /
• 2000

$UO_2$분말에 PVA-Al(III) 착물과 AlOOH를 각각 0.03~1.0wt%를 첨가하여 소결체를 제조한 후 소결체 특성을 비교하였다. PVA-Al(III) 착물과 AlOOH는 $1000^{\circ}C$의 소소분위기에서 열분해하는 경우 생성상은 $\theta-Al_2O_3$이었다. 순수 $UO_2$분말에 비해 AlOOH가 첨가된 혼합 분말의 곁보기 밀도는 더 높았고, PVA-Al(III) 착물이 첨가된 혼합 분말은 더 낮은 겉보기 밀도를 보였다. AlOOH가 첨가된 소결체의 경우 약 $800^{\circ}C$ 부근에서 치밀화가 시작되었지만, PVA-Al(III) 착물이 첨가된 소결체의 경우에는 약 $900^{\circ}C$에서 치밀화가 시작되었다. 기공 크기 분포는 AlOOH가 첨가된 $UO_2$소결체의 경우에 monomodal 형태로, 그리고 PVA-Al(III) 착물의 첨가된 소결체의 경우에는 bimodal 형태로 나타났다. 결정립 크기는 1wt% 첨가시 AlOOH가 첨가된 $UO_2$소결체의 경우에 약 $13\mu\textrm{m}$이었지만, PVA-Al(III) 착물의 첨가된 소결체의 경우에는 약 $\36mu\textrm{m}$까지 성장하는 현저한 효과를 확인하였다.