고온에서 이산화우라늄 소결체와 지르코늄 피복관의 상호반응 연구

이건용 경희대학교 대학원 2019 국내석사

원자력발전소는 정상상태에서 노심을 냉각하기 위하여 냉각재를 이용한다. 그러나 LOCA(Loss of Coolant Accident)와 같이 냉각재가 유출되는 사고에서는 냉각재가 노심에 제대로 공급되지 않아 핵연료가 용융될 수 있다. 온도의 증가에 따라 핵연료의 용융이 시작될 경우 이산화우라늄과 지르코늄 피복관이 액화될 수 있으며 반응이 지속적으로 진행될 경우 서로 혼합되게 된다. 실제로 일본의 후쿠시마 원전과 미국의 TMI 원전 사고에서는 중대사고 발생으로 핵연료의 급격한 온도상승으로 노심 용융사고가 발생하였다. 증대사고로 인한 노심용융물은 사고과정 및 이를 처리하기 위한 연구가 중요하다. 향후 노심용융물의 냉각, 재처리, 영구처분 등 중대사고 이후 대처를 위해서는 온도에 따른 이산화우라늄, 지르코늄 피복관의 상호반응과 생성물에 대한 분석이 필수적이며 이러한 자료를 바탕으로 향후 중대사고 발생 시 발생되는 용융물의 형태와 용융 거동에 대한 예측이 가능하다. 본 연구에서는 이산화우라늄과 지르코늄 피복관의 1200℃ 이상의 온도에서 반응으로 생성되는 생성층을 규명하고 이산화우라늄과 지르코늄 산화물의 반응으로 생성되는 (U,Zr)O2 형성에 영향을 미치는 인자를 도출하고자 하였다. 이산화우라늄과 지르코늄 피복관의 상호반응으로 인한 생성층 분석을 위하여 [UO2.0 디스크 – Zry4 피복관] , [UO2+x 디스크 – Zry4 피복관]을 각각 접합하여 1300℃에서 반응시켜 실험을 진행하였으며 XRD, SEM/EDS를 이용하여 생성층에 대한 분석을 진행하였다. 실험 결과 [UO2.0 디스크 – Zry4 피복관] 의 1300℃ 하에서 반응 시 [α-Zr+(U,Zr)],[(U,Zr)],[α-Zr],[β-Zr] 순으로 생성층이 형성되는 것을 확인 할 수 있었으며 [UO2+x 디스크 – Zry4 피복관] 반응의 경우 [(U,Zr)O2],[α-Zr],[β-Zr] 순으로 생성층이 형성되는 것을 확인 할 수 있었다. UO2 디스크에서 지르코늄 피복관 으로의 산소 확산으로 [β-Zr] 층이 [α-Zr]으로 산화되는 현상을 관찰하였으며 이에 따라 MATLAB과 기존문헌자료를 이용하여 두 가지 디스크의 종류에 따른 [α-Zr]층의 성장 속도를 계산, 비교하였다. 이산화우라늄과 지르코늄 산화물의 상호반응으로 인한 (U,Zr)O2 생성의 영향을 미치는 인자를 파악하기 위하여 UO2.0 분말시료와 ZrO2 분말시료를 혼합하여 실험에 사용하였으며 각각 Zr의 몰분율, 산화/환원분위기, 산소분압에 따른 영향을 분석하였다. 실험 결과 Zr의 몰분율이 증가함에 따라 (U,Zr)O2의 격자상수가 감소함을 확인하였으며 산화분위기에서는 산화반응으로 인하여 생성된 U3O8이 1300℃ 이상의 온도에서 UO2로 분해되는 특성으로 인해 (U,Zr)O2, U3O8 이 함께 생성되며 환원분위기에서와 다른 생성물을 가지는 것으로 관측되었다. 산소분압에 따른 영향을 알아보기 위하여 진행한 실험에서는 UO2와 ZrO2의 반응으로 인하여 생성된 (U,Zr)O2의 격자상수가 산소 분압의 증가에 따라 감소하는 결과를 얻을 수 있었다. 본 연구에서는 이산화우라늄과 지르코늄 피복관의 상호반응을 예측하였으며, 이를 통하여 생성물 형성에 영향을 미치는 인자를 도출하였다. 해당 결과는 중대사고시 노심의 용융반응을 예측하는 기초자료로 사용될 것으로 기대되며, 아울러 사고 이후 노심용융물의 재처리, 처분에 필요한 기초자료로 사용 될 수 있다. The interaction between UO2 and Zirconium cladding is very important in normal and off-normal sates. At high fuel burnup, there is a possibility that UO2 pellets will expand and contact the cladding. When Zr cladding contacts UO2 pellets in this way, it absorbs the oxygen from the UO2 and becomes oxidized. Meanwhile, the UO2 is reduced to UO2-x and partially produces a eutectic phase. Above 800°C, β-Zr is in a stable phase but if it is then oxidized to α-Zr, the cladding becomes brittle. In addition, in the case of defective nuclear fuel, the inside of the cladding is oxidized by steam, and the oxide film may grow and react with the UO2 pellet. In the situation mentioned above, ZrO2 and UO2 form a bonding layer between the cladding and the pellet. Due to the long-term storage of the high burnup spent nuclear fuel, an analysis of the formation and behavior of theses bonding layer is important. UO2 interaction with Zr cladding above 1200°C has been reported to form five layers, namely [UO2],[α-Zr+(U-Zr)],[(U-Zr)],[α-Zr], and [β-Zr] [1]. However, hyper-stoichiometric UO2 (UO2+x) forms a different layer than stoichiometric UO2 (UO2.0) because UO2+x is more oxidized than UO2.0 and is not reduced to UO2-x through contact with the Zr cladding. So, hyper-stoichiometric UO2 (UO2+x) and Zry4 forms four layers. Namely [UO2], [(U-Zr)O2], [α-Zr], and [β-Zr] In our study, we conducted an experiment concerning the interaction between UO2 and Zry4. The temperature was increased from room temperature to 1300°C using Ar+4%H2 gas. After the heat treatment, we used X-ray diffraction (XRD) to analyze the structure of the material and an optical microscope (OM) to analyze the thickness of the interaction layer. we conducted second experiment which is related to the interaction between UO2 and ZrO2. The temperature was heated from R.T to 1350°C and 1400°C using gas with various oxygen pressures. After heat treatment, we used X-ray diffraction(XRD) to analyze the structure of the material. To begin of the UO2 interaction with Zry4. A UO2 disk was stacked on top of Zry4. A vacuum was created in the furnace using a vacuum motor, and Ar+4%H2 gas was then introduced into the furnace to prevent oxidation. For conducting the (U,Zr)O2 experiment. UO2.0 and ZrO2 powders were blended for 20 minutes with the same molar ratio. After blending, the sample was put into the furnace for heat treatment. Heat treatment followed. first, sample was heated up to 1350 °C and 1400°C at 10°C/min and then maintained for three hour. After heat treatment, it was cooled down to room temperature. In this experiment. we use a various gas and condition. For a comparison of the X-ray diffraction patterns in the various partial oxygen pressure and various molar rate of rhe Zr. As a result of the two experiments. We found some result. Unlike the model of UO2.0 and Zry4 interaction. the UO2+x and Zry4 interaction has four layers: [UO2], [(U-Zr)O2], [α-Zr], and [β-Zr]. As the interaction time increases at high temperatures, β-Zr becomes α-Zr. UO2+x is more oxidized than UO2.0 and so it is not reduced to UO2-x by interaction with Zr cladding. As a result, it forms an additional layer to UO2.0-Zr cladding. UO2 produced a (U-Zr)O2 layer on the outside of the disk because Zr diffuses to the UO2 side. Therefore, the XRD peak shifted to a higher point and the unit cell shrank. We conducted experiments using Ar, Ar+H2O and air according to the oxygen pressure.The lattice parameter had a decrease tendency as increasing Zr. Thus Shrinking of unit cell may be due to the intrusion of Zr. As the partial pressure of oxygen increased, And Zr reacted more easily to uranium. Therefore, the peak shifted to a high angle and the lattice parameter was decreased

이산화우라늄(UO2)의 低溫燒結에 관한 硏究

이규암 忠南大學校 1988 국내박사

UO₂소결체는 中性子를 흡수하면 높은 일을 발생하는 核 分裂性 物質로써 용융점이 높고 高溫水에 대한 내식성이 우수하고, 放射線 照射時 고온에서도 칫수가 안정하기 때문에 원자력 발전소의 핵연료로서 가장 많이 사용되고 있다. 일반적으로 UO₂소결체를 제조하는 방법으로서는 이산화 우라늄 압분체를 1700℃ 이상의 高溫에서 長時間 소결하여 高密度의 소결체를 만드는 방법이 널리 사용되고 있으며, 이때 원자로 내에서의 열전도도를 고려하여 O/U비가 2.00인 UO₂소결체를 만들고자 수소간스 분위기에서 소결한다. 그러나 이와같은 방법은 1,700℃의 고온에서 장시간 燒結하기 때문에 전력이 많이 소모되고 고온용 내화벽돌 및 발영체가 내장된 값비싼 소결장비를 사용해야 하는 단점이 있어, 본 연구에서는 저온에서 단시간 내에 소결하여 소결밀도를 低下시키지 않고 고밀도의 소결체를 만드는 방법을 연구하였다. 한국 에너지 연구소(Korea Advanced Energy Research Institute)에서 AUC공정법으로 만든 이산화 우라늄분말(이는 국내에서 최초로 만든 순수한 國産 核燃料 분말임)을 사용하여 만든 UO₂압분체를 CO₂가스 분위기에서 1,100℃에서 1시간 소결한 후 수소 가스 분위기에서 1100℃에서 1시간 환원한 결과, 1700℃ 이상의 고온에서 장시간 소결하여 얻을 수 있는 高密度의 소결체를 만들 수 있었으며 상세한 내용은 다음과 같다. 제1차 연구로써 UO₂압분체를 CO₂가스 분위기에서 1000~1300℃에서 1~8시간 동안 환원하여 만든 소결체의 밀도 및 조직변화를 관찰한 결과 다음과 같다. 1000℃에서 1시간 소결한 소결체의 소결밀도는 10.50g/cc (95.8% T.D)이었고 1100℃에서 소결한 소결체의 소결밀도는 10.66g/cc(97.3% T.D)로 최고치에 이르렀으며, 이러한 소결밀도는 통상적인 소결방법인 수소가스 분위기에서 1700℃ 이상의 고온에서 장시간 소결하여 얻을 수 있는 고밀도이었다. 1000℃에서 1시간 동안 소결한 소결체의 조직은 2~3㎛의 微細粒子와 50㎛의 粗大結晶粒子가 섞여져있는 2중 구조이었다. 이와같이 CO₂가스 분위기에서 소결된 UO₂소결체의 소결밀도는 일반적으로 原子爐의 核燃料로써 요구되는 규제치(93~98%, T.D)에 비하면 상당히 높은편에 속하므로 원자로의 종류에 따라 요구되는 低密度의 소결체를 제조할 필요성이 대두되었다. 따라서 제2차 연구로써 원료분말에 불량소결체를 공기중에서 500℃에서 1시간 동안 산화시켜 만든 U₃O_8분말을 4%, 8% 및 12%를 첨가하여 압분체를 만든 후, 제1차 연구방법과 동일한 방법으로 소결한 결과 다음과 같다. 同一 온도에서 동일시간 소결할 때 U₃O_8의 첨가량이 4%씩 증가할수록 소결밀도는 0.06g/cc씩 감소하였으며 이는 내부 기공도가 증가하였기 때문이다. CO₂가스 분위기에서 소결된 UO₂소결체의 미세조직은 50㎛의 조대입자에 2~3㎛의 미세입자가 다량 섞여있는 2중 구조로써, 원자로의 핵연료로서 UO₂소결체의 결정입자가 5~25㎛범위로 규제되어야하는 입도 요구사항에는 어긋나는 실정이다. 따라서 제3차 연구로써 분말의 O/U비가 2.17인 원료분말과 이와 같은 원료분말을 300℃ 및 500℃에서 각각 1분간씩 산화시켜 O/U비가 2.19 및 2.49분말을 사용하여 만든 압분체를 제1차 연구방법과 동일한 방법으로 소결한 결과 다음과 같다. 1100℃에서 1시간동안 소결한 소결체의 조직을 조사한 결과 UO_2.17경우 2~3㎛의 미세입자와 50㎛의 조대입자가 섞여져 있었으며, UO_2.19경우 30㎛의 조대입자가 주종을 이루었고 2~3㎛의 미세입자가 국부적으로 존재하였으며 UO_2.49경우 30㎛의 조대입자가 주종인 균일한 조직이었다. 이와같이 제1차, 2차 및 3차 연구방법으로 제조한 UO₂소결체를 원자력 발전소의 핵연료로써 사용할 수 있느냐의 여부는 실제로 발전소 내에서 放射線을 照射시켰을때 칫수변화가 작은 안정한 소결체임이 확인되어야 한다. 방사선조사시 소결체가 高密化되는 현상과 소결체 제조시 소결의 말기단계에서 소결체가 고밀화되는 현상은 서로 유사성이 있으므로 본 연구에서 제조한 소결체를 170℃의 수소가스 분위기에서 4시간 동안 재소결하여 소결밀도를 관찰한 결과 다음과 같다. 1000℃에서 1시간, 2시간 소결한 경우에만 0.01~0.15 g/cc 범위로 증가하였으며, 나머지 온도 및 시간 구간에서는 0.05 g/cc 이하로 감소하는 등 고밀화 (densification) 및 swelling 현상이 크게 나타나지 않았다. 따라서 방사선 조사시 安定性있는 소결체임을 예측할 수 있었다. The green pellets using the KAERI (Korea Advanced Energy Research Institute) UO₂powder made by AUC process were sintered in CO₂atmosphere (1~8hrs, 100~1300℃) and were subsequently reduced at 1100℃ for 1 hr in the H₂atmosphere. this study was aimed at investigating the effects of the variation of sintering time (l~8 hrs), U₃O_8(4,8,12%) addition and O/U ratio (2.17,2.19,2.49)of powder on the density, microstructure and thermal stability of UO₂sintered pellets. 1. When green pellets (0/U ratio=2.17) were sintered from 1000℃ to 1300℃ by changing sintering time (1~8 hrs) in CO₂atmosphere and reduced for 1 hr at 1100℃ in hydrogen atmosphere, the following results were obtained. 1) The UO₂pellets sintered at 1000℃ for 1 hr showed a density 10.50 q/cc (95.8% T.D). whereas those at 1100℃ showed the maximum density of 10.66 g/cc (97.3% T.D). this density corresponds to the sintered density which can be obtained by the conventional sintering process over 1700℃ in hydrogen atmosphere. 2) The density of UO₂pellet sintered at 1000℃ was increased from 10.50 g/cc to 10.59 g/cc as a function of sintering time (1~8 hrs). On the contrary, the sintered density was decreased with the sintering time over 1100℃, which was attributable to the increase of the closed porosity. 3) UO₂pellets sintered at 1300℃ for 1 hr in hydrogen atmosphere showed density of 10.30 g/cc (93.98%), which was noticeablv lower than the density of UO₂pellets sintered in CO₂atmosphere. It was found that sintering at the temperature was underway. 4) UO₂pellets sintered at 1000℃ for 1 hr showed a bimodal structure containing small grains of 2~3μm and large grains of 50μm, whereas the number of small grains in UO₂ pellets sintered at 1300℃ for 8 hrs were greatly reduced and the average size of grains was 20μm. With sintering time and temperature, thus, the small qrain qrew while large grains did not qrow. 2. The results of U₃O_8(4.8 and 12%) addition to UO₂powder are the followings. 1) At a sinterinq temperature and time, the sintered density was reduced by 0.06 g/cc with an increase of every 4% of U₃O_8 addition due to the increase of internal porosity. 2) The sintered density at 1000℃ was increased in the range of 0.16 g/cc to 0.19 q/cc with sintering time from 1 to 8 hrs regardless of U₃O_8 addition, while the sintered density was decreased in the range of 0.26 g/cc~0.30 g/cc with sintering time when sintering temperature was above 1100℃. This tendency was resulted from the increase of the number and size of large pores. 3) The large grain size decreased with the addition of U₃O_8 when UO₂pellets were sintered at 1000℃ for 1 hr, and the large grain size of UO_2 pellets sintered at 1300℃for 8 hrs was under 25μm. 4) It was found that the addition of U₃O_8 acted not only as a pore-former but also as an inhibitor of grain coarsening. 3. The effect of 0/U ratio (2.17,2.19 and 2.49) of powder on the microstructure and density of the sintered pellets were investigated. 1) The resultant microstructure of UO₂pellets sintered at 1100℃ for 1 hr in CO₂atmosphere was as follows. UO_2.17 : mixture of fine grains of 2~3 μm and large grains of 50μm. UO_2.19 : grain sizes of 30μm were predominant and small grains were locally existed. UO_2.19 : a homogeneous microstructure containing 30 μm of grains size predominantly. 2) In the case of pellets sintered at 1300℃ for 8 hrs, the sintered pellets using UO_2.17 powder showed mainly large grains and the average grain size was 28μm, while sintered pellets using UO_2.19 and UO_2.49 powder showed a decrease in the number of grain size. 3) Above 1100℃, the sintered density was decreased with temperature and with O/U ratio as well. 4. To investigate the thermal stability of UO℃ pellets manufactured for this study, the pellets were sintered again at 1700℃ for 4 hrs in hydrogen atmosphere. The pellets sintered at 1000℃ for 1 hr and 2 hrs in CO₂atmosphere only showed an increase in density in the range of 0.01~0.15 q/cc. For other ranges of sintering temperature and time, UO₂sintered pe1lets decreased in density with in 0.05 g/cc range and did not show any remarkable densification and swelling phenomena. Therefore, it could be predicted that the UO₂pellets should stable under irradiation in nuclear reactors.

이산화우라늄 분말의 특성조절 연구

안창모 韓南大學校 1995 국내석사

이산화우라늄의 염소화반응에 의한 사염화우라늄의 제조

김상중 전북대학교 대학원 1996 국내석사

Stainless steel 304 피복 이산화우라늄 핵연료의 발전용 원자로내 열적, 기계적 성질에 관한 연구

박진영 서울大學校 大學院 1975 국내석사

소결 분위기와 Cr₂O₃의 첨가량에 따른 이산화 우라늄의 치밀화 및 입자성장거동

조영철 忠南大學校 大學院 2005 국내석사

In order to enhance the economical efficiency of the nuclear fuel, which can be used in high burn up and long term cycle plant, is required. Increase in fission gas release can be a major obstacle to increasing the burnup. A large grain size fuel is seen as desirable for the reduction of fission gas release. In the present study, the effect of additives on the sintering behavior has been investigated. Sintering and grain growth of compacted uranium dioxide powder pellets doped with Cr_(2)O_(3). The influence of parameters such as the oxygen potential of the sintering atmosphere and pellet green density on the final microstructure was studied. Dilatometric analysis and monitoring of microstructural development revealed a phenomenon of grain growth promoting densification. The existence of a eutectic between Cr and Cr_(2)O_(3) is also discussed. The function of green density, so that it is possible to propose a solubility limit for Cr_(2)O_(3) in stoichiometric UO_(2) at 1700℃.

紫外線에 의한 二酸化우라늄의 溶解 및 몰리브덴-99의 分離 特性

황두성 충남대학교 대학원 2002 국내박사

This study suggested new dissolution technique of uranium dioxide target by using UV and investigated the separation characteristic of molybdenum-99 to improve production technology of fission molybdenum-99. In the dissolution of UO_2 target, the study was to dissolve UO_2 under conditions of low temperature and low nitric acid concentration by using UV. Dissolution tests of UO_2 powder, green pellet, and sintered pellet were carried out in a 2 M nitric acid solution under UV irradiation. Hydrogen peroxide and metal nitrate salts were used as oxidant in order to accelerate the dissolution rate of UO_2 sintered pellet. The light source is a mercury lamp emitting 254: nm wavelength. In the precipitation of molybdenum by α -benzoinoxime, the study investigated the formation property of precipitate and improved the demerit of existing process using hydrogen peroxide that acts as a disturbing compound in a subsequent purification process. The property of molybdenum precipitate was investigated by SEM, FTIR, TG-DTA, and XRD. The precipitate was produced by adding α -benzoinoxime solution of 2 wt% dissolved in 0.4 M NaOH to 1 M nitric acid solution containing molybdenum and other elements. Tracers of radioactive isotopes were used to examine closely the removal efficiency of other elements containing with a very small amount. The results are as follows. From the result of UV spectra, it was found the fact that nitric acid solution absorbs wavelengths below 350 nm and thereby nitrate ion is converted to excited nitrate ion. And then, the excited nitrate ion was considered to dissolve uranium dioxide. The dissolution rate of uranium dioxide powder with high specific surface area was about five times faster with UV irradiation than without UV irradiation in 2 M nitric acid solution. The sequence in the dissolution rate of uranium dioxide was uranium dioxide powder > green pellet > sintered pellet in 2 M nitric acid solution. In particular, uranium dioxide sintered pellet was hardly dissolved even under UV irradiation. This was interpreted to be difficult for sintered pellet to be oxidized because its surface is very dense. However, this was considerably improved by adding hydrogen peroxide as oxidant, which could increase tile surface area with etching the surface of sintered pellet. The dissolution rate of uranium dioxide sintered pellets in the simulated solution containing the elements such as cesium, strontium, molybdenum, zirconium, ruthenium and neodymium was considerably increased. When metal ions such as molybdenum and ruthenium in these elements were included in the solution, the dissolution rate of sintered pellet was increased to more than two times. This result indicates that uranium dioxide can be oxidized by these elements activated with absorbing the light. Additionally, when the pulverized sintered pellets were dissolved in the simulated solution under UV irradiation, dissolution rate was increased about five times more than that of sintered pellets without pulverization. The chemicals such as nitrite ion and hydrogen peroxide as reaction products were detected during dissolution of uranium dioxide in nitric acid solution under UV irradiation. Based on this result, a new dissolution mechanism of uranium dioxide in nitric acid solution by photochemical reaction was suggested. The precipitate was composed of α -benzoinoxime-Mo precipitate and re-precipitate of α -benzoinoxime added excessively for increasing precipitation efficiency of molybdenum. As results of FTIR and TG-DTA, it was confirmed that α -benzoinoxime-Mo precipitate was formed by reaction of two α -benzoinoxime molecules and one MoO_2^+2. α - Benzoinoxime-Mo precipitate of these precipitates was dissolved in 0,4 M NaOH solution within 5 minutes without hydrogen peroxide, which affects a subsequent alumina adsorption process badly. The recovery yield of molybdenum was similar to the case that added hydrogen peroxide. Hydrogen peroxide induced only the rapid dissolution of α -benzoinoxime re-precipitate. Besides, the dissolution method without hydrogen peroxide was favorable in the purification aspect because zirconium and ruthenium were contained as a small fraction of 1.3 and 7.7%, respectively, in dissolving solution. Such a selective dissolution of molybdenum made possible to decrease of about 50% organic quantity of the dissolving solution to be fed in a subsequent silver coated activated carbon adsorption process. As results of tracer experiments, physical treatment such as changing the adding method of α -benzoinoxime didn't affect at the precipitation behavior of molybdenum-99 and the other nuclides. It was optimum conditions of molybdenum-99 precipitation process by α -benzoinoxime that precipitate formed by the batch type addition of α -benzoinoxime was dissolved during 20 minutes in 0.4 M sodium hydroxide solution. In this condition, yield of molybdenum-99 was 97.1% and decontamination factor of iodine-131, ruthenium-103, and zirconium-95 was 4.8, 45.5, and 27.8, respectively. And also, the other nuclides could be removed almost.

Ab initio modeling of uranium dioxide on vacancy defects, their complexes, and Xe diffusion

윤연숙 Kyung Hee Univ. 2006 국내박사

추출 크로마토그래피와 유도 결합 플라스마 원자 방출 분광기를 이용한 이산화 우라늄 분말 중 불순물 분석

최광순 忠南大學校 1993 국내석사

Hydrogeochemical and isotope geochemical features of the Han River system

유종식 서울대학교 대학원 2008 국내박사