• Nuclear Engineering and Technology
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• 제40권3호
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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.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• 제28권4호
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• pp.366-372
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• 1996

The simulated spent PWR fuel pellet which is corresponding to the turnup of 33,000 MWD/MTU is prepared by adding 11 fission-product elements to UO$_2$. The simulated spent fuel pellet is treated at 40$0^{\circ}C$ in air (oxidation), at 110$0^{\circ}C$ in air (high-temperature treatment), and at $600^{\circ}C$ in hydrogen (reduction). The product is treated through additional addition and reduction up to 3 cycles. Pellets are completely pulverized by the first oxidation, and the high-temperature treatment causes particle and crystallite to grow and surface to be smooth, and thus particle size significantly increases and surface area decreases. The reduction following the high-temperature treatment decreases much the particle size by means of the formation of intercrystalline cracks. The particle size decreases a little during the second oxidation and reduction cycle and then remains nearly constant during the third and fourth cycles. Surface area of pounder increases progressively with the repetition of oxidation and reduction cycles, mainly due to the formation of Surface cracks. The degradation of surface area resulting from high-temperature treatment is restored by too subsequent resulting oxidation and reduction cycles.

• 한국재료학회지
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• 제3권3호
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• pp.245-252
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• 1993

순수 $UO_2$에 첨가량 변화 및 ball-milling 시간에 따른 (U, Ce)$O_2$ 분말의 특성과 각 조건별로 제조된 분말을 압분 및 소결하여 (U, Ce)$O_2$ 분말 특성에 따른 소결성을 비교 조사하였다. 실험 결과로 부터 ball-milling시간이 길어짐에 따라 입자들은 미세화되고, Ce$O_2$ 함량이 증가할수록 압분, 소결밀도는 저하 하였으며, $CeO_2$는 소결성을 저하시키는 산화물임을 확인하였다. 10wt%,$CeO_2$ 가 첨가된 (U, Ce)$O_2$ 분말의 경우, ball-milling 4시간 수행한 분말의 소결체가 기공의 수도 적고, 구형에 가까왔으며, 소결밀도가 가장 높았다. 이는 4시간 ball-milling한 (U, Ce)$O_2$분말이 비표면적이 크로 그의 packing ratio가 적절하였기 때문이다.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.335-343
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• 1999

The effects of N $b_2$ $O_{5}$ and oxygen potential on the densification and grain growth of U $O_2$ fuel have been investigated.0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$fuel pellets were sintered at 1$700^{\circ}C$ for 4 hours in sintering atmospheres which have various ratios of $H_2O$ to $H_2$ gas. Compared with those of undoped U $O_2$ pellets, the sintered density and grain size of the 0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$ pellet increase under the $H_2O$/ $H_2$ gas ratio of 5.0$\times$10$^{-3}$ to 1.0$\times$10$^{-2}$ and under the $H_2O$/ $H_2$gas ratio of 5.0$\times$10$^{-3}$ to $1.5\times$10$^{-2}$ , respectively. The sintering of U $O_2$fuel pellets containing 0.1 wt% to 0.5 wt% N $b_2$ $O_{5}$ was carried out at 168$0^{\circ}C$ for 4 hours. The enhancing effect of N $b_2$ $O_{5}$ on the sintered density and grain size becomes larger as the N $b_2$ $O_{5}$ content increases. The solubility limit of N $b_2$ $O_{5}$ in U $O_{2}$ seems to be between 0.3 wt% and 0.5 wt%, and beyond the solubility limit the second phase whose composition corresponds near to N $b_2$U $O_{6}$ is precipitated on grain boundary. The enhancement of densification and grain growth in U $O_2$ is attributed to the increased concentration of a uranium vacancy which is formed by the interstitial N $b^{4+}$ ion in the U $O_2$ lattice.

• 한국세라믹학회지
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• 제40권9호
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• pp.916-920
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• 2003

무게 비로 6%의 Gd가 치환된 이산화 우라늄, ( $U_{0.913}$G $d_{0.087}$) $O_2$를 475$^{\circ}C$ 공기 분위기에서 산화시키고 130$0^{\circ}C$ 공기 분위기에서 열처리시킬 때 변화하는 결정 구조, 형상 등을 XRD, SEM 및 EPMA 등을 이용하여 관찰하였다. 입방계 구조의 ( $U_{0.913}$G $d_{0.087}$) $O_2$는 475$^{\circ}C$ 공기 분위기에서 사방정게 구조의 ( $U_{0.913}$G $d_{0.087}$)$_3$ $O_{8}$로 산화되었다. 저온 산화에 의해 생성된 사방정계 130$0^{\circ}C$의 고온에서 열처리하는 동안 사방정계 상과 압방정계 상으로 다시 분리되었다. XRD와 EPMA 관찰결과, 분리된 사방정계 상과 입방정계 상은 각각 $U_3$ $O_{8}$과 ( $U_{0.67}$G $d_{0.33}$) $O_{2+}$x/인 것을 확인하였다. 열처리 동안 일어나는 일련의 산화와 상 분리 과정은 상 반응식으로 나타낼 수 있다. 각 열처리 단계에서의 무게 변화비를 측정하고 상 반응식을 이용하면 (U,Gd) $O_2$에 고용되어 있는 초기 Gd 함량을 정확히 계산할 수 있다.

• 한국정밀공학회지
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• 제30권1호
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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.

• 분석과학
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• 제19권5호
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• pp.407-414
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• 2006

An X-ray diffractometer for spatially resolved X-ray diffraction measurements was developed to identify phase in the narrow (micron-scaled) region of high burn-up fuels and some nuclear materials. The micro-XRD was composed of an X-ray microbeam alignment system and a sample micro translation system instead of a normal slit and a fixed sample stage in a commercial XRD. The X-ray microbeam alignment system was fabricated with a microbeam concentrator having two Ni deposited mirrors, a vertical positioner, and a tilt table for the generation of a concentrated microbeam. The sample micro translation system was made with a sample holder and a horizontal translator, allowing movement of a specimen at $5{\mu}m$ steps. The angular intensity profile of the microbeam generated through a concentrator was symmetric and not distorted. The size of the microbeam was $4,000{\times}20{\mu}m$ and the spatial resolution of the beam was $47{\mu}m$ at the sample position. When the diffraction peaks were measured for a $UO_2$ pellet specimen by this system, the reproducibility ($2{\Theta}={\pm}0.01^{\circ}$) of the peaks was as good as a conventional X-ray diffractometer. For the cross section of oxidized titanium metal, not only $TiO_2$ in an outer layer but also TiO near an oxide-metal interface was observed.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.