• Nuclear Engineering and Technology
• /
• v.28 no.4
• /
• pp.366-372
• /
• 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.

• Proceedings of the Korean Society for Technology of Plasticity Conference
• /
• 2004.05a
• /
• pp.109-113
• /
• 2004

Slitting device is equipment to separate spent fuel of 250 mm rod cut pellets and hull in order to supply required $UO_2$ pellets through the dry pulverizing/mixing device. For development of its device, We have analyzed slitting programs so that the existing device is modified an appropriate scale in the advanced spent fuel conditioning process. The results of the analysis, we added the automatic separation function of pellets and hull, After slitting. Also, we have concentrated on reducing the operation time so that the support and the body of a slitting blade could have been established in the single structure to be easily maintained. It is based on a design and manufacture of a testing device and we have performed an efficiency evaluation. We have analyzed the results of efficiency tests of the slitting device and get the specification of the slitting device. we complete the basic design of the slitting device by using of these data. Therefore, We apply to a basic data when manufacturing a slitting device.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1714-1723
• /
• 2020

This study aims at characterizing the rheological properties of uranium oxide powders for nuclear fuel pellets manufacturing. The flowability of these powders must be compatible with a reproducible filling of press molds. The particle size distribution is known to have an impact on the rheological properties and fine particles (<100 ㎛) are suspected to have a detrimental effect. In this study, the impact of the particle size distribution on the rheological properties of UO₂ powders was quantified, focusing on the influence of fine particles. Two complementary approaches were used. The first approach involved characterizing the powder in a static state: density, compressibility and shear test measurements were used to understand the behavior of the powder when it is transitioned from a static to a dynamic state (i.e., incipient flow conditions). The second approach involved characterizing the behavior of the powder in a dynamic state. Two zones, corresponding to two characteristic behaviors, were demonstrated for both types of measurements. The obtained results showed the amount of fines should be kept below 10 % wt to ensure a robust mold filling operation (i.e., constant mass and production rate).

• Nuclear Engineering and Technology
• /
• v.41 no.1
• /
• pp.39-52
• /
• 2009

Theory-based models and high performance simulations are briefly reviewed starting with atomistic methods, such as Electronic Structure calculations, Molecular Dynamics, and Monte Carlo, continuing with meso-scale methods, such as Dislocation Dynamics and Phase Field, and ending with continuum methods that include Finite Element and Finite Volume. Special attention is paid to relating thermo-mechanical and chemical properties of the fuel to reactor parameters. By inserting atomistic models of point defects into continuum thermo-chemical calculations, a model of oxygen diffusivity in $UO_{2+x}$ is developed and used to predict point defect concentrations, oxygen diffusivity, and fuel stoichiometry at various temperatures and oxygen pressures. The simulations of coupled heat transfer and species diffusion demonstrate that including the dependence of thermal conductivity and density on composition can lead to changes in the calculated centerline temperature and thermal expansion displacements that exceed 5%. A review of advanced nuclear fuel performance codes reveals that the many codes are too dedicated to specific fuel forms and make excessive use of empirical correlations in describing properties of materials. The paper ends with a review of international collaborations and a list of lessons learned that includes the importance of education in creating a large pool of experts to cover all necessary theoretical, experimental, and computational tasks.

• Korean Journal of Veterinary Research
• /
• v.38 no.3
• /
• pp.652-658
• /
• 1998

The efficiency of nuclear transfer using donor embryos originated from ovum pickup(OPU) and activated recipient cytoplasts were examined for induction of twinning in Korean native cattle(KNC). After aspiration of follicle by OPU, regardless of the vacuum applied, we obtained same result in proportions of recovered cumulus-oocyte complex (COCs) with compact cumulus. Under electric stimulation(1.0kV/cm DC for $40{\mu}s$), most of activated oocytes proceed to anaphase II/telophase II within 3h(84.7%). In the treatment of oocyte activation, the preactivation which was performed before fusion had significant effect on the developmental rates to morula/blastocyst stage(9.4 vs 4.0%). In embryo transfer of nuclear transferred embryos, we obtained 2 twins from KNC recipients and 1 twin from a Holstein recipient. Our results showed that it is possible to obtain twins using nuclear transfer technique in KNC.

• Water for future
• /
• v.27 no.2
• /
• pp.121-128
• /
• 1994

Reduced trend of surface velocity, length of the separated drop area and width of potential core have been verified through experimental study of flow characteristics at the step down. To inverstigate sediment behaviors, experimental study which involved accumulated sediment transport reducing water velocity in the separated drop area was performed. From the experimental results, surface velocity, length of the separated drop area and width of potential core were formulated, and calculated output was corroborated by experimental outcome. Furthermore an examination of the parameter which is defined by $q_{sf}$/$q_{uo}$ was performed by detecting sediment in the separated drop area. Therefore these experiments can express the phenomena of flow characteristics and sediment behaviors at the step down.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2007.11a
• /
• pp.305-306
• /
• 2007

ACP(Advanced Spent Fuel Conditioning Process)의 금속전환로에 $U_3O_8$을 공급하기 위하여 20 kgHM/batch의 $UO_2$ 펠릿(pellets)을 처리할 수 있는 건식분말화 장치가 개발되고있다. 건식분말화 장치는 500 $^{\circ}C$온도에서 공기를 공급하여 일정한 입도범위의 균질한 $U_3O_8$을 만든다. 이런 건식 분말화 장치의 효율을 높이기 위해서는 반웅로에 불어 넣어주는 공기의 유량을 증가시킬 필요가 있다. 하지만 공기와 반응하여 생성되는 $U_3O_8$ 입자는 그 크기가 최소 3 ${\mu}$m 정도로 매우 미세하여，반응로 출구를 통해 외부로 빠져나갈 가능성 이있다. 이를 방지하기 위해 분말화 장치 출구 바깥에는 필터가 설치되어 있으나 공기와 함께 $U_3O_8$ 입자가 계속해서 빠져 나갈 경우 입자로 인해 필터가 막혀 제 기능을 할 수 없게 된다. 따라서 건식 분말화 장치는 미세한 $U_3O_8$ 입자가 반응로 밖으로 빠져나가지 않도록 입구에서의 공기 유량을 일정 수준 이하로 조절해주는 것이 필요하다. 이 연구의 목적은 초기 유량으로부터 유량을 점점 증가시키면서 시간변화에 따른 입자 거동 특성을 해석하며， 결과로부터 주어진 크기의 타원입자에 대해 최대 허용 공기 유량을 결정하고자한다. 이 해석을 위해 유동과 입자를 동시에 해석할 수 있는 ANSYS-CFX 5.7.1과 ANSYS-CFX 10.0 두 가지의 소프트웨어가 사용되었다. 해석 결과를 바탕으로 좀더 정확한 유량 한계치 계산을 위해 추가로 수행되어야 할 해석에 대해 제안하였다.

• Nuclear Engineering and Technology
• /
• v.29 no.3
• /
• pp.201-210
• /
• 1997

A thermal conductivity correlation has been proposed which can be applied to high turnup fuel by considering both of thermal conductivity with turnup across fuel pellet and additional degradation at pellet rim due to very high porosity. In addition, a correlation has been developed that can estimate the porosity of rim region as a function of rim burnup under the assumptions that all the produced fission gases are retained in the in porosity and threshold pellet average burnup required for the formation of rim region is 40 MWD/㎏U. Rim width is correlated to rim burnup using measured data. For the RISO experimental data obtained at pellet average turnup of 43.5 MWD/㎏U for three linear heat generation rates of 30, 35 and 40 ㎾/m, radial temperature distributions ore calculated using the present correlation and compared with the measured ones. This comparison shows that the present correlation gives the best agreement with the measured data when it is combined with the HALDEN's correlation for thermal conductivity considering its degradation with burnup. Another comparison with the HALDEN's measured fuel centerline temperature as a function of burnup at 25 ㎾/m up to about 44 MWD/㎾U also suggest that the present correlation yields the best agreement when it is combined with the HALDEN's thermal conductivity.

• Nuclear Engineering and Technology
• /
• v.37 no.4
• /
• pp.317-326
• /
• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2830-2838
• /
• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.