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Aleksey L. Izhutov,Valeriy V. Iakovlev,Andrey E. Novoselov,Vladimir A. Starkov,Aleksey A. Sheldyakov,Valeriy Yu Shiishin,Vladimir M. Kosenkov,Aleksandr V. Vatulin,Irina V. Dobrikova,Vladimir B. Suprun 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.7
The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel thatwas irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods andmini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrixand ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ~ 60%235U; the mini-rods were irradiated to an averageburnup of ~ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard tothe formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%.
Izhutov, Aleksey.L.,Iakovlev, Valeriy.V.,Novoselov, Andrey.E.,Starkov, Vladimir.A.,Sheldyakov, Aleksey.A.,Shishin, Valeriy.Yu.,Kosenkov, Vladimir.M.,Vatulin, Aleksandr.V.,Dobrikova, Irina.V.,Suprun, V Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.7
The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ${\sim}60%^{235}U$; the mini-rods were irradiated to an average burnup of ${\sim}85%^{235}U$. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%.
Toward Zero Micro/Macro-Scale Wear Using Periodic Nano-Layered Coatings
Penkov, Oleksiy V.,Devizenko, Alexander Yu.,Khadem, Mahdi,Zubarev, Evgeniy N.,Kondratenko, Valeriy V.,Kim, Dae-Eun American Chemical Society 2015 ACS APPLIED MATERIALS & INTERFACES Vol.7 No.32
<P>Wear is an important phenomenon that affects the efficiency and life of all moving machines. In this regard, extensive efforts have been devoted to achieve the lowest possible wear in sliding systems. With the advent of novel materials in recent years, technology is moving toward realization of zero wear. Here, we report on the development of new functional coatings comprising periodically stacked nanolayers of amorphous carbon and cobalt that are extremely wear resistant at the micro and macro scale. Because of their unique structure, these coatings simultaneously provide high elasticity and ultrahigh shear strength. As a result, almost zero wear was observed even after one million sliding cycles without any lubrication. The wear rate was reduced by 8–10-fold compared with the best previously reported data on extremely low wear materials.</P><P><B>Graphic Abstract</B> <IMG SRC='http://pubs.acs.org/appl/literatum/publisher/achs/journals/content/aamick/2015/aamick.2015.7.issue-32/acsami.5b05599/production/images/medium/am-2015-05599v_0008.gif'></P><P><A href='http://pubs.acs.org/doi/suppl/10.1021/am5b05599'>ACS Electronic Supporting Info</A></P>
Reprocessing of simulated voloxidized uranium–oxide SNF in the CARBEX process
Alexander V. Boyarintsev,Sergei I. Stepanov,Galina V. Kostikova,Valeriy I. Zhilov,Alexander M. Chekmarev,Aslan Yu. Tsivadze 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.7
The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of Na2CO3 or (NH4)2CO3 and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or Na2CO3–H2O2 and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values 103–105. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.
Separation and purification of elements from alkaline and carbonate nuclear waste solutions
Boyarintsev Alexander V.,Stepanov Sergei I.,Kostikova Galina V.,Zhilov Valeriy I.,Safiulina Alfiya M.,Tsivadze Aslan Yu 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.2
This article provides a survey of wet (aqueous) methods for recovery, separation, and purification of uranium from fission products in carbonate solutions during the reprocessing of spent nuclear fuel and methods for removal of radionuclides from alkaline radioactive waste. The main methods such as selective direct precipitation, ion exchange, and solvent extraction are considered. These methods were compared and evaluated for reprocessing of spent nuclear fuel in carbonate media according to novel alternative non-acidic methods and for treatment processes of alkaline radioactive waste.