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Published: 30 May 2022
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1

Yapıcı, Hüseyin. "Burning and/or transmutation of transuraniums discharged from PWR-UO2 spent fuel and power flattening along the operation period in the force free helical reactor." Energy Conversion and Management 44, no. 18 (November 2003): 2893–913. http://dx.doi.org/10.1016/s0196-8904(03)00068-2.

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2

Brolly, Á., and P. Vértes. "Transmutation: Towards Solving Problem of Spent Nuclear Fuel." Acta Physica Hungarica A) Heavy Ion Physics 19, no. 3-4 (April 1, 2004): 263–71. http://dx.doi.org/10.1556/aph.19.2004.3-4.19.

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3

Tikhomirov, G. V., and A. S. Gerasimov. "THE MAIN PROBLEMS OF THE MANAGEMENT OF RADIOACTIVE WASTE FROM NPP SPENT FUEL USING NUCLEAR TRANSMUTATION." Professor’s Journal. Series: Technical science 3 (September 1, 2019): 41–56. http://dx.doi.org/10.18572/2686-8598-2019-3-3-41-56.

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the main problems associated with research on transmutation, and whichshould be paid attention to by today's young researchers, are formulated. The processes of formation of hazardous nuclides during transmutation in reactor facilities are considered. The goals of transmutation and the choice of nuclides to be transmuted are discussed. The concept of radiotoxicity is explained as a measure of the radiological hazard of radio-active nuclides, based on the maximum permissible concentration of nuclides according to the IAEA standards. The problem of the formation of secondary radioactive nuclides in nuclear fuel during generation of neutrons for transmutation is discussed. The advantages and disadvantages of various methods of transmutation in nuclear installations are con-sidered: inclusion of transmutable nuclides in nuclear fuel in fast reactors, transmutation in specialized thermal and fast transmutation reactor installations and ADS systems. The problem of the accumulation of highly radioactive actinides in a transmutation installa-tion during long-term transmutation and potential hazard of the transmutation instal-lation itself is discussed. The unacceptability of the use of serial nuclear reactors for the transmutation of long-lived fission products has been shown.
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4

Durmaz, Busra, Gizem Bakir, Bugra Arslan, and Huseyin Yapici. "Neutronic analysis of an ads fuelled with minor actinide and designed for spent fuel enrichment and fissile fuel production." Nuclear Technology and Radiation Protection 36, no. 4 (2021): 299–314. http://dx.doi.org/10.2298/ntrp2104299d.

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This paper presents analyses of enrichments of uranium taken out from Canada Deuterium Uranium and pressurized water reactors spent fuels and fissile fuel breeding from thorium in two different helium cooled-accelerator driven system designs, DESIGN A and DESIGN B. In the beginning, the 235U percentages in the uranium fuels taken out from the reactors spent fuels are 0.17% and 0.91%, respectively. Both system cores are fuelled with two different minor actinides compositions extracted from PWR-MOX spent fuels. The DESIGN A has one transmutation zone (enrichment zone) surrounding the fuel core and containing thorium or spent uranium fuels, while DESIGN B has a second transmutation zone (fissile fuel breeding zone) surrounding the first transmutation zone and containing only thorium fuel. In brief, a total of ten cases formed by the combinations of accelerator driven system designs, minor actinides components, and spent uranium with thorium fuels are analysed, which are six in DESIGN A containing one transmutation zone and four in DESIGN B containing two transmutation zones. Lead-bismuth eutectic alloy, a liquid heavy metal, consisting of 45% lead and 55 % bismuth is used as target material in the investigated accelerator driven system. It is assumed that the target is bombarded with 1.2383?1017 protons per second and that the energy of each proton is 1000 MeV. This means a proton beam power of 20 MW. The 3-D and time-dependent neutronic analyses are conducted by using the MCNPX 2.7 and CINDER 90 nuclear code. Both accelerator driven system designs are operated until the values of keff rise to 0.985 to determine the longest operation times that are the effective burn times in all cases. Depending on the design, minor actinide composition, and fuel type (spent UO2 and ThO2), the results obtained at the end of cycle exhibit the effective burn times vary from 300 days to 2050 days, the fuel enrichments can reach up to 2.49-4.23% and the values of gain reach up to 10.8-25.1.
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5

Sadighi, S. K., and R. Sadighi-Bonabi. "The evaluation of transmutation of hazardous nuclear waste of 90Sr, into valuable nuclear medicine of 89Sr by ultraintense lasers." Laser and Particle Beams 28, no. 2 (April 14, 2010): 269–76. http://dx.doi.org/10.1017/s0263034610000145.

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AbstractThe analytical evaluation of the capability of Bremsstrahlung highly directional energetic γ-beam to induce photo transmutation of 90Sr (γ,n) 89Sr is presented. Photo transmutation of hazardous nuclear waste of 90Sr, one of the two main sources of heat and radioactivity in spent fuel into valuable nuclear medicine radioisotope of 89Sr is explained. Based on the calculations, a fairly decent fraction of gamma rays in this range are used in transmuting of 90Sr into 89Sr where according to the available experimental data it is shown that by irradiating a 1-cm thick 90Sr sample with lasers of intensity of 1021 W/cm2 and repletion rate of 100 Hz for an hour, the reaction activity would be 1.45 kBq. It is shown that there is not a linear relationship between the growth of the activity and increasing the laser intensity, but there is a dramatic increase in the growth rate especially between 1020and 1021 W/cm2. In this work, the advantage of photonuclear transmutation over the neutron capture transmutation for 90Sr isotope is also discussed.
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6

Tran, Vinh Thanh, Thanh Mai Vu, Van Khanh Hoang, and Viet Ha Pham Nhu. "Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions." Nuclear Science and Technology 9, no. 4 (September 3, 2021): 16–26. http://dx.doi.org/10.53747/jnst.v9i4.134.

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The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The comparative analysis was conducted for the two models of minor actinide loading in the LEU fuel assembly: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The parameters to be analysed and compared include the reactivity of the LEU fuel assembly versus burnup and the transmutation of minor actinide nuclides when loading different minor actinide vectors into the LEU fuel assembly.
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7

Abderrahim, Hamid Aït. "Realization of a new large research infrastructure in Belgium: MYRRHA contribution for closing the nuclear fuel cycle making nuclear energy sustainable." EPJ Web of Conferences 246 (2020): 00012. http://dx.doi.org/10.1051/epjconf/202024600012.

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In order to provide an appropriate level of energy to the whole world, nuclear energy is still going to play an important role. Nuclear energy can help reducing the CO2 emissions, which today are excessive. The problematics of nuclear waste can be solved using long-term geological storage in deep suitable formations. Partitioning and transmutation can help reducing the radiotoxicity of spent fuel to more acceptable durations of time. The MYRRHA project investigates since more than 20 years the possibility to demonstrate transmutation at a reasonable power level. In this paper we present the current state of the MYRRHA reactor design and the associated research and development activities.
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8

McDeavitt, S. M., A. Parkison, A. R. Totemeier, and J. J. Wegener. "Fabrication of Cermet Nuclear Fuels Designed for the Transmutation of Transuranic Isotopes." Materials Science Forum 561-565 (October 2007): 1733–36. http://dx.doi.org/10.4028/www.scientific.net/msf.561-565.1733.

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The Uranium Extraction (UREX) family of processes uses solvent extraction techniques designed to partition spent uranium and transuranic (TRU) isotopes from fission product waste. Once separated, the collective TRU elements (Np, Pu, Am, and Cm) can be recycled in advanced nuclear energy systems. A zirconium matrix cermet is proposed as a fuel form for this application. Processing methods have been designed to convert the TRU product and spent Zircaloy cladding into feed materials for the hot extrusion of the cermet fuel pins. The TRU conversion process is being developed using a surrogate mixture of uranium and cerium nitrate solutions to generate mixed oxide microspheres. The Zircaloy recovery process is a hydride-dehydride method that is being demonstrated at the bench scale. The powder products from these methods may be combined through hot extrusion into a cermet composite; demonstration experiments using zirconium powder and zirconia microspheres have been completed.
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9

Şahin, Sümer, and Mustafa Übeyli. "LWR spent fuel transmutation in a high power density fusion reactor." Annals of Nuclear Energy 31, no. 8 (May 2004): 871–90. http://dx.doi.org/10.1016/j.anucene.2003.11.003.

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10

Arslan, Alper Buğra, İlayda Yilmaz, Gizem Bakir, and Hüseyin Yapici. "Transmutations of Long-Lived and Medium-Lived Fission Products Extracted from CANDU and PWR Spent Fuels in an Accelerator-Driven System." Science and Technology of Nuclear Installations 2019 (October 20, 2019): 1–13. http://dx.doi.org/10.1155/2019/4930274.

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This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.
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11

François, J. L., J. J. Dorantes, C. Martín-del-Campo, and J. J. E. Herrera. "LWR spent fuel transmutation with fusion-fission hybrid reactors." Progress in Nuclear Energy 65 (May 2013): 50–55. http://dx.doi.org/10.1016/j.pnucene.2013.02.005.

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12

Maltseva, T., А. Shyshuta, and S. Lukashyn. "Modern Methods of Radiochemical Reprocessing of Spent Nuclear Fuel." Nuclear and Radiation Safety, no. 1(81) (March 12, 2019): 52–57. http://dx.doi.org/10.32918/nrs.2019.1(81).09.

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The paper is devoted to the history of development and the current state of technological and scientific advances in radiochemical reprocessing of spent nuclear fuel from water-cooled power reactors. Regarding spent nuclear fuel (SNF) of NPP power reactors, long-term energy security involves adopting a version of its radiochemical treatment, conditioning and recirculation. Recycling SNF is required for the implementation of a closed fuel cycle and the re-use of regeneration products as energy reactor fuels. The basis of modern technological schemes for the reprocessing of the spent nuclear fuel is the “Purex” process, developed since the 60s in the USA. The classic approach to the use of U and Pu nuclides contained in spent nuclear fuel is to separate them from fission products, re-enrich regenerated uranium and use plutonium for the production of mixed-oxide (MOX) fuel with depleted uranium. The modern reprocessing plants are able to deal with fuel with further increase of its main characteristics without significant changes in the initial project. In order to close the fuel cycle, it is needed to add the following technological steps: (1) removal of high-level and long-lived components and minor actinides; (2) return of actinides to the technological cycle; (3) safe disposal of unused components. Each of these areas is under investigation now. Several new promising multi-cycle hydrometallurgical processes based on the joint extraction of trivalent lanthanides and minor actinides with their subsequent separation have been developed. A number of promising materials is suggested to be potential matrices for the immobilization of high-level components of radioactive wastes. To improve the compatibility of fuel processing with the environment, non-aqueous technologies are being developed, for instance, pyro-chemical methods for the reprocessing of various types of highly active fuels based on metals, oxides, carbides, or nitrides. An important scientific and technological task under investigation is transmutation of actinides. The results of international large-scale experiments on the partitioning and transmutation of fuel with various minor actinides and long-lived fission products confirm the real possibility and expediency of closing the nuclear fuel cycle.
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13

Wiss, Thierry, Oliver Dieste, Emanuele De Bona, Alessandro Benedetti, Vincenzo Rondinella, and Rudy Konings. "SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal." Materials 14, no. 21 (October 30, 2021): 6538. http://dx.doi.org/10.3390/ma14216538.

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The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate.
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14

Tran, Vinh Thanh, Hoai-Nam Tran, Huu Tiep Nguyen, Van-Khanh Hoang, and Pham Nhu Viet Ha. "Study on Transmutation of Minor Actinides as Burnable Poison in VVER-1000 Fuel Assembly." Science and Technology of Nuclear Installations 2019 (November 3, 2019): 1–12. http://dx.doi.org/10.1155/2019/5769147.

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Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feasibility study on transmutation of minor actinides as burnable poison in the VVER-1000 LEU (low enriched uranium) fuel assembly. The VVER-1000 LEU fuel assembly was modeled using the SRAC code system, and the SRAC calculation model was verified against the MCNP6 calculations and the available published benchmark data. Two models of minor actinide loading in the LEU fuel assembly have been investigated: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The consequent negative reactivity insertion by minor actinides was compensated by reducing the gadolinium content and boron concentration. The reactivity of the LEU assembly versus burnup and the transmutation of minor actinide nuclides were examined in comparison with the reference case. The results demonstrate that transmutation of minor actinides as burnable poison in the VVER-1000 reactor is feasible as minor actinides could partially replace the functions of gadolinium and boric acid for excess reactivity control.
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15

Stacey, W. M., Z. Abbasi, C. J. Boyd, A. H. Bridges, E. A. Burgett, M. W. Cymbor, S. W. Fowler, et al. "A Subcritical, Helium-Cooled Fast Reactor for the Transmutation of Spent Nuclear Fuel." Nuclear Technology 156, no. 1 (October 2006): 99–123. http://dx.doi.org/10.13182/nt06-a3777.

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16

Egorov, Alexander V., Yurii S. Khomyakov, Valerii I. Rachkov, Elena A. Rodina, and Igor R. Suslov. "Minor actinides transmutation in equilibrium cores of next generation FRs." Nuclear Energy and Technology 5, no. 4 (December 10, 2019): 353–59. http://dx.doi.org/10.3897/nucet.5.46517.

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The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and МА transmutation. This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control. The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. Using МА reprocessed from thermal reactor spent fuel in initial fuel loads in FR ensures a minimal reactivity margin during the entire fast reactor fuel operating period, comparable to the levels achieved in equilibrium state with any kind of relevant Pu isotopic composition. This should be combined with using reactivity compensators in the first fuel micro-campaigns. In the paper presented are the results of simulation of the overall life cycle of a 1200 MWe fast reactor, reaching equilibrium fuel composition, and respective changes in spent fuel nuclide and isotopic composition. It is shown that МА from thermal and fast reactors spent fuel can be completely utilized in the new generation FRs without using special actinide burners.
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17

Bakir, Gizem, Gamze Genc, and Huseyin Yapici. "Study of a conceptual accelerator driven system loaded with thorium dioxide mixed with transuranic dioxides in TRISO particles." Nuclear Technology and Radiation Protection 31, no. 3 (2016): 197–206. http://dx.doi.org/10.2298/ntrp1603197b.

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Nuclear spent fuel management is one of the top major subjects in the utilization of nuclear energy. Hence, solutions to this problem have been increasingly researched in recent years. The basic aim of this work is to examine the fissile breeding and transuranic fuel transmutation potentials of a gas cooled accelerator-driven system. In line with this purpose, firstly, the conceptually designed system is optimized by using several target materials and fuel mixtures, from the point of neutronic. Secondly, three different material compositions, namely, pure lead bismuth eutectic (LBE), LBE+natural UO2, and LBE+15 % enrichment UO2, are considered as target material. The target zone is separated to two sub-zones but as one within the other. The outer sub-zone is pure LBE target, and the inner sub-zone is either UO2 or pure LBE target. The UO2 target sub-zone is cooled with helium gas. Finally, the thorium dioxide mixed with transuranic dioxides, discharged from PWR-MOX spent fuel, in pebbles composed of graphite and TRISO-coated spherical fuel particles, is used for breeding fissile fuel and transmuting transuranic fuels. Three different thorium-transuranic mixtures, (Th, Pu)O2, (Th, Cm)O2, (Th, Pu, MA)O2, are examined with various mixture fractions. The packing fractions of the fuel pebbles in the transmutation core and the tristructural-isotropic coated fuel particles in a pebble are assumed as 60 % and 29 %, respectively. The transmutation core is also cooled with a high-temperature helium coolant. In order to produce high-flux neutrons that penetrate through the transmutation core, the target is exposed to the continuous beams of 1 GeV protons. The computations have been carried out with the high-energy Monte Carlo code MCNPX using the LA150 library. The numerical outcomes show that the examined accelerator-driven system has rather high neutronic data in terms of the energy production and fissile fuel breeding.
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18

Acır, Adem. "Numerical and Statistical Analysis of FR Spent Fuel Transmutation in a Thorium Fusion Breeder." Journal of Fusion Energy 28, no. 3 (October 16, 2008): 258–67. http://dx.doi.org/10.1007/s10894-008-9163-0.

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19

Narbutt, Jerzy. "New trends in the reprocessing of spent nuclear fuel. Separation of minor actinides by solvent extraction." Annales Universitatis Mariae Curie-Sklodowska, sectio AA – Chemia 71, no. 1 (May 24, 2016): 123. http://dx.doi.org/10.17951/aa.2016.71.1.123.

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<p>Recycling of actinides from spent nuclear fuel by their selective separation followed by transmutation in fast reactors will optimize the use of natural uranium resources and minimize the long-term hazard from high-level nuclear waste. This paper describes solvent extraction processes recently developed, aimed at the separation of americium from lanthanide fission products as well as from curium present in the waste. Depicted are novel poly-N-heterocyclic ligands used as selective extractants of actinide ions from nitric acid solutions or as actinide-selective hydrophilic stripping agents.</p>
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20

Plukienė, R., A. Plukis, L. Juodis, V. Remeikis, O. Šalkauskas, D. Ridikas, and W. Gudowski. "Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system." Nuclear Engineering and Design 330 (April 2018): 241–49. http://dx.doi.org/10.1016/j.nucengdes.2018.01.046.

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21

Bourg, Stéphane, Andreas Geist, and Jerzy Narbutt. "SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels." Nukleonika 60, no. 4 (December 1, 2015): 809–14. http://dx.doi.org/10.1515/nuka-2015-0152.

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Abstract Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes). Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.
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22

Şahi̇n, Sümer, Hacı Mehmet Şahin, and Güven Tunç. "Monte Carlo analysis of LWR spent fuel transmutation in a fusion-fission hybrid reactor system." Nuclear Engineering and Technology 50, no. 8 (December 2018): 1339–48. http://dx.doi.org/10.1016/j.net.2018.08.006.

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23

Hoggett-Jones, C., C. Robbins, G. Gettinby, and S. Blythe. "Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management." Annals of Nuclear Energy 29, no. 5 (March 2002): 491–508. http://dx.doi.org/10.1016/s0306-4549(01)00065-2.

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24

Stacey, W. M. "Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor." Nuclear Fusion 41, no. 2 (February 2001): 135–54. http://dx.doi.org/10.1088/0029-5515/41/2/301.

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25

Stacey, W. M. "Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor." Nuclear Fusion 41, no. 4 (April 2001): 467. http://dx.doi.org/10.1088/0029-5515/41/4/512.

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26

Bourg, Stéphane, Andreas Geist, Jean-Marc Adnet, Chris Rhodes, and Bruce C. Hanson. "Partitioning and transmutation strategy R&D for nuclear spent fuel: the SACSESS and GENIORS projects." EPJ Nuclear Sciences & Technologies 6 (2020): 35. http://dx.doi.org/10.1051/epjn/2019009.

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Processes such as PUREX allow the recovery and reuse of the uranium and the plutonium of GEN II/GEN III reactors and are being adapted for the recycling of the uranium and the plutonium of GEN IV MOX fuels. However, it does not fix the sensitive issue of the long-term management of the high active nuclear waste (HAW). Indeed, only the recovery and the transmutation of the minor actinides can reduce this burden down to a few hundreds of years. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS focuses on the reprocessing of MOX fuel containing minor actinides, taking into account safety issues under normal and mal-operation. By implementing a three-step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will provide more science-based strategies for nuclear fuel management in the EU.
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27

Nicolaou, G., and N. Tsagas. "Criticality safety of spent nuclear fuel assemblies from the transmutation of minor actinides in fast reactors." Annals of Nuclear Energy 33, no. 4 (March 2006): 305–9. http://dx.doi.org/10.1016/j.anucene.2005.11.006.

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28

Shlenskii, Mikhail, and Boris Kuteev. "System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle." Applied Sciences 10, no. 23 (November 26, 2020): 8417. http://dx.doi.org/10.3390/app10238417.

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This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a fleet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130.
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29

Aït Abderrahim, Hamid, Peter Baeten, Alain Sneyers, Marc Schyns, Paul Schuurmans, Anatoly Kochetkov, Gert Van den Eynde, and Jean-Luc Biarrotte. "Partitioning and transmutation contribution of MYRRHA to an EU strategy for HLW management and main achievements of MYRRHA related FP7 and H2020 projects: MYRTE, MARISA, MAXSIMA, SEARCH, MAX, FREYA, ARCAS." EPJ Nuclear Sciences & Technologies 6 (2020): 33. http://dx.doi.org/10.1051/epjn/2019038.

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Today, nuclear power produces 11% of the world's electricity. Nuclear power plants produce virtually no greenhouse gases or air pollutants during their operation. Emissions over their entire life cycle are very low. Nuclear energy's potential is essential to achieving a deeply decarbonized energy future in many regions of the world as of today and for decades to come, the main value of nuclear energy lies in its potential contribution to decarbonizing the power sector. Nuclear energy's future role, however, is highly uncertain for several reasons: chiefly, escalating costs and, the persistence of historical challenges such as spent fuel and radioactive waste management. Advanced nuclear fuel recycling technologies can enable full use of natural energy resources while minimizing proliferation concerns as well as the volume and longevity of nuclear waste. Partitioning and Transmutation (P&T) has been pointed out in numerous studies as the strategy that can relax constraints on geological disposal, e.g. by reducing the waste radiotoxicity and the footprint of the underground facility. Therefore, a special effort has been made to investigate the potential role of P&T and the related options for waste management all along the fuel cycle. Transmutation based on critical or sub-critical fast spectrum transmuters should be evaluated in order to assess its technical and economic feasibility and capacity, which could ease deep geological disposal implementation.
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Yano, Toyohiko, Junichi Yamane, and Katsumi Yoshida. "Low Temperature Sintering of Si3N4 Ceramics and its Applicability as an Inert Matrix of the Transuranium Elements for Transmutation of Minor Actinides." Key Engineering Materials 403 (December 2008): 23–26. http://dx.doi.org/10.4028/www.scientific.net/kem.403.23.

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For the transmutation of the very long half-lived isotopes which are separated from the spent nuclear fuels, it is necessary to find proper inert matrices these are stable under heavy neutron irradiation at high temperature. Silicon nitride ceramics is a candidate since it is very tolerant for heavy neutron irradiation and keeps relatively high thermal conductivity. For these reasons, we try to sinter Si3N4 ceramics containing large amounts of CeO2 as a simulant for Am2O3, a typical transuranium element. The low-temperature pressureless-sintering behavior of the ceramics and chemical and thermal properties of the obtained sintered bodies are reported.
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31

Eleme, Z., N. Patronis, A. Stamatopoulos, A. Tsinganis, M. Kokkoris, V. Michalopoulou, M. Diakaki, et al. "First results of the 241Am(n,f) cross section measurement at the Experimental Area 2 of the n_TOF facility at CERN." EPJ Web of Conferences 239 (2020): 05014. http://dx.doi.org/10.1051/epjconf/202023905014.

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Feasibility, design and sensitivity studies on innovative nuclear reactors that could address the issue of nuclear waste transmutation using fuels enriched in minor actinides, require high accuracy cross section data for a variety of neutron-induced reactions from thermal energies to several tens of MeV. The isotope 241Am (T1/2= 433 years) is present in high-level nuclear waste (HLW), representing about 1.8 % of the actinide mass in spent PWR UOx fuel. Its importance increases with cooling time due to additional production from the β-decay of 241Pu with a half-life of 14.3 years. The production rate of 241 Am in conventional reactors, including its further accumulation through the decay of 241Pu and its destruction through transmutation/incineration are very important parameters for the design of any recycling solution. In the present work, the 241 Am(n,f) reaction cross-section was measured using Micromegas detectors at the Experimental Area 2 of the n_TOF facility at CERN. For the measurement, the 235U(n,f) and 238U(n,f) reference reactions were used for the determination of the neutron flux. In the present work an overview of the experimental setup and the adopted data analysis techniques is given along with preliminary results.
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32

Yapıcı, Hüseyin, Gamze Genç, and Nesrin Demir. "Transmutation–incineration potential of transuraniums discharged from PWR-UO2 spent fuel in modified PROMETHEUS fusion reactor." Fusion Engineering and Design 81, no. 18 (August 2006): 2093–108. http://dx.doi.org/10.1016/j.fusengdes.2005.12.003.

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33

Di Sanzo, Christian, Mohamed Abdou, and Mahmoud Youssef. "Transuranic transmutation efficiency of a small fusion–fission facility for spent uranium-oxide and inert matrix fuels." Fusion Engineering and Design 85, no. 7-9 (December 2010): 1488–91. http://dx.doi.org/10.1016/j.fusengdes.2010.04.023.

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34

Bakır, Gizem, Gamze Genç, and Hüseyin Yapıcı. "Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles." Science and Technology of Nuclear Installations 2016 (2016): 1–11. http://dx.doi.org/10.1155/2016/2612459.

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This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.
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35

Setiawan, M. Budi, P. Made Udiyani, S. Kuntjoro, I. Husnayani, and T. Surbakti. "Analysis on Transmutation of Long-Lived Fission Products from PWR Spent Fuel Using the 30-MW(thermal) RSG-GAS Reactor." Nuclear Technology 206, no. 12 (March 10, 2020): 1945–50. http://dx.doi.org/10.1080/00295450.2020.1720558.

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36

Yapıcı, Hüseyin. "Study on transmutation of minor actinides discharged from high burn-up PWR-MOX spent fuel in the force-free helical reactor." Annals of Nuclear Energy 30, no. 4 (March 2003): 413–36. http://dx.doi.org/10.1016/s0306-4549(02)00078-6.

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37

Ruskov, Ivan, Andrei Goverdovski, Walter Furman, Yury Kopatch, Oleg Shcherbakov, Franz-Josef Hambsch, Stephan Oberstedt, and Andreas Oberstedt. "Neutron induced fission of 237Np – status, challenges and opportunities." EPJ Web of Conferences 169 (2018): 00021. http://dx.doi.org/10.1051/epjconf/201816900021.

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Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated) in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel (“waste”), the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ), has not been updated for decades.
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38

Trellue, Holly R. "Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors." Nuclear Technology 147, no. 2 (August 2004): 171–80. http://dx.doi.org/10.13182/nt04-a3523.

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39

Bakır, Gizem, Saltuk Buğra Selçuklu, and Hüseyin Yapıcı. "Medical Radioisotope Production in a Power-Flattened ADS Fuelled with Uranium and Plutonium Dioxides." Science and Technology of Nuclear Installations 2016 (2016): 1–11. http://dx.doi.org/10.1155/2016/5302176.

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This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2in the mixture of UO2and PuO2in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power (Pth) for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28.
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40

Milian Lorenzo, Daniel Evelio, Daniel Milian Pérez, Lorena Pilar Rodríguez García, Jesús Salomón Llanes, Carlos Alberto Brayner de Oliveira Lira, Manuel Cadavid Rodríguez, and Carlos Rafael García Hernández. "Study of Thorium Fuel Cycles for Light Water Reactor VBER-150." International Journal of Nuclear Energy 2013 (December 23, 2013): 1–9. http://dx.doi.org/10.1155/2013/491898.

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The main objective of this paper is to examine the use of thorium-based fuel cycle for the transportable reactors or transportable nuclear power plants (TNPP) VBER-150 concept, in particular the neutronic behavior. The thorium-based fuel cycles included Th232+Pu239, Th232+U233, and Th232+U and the standard design fuel UOX. Parameters related to the neutronic behavior such as burnup, nuclear fuel breeding, MA stockpile, and Pu isotopes production (among others) were used to compare the fuel cycles. The Pu transmutation rate and accumulation of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The Th232+U233 fuel cycle proved to be the best cycle for minimizing the production of Pu and MA. The neutronic calculations have been performed with the well-known MCNPX computational code, which was verified for this type of fuel performing calculation of the IAEA benchmark announced by IAEA-TECDOC-1349.
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41

Barros, Graiciany de Paula, Claubia Pereira, Maria A. F. Veloso, and Antonella L. Costa. "Study of an ADS Loaded with Thorium and Reprocessed Fuel." Science and Technology of Nuclear Installations 2012 (2012): 1–12. http://dx.doi.org/10.1155/2012/934105.

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Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods wasT232hO2forU233production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use ofT232hO2without the initial requirement ofU233enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor (keff) evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use ofT232hO2and reprocessed fuel allowedU233production without the initial requirement ofU233enrichment.
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42

Stacey, W. M. "Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel." Journal of Fusion Energy 28, no. 3 (March 19, 2009): 328–33. http://dx.doi.org/10.1007/s10894-009-9195-0.

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43

Cetnar, Jerzy, Grażyna Domańska, Paweł Gajda, and Jerzy Janczyszyn. "Assessment of the control rods shadow effect in the VENUS-F core." Nukleonika 59, no. 4 (December 1, 2014): 137–43. http://dx.doi.org/10.2478/nuka-2014-0020.

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Abstract The partitioning and transmutation (P&T) of spent nuclear fuel is an important field of present development of nuclear energy technologies. One of the possible ways to carry out the P&T process is to use the accelerator driven systems (ADS). This technology has been developed within the EURATOM Framework Programmes for several years now. Current research in this field is carried out within the scope of 7th FP project FREYA. Important parts of the project are experiments performed in the GUINEVERE facility devoted to characterising the subcritical core kinetics and development of reactivity monitoring techniques. The present paper considers the effects of control rods use on the core reactivity. In order to carry out the evaluation of the experimental results, it is important to have detailed core characteristics at hand and to take into consideration the differences in the effect of control rods acting separately or together (the so-called shadow effect) on both the reactivity value and the measured neutron flux. Also any core asymmetry should be revealed. This goal was achieved by both MCNP simulations and the experimental results. However, in the case of experimental results, the need for calculating respective correction factors was unavoidable.
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44

Yapıcı, Hüseyin. "Determination of the optimal plutonium fraction in transuranium discharged from pressured water reactor (PWR) spent fuel for a flat fission power generation in the force-free helical reactor (FFHR) along the transmutation period." Annals of Nuclear Energy 30, no. 6 (April 2003): 633–49. http://dx.doi.org/10.1016/s0306-4549(02)00114-7.

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45

Ivanov, V. K., E. O. Adamov, E. V. Spirin, V. M. Solomatin, S. Yu Chekin, and A. N. Menyajlo. "Evaluation of optimal amount of americium that should be extracted from spent nuclear fuel of the BREST-OD-300 reactor for transmutation to ensure radiological equivalence of radioactive waste and natural uranium." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 29, no. 1 (2020): 5–17. http://dx.doi.org/10.21870/0131-3878-2020-29-1-5-17.

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46

Chen, Shengli, and Cenxi Yuan. "Transmutation Study of Minor Actinides in Mixed Oxide Fueled Typical Pressurized Water Reactor Assembly." Journal of Nuclear Engineering and Radiation Science 4, no. 4 (September 10, 2018). http://dx.doi.org/10.1115/1.4040423.

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The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. The partitioning-transmutation method is supposed to efficiently treat the long-lived radionuclides. Accordingly, the transmutation of long-lived minor actinides (MAs) is significant for the postprocessing of spent fuel. In the present work, the transmutations in pressurized water reactor (PWR) mixed oxide (MOX) fuel are investigated through the Monte Carlo neutron transport method. Two types of MAs are homogeneously incorporated into MOX fuel assembly with different mixing ratios. In addition, two types of design of semihomogeneous loading of 237Np in MOX fuels are studied. The results indicate an overall nice efficiency of transmutation in PWR with MOX fuel, especially for 237Np and 241Am, which are primarily generated in the current uranium oxide fuel. In addition, the transmutation efficiency of 237Np is excellent, while its inclusion has no much influence on other MAs. The flattening of power and burnup are achieved by semihomogeneous loading of MAs. The uncertainties of Monte Carlo method are negligible, while those due to nuclear data change little the conclusions of the transmutation of MAs. The transmutation of MAs in MOX fuel is expected to be an efficient method for spent fuel management.
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47

Bresee, James. "Transmutation and the Global Nuclear Energy Partnership." MRS Proceedings 985 (2006). http://dx.doi.org/10.1557/proc-985-0985-nn14-02.

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AbstractIn the January 2006 State of the Union address, President Bush announced a new Advanced Energy Initiative, a significant part of which is the Global Nuclear Energy Initiative. Its details were described on February 6, 2006 by the U.S. Secretary of Energy. In summary, it has three parts: (1) a program to expand nuclear energy use domestically and in foreign countries to support economic growth while reducing the release of greenhouse gases such as carbon dioxide. (2) an expansion of the U.S. nuclear infrastructure that will lead to the recycling of spent fuel and a closed fuel cycle and, through transmutation, a reduction in the quantity and radiotoxicity of nuclear waste and its proliferation concerns, and (3) a partnership with other fuel cycle nations to support nuclear power in additional nations by providing small nuclear power plants and leased fuel with the provision that the resulting spent fuel would be returned by the lessee to the lessor. The final part would have the effect of stabilizing the number of fuel cycle countries with attendant non-proliferation value. Details will be given later in the paper.
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48

Zuhair, R. Andika Putra Dwijayanto, Sriyono, Suwoto, and Zaki Su’ud. "Preliminary study on TRU transmutation in VVER-1000 fuel assembly using MCNP6." Kerntechnik, February 25, 2022. http://dx.doi.org/10.1515/kern-2021-1017.

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Abstract Transmutation technology has been developed to address the issue of long-term safety of nuclear waste disposal in geological repository. Transuranic (TRU) transmutation in thermal neutron spectrum can be considered more beneficial at the present time due to the larger fission cross section of TRU elements than in fast neutron spectrum. This paper discusses preliminary study on TRU transmutation in Vodo-Vodyanoi Energetichesky Reactor (VVER)-1000 fuel assembly using MCNP6 code. The fuel assembly is configured by 312 fuel cells consist of 300 UO2 fuel rods with 3.7 wt% 235U and 12 UGD fuel rods containing a mixture of UO2 with 3.6 wt% 235U and 4.0 wt% Gd2O3. The calculation results show that the k inf value is higher with the increase in the TRU thickness at the beginning of cycle and at the end of cycle, which means that the addition of TRU will increase the fuel cycle length. The total temperature coefficient of reactivity is negative for all fuel assemblies both without TRU and with TRU. In general, the computed β eff values of the assembly with and without TRU addition are not significantly different. It shows that the coating of the TRU layer in the UGD fuel cell will not complicate the reactor control. The transmutation of TRU recycled from spent nuclear fuel in the VVER-1000 fuel assembly can be considered feasible from the viewpoint of excess reactivity and control safety characteristics. The total transmutation rate of ∼55.73% can be achieved with 239Pu and 241Am isotopes are taking the largest portion of transmuted nuclides.
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49

Andersson, Sofie, C. Ekberg, J.-O. Liljenzin, M. Nilsson, and G. Skarnemark. "Study of nitrate complex formation with trivalent Pm, Eu, Am and Cm using a solvent extraction technique." Radiochimica Acta 92, no. 12 (January 1, 2004). http://dx.doi.org/10.1524/ract.92.12.863.55109.

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SummaryThe separation of actinides and lanthanides is an important question in the treatment of spent nuclear fuel in the transmutation concept. To find an efficient and well functioning separation process it is necessary to study the chemistry of the elements in the two groups in different aqueous media. The stability constants of the nitrate complex formation with Pm, Eu, Am and Cm were determined using solvent extraction. The extraction was studied using the synergistic system of 2,6-bis-(benzoxazolyl)-4-dodecyloxylpyridine and 2-bromodecanoic acid in
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