Готові списки джерел за темами / Transmutation of spent nuclear fue

Добірка наукової літератури з теми "Transmutation of spent nuclear fue"

Автор: Grafiati

Опубліковано: 30 травня 2022

Оновлено: 31 травня 2022

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Статті в журналах з теми "Transmutation of spent nuclear fue"

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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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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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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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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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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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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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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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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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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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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Більше джерел

Дисертації з теми "Transmutation of spent nuclear fue"

Sommer, Christopher Michael. "Subcritical transmutation of spent nuclear fuel." Diss., Georgia Institute of Technology, 2011. http://hdl.handle.net/1853/41205.

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A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

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Yee, Shannon K. "Nuclear Fuel Cycle Modeling Approaches For Recycling And Transmutation Of Spent Nuclear Fuel." The Ohio State University, 2008. http://rave.ohiolink.edu/etdc/view?acc_num=osu1213905425.

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Hoggett-Jones, Craig. "Modelling and assessment of partitioning and transmutation approaches to spent nuclear fuel management." Thesis, University of Strathclyde, 2001. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.248302.

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Szakaly, Frank Joseph. "Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems." Thesis, Texas A&M University, 2003. http://hdl.handle.net/1969.1/31.

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The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca Mountain Repository. A two tier recycling strategy is proposed. Thermal spectrum transmutation systems converted from the existing LWR fleet were modeled for the first tier, and the Japanese fast reactor MONJU was used for the fast-spectrum transmutation. The modeling was performed with the Monteburns code. Transmutation performance was investigated as well as delayed neutron fraction, heat generation rates, and radioactivity of the spent material in the short and long term for the different transmutation fuel cycles. A two-tier recycling strategy incorporating fast and thermal transmutation with uranium-free nitride fuel was shown to reduce the long-term heat generation rates and radioactivity of the spent nuclear fuel inventory.

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Онисимчук, Тетяна Михайлівна. "Трансмутація радіоактивних відходів з удосконаленням системи сповільнення швидких нейтронів". Master's thesis, Київ, 2018. https://ela.kpi.ua/handle/123456789/25799.

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Дисертацію присвячено впровадженню системи сповільнення швидких нейтронів в підкритичних ядерних установках, керованих зовнішнім джерелом, виконаної із вторинно переробленого поліетилентерефталату. У дисертації визначено показники безпеки функціонування підкритичних систем з пластиковим сповільнювачем, доведено можливість застосування вторинного полімеру в якості сповільнювача швидких нейронів та встановлено залежність коефіцієнту пропускання іонізуючого випромінювання від циклу переробки матеріалу захисного шару. Визначено, що економічна ефективність використання вторинної сировини з врахуванням збільшення кількості полімеру в 1,3 рази в порівнянні з еталонним поліетиленом, окреслюється економією капіталовкладень у розмірі 1 197 683 грн на момент проведення розрахунку. Отримано залежністькоефіцієнту пропускання іонізуючого випромінювання від товщини зовнішнього шару сповільнювача після циклічних етапів переробки, яка описується поліномом 3-го порядку. Виявлено, що допустима кількість циклів переробки для поліетилентерефталату становить три, при якій товщина сповільнювача становитиме 750 мм. При подальших циклах переробки використання вторинної сировини економічно недоцільне. Розроблений стартап-проект реалізації технології на вітчизняному ринку прогнозує отримання разового прибутку у розмірі близько 800 тис. грн протягом 1 року від дати отримання сертифікату відповідності.
The dissertation is devoted to the introduction of a fast neutron slowdown system in subcritical nuclear installations controlled by an external source made from recycled polyethylene terephthalate. The dissertation defines the performance indicators of subcritical systems with plastic moderator, proved the possibility of using the secondary polymer as a moderator of fast neurons, and the dependence of the ionizing radiation transmittance coefficient on the cycle of processing the material of the protective layer has been established. It is determined that the economic efficiency of the use of secondary raw materials, taking into account the increase in the amount of polymer in 1,3 times compared with the reference polyethylene, is defined by the savings of investments in the amount of 1 197 683 UAH at the time of calculation. The dependence of transmittance coefficient of ionizing radiation on the thickness of the outer layer of the retarder after the cyclic stages of processing, which is described by the polynomial of the 3rd order, is obtained.It was found that the permissible number of recycling cycles for polyethylene terephthalate is three, in which the thickness of the moderator will be 750 mm. In subsequent cycles of recycling, the use of secondary raw materials will be economically impractical. The developed start-up project of technology implementation on the domestic market predicts a one-time profit of about 800 thousand UAH for 1 year from the date of receipt of the certificate of conformity.
Диссертация посвящена внедрению системы замедления быстрых нейтронов в подкритических ядерных установках, управляемых внешним источником, выполненной из вторично переработанного полиэтилентерефталата. В диссертации определены показатели безопасности функционирования подкритических систем с пластиковым замедлителем, доказана возможность применения вторичного полимера в качестве замедлителя быстрых нейронов и установлена зависимость коэффициента пропускания ионизирующего излучения от цикла переработки материала защитного слоя. Определено, что экономическая эффективность использования вторичного сырья с учетом увеличения количества полимера в 1,3 раза в сравнении с эталонным полиэтиленом, определяется экономией капиталовложений в размере 1 197 683 грн на момент проведения расчета. Получена зависимость коэффициента пропускания ионизирующего излучения от толщины внешнего слоя замедлителя после циклических этапов переработки, которая описывается полиномом 3-го порядка. Выявлено, что допустимое количество циклов переработки для полиэтилентерефталата составляет три, при котором толщина замедлителя составит 750 мм. При последующих циклах переработки использование вторичного сырья экономически нецелесообразно. Разработанстартап-проект реализации технологии на отечественном рынке прогнозирует получение разового дохода в размере около 800 тыс. грн в течение 1 года от даты получения сертификата соответствия.

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Jarchovský, Petr. "Výpočetní simulace urychlovačem řízeného jaderného reaktoru pro transmutaci vyhořelého jaderného paliva." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2015. http://www.nusl.cz/ntk/nusl-221209.

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This master’s thesis deals with usage of burn-up (spent) nuclear fuel in nuclear power plants of next generation – accelerator driven transmutation plants which is produced in current nuclear power plants. This system could significantly reduce the volume of dangerous long-lived radioisotopes and moreover they would be able to take advantage of its great energy potential due to fast neutron spectrum. In the introduction are listed basic knowledge and aspects of spent nuclear fuel along with its reprocessing and the possibility of further use while minimizing environmental impact. As another point detailed description of accelerator driven systems is described together with its basic components. In addition this search is followed by individual chronological enumeration of projects of global significance concerning their current development. Emphasis is placed on SAD and MYRRHA projects, which are used like base for calculations. This last, computational part, deals with the creation of the geometry of subcritical transmutation reactor driven by accelerator and subsequent evaluation which assembly is the most effective for transmutation and energy purposes along with changing of target, nuclear fuel and coolant/moderator.

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Alajo, Ayodeji Babatunde. "Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios." Thesis, 2010. http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7809.

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The closure of the nuclear fuel cycle is a topic of interest in the sustainability context of nuclear energy. The implication of such closure includes considerations of nuclear waste management. This originates from the fact that a closed fuel cycle requires recycling of useful materials from spent nuclear fuel and discarding of non-usable streams of the spent fuel, which are predominantly the fission products. The fission products represent the near-term concerns associated with final geological repositories for the waste stream. Long-lived fission products also contribute to the long-term concerns associated with such repository. In addition, an ultimately closed nuclear fuel cycle in which all actinides from spent nuclear fuels are incinerated will result in fission products being the only source of radiotoxicity. Hence, it is desired to develop a transmutation strategy that will achieve reduction in the inventory and radiological parameters of significant fission products within a reasonably short time. In this dissertation, a transmutation strategy involving the use of the VHTR is developed. A set of specialized metrics is developed and applied to evaluate performance characteristics. The transmutation strategy considers six major fission products: 90Sr, 93Zr, 99Tc, 129I, 135Cs and 137Cs. In this approach, the unique core features of VHTRs operating in equilibrium fuel cycle mode of 405 effective full power days are used for transmutation of the selected fission products. A 30 year irradiation period with 10 post-irradiation cooling is assumed. The strategy assumes no separation of each nuclide from its corresponding material stream in the VHTR fuel cycle. The optimum locations in the VHTR core cavity leading to maximized transmutation of each selected nuclides are determined. The fission product transmutation scenarios are simulated with MCNP and ORIGEN-S. The results indicate that the developed fission product transmutation strategy offers an excellent potential approach for the reduction of inventories and radiological parameters, particularly for long-lived fission products (93Zr, 99Tc, 129I and 135Cs). It has been determined that the in-core transmutation of relatively short-lived fission products (90Sr and 137Cs) has minimal advantage over a decay-only scenario for these nuclides. It is concluded that the developed strategy is a viable option for the reduction of radiotoxicity contributions of the selected fission products prior to their final disposal in a geological repository. Even in the cases where the transmutation advantage is minimal, it is deemed that the improvement gained, coupled with the virtual storage provided for the fission products during the irradiation period, makes the developed fission product transmutation strategy advantageous in the spent fuel management scenarios. Combined with the in-core incineration options for TRU, the developed transmutation strategy leads to potential achievability of engineering time scales in the comprehensive nuclear waste management.

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Частини книг з теми "Transmutation of spent nuclear fue"

Yim, Man-Sung. "Spent Fuel Reprocessing and Nuclear Waste Transmutation." In Lecture Notes in Energy, 341–84. Dordrecht: Springer Netherlands, 2021. http://dx.doi.org/10.1007/978-94-024-2106-4_8.

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Verma, Vinod Kumar, and Karel Katovsky. "Spent Nuclear Fuel and Alternative Methods of Transmutation." In Spent Nuclear Fuel and Accelerator-Driven Subcritical Systems, 1–19. Singapore: Springer Singapore, 2018. http://dx.doi.org/10.1007/978-981-10-7503-2_1.

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Verma, Vinod Kumar, and Karel Katovsky. "Transmutation of Spent Nuclear Fuel and Extension of a Fuel Cycle." In Spent Nuclear Fuel and Accelerator-Driven Subcritical Systems, 67–80. Singapore: Springer Singapore, 2018. http://dx.doi.org/10.1007/978-981-10-7503-2_5.

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Suyama, Kenya, Gunzo Uchiyama, Hiroyuki Fukaya, Miki Umeda, Toru Yamamoto, and Motomu Suzuki. "Development of the Method to Assay Barely Measurable Elements in Spent Nuclear Fuel and Application to BWR 9 × 9 Fuel." In Nuclear Back-end and Transmutation Technology for Waste Disposal, 47–56. Tokyo: Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55111-9_6.

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Abbasi, Akbar. "Nuclear Fuel Transmutation." In Nuclear Power Plants - The Processes from the Cradle to the Grave. IntechOpen, 2021. http://dx.doi.org/10.5772/intechopen.94065.

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Nuclear power plants to generates electric energy used nuclear fuel such as Uranium Oxide (UOX). A typical VVER−1000 reactor uses about 20–25 tons of spent fuel per year. The fuel transmutation of UOX fuel was evaluated by VISTA computer code. In this estimation the front end and back end components of fuel cycle was calculated. The front end of the cycle parameter are FF requirements, enrichment value requirements, depleted uranium amount, conversion requirements and natural uranium requirements. The back-end component is Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes.

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Salvatores, M. "Partitioning and transmutation of spent nuclear fuel and radioactive waste." In Nuclear Fuel Cycle Science and Engineering, 501–30. Elsevier, 2012. http://dx.doi.org/10.1533/9780857096388.4.501.

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Hill, C. "Development of highly selective compounds for solvent extraction processes: partitioning and transmutation of long-lived radionuclides from spent nuclear fuels." In Advanced Separation Techniques for Nuclear Fuel Reprocessing and Radioactive Waste Treatment, 311–62. Elsevier, 2011. http://dx.doi.org/10.1533/9780857092274.3.311.

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Тези доповідей конференцій з теми "Transmutation of spent nuclear fue"

Chen, Shengli, Cenxi Yuan, Jingxia Wu, and Yaolei Zou. "Study of Minor Actinides Transmutation in PWR MOX Fuel." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66250.

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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. Partitioning-Transmutation is supposed to treat efficiently the long-lived radionuclides. Accordingly, the study of transmutation for long-lived Minor Actinides (MAs) is a significant work for the post-processing of spent fuel. In the present work, the transmutations in Pressurized Water Reactor (PWR) Mixed OXide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs are incorporated homogeneously into two initial concentrations MOX fuel assembly. The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for two MAs primarily generated in the UOX fuel, 237Np and 241Am. In addition, the inclusion of 237Np has no large influence on other MAs, while the transmutation efficiency of 237Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be an efficient nuclear spent fuel management method.

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Bergelson, B. R., A. S. Gerasimov, and G. V. Tikhomirov. "Application of Power Reactors for Transmutation of Actinides." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49049.

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Heavy-water and light-water power reactors can be used for partial transmutation of radwaste. Such transmutation allows us to limit on relatively low level the radiotoxicity accumulated in long-term storage of spent fuel. At transmutation in PHWR-880 reactor operating in the mode of self-service, equilibrium radiotoxicity in storage facility and the time of its achievement are 4–5 times less than for the transmutation in power reactor VVER-1000.

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Zhang, Wenxin, Haoyang Yu, Bin Liu, Jin Cai, and Shuangshuang Cui. "The Effect of Minor Actinide Transmutation on Temperature Coefficient in PWR." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67649.

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Minor actinides in the spent fuel have strong radiotoxicity and very long half-life, the the properly dispose of spent fuel is indispensible to the development of nucler energy. Generally,we dispose the spent fuel by geological burying. But it can not compeletly solve the problem. Neutron transmutation is the only way to shorten the half-life of radioactive nuclides, under the irradiation of neutron MA nuclide will capture neutron or fission, and translate into the short lived nuclide or something valued nuclide. Reactivity temperature coefficient is an improtant safety parameter in nuclear reactor physics.In the reactor design, for the safely operation of reactor, reactivity temperature coefficient must be be negative. The introduction of MA in the PWR must have interference to the temperature coefficient. This paper mainly studied the influence of PWR transmutation minor actinide on the temperature coefficient.

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Gerasimov, Aleksander S., Gennady V. Kiselev, Lidia A. Myrtsymova, and Tamara S. Zaritskaya. "Cyclic Mode of Transmutation of Minor Actinides in Heavy-Water Reactor." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22668.

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Characteristics of process of transmutation of americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, 244Cm and 238Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of a transmutation, the radiotoxicity is increased by 11 times in comparison with initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute americium and curium extracted from 7–8 VVER-1000 type reactors. It means that the required power of transmutation reactor makes about 4% of thermal power of VVER-1000 type reactors.

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Balas (Ghizdeanu), Nineta, and Petre Ghitescu. "Transmutation Efficiency of Plutonium and Minor Actinides in PHWR." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48570.

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PHWRs use natural uranium as fuel and consequently the burn-up coefficient is relatively small compared to PWRs or other existing power reactors. The small burn-up coefficient results in a high volume of irradiated fuel to be disposed, with a high concentration of plutonium and minor actinides. In Romania the irradiated fuel from the existing CANDU 6 spent fuel pool is currently transferred in the Dry Intermediate Fuel Storage Facility existing at the NPP site. Partitioning and Transmutation (P&T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. The use for this purpose of ADS and FBR was more studied, but HWR were not. Therefore, the paper presents different theoretical possibilities to transmute/burn the Plutonium and minor actinides in two different PHWRs — CANDU and ACR, using WIMSD code. Different types of MOX alternative fuel, with variable initial Pu content are analyzed. The results present the reactivity effects along with the isotopes concentration in spent alternative fuel and determine the optimal solution for the fuel type/composition. Thus is indicated the most suitable PHWR type of reactor for possible Plutonium and minor actinides transmutation. The simulations showed that Pu content for an irradiation period of 200 days decreases from the initial value up to 11% in a CANDU reactor and 29% in an ACR. Thus ACR can reduce the plutonium inventory from MOX fuel and could be a transmutation solution. From the economic/technical point of view this analysis also provides input for a study yet to be conducted.

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Andrello, Concettina, Daniel Freis, Rosa Lo Frano, Dimitri Papaioannou, and Fabienne Delage. "Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67252.

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The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation. MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option. In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds. Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination. After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.

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Hyland, Bronwyn, and Brian Gihm. "Scenarios for the Transmutation of Actinides in CANDU Reactors." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30123.

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With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU® reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past [1–4]. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100 to 1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

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Yu, Haoyang, Bin Liu, Wenxin Zhang, and Jin Cai. "The Effect of MA Transmutation in the PWR on Fuel Cycle." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67787.

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The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

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Sun, Lijian, Haritha Royyuru, Hsuan-Tsung Hsieh, Yitung Chen, George Vandegrift, Jackie Copple, and James Laidler. "Development of Systems Engineering Model for Spent Fuel Extraction Process." In ASME 2004 International Mechanical Engineering Congress and Exposition. ASMEDC, 2004. http://dx.doi.org/10.1115/imece2004-60178.

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The mission of the Transmutation Research Program (TRP) at University of Nevada, Las Vegas (UNLV) is to establish a nuclear engineering test bed that can carry out effective transmutation and advanced reactor research and development effort. Chemical Engineering Division, Argonne National Laboratories (ANL) is in charge the design, modeling, and demonstration of countercurrent solvent-extraction process for treating high-level liquid waste, such as U and Tc. The Nevada Center for Advanced Computational Methods (NCACM) at UNLV is developing a systems engineering model that provides process optimization through the automatic adjustment on input parameters, such as feed compositions, stages, flow rates, etc., based on the extraction efficiency of components and concerned output factors. An object-oriented programming (OOP) is considered. Previously designed Microsoft (MS) Excel macro-based program, Argonne Model for Universal Solvent Extraction (AMUSE) code, based on firm understanding of the chemistry and thermodynamics, is the core module for Uranium Extraction process (UREX). Currently AMUSE is the only available module. The Transmutation Research Program System Engineering Model Project (TRPSEMPro) consists of task manager, task integration and solution/monitor modules. A MS SQL server database is implemented for managing large data flow from optimization processing. Task manager coordinates and interacts with other two modules. Task integration module works as a flowsheet constructor that builds task hierarchy, input parameter values and constrains. Task solution/monitor component presents both final and in-progress outputs in tabular and graphical formats. The package also provides a multiple-run process that executes a design matrix without invoking the optimization module. Experimental reports can be generated through database query and formatting.

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Gerasimov, Aleksander S., Boris R. Bergelson, and Tamara S. Zaritskaya. "Two Periods of Long-Term Storage of Thorium Spent Fuel." In ASME 2001 8th International Conference on Radioactive Waste Management and Environmental Remediation. American Society of Mechanical Engineers, 2001. http://dx.doi.org/10.1115/icem2001-1219.

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Abstract Radiotoxicity and decay heat power of actinides from spent thorium-uranium nuclear fuel of VVER-1000 type reactor during 100 000 year storage are discussed. Actinide accumulation in thorium fuel cycle is much less than in uranium fuel cycle. The radiotoxicity of actinides of thorium-uranium fuel by air is 5.5 times less and radiotoxicity by water is 3.5 times less than radiotoxicity of actinides of uranium fuel. Extraction of most important nuclides for transmutation permits to reduce radiologic danger of wastes remaining in storage.

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