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Journal articles on the topic 'Sodium cooled fast reactors'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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1

Dawn, William C., and Scott Palmtag. "A MULTIPHYSICS SIMULATION SUITE FOR SODIUM COOLED FAST REACTORS." EPJ Web of Conferences 247 (2021): 06019. http://dx.doi.org/10.1051/epjconf/202124706019.

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A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P3 (SP3) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors. Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity. This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models sufficiently describe the multiphysics feedback phenomena and can be used to estimate multiphysics reactivity feedback coefficients.

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2

Mishra, Govind Kumar, M. Sakthivel, S. L. N. Swamy, and K. Madhusoodanan. "Instrumentation for Sodium-Cooled Fast Breeder Reactors." Nuclear Science and Engineering 174, no. 1 (May 2013): 96–102. http://dx.doi.org/10.13182/nse10-74.

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3

Konomura, Mamoru, and Masakazu Ichimiya. "Design challenges for sodium cooled fast reactors." Journal of Nuclear Materials 371, no. 1-3 (September 2007): 250–69. http://dx.doi.org/10.1016/j.jnucmat.2007.05.012.

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4

Bagdasarov, Yu E. "Unanticipated accidents in sodium-cooled fast reactors." Atomic Energy 108, no. 3 (June 25, 2010): 165–69. http://dx.doi.org/10.1007/s10512-010-9272-5.

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5

Horowitz, Emmanuel. "ICONE19-43335 FURTHER BASIC STUDIES NEEDED TO SPECIFY MATERIALS FOR SODIUM COOLED FAST REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1943. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1943_149.

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6

Zhang, Zhen Guo, and Jie Xie. "The Application of New Materials in New Nuclear Reactors." Advanced Materials Research 1030-1032 (September 2014): 197–200. http://dx.doi.org/10.4028/www.scientific.net/amr.1030-1032.197.

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This paper combines the technical features and operation condition of the new nuclear reactors, such as sodium cooled fast reactor (SFR), supercritical water cooled reactor (SCWR), very high temperature reactor (VHTR), molten salt reactor (MSR) and nuclear fusion reactor, and analyzes the performance requirements of materials, it introduces the current application and prospects of new materials in different new nuclear reactors.

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7

Tenchine, D., C. Fournier, and Y. Dolias. "Gas entrainment issues in sodium cooled fast reactors." Nuclear Engineering and Design 270 (April 2014): 302–11. http://dx.doi.org/10.1016/j.nucengdes.2014.02.002.

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8

Crawford, Douglas C., Douglas L. Porter, and Steven L. Hayes. "Fuels for sodium-cooled fast reactors: US perspective." Journal of Nuclear Materials 371, no. 1-3 (September 2007): 202–31. http://dx.doi.org/10.1016/j.jnucmat.2007.05.010.

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9

Le coz, Pierre, Jean-François Sauvage, and Jean-Pol Serpantie. "Sodium-Cooled Fast Reactors: the ASTRID Plant Project." Revue Générale Nucléaire, no. 5 (September 2011): 39–44. http://dx.doi.org/10.1051/rgn/20115039.

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10

Bagdasarov, Yu E., and A. A. Kamaev. "Successive development phases for sodium-cooled fast reactors." Atomic Energy 111, no. 6 (March 23, 2012): 389–97. http://dx.doi.org/10.1007/s10512-012-9509-6.

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11

Raj, Baldev. "Materials science research for sodium cooled fast reactors." Bulletin of Materials Science 32, no. 3 (June 2009): 271–83. http://dx.doi.org/10.1007/s12034-009-0041-9.

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12

Tenchine, D. "Some thermal hydraulic challenges in sodium cooled fast reactors." Nuclear Engineering and Design 240, no. 5 (May 2010): 1195–217. http://dx.doi.org/10.1016/j.nucengdes.2010.01.006.

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13

Coulon, R., S. Normand, G. Ban, E. Barat, T. Montagu, T. Dautremer, H. P. Brau, et al. "Delayed gamma power measurement for sodium-cooled fast reactors." Nuclear Engineering and Design 241, no. 1 (January 2011): 339–48. http://dx.doi.org/10.1016/j.nucengdes.2010.10.002.

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14

Tenchine, D., V. Barthel, U. Bieder, F. Ducros, G. Fauchet, C. Fournier, B. Mathieu, F. Perdu, P. Quemere, and S. Vandroux. "Status of TRIO_U code for sodium cooled fast reactors." Nuclear Engineering and Design 242 (January 2012): 307–15. http://dx.doi.org/10.1016/j.nucengdes.2011.10.026.

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15

Tenchine, D., R. Baviere, P. Bazin, F. Ducros, G. Geffraye, D. Kadri, F. Perdu, D. Pialla, B. Rameau, and N. Tauveron. "Status of CATHARE code for sodium cooled fast reactors." Nuclear Engineering and Design 245 (April 2012): 140–52. http://dx.doi.org/10.1016/j.nucengdes.2012.01.019.

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16

Sidorenko, V. A. "Comparison of lead-and sodium-cooled fast-neutron reactors." Atomic Energy 86, no. 3 (March 1999): 170–72. http://dx.doi.org/10.1007/bf02672941.

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17

Murogov, V. M., V. I. Subbotin, V. S. Kagramanyan, V. M. Poplavskii, N. S. Rabotonov, and M. F. Troyanov. "Stimulation of the development of sodium-cooled fast reactors." Atomic Energy 74, no. 4 (April 1993): 268–71. http://dx.doi.org/10.1007/bf00739010.

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18

Abdelhameed, Ahmed Amin E., Chihyung Kim, and Yonghee Kim. "Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors." Energies 14, no. 15 (July 29, 2021): 4610. http://dx.doi.org/10.3390/en14154610.

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The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

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19

Cui, Manman, Yun GUO, and Zhijian Zhang. "ICONE19-43225 SIMULATION ON PRIMARY COOLANT SYSTEM OF SODIUM COOLED FAST REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1943. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1943_93.

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20

Wessels, J. "Inservice inspection of the reactor block of sodium-cooled fast breeder reactors." Nuclear Engineering and Design 130, no. 1 (September 1991): 33–42. http://dx.doi.org/10.1016/0029-5493(91)90190-s.

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21

Babushkin, S. V., B. A. Vasil’ev, A. V. Vasyaev, V. E. Vorontsov, S. A. Dushev, D. L. Zverev, A. I. Kiryushin, et al. "Reactor Installations with Sodium-Cooled Fast Reactors for Two-Component Nuclear Energy." Atomic Energy 129, no. 1 (November 2020): 8–17. http://dx.doi.org/10.1007/s10512-021-00705-y.

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22

Maslov, N. V., E. I. Grishanin, and P. N. Alekseev. "Calculation studies of coated particles performance in sodium-cooled fast reactor." Kerntechnik 86, no. 1 (March 1, 2021): 45–49. http://dx.doi.org/10.1515/kern-2019-0054.

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Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

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23

Uchibori, Akihiro, and Hiroyuki Ohshima. "ICONE23-1502 APPLICABILITY OF A MECHANISTIC NUMERICAL METHOD FOR SODIUM-WATER REACTION PHENOMENA IN STEAM GENERATORS OF SODIUM-COOLED FAST REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–1—_ICONE23–1. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-1_230.

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24

EZURE, Toshiki, Takamitsu ONOJIMA, Yuri KAMEYAMA, and Masaaki TANAKA. "Study on Decay Heat Removal in Sodium-cooled Fast Reactors." Proceedings of the National Symposium on Power and Energy Systems 2019.24 (2019): B112. http://dx.doi.org/10.1299/jsmepes.2019.24.b112.

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25

Zhao, Haihua, and Per F. Peterson. "Multiple reheat helium Brayton cycles for sodium cooled fast reactors." Nuclear Engineering and Design 238, no. 7 (July 2008): 1535–46. http://dx.doi.org/10.1016/j.nucengdes.2007.12.002.

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26

MORRIS, Edgar E., and W. Mark NUTT. "Uncertainty Analysis for Unprotected Accidents in Sodium-Cooled Fast Reactors." Journal of Nuclear Science and Technology 48, no. 4 (April 2011): 532–37. http://dx.doi.org/10.1080/18811248.2011.9711730.

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27

Buiron, L., Ph Dufour, G. Rimpault, G. Prulhiere, C. Thevenot, J. Tommasi, F. Varaine, and A. Zaetta. "Innovative Core Design for Generation IV Sodium-Cooled Fast Reactors." Revue Générale Nucléaire, no. 4 (July 2007): 48–62. http://dx.doi.org/10.1051/rgn/20074048.

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28

Rachkov, V. I., Yu S. Khomyakov, and Yu E. Shvetsov. "Russian Codes for Safety Analysis of Sodium-Cooled Fast Reactors." Atomic Energy 116, no. 4 (July 27, 2014): 265–70. http://dx.doi.org/10.1007/s10512-014-9852-x.

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29

Hombourger, Boris, Jiři Křepel, and Andreas Pautz. "Breed-and-burn fuel cycle in molten salt reactors." EPJ Nuclear Sciences & Technologies 5 (2019): 15. http://dx.doi.org/10.1051/epjn/2019026.

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The operation of a reactor on an open but self-sustainable cycle without actinide separation is known as breed-and-burn. It has mostly been envisioned for use in solid-fueled fast-spectrum reactors such as sodium-cooled fast reactors. In this paper the applicability of breed-and-burn to molten salt reactors is investigated first on a cell level using a modified neutron excess method. Several candidate fuel salts are selected and their performance in a conceptual three-dimensional reactor is investigated. Chloride-fueled single-fluid breed-and-burn molten salt reactors using enriched chlorine are shown to be feasible from a neutronics and fuel cycle point of view at the cost of large fuel inventories.

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30

Jayakumar, Tammana, A. K. Bhaduri, M. D. Mathew, Shaju K. Albert, and U. Kamachi Mudali. "Nitrogen Enhanced 316LN Austenitic Stainless Steel for Sodium Cooled Fast Reactors." Advanced Materials Research 794 (September 2013): 670–80. http://dx.doi.org/10.4028/www.scientific.net/amr.794.670.

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For the future sodium-cooled fast reactors (SFRs), which are envisaged with a design life of 60 years, nitrogen-enhanced 316LN austenitic stainless steel (SS) with improved high-temperature properties is being developed. To optimize the enhanced nitrogen content in 316LN SS, the effect of nitrogen on its tensile, creep and low cycle fatigue behavior has been investigated. For different heats of 316LN SS containing 0.07-0.22 wt% nitrogen, the tensile and creep properties increased with increase in nitrogen content, while low cycle fatigue properties peaked at 0.14 wt% nitrogen. Finally, based on the evaluation of the hot cracking susceptibility of the different heats of 316LN SS with varying nitrogen content, using the Varestraint and Gleeble hot-ductility tests, the nitrogen content for the nitrogen-enhanced 316LN SS has been optimized at a level of 0.14 wt%. The 0.14 wt% nitrogen content in this optimised composition shifts the solidification mode of the weld metal to fully austenitic region, including that due to dilution of nitrogen from the base metal, thereby increasing its hot cracking susceptibility. This necessitated development and qualification of welding electrodes for obtaining weld metal with 0.14 wt% nitrogen by optimising the weld metal chemistry so as to obtain the requisite delta ferrite content, tensile properties, and very importantly impact toughness both in the as-welded and aged conditions. Studies on localised corrosion behaviour of nitrogen-enhanced 316LN SS indicated the beneficial effect of nitrogen addition to sensitization, pitting, intergranular corrosion, stress corrosion cracking and corrosion fatigue.

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31

Bhaduri, A. K., K. Laha, V. Ganesan, T. Sakthivel, M. Nandagopal, G. V. Prasad Reddy, J. Ganesh Kumar, et al. "Advanced materials for structural components of Indian sodium-cooled fast reactors." International Journal of Pressure Vessels and Piping 139-140 (March 2016): 123–36. http://dx.doi.org/10.1016/j.ijpvp.2016.02.027.

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32

Chellapandi, P., S. C. Chetal, and Baldev Raj. "Thermal striping limits for components of sodium cooled fast spectrum reactors." Nuclear Engineering and Design 239, no. 12 (December 2009): 2754–65. http://dx.doi.org/10.1016/j.nucengdes.2009.08.014.

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33

Kazachkovskii, O. D., V. A. Eliseev, V. I. Matveev, and A. M. Tsibulya. "Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors." Atomic Energy 96, no. 5 (May 2004): 332–37. http://dx.doi.org/10.1023/b:aten.0000038099.30364.6a.

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34

Raj, Baldev, and M. Vijayalakshmi. "Ferritic steels for sodium-cooled fast reactors: Design principles and challenges." JOM 62, no. 9 (September 2010): 75–83. http://dx.doi.org/10.1007/s11837-010-0143-2.

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35

Poplavskii, V. M., L. M. Zabudko, I. A. Shkabura, M. V. Skupov, A. V. Bychkov, V. A. Kislyi, F. N. Kryukov, and B. A. Vasiliev. "Fuel for advanced sodium-cooled fast reactors: current status and plans." Atomic Energy 108, no. 4 (August 2010): 267–73. http://dx.doi.org/10.1007/s10512-010-9288-x.

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36

Kochetkov, L. A., A. I. Kiryushin, and N. N. Oshkanov. "Sodium-cooled fast reactors in Russia: Looking beyond the year 2000." Atomic Energy 74, no. 4 (April 1993): 265–67. http://dx.doi.org/10.1007/bf00739009.

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37

Raj, Baldev, Divakar Ramachandran, and M. Vijayalakshmi. "Development of cladding materials for sodium-cooled fast reactors in India." Transactions of the Indian Institute of Metals 62, no. 2 (April 2009): 89–94. http://dx.doi.org/10.1007/s12666-009-0012-2.

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38

EZURE, Toshiki, Kazuyoshi NAGASAWA, and Masaaki TANAKA. "Numerical Analysis on Thermal Stratification Phenomena in Sodium-cooled Fast Reactors." Proceedings of Mechanical Engineering Congress, Japan 2020 (2020): S08110. http://dx.doi.org/10.1299/jsmemecj.2020.s08110.

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39

NISHI, Yoshihisa, Nobuyuki UEDA, Tomonari KOGA, and Hisato MATSUMIYA. "ICONE15-10603 HEAT REMOVAL CHARACTERISTICS OF THE 10MWE SODIUM COOLED SMALL FAST REACTOR (4S)." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_327.

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40

Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, and Anatoliy Blokhin. "Lead reactor of small power with metallic fuel." Nuclear Energy and Technology 4, no. 2 (November 26, 2018): 99–102. http://dx.doi.org/10.3897/nucet.4.30527.

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The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.

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41

UCHIBORI, Akihiro, Hiroyuki OHSHIMA, Takashi TAKATA, and Akira YAMAGUCHI. "OS8-9 Numerical Analysis of Sodium-Water Reaction Phenomena in Sodium Cooled Fast Reactors." Proceedings of the National Symposium on Power and Energy Systems 2007.12 (2007): 211–12. http://dx.doi.org/10.1299/jsmepes.2007.12.211.

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42

Holdampf, Sydney A., Andrew G. Osborne, and Mark R. Deinert. "VALIDATION OF COOLANT THERMAL RESPONSE IN A TRANSIENT FINITE DIFFERENCE THERMAL TRANSPORT MODEL WITH APPLICATIONS TO FAST SPECTRUM REACTORS." EPJ Web of Conferences 247 (2021): 07018. http://dx.doi.org/10.1051/epjconf/202124707018.

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Transient behavior in nuclear reactors is important in accidents and with reactivity control systems that are driven by thermal feedback. Here, we describe a transient finite difference model for a pin cell system. The fidelity of the model is shown by validation against the thermocouple measurements of the CABRI BI1 experiment and the Safety Analysis System-Sodium Fast Reactor model of the experiment. In the BI1 experiment, a sodium-cooled mixed oxide fuel pin was subject to a loss of flow transient to coolant boiling within a sodium test loop positioned in the center of the CABRI research reactor. Comparisons to the initial steady-state coolant temperature profile, coolant temperature profile at twenty seconds into the transient, and at four axial locations within the coolant show agreement of the simple model with the experimental results better than or similar to those of the Safety Analysis System-Sodium Fast Reactor model. The model can be used to determine the thermal response times of coolant in fast reactors currently operating or in the design phase when subject to loss of flow accidents or other transients. Here, we investigate the difference in coolant thermal response for metal fueled and mixed oxide fueled sodium fast reactors when subject to transient overpower and loss of flow events. Additionally, we determine the effect of pin pitch on outlet coolant temperatures during the overpower event. Finally, we return to the CABRI experiment and show the importance of porosity in fuel temperature calculations.

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43

Alekseev, Viktor V., Yuliya A. Kuzina, and Aleksandr P. Sorokin. "Conceptual issues of the cold filter trap development for the sodium coolant purification in fast-neutron reactors." Nuclear Energy and Technology 6, no. 2 (June 25, 2020): 105–11. http://dx.doi.org/10.3897/nucet.6.55220.

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The paper presents the results of studying the peculiarities of heat and mass exchange in cold traps for the sodium purification of impurities in fast reactor circuits both in dedicated test areas simulating various trap components (isothermal sump, nonisothermal sump, filters, final cooling area) and in trap prototype models. As a result, a scientific rationale has been formed for developing traps of a unique design for various reactors. The impurity capacity of the traps is three to four times as high as that of the best foreign counterparts. Tests have shown these to be highly efficient in purifying sodium of oxygen and hydrogen and much less efficient in sodium purification of corrosion products and carbon. Taking into account the leakage of radioactive sodium during operation of the BN-600 reactor primary circuit traps, a decision was made to install the purification system in the reactor tank to improve the safety of the large fast reactor. It was resolved to exclude the accumulation of hydrogen in the primary circuit traps in nominal conditions. Two trap designs, with argon and sodium cooling, are discussed. It has been shown that operation of the reactor purification system with argon cooling will require 20 trap replacements during the reactor operating life and seven replacements if the deposition of hydrogen into the primary circuit cold traps is excluded. The sodium-cooled version of the trap built in the reactor tank has the same overall dimensions as the argon-cooled trap. The cooling sodium circulates in two trains: outside the jacketed working space body (up to 30% of the flow rate) and in the coil inside of the working space (up to 70% of the flow rate). Updates have been proposed to the trap design based on the calculations using the codes simulating the in-trap processes of heat and mass exchange.

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44

Usanov, Vladimir I., Stepan A. Kvyatkovskiy, Andrey A. Andrianov, and Ilya S. Kuptsov. "Multi-criteria evaluation and ranking of potential scenarios for the development of Russian two-component nuclear energy system with thermal and sodium-cooled fast reactors." Nuclear Energy and Technology 7, no. 3 (September 14, 2021): 165–72. http://dx.doi.org/10.3897/nucet.7.72391.

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The paper presents the results from a multi-criteria comparative evaluation of potential deployment scenarios for Russian nuclear power with thermal and sodium-cooled fast reactors in a closed nuclear fuel cycle (the so-called two-component nuclear energy system). The comparison and the ranking were performed taking into account the recommendations and using the IAEA/INPRO software tools for comparative evaluation of nuclear energy systems, including tools for sensitivity/uncertainty analysis with respect to weighting factors. Ten potential Russian nuclear power deployment scenarios with different shares of thermal and sodium-cooled fast reactors were considered, including options involving the use of MOX fuel in VVER reactors. Eight key indicators were used, estimated as of 2100 and structured into a three-level objectives tree. The comparative evaluation and the ranking were carried out based on the multi-attribute value theory. The model for assessing the key indicators was developed using the IAEA/INPRO MESSAGE-NES energy system planning software tool. The information base for the study was formed by publications of experts from JSC SSC RF-IPPE, NRC Kurchatov Institute and NRNU MEPhI. The presented results show that it is possible to enhance significantly the sustainability of the Russian nuclear energy system, when considering multiple performance indicators, through the intensive deployment of sodium-cooled fast reactors and the transition to a closed nuclear fuel cycle. Tasks have been outlined for the follow-up studies to make it possible to obtain more rigorous conclusions regarding the preferred options for the evolution of a two-component nuclear energy system.

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45

Lee, Je-Whan, Yong-Hoon Jeong, Yoon-Il Chang, and Soon-Heung Chang. "LINEAR PROGRAMMING OPTIMIZATION OF NUCLEAR ENERGY STRATEGY WITH SODIUM-COOLED FAST REACTORS." Nuclear Engineering and Technology 43, no. 4 (August 31, 2011): 383–90. http://dx.doi.org/10.5516/net.2011.43.4.383.

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46

Huang, Xiaoxue, and Shuisheng He. "Numerical modelling of cover gas thermal hydraulics in Sodium-cooled Fast Reactors." Nuclear Engineering and Design 355 (December 2019): 110347. http://dx.doi.org/10.1016/j.nucengdes.2019.110347.

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47

Takeda, Toshikazu, Yoichiro Shimazu, Basma Foad, and Katsuhisa Yamaguchi. "Review of safety improvement on sodium-cooled fast reactors after Fukushima accident." Natural Science 04, no. 11 (2012): 929–35. http://dx.doi.org/10.4236/ns.2012.431121.

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Bhaduri, A. K., and K. Laha. "Development of Improved Materials for Structural Components of Sodium-Cooled Fast Reactors." Procedia Engineering 130 (2015): 598–608. http://dx.doi.org/10.1016/j.proeng.2015.12.276.

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Tenchine, D., S. Vandroux, V. Barthel, and O. Cioni. "Experimental and numerical studies on mixing jets for sodium cooled fast reactors." Nuclear Engineering and Design 263 (October 2013): 263–72. http://dx.doi.org/10.1016/j.nucengdes.2013.06.001.

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Saraev, O. M., N. N. Oshkanov, A. V. Zrodnikov, V. M. Poplavskii, Yu M. Ashurko, M. V. Bakanov, B. A. Vasiliev, et al. "Operating experience and prospects for future development of sodium-cooled fast reactors." Atomic Energy 108, no. 4 (August 2010): 240–47. http://dx.doi.org/10.1007/s10512-010-9284-1.

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