Accessible Requires Authentication Published by De Gruyter August 18, 2021

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Eine Monte-Carlo-Studie zum Abbrand von MOX-Brennstoff in natriumgekühlten Reaktoren mit Th-Brennstoff
M. E. Korkmaz and N. K. Arslan
From the journal Kerntechnik

Abstract

Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Abstract

Natriumgekühlte Reaktoren gehören zu den Anlagen der Generation-IV, die für die Entsorgung der langlebigen minoren Aktiniden und die Transmutation der langlebigen radioaktiven Elemente ausgewählt wurden. Diese Studie stellt den Vergleich zwischen zwei konzipierten SFR-Kernen mit 600 und 800 MWth Gesamtheizleistung vor. Wir haben ein konzeptionelles Kerndesign und die nukleare Charakteristik des SFR analysiert. Es wurden Monte-Carlo-Berechnungen durchgeführt, um die wesentlichen Eigenschaften des SFR-Kerns zu untersuchen. Die Kernberechnungen wurden mit dem Monte-Carlo-Code Serpent unter Verwendung der Querschnittsbibliothek ENDF-7 durchgeführt, dabei wurde das Abbrandverhalten des SFR, die Leistungsverteilung und den effektiven Multiplikationsfaktor bestimmt. Der Kern des natriumgekühlten schnellen Reaktors wurde als Referenzkern für die Th-232-Abbrandberechnungen ausgewählt. Die Ergebnisse zeigen, dass der SFR eine wichtige Option für die Abreicherung der minoren Aktiniden sowie für die Transmutation von Th-232 zu U-233 ist.


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References

1 Ochoa, R.; Jimenez, G.; Perez-Martin,S.: Analysis of a Spanish energy scenario with Generation IV nuclear reactors, Energy Conversion and Management 75 (2013) 389–397, DOI:10.1016/j.enconman.2013.06.038 Search in Google Scholar

2 Mukaida, K.; Katoh, A.; Shiotani, H.; Hayafune, H.; Ono, K.: Benchmarking of economic evaluation models for an advanced loop-type sodium cooled fast reactor, Nuclear Engineering and Design, 324 (2017) 35–44, DOI:10.1016/j.nucengdes.2017.08.011 Search in Google Scholar

3 Al Qaaod, A. A.; Shahbunder, H.; Refeat, R. M.; Amin, E. A.; El-Kameesy, S. U.: Transmutation performance of uniform and nonuniform distributions of plutonium and minor actinides in TRIGA Mark II ADS reactor, Annals of Nuclear Energy, 121 (2018) 101– 107, DOI:10.1016/j.anucene.2018.07.012 Search in Google Scholar

4 Il Kim, Y.; Lee, Y. B.; Lee, C. B.; Chang, J.; Choi, C.: Design concept of advanced sodium-cooled fast reactor and related R&Din Korea, Science and Technology of Nuclear Installations, 2013 (2013), DOI:10.1155/2013/290362 Search in Google Scholar

5 Song, P.; Zhang, D.; Feng, T.; Wang, S.; Chen, J.; Wang, X.; Xue, X.; Zhang, Y.; Wang, M.; Qiu, S.; Su, G. H.: Numerical approach to study the thermal-hydraulic characteristics of Reactor Vessel Cooling system in sodium-cooled fast reactors, Progress in Nuclear Energy, 110 (2019) 213–223, DOI:10.1016/j.pnucene.2018.09.021 Search in Google Scholar

6 Kim, C.; Hartanto, D.; Kim, Y.: Neutronics feasibility of simple and dry recycling technologies for a self-sustainable breed-and-burn fast reactor, Annals of Nuclear Energy, 110 (2017) 847–855, DOI:10.1016/j.anucene.2017.07.036 Search in Google Scholar

7 Cui, D. Y.; Xia, S. P.; Li, X. X.; Cai, X. Z.; Chen, J. G.: Transition toward thorium fuel cycle in a molten salt reactor by using plutonium, Nuclear Science and Techniques, 28 (2017) 1–10, DOI:10.1007/s41365-017-0303-y Search in Google Scholar

8 Ault, T.; Krahn, S.; Worrall, A.; Croff, A.: Applications for Thorium in Multistage Fuel Cycles with Heavy Water Reactors, Nuclear Technology, 204 (2018) 41 –58, DOI:10.1080/00295450.2018.1468702 Search in Google Scholar

9 György, H.; Czifrus, S.: Investigation on the potential use of thorium as fuel for the Sodium-cooled Fast Reactor, Annals of Nuclear Energy, 103 (2017) 238–250, DOI:10.1016/j.anucene.2017.01.030 Search in Google Scholar

10 Hartanto, D.; Kim, C.; Kim, Y.: A comparative physics study for an innovative sodium-cooled fast reactor (iSFR), in: International Journal of Energy Research, 2018, DOI:10.1002/er.3612 Search in Google Scholar

11 Park, T.; Lin, C. S.; Yang,W. S.:A moderated target design for minor actinide transmutation in sodium-cooled fast reactor, Annals of Nuclear Energy, 98 (2016) 178–190, DOI:10.1016/j.anucene.2016.08.003 Search in Google Scholar

12 Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Brovchenko, M.; Ghetta, V.; Rubiolo, P.: Towards the thorium fuel cycle with molten salt fast reactors, Annals of Nuclear Energy, 64 (2014) 421–429, DOI:10.1016/j.anucene.2013.08.002 Search in Google Scholar

13 Korkmaz, M. E.; Agar, O.: The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor, Nuclear Engineering and Technology, 46 (2014) 407–412, DOI:10.5516/NET.07.2013.050 Search in Google Scholar

14 Korkmaz, M. E.; Agar, O.; Büyüker, E.: Burnup analysis of the VVER-1000 reactor using thorium-based fuel, Kerntechnik. 79 (2014) 478–483, DOI:10.3139/124.110449 Search in Google Scholar

15 Şarer, B.; Korkmaz, M. E.; Günay, M.; Aydin, A.: Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste, 13th Int. Conf. Emerg. Nucl. Energy Syst. 2007, ICENES 2007. 1 (2007) 129–141, DOI:10.1016/j.enconman.2007.09.029 Search in Google Scholar

16 Qvist, S.; Greenspan, E.: The ADOPT code for automated fast reactor core design, Annals of Nuclear Energy, 71 (2014) 23 –36, DOI:10.1016/j.anucene.2014.03.013 Search in Google Scholar

17 Mochizuki, H.: Analyses of decay heat removal tests with PRACS and DRACS of a sodium loop using the NETFLOW++ code, Nuclear Engineering and Design, 322 (2017) 345–359, DOI:10.1016/j.nucengdes.2017.07.013 Search in Google Scholar

18 Vidal, J. F.; Archier, P.; Faure, B.; Jouault, V.; Palau, J. M.; Pascal, V.; Rimpault, G.; Auffret, F.; Graziano, L.; Masiello, E.; Santandrea, S.: APOLLO3 homogenization techniques for transport core calculations–application to the ASTRID CFV core, Nuclear Engineering and Technology, 49 (2017) 1379–1387, 10.1016/j.net.2017.08.014 Search in Google Scholar

19 Ohshima, H.; Kubo, S.: 5 – Sodium-cooled fast reactor, Elsevier Ltd, 2016, DOI:10.1016/B978-0-08-100149-3.00005-7 Search in Google Scholar

20 Scherr, J.; Tsvetkov, P.: Annals of Nuclear Energy Reactor design strategy to support spectral variability within a sodium-cooled fast spectrum materials testing reactor, Annals of Nuclear Energy, 113 (2018) 15–24. , DOI:10.1016/j.anucene.2017.10.049 Search in Google Scholar

21 Aoto, K.; Dufour, P.; Hongyi, Y.; Glatz, J. P.; Il Kim, Y.; Ashurko, Y.; Hill, R.; Uto, N.: A summary of sodium-cooled fast reactor development, Progress in Nuclear Energy, 77 (2014) 247–265, DOI:10.1016/j.pnucene.2014.05.008 Search in Google Scholar

22 Loewen, E.; DeSilva, S.; Stachowski, R.: PRISM reference fuel design, Nuclear Engineering and Design, 340 (2018) 40 –53, DOI:10.1016/j.nucengdes.2018.09.016 Search in Google Scholar

23 Shin, S. H.; Kim, J. H.; Ryu, W. S.; Heo, H. M.; Kim, S. H.: Aging effect on the thermal transient behavior of the fuel cladding of a sodium-cooled fast reactor, Journal of Nuclear Materials, 495 (2017) 225–233, DOI:10.1016/j.jnucmat.2017.08.022 Search in Google Scholar

24 Silva Pinto Wahnon, S.; Ammirabile, L.; Kloosterman, J. L.; Lathouwers, D.: Multi-physics models for design basis accident analysis of sodium fast reactors. Part I: Validation of three-dimensional TRACE thermal-hydraulics model using Phenix end-of-life experiments, Nuclear Engineering and Design, 331 (2018) 331–341, DOI:10.1016/j.nucengdes.2018.02.038 Search in Google Scholar

25 Leppänen, J.: Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code, VTT Technical Research Centre of Finland (2013). http://www.polymtl.ca/merlin/Serpent_Dragon/Serpent_manual_2013.pdf Search in Google Scholar

26 Macdonald, R.; Driscoll,M. J.: Magnesium oxide: An improved reflector for blanket-free fast reactors, Transactions of the American Nuclear Society, 102 (2010) 488–489. Search in Google Scholar

27 Zhang, G.; Fratoni, M.; Greenspan,E.: Advanced burner reactors with breed-and-burn thorium blankets for improved economics and resource utilization, Nuclear Technology, 199 (2017) 187–218, DOI:10.1080/00295450.2017.1337408 Search in Google Scholar

28 Vauchy, R.; Belin, R. C.; Robisson, A. C.; Lebreton, F.; Aufore, L.; Scheinost, A. C.; Martin, P. M.: Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides, Inorganic Chemistry, 55 (2016) 2123–2132, DOI:10.1021/acs.inorgchem.5b02533 Search in Google Scholar

29 Tanaka, K.; Osaka, M.; Miwa, S.; Hirosawa, T.; Kurosaki, K.; Muta, H.; Uno, M.; Yamanaka, S.: Preparation and characterization of the simulated burnup americium-containing uranium-plutonium mixed oxide fuel, Journal of Nuclear Materials, 420 (2012) 207–212, DOI:10.1016/j.jnucmat.2011.10.004 Search in Google Scholar

30 World Nuclear Association, 2018. \Plutonium," (2018). https://www.world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/plutonium.aspx (accessed July 23, 2019). Search in Google Scholar

31 Zhang, G.; Fratoni, M.; Greenspan, E.: Fuel cycle analysis of Advanced Burner Reactor with breed-and-burn thorium blanket, Annals of Nuclear Energy,112 (2018) 383–394, DOI:10.1016/j.anucene.2017.10.016 Search in Google Scholar

32 György, H.; Czifrus, S.: The utilization of thorium in Generation IV reactors, Progress in Nuclear Energy, 93 (2016) 306–317, DOI:10.1016/j.pnucene.2016.09.007 Search in Google Scholar

Received: 2019-12-18
Published Online: 2021-08-18
Published in Print: 2021-08-31

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