ТОРИЙ ОТЫНЫ ҚОЛДАНЫЛАТЫН ҚЫСЫМДАҒЫ СУ РЕАКТОРЫНЫҢ НЕЙТРОНДЫҚ-ФИЗИКАЛЫҚ ЖӘНЕ ЖЫЛУ-ГИДРАВЛИКАЛЫҚ СИПАТТАМАЛАРЫН ЗЕРТТЕУ

Жұмыста жылу-гидравликалық үдерістердің және қатты дене механикасы құбылыстарының динамикасы COMSOL Multiphysics бағдарламалық кешенінде соңғы элементтер әдісі негізінде модельденді. Нейтрондық-физикалық сипаттамаларды зерттеу үшін OpenMC бағдарламалық пакеті стационарлық нейтроника жуықтауында қолданылды. Торий негізіндегі реактордың белсенді аймағын оңтайландыру мақсатында салқындатқыш ретінде жоғары қысымды жеңіл су пайдаланылды. Белсенді аймақта торий қосылыстарымен толтырылған 49 отын штангасы орналастырылды. OpenMC пакетінде жүзеге асырылған Монте-Карло әдісі арқылы нейтрондардың тасымалдануын модельдеу нәтижесінде қарастырылған реактор үлгісінің тиімді көбейту коэффициенті (kₑff) мен қуаттың таралуы анықталды. Жылу-гидравликалық есептеулер нәтижесінде температуралық профильдер мен ағын жылдамдығының мәндері алынды, ал қатты дене механикасы бойынша жүргізілген талдау нәтижесінде кернеулер мен деформациялардың кеңістіктік таралуы анықталды. Алынған нәтижелер стационарлық нейтроника мен өтпелі жылу-гидравликаны байланыстыра есептеу әдісін қолданудың тиімділігін растады.

1. World Nuclear Association. Thorium. URL: https://world-nuclear.org/information-library/currentand-future-generation/thorium

2. Oak Ridge National Laboratory. Molten-Salt Reactor Experiment 1965–1972 (Oak Ridge, TN: ORNL, 2015). URL: https://web.archive.org/web/20160303211133if_/https://dl.dropboxusercontent.com/u/15726934/Historic_Molten_Salt_Reactor_Experiment_Brochure_ORNL_1965-1972.pdf

3. Carter, L.L. Monte Carlo Code Development in Los Alamos. Los Alamos Scientific Laboratory Report (March 1975).

4. James, M.R. MCNPX 2.7.x – New Features Being Developed. Los Alamos National Laboratory Technical Report.

5. Romano, P.K., Horelik, N.E., Herman, B.R., Nelson, A.G., Forget, B., and Smith, K. OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development. Annals of Nuclear Energy, 82, 90–97 (2015). https://doi.org/10.1016/j.anucene.2014.07.048

6. Brown, D.A., Chadwick, M.B., Capote, R., Kahler, A.C., Trkov, A., Herman, M.W., Sonzogni, A.A., Danon, Y., Carlson, A.D., Dunn, M., et al. ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-Project Cross Sections, New Standards and Thermal Scattering Data. Nuclear Data Sheets, 148, 1–142 (2018). https://doi.org/10.1016/j.nds.2018.02.001

7. Plompen, A.J.M., Cabellos, O., De Saint Jean, C., Fleming, M., Algora, A., Angelone, M., Archier, P., et al. The Joint Evaluated Fission and Fusion Nuclear Data Library, JEFF-3.3. European Physical Journal A, 56 (7), 181 (2020). https://doi.org/10.1140/epja/s10050-020-00141-9

8. Tsige-Tamirat, H. Neutronics Assessment of the Use of Thorium Fuels in Current Pressurized Water Reactors. Progress in Nuclear Energy, 53 (6), 717–721 (2011). https://doi.org/10.1016/j.pnucene.2010.12.001

9. Trellue, H.R., Bathke, C.G., and Sadasivan, P. Neutronics and Material Attractiveness for PWR Thorium Systems Using Monte Carlo Techniques. Progress in Nuclear Energy, 53 (6), 698–707 (2011). https://doi.org/10.1016/j.pnucene.2011.02.002

10. Chaudri, K.S., Tian, W., Su, G., and Qiu, S. Coupled Neutronics/Thermal Hydraulics Evaluation for Thorium Based Fuels in Thermal Spectrum SCWR. Progress in Nuclear Energy, 68, 55–64 (2013). https://doi.org/10.1016/j.pnucene.2013.04.005

11. Arzhannikov, A.V., Shmakov, V.M., Modestov, D.G., Bedenko, S.V., Prikhodko, V.V., Lutsik, I.O., and Shamanin, I.V. Facility to Study Neutronic Properties of a Hybrid Thorium Reactor with a Source of Thermonuclear Neutrons Based on a Magnetic Trap. Nuclear Engineering and Technology, 52 (11), 2460– 2470 (2020). https://doi.org/10.1016/j.net.2020.04.017

12. Kumar, E.R.S.A., Pancholi, M.K., Darnowski, P., and Dzido, A. Neutronic Performance of a Thorium Based Mixed Oxide Fuel in a Burner Sodium-Cooled Fast Reactor. Energy, 212, 118744 (2020). https://doi.org/10.1016/j.energy.2020.118744

13. Galahom, A.A., Mohsen, M.Y., and Amrani, N. Explore the Possible Advantages of Using ThoriumBased Fuel in a Pressurized Water Reactor (PWR). Part 1: Neutronic Analysis. Nuclear Engineering and Technology, 54 (1), 1–10 (2022). https://doi.org/10.1016/j.net.2021.07.012

14. Li, J., Li, X., and Cai, J. Neutronic Characteristics and Feasibility Analysis of Micro-Heterogeneous Duplex ThO₂-UO₂ Fuel Pin in PWR. Nuclear Engineering and Design, 382, 111382 (2021). https://doi.org/10.1016/j.nucengdes.2021.111382

15. Kabach, O., Chetaine, A., Benchrif, A., Amsil, H., and El Banni, F. A Comparative Analysis of the Neutronic Performance of Thorium Mixed with Uranium or Plutonium in a High-Temperature Pebble-Bed Reactor. International Journal of Energy Research, 45 (11), 16824–16841 (2021). https://doi.org/10.1002/er.6863

16. Mirvakili, S.M., Kavafshary, M.A., and Vaziri, A.J. Comparison of Neutronic Behavior of UO₂, (Th233U)O₂ and (Th-235U)O₂ Fuels in a Typical Heavy Water Reactor. Nuclear Engineering and Technology, 47 (3), 315–322 (2015). https://doi.org/10.1016/j.net.2014.12.005

17. Tucker, L.P., Alajo, A., and Usman, S. Thorium-Based Mixed Oxide Fuel in a Pressurized Water Reactor: A Beginning-of-Life Feasibility Analysis with MCNP. Annals of Nuclear Energy, 76, 323–334 (2015). https://doi.org/10.1016/j.anucene.2014.10.020

18. Pillai, C.G.S., and Raj, P. Thermal Conductivity of ThO₂ and Th0.98U0.02O₂. Journal of Nuclear Materials, 277 (1), 116–119 (2000). https://doi.org/10.1016/S0022-3115(99)00237-8

19. Kutty, T.R.G., Kulkarni, R.V., Sengupta, P., Khan, K.B., Bhanumurthy, K., Sengupta, A.K., Panakkal, J.P., Kumar, A., and Kamath, H.S. Development of CAP Process for Fabrication of ThO₂–UO₂ Fuels. Part II: Characterization and Property Evaluation. Journal of Nuclear Materials, 373 (1–3), 309–318 (2008). https://doi.org/10.1016/j.jnucmat.2007.06.011

20. Bakker, K., Cordfunke, E.H.P., Konings, R.J.M., and Schram, R.P.C. Critical Evaluation of the Thermal Properties of ThO₂ and Th1−yUyO₂ and a Survey of the Literature Data on Th1−yPuyO₂. Journal of Nuclear Materials, 250 (1), 1–12 (1997). https://doi.org/10.1016/S0022-3115(97)00241-9

21. Tyagi, A.K., Mathews, M.D., Ambekar, B.R., and Ramachandran, R. Thermal Expansion of ThO₂–2, 4 and 6 wt.% UO₂ by HT-XRD. Thermochimica Acta, 421 (1–2), 69–71 (2004). https://doi.org/10.1016/j.tca.2004.04.025

22. Tyagi, A.K., and Mathews, M.D. Thermal Expansion of ThO₂–2 wt.% UO₂ by HT-XRD. Journal of Nuclear Materials, 278 (1), 123–125 (2000). https://doi.org/10.1016/S0022-3115(99)00277-9

23. Anthonysamy, S., Panneerselvam, G., Bera, S., Narasimhan, S.V., and Vasudeva Rao, P.R. Studies on Thermal Expansion and XPS of Urania–Thoria Solid Solutions. Journal of Nuclear Materials, 281 (2–3), 308–316 (2000). https://doi.org/10.1016/S0022-3115(00)00185-9

24. Ghosh, P.S., Somayajulu, P.S., Krishnan, K., Pathak, N., Arya, A., and Dey, G.K. Thermal Expansion and Thermal Conductivity of (Th,U)O₂ Mixed Oxides: A Molecular Dynamics and Experimental Study. Journal of Alloys and Compounds, 650, 165–177 (2015). https://doi.org/10.1016/j.jallcom.2015.07.260

25. Xiao, H., Long, C., Tian, X., Chen, H., and Duan, X. Effect of Thorium Addition on the Thermophysical Properties of Uranium Dioxide: Atomistic Simulations. Materials & Design, 96, 335–340 (2016). https://doi.org/10.1016/j.matdes.2016.02.019

26. Griggs, D.P., Henry, A.F., and Kazimi, M.S. Development of a Three-Dimensional Two-Fluid Code with Transient Neutronic Feedback for LWR Applications. Massachusetts Institute of Technology Technical Report (1981).

27. Mahadevan, V.S. High Resolution Numerical Methods for Coupled Non-Linear Multi-Physics Simulations with Applications in Reactor Analysis. PhD Dissertation, Texas A&M University, College Station, TX, USA (2010).

28. Pijls, T. Developing a GPU-Accelerated Filter-Matrix Lattice Boltzmann Multiphysics Tool for the Transient Thermal-Hydraulics and Neutronics of a Molten Salt Fast Reactor Core. MSc Thesis, Delft University of Technology, Delft, The Netherlands (2025). URL: https://martinrohde.nl/theses/msc/MSc_Thomas_Pijls. pdf

29. Hanumantharao, G., Vijay, S., and Venkateswara Rao, M. Transient Heat Transfer Analysis for Optimum Temperature Distribution to Reduce Thermal Stresses. International Journal of Engineering Research and Technology, 1, 1–7 (2012).

30. Fuertes-Miquel, V.S. Transient Flows: Mathematical Models, Laboratory Tests, Protection Elements and Systems. Water, 17, 1454 (2025). https://doi.org/10.3390/w17111454

31. Hu, R., Fang, J., Nunez, D., Tano, M., Giudicelli, G., and Salko, R. Development of Integrated Thermal Fluids Modeling Capability for MSRs. Argonne National Laboratory Report ANL/NSE-22/56 (2022).

32. Mohsen, M.Y.M., Abdel-Rahman, M.A.E., and Galahom, A.A. Ensuring the Possibility of Using Thorium as a Fuel in a Pressurized Water Reactor (PWR). Nuclear Science and Techniques, 32, 137 (2021). https://doi.org/10.1007/s41365-021-00981-0

33. Hammer, R.R. Zircaloy-4, Uranium Dioxide, and Materials Formed by Their Interaction: A Literature Review with Extrapolation of Physical Properties to High Temperatures. Idaho Nuclear Corporation Report IN-1093 (Idaho Falls, ID, USA, 1967).

34. Shields, A.E., Hernandez, S.E.R., and de Leeuw, N.H. Theoretical Analysis of Uranium-Doped Thorium Dioxide: Introduction of a Thoria Force Field with Explicit Polarization. AIP Advances, 5, 087118 (2015). https://doi.org/10.1063/1.4928438

35. Jiao, G., Xia, G., Zhu, H., Zhou, T., and Peng, M. Thermal-Mechanical Coupling Characteristics and Heat Pipe Failure Analysis of Heat Pipe Cooled Space Reactor. Annals of Nuclear Energy, 192, 110025 (2023). https://doi.org/10.1016/j.anucene.2023.110025

36. Mohsen, M.Y.M., Soliman, A.Y., and Abdel-Rahman, M.A.E. Thermal-Hydraulic and Solid Mechanics Safety Analysis for VVER-1000 Reactor Using Analytical and CFD Approaches. Progress in Nuclear Energy, 130, 103568 (2020). https://doi.org/10.1016/j.pnucene.2020.103568

37. Mohsen, M.Y.M., Abdel-Rahman, M.A.E., Hassan, M.S., and Galahom, A.A.G. Searching for the Most Optimum Burnable Absorbers (BAs) for AP-1000 from the Neutronic, Thermal-Hydraulic, and Solid Mechanics Points of View. Nuclear Engineering and Design, 391, 111728 (2022). https://doi.org/10.1016/j.nucengdes.2022.111728