Investigation of Neutron Attenuation Properties of Boron Carbide-Silicon Carbide Composite Materials by MCNP6 Simulation

Öz

The macroscopic neutron cross-sections of boron carbide-silicon carbide composite materials used in critical fields such as nuclear technology, nanotechnology and space technology were calculated using the MCNP6 program. The interaction of neutrons with matter depends especially on the neutron energy and the density of the shielding material. In this study, neutron intensities of 4 different B4C-SiC composites which were coded as 100_B4C, 8200, 7300 and 6400. Experimental neutron results obtained using Neutron Howitzer (Pu-Be) counting system in ITU TRIGA Mark-II nuclear reactor were compared with Monte Carlo simulations. Neutron attenuation properties of the samples were evaluated by using theoretical obtained results and compared with the experimental results in the literature. Composites with the highest neutron radiation shielding were found to be 100_B4C>8200>7300>6400, respectively. In addition, the total macroscopic cross-sections of neutrons Σ_tot (cm-1), were calculated theoretically using the mass separation cross-section values of the elements (Σ_tot/ρ) (cm2g-1), and the difference was determined.

Anahtar Kelimeler

• B4C
• Neutron radiation
• TRIGA Mark-II reactor
• MCNP6

Bor karbür-silisyum karbür kompozit malzemelerin nötron zayıflatma özelliklerinin MCNP6 simülasyonu ile incelenmesi

Öz

Nükleer teknoloji, nanoteknoloji ve uzay teknolojisi gibi kritik alanlarda kullanılan bor karbür-silisyum karbür kompozit malzemelerinin makroskobik nötron tesir kesitleri, MCNP6 programı kullanılarak hesaplanmıştır. Nötronların madde ile etkileşimi, özellikle nötron enerjisine ve koruyucu malzemenin yoğunluğuna bağlıdır. Bu çalışmada, farklı oranlarda B4C ve SiC bileşenleri kullanılarak oluşturulan 100_B4C, 8200, 7300 ve 6400 kodlu 4 farklı B4C-SiC içerikli kompozit kullanılmıştır. İTÜ TRIGA Mark-II nükleer reaktöründe Nötron Howitzer (Pu-Be) sayım sistemi kullanılarak elde edilen deneysel nötron sonuçları Monte Carlo simülasyonları ile karşılaştırılmıştır. Bu karşılaştırma, kompozit malzemelerin nötron zayıflatma özelliklerini deneysel ve teorik sonuçların birlikte değerlendirilmesini sağlamıştır. En yüksek nötron radyasyon zayıflatma özelliğine sahip olan kompozitlerin sırasıyla 100_B4C>8200>7300>6400 şeklinde olduğu görülmüştür. Ayrıca, nötronların toplam makroskobik tesir kesitleri Σ_tot (cm-1), elementlerin kütlesel ayırma tesir kesiti değerleri (Σ_tot/ρ) (cm2g-1), teorik olarak da hesaplanmış ve aradaki fark belirlenmiştir.

Anahtar Kelimeler

• B4C
• Nötron radyasyonu
• TRIGA Mark-II araştırma reaktörü
• MCNP6

Kaynakça

1. Gencel, O., Brostow, W., Ozel, C., & Filiz, M. (2010). An Investigation on the concrete properties containing colemanite. International Journal of Physical Sciences, 5(3), 216-225. http://www.academicjournals.org/IJPS.
2. Barbieri, S. (2019). DNA damage by charged and neutral radiation at different spatial and temporal scales: integrating Monte Carlo simulations with in vitro experiments [Doctoral dissertation, University of Pavia].
3. Rinard, P. (1991). Neutron interactions with matter. In D. Reilly, N. Ensslin, H. Smith Jr., & S. Kreiner (Eds.) Passive Nondestructive Assay of Nuclear Materials (pp.375-377). ISBN 0160327245.
4. Singh, V. P., Medhat, M. E., & Badiger, N. M. (2014). Utilization of Geant4 Monte Carlo simulation method for studying attenuation of photons in normal and heavy concretes at high energy values. Journal of Radioanalytical and Nuclear Chemistry, 300, 325-331. https://doi.org/10.1007/s10967-014-2984-6.
5. Buyuk, B., Tugrul, A. B., Demir, E., Aktop, S., & AddemirA, O. (2017). Boron carbide particle size effects on thermal neutron attenuation behavior of boron carbidetitanium diboride composites. In : V. E. Borisenko, S. V. Gaponenko, V. S. Gurin, and C. H. Kam (Eds.), Physics, Chemistry and Application of Nanostructures (pp. 106- 109). World Scientific Publishing Co. Pte. Ltd. https://doi. org/10.1142/9789813224537_0025.
6. Martz, R. L. (2017). The MCNP6 Book On Unstructured Mesh Geometry: User's Guide for MCNP 6.2.1. U.S. Department of Energy, Office of Scientific and Technical Information. https://doi.org/10.2172/1467189.
7. Jaeger, T. (1965). Principles of Radiation Protection Engineering. McGraw-Hill Book Company. ISBN 1114788317.
8. Cierjacks, S, & Smith, A. B. (1988). Neutron sources. In S. Igarashi (Ed.), Nuclear data for science & technology. Saikon Publishing Co, Ltd.

9. Holmes, R. J. (1982). Gamma ray and neutron sources (Report No. AAEC/S-24). Australian Atomic Energy Commission Research Establishment. Retrieved from https://inis.iaea.org/search/search. aspx?orig_q=RN:14792880.
10. Bauer, G.R. (1993). Neutron sources. In A. Furrer (Ed.), Neutron Scattering (pp. 331-357). PSI-Proceedings No. 93-01, ISSN 1019-6447, Paul Scherrer Institute, Villigen.
11. Gaylan, Y., Bozkurt, A., & Avar, B., (2021). Investigating thermal and fast neutron shielding properties of B4C, B2O3, Sm2O3, and Gd2O3 doped polymer matrix composites using Monte Carlo simulations. Süleyman Demirel University Faculty of Arts and Science Journal of Science, 16(2), 490-499. https://doi.org/10.29233/ sdufeffd.933338.
12. Mughabghab, S. F. (2003). Thermal neutron captures cross sections resonance integrals and g-factors (Report No. INDC(NDS)-440). Retrieved from https://www.osti.gov/etdeweb/biblio/20332542
13. Auden, E. C., Quinn, H. M., Wender, S. A., O’Donnell, J. M., Lisowski, P. W., George, J. S., … & Black, J. D. (2020). Thermal neutron-induced single-event upsets in microcontrollers containing boron-10. IEEE Transactions on Nuclear Science, 67(1), 29-37. Doi 10.1109/TNS.2019.2951996.
14. Buyuk, B., Goncu, Y., Tugrul, A. B., & Ay, N. (2020). Swelling on neutron induced hexagonal boron nitride and hexagonal boron nitride-titanium diboride composites. Vacuum, 177, 109350. doi.org/10.1016/j. vacuum.2020.109350.
15. Werner, C. J., Bull, J. S., Solomon, C. J., Brown, F. B., McKinney, G. W., Rising, M. E., ... & Casswell, L. (2018). MCNP version 6.2 release notes (Report No. LA-UR-18-20808). Los Alamos National Laboratory (LANL). doi.org/10.2172/1419730.
16. Los Alamos National Laboratory. (2000). MCNPTM-A General Monte Carlo N-Particle Transport Code (Version 6). https://s3.cern.ch/inspire-prod-files-7/78c669e8d3bb59ccf6fb868a6061450c.
17. Buyuk, B., & Tugrul, A. B. (2014). Gamma and neutron attenuation behaviours of boron carbide-silicon carbide composites. Annals of Nuclear Energy, 71, 46-51. doi.org/10.1016/j.anucene.2014.03.026.
18. Orak, S., & Baysoy, D. Y. (2013). Neutron shielding properties of concrete with boron and boron containing mineral. Academic Platform-Journal of Engineering and Science, 1(1), 15-19. doi:10.5505/apjes.2013.99609.
19. Büyük, B., & Tuğrul, B. (2015). Investigation on the Behaviours of TiB2 Reinforced B4C-SiC Composites Against Co-60 Gamma Radioisotope Source. Pamukkale University Journal of Engineering Sciences, 21(1), 24-29. doi: 10.5505/pajes.2014.29052.
20. Topuz, A. I., & Reyhancan, I. A. (2018). 3D Surface Radiation Dosimetry of a Nuclear-Chicago NH3 Neutron Howitzer. arXiv preprint, arXiv:1806.02387. https://doi. org/10.48550/arXiv.1806.02387.
21. Topuz, A. I., & Reyhancan, I. A. (2018). Neutronic Analysis of a Nuclear-Chicago NH3 Neutron Howitzer. arXiv preprint, arXiv:1806.05255. https://doi. org/10.48550/arXiv.1806.05255.
22. Tekin, H. O., Kavaz, E., Papachristodoulou, A., Kamislioglu, M., Agar, O., Altunsoy Guclu, E. E., Kilicoglu, O., & Sayyed, M. I. (2019). Characterization of SiO2- PbO–CdO–Ga2O3 glasses for comprehensive nuclear shielding performance: Alpha, proton, gamma, neutron radiation. Ceramics International, 45(15), 19206-19222. https://dx.doi.org/10.1016/j.ceramint.2019.06.168.