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

Year 2023, Özel Sayı, 40 - 45, 30.09.2023

Bülent Büyük Mehmet Murat Yaşar Nuri Yorulmaz Miraç Kamışlıoğlu

https://doi.org/10.30728/boron.1267320

Abstract

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.

Keywords

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

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

Abstract

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.

Keywords

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

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