İyonizan radyasyon zırhlaması için polianilin içeren kompozitlerin kullanımının incelenmesi
Investigation of the use of polyaniline containing composites for ionizing radiation shielding
Yazar: DENİZ AGEHAN YALÇINKAYA
Danışman: PROF. DR. SEVİLAY HACIYAKUPOĞLU
İstanbul Teknik Üniversitesi / Enerji Enstitüsü / Nükleer Araştırmalar Ana Bilim Dalı / Radyasyon Bilim ve Teknoloji Bilim Dalı
Derece: Yüksek Lisans, 2022
Konu: Nükleer Mühendislik = Nuclear Engineering
Dizin: Gama radyasyonu = Gamma radiation ; Kompozit polimerler = Composite polymers ; Nötron zırhlama = Neutron shielding ; Nükleer radyasyon = Nuclear radiation

Özet:
Günümüzde doğal kaynakların yanı sıra medikal tıpta ve sanayide insan yapımı radyasyona maruz kalınmakta. Herhangi bir duyu organıyla hissedilmeyen radyasyonun, belirli prensiplere ve doz sınırına uyulmaz ise insan sağlığı için zararlı etkileri gözlemlenir. Radyasyonun zararlı etkilerinden korunmak için Uluslarası Atom Enerji Ajansı (IAEA), Uluslararası Radyasyondan Korunma Komisyonu 'International Commission on Radiological Protection'(ICRP) gibi kuruluşlar radyasyon çalışanları ve halk için acil durumlarda ve yıllık olarak alınabilecek doz sınırları belirlemiştir. Bu doz sınırlarına uymak için radyasyonun zırhlamansı hayati önem taşımaktadır.
Geant4 parçacık fiziği, nükleer fizik, medikal fizik, hızlandırıcı tasarımı ve uzay mühendisliği çalışmaları için fizikçiler ve yazılım mühendisleri tarafından geliştirilen C++ diline uyarlanmış, Monte Carlo yöntemini kullanan bir araç setidir. Bu araç seti farklı fizik süreçlerinden yararlanıp, parçacıkların malzemeden geçişlerini simüle edilmesine olanak sağlar.
Bu yüksek lisans tez kapsamında; polianilin, formik asit, borik asit ve bor nitrür'ün farklı oranlarında oluşturulan kompozitlerinin gama ve nötron radyasyonunu zırhlama özellikleri, gamalar için deneysel ve Geant4 simülasyon uygulaması, nötronlar için ise teorik hesaplama ve Geant4 simülasyon uygulaması yardımıyla incelenmiştir.
Kimyasal formülü (C6H5NH2)n olan polianilinin ve kimyasal formülü CH₂O₂ olan formik asitin %50 oranlarla karıştırılmasıyla elde edilen örneğe  %0, %1, %2 ve %5 oranlarında kimyasal formülü BN olan bor nitrür eklenen dört örneğin yanı sıra, polianilin ve borik asitin %50 oranlarla karıştırılan bir örnek daha incelenmiştir.
Gama zırhlama incelemeleri için enerjisi 662 keV olan 137Cs kaynağı ve enerjisi 1253 keV 60Co kaynağı kullanılmıştır. Nötron tesir kesiti hesaplamaları ise 0,025 eV, 1 keV ve 10 MeV enerji değerleri için yapılmıştır.
Geant4 simülasyonunda gama kaynağı ile yapılan uygulamalarda, örneklerin lineer zayıflatma katsayısını tespit etmek için 0,5 cm, 1 cm 1,5 cm ve 2 cm kalınlıklarında da simülasyon çalıştırılıp, etkileşim yapmadan geçen gama fotonlarının sayısı tespit edilmiştir. Elde edilen sonuçlardan kalınlığa karşı, I/I0 oranının grafiği çizilip grafik denkleminden lineer zayıflatma katsayıları bulunmuştur. 
Deneysel uygulamalarda dedektör olarak NaI sintilasyon dedektörü kullanılmıştır. Her örnek için deneyler üç kez tekrarlanıp ortalamaları alınarak hesap yapılmıştır. Deneyler sonucunda örneklerin lineer zayıflatma katsayıları, kütlesel zayıflatma katsayısı, yüzde zayıflatma oranlar ve yarı değer kalınlıkları tespit edilmiştir. Elde edilen sonuçlar simülasyon sonuçlarıyla karşılaştırılıp, yüzde farklılık oranları hesaplanmıştır.
Geant4 nötron simülasyonlarında, nötronların malzeme içinde aldığı ortalama serbest yolun bulunması ve malzemeyi geçen nötronların enerji spekturumunda ilgili enerjideki parçacık sayısının bulunması olmak üzere iki farklı yöntemle nötron tesir kesiti hesaplanmıştır. Sonuçlar teorik hesaplamalarla karşılaştırılmıştır. 
Geant4 ile simülasyonları ve teorik hesaplamalar sonucunda bor elementinin yüzdesinin fazla olduğu örneğin termal enerjili nötronlar için (0,025 eV) zırhlama özelliklerinin oldukça iyi olduğu gözlemlenmiştir. 0.025 eV enerji için S4 örneğinin 0,11 cm kalınlığının diğer örneklerin ise yaklaşık 0,4 cm kalınlığının, 1 keV için tüm örneklerin yaklaşık 0,6 cm kalınlığının, 10 MeV için ise tüm örneklerin yaklaşık 6 cm kalınlığının radyasyon şiddetini yarıya değerine indirmeye yeterli olduğu gözlemlenmiştir.
Gama radyasyonu incelemelerinde Geant 4 ile deneysel çalışmaların büyük oranda uyumlu olduğu gözlenmiştir. Oluşan farklılıkların sebebi gerçekte karışımın simülasyonda hesaplandığı kadar homojen olmaması, bileşiklerde safsızlık bulunması, kimyasal bileşenlerin öngörülmeyen etkileşimlere girebilmesi, radyasyon kaynağının şiddetinde salınımlar olması ile açıklanabilir. 60Co ölçümlerinde farkın artmasının sebebi ise orta enerjilerde Compton etkileşiminden kaynaklanan saçılmaların olduğu düşünülebilir. Yukarıda belirtilen eksiklikler gelecekteki araştırmalarda ele alınarak genişletilebilir.

Abstract:
Nowadays, in addition to natural resources, man-made radiation is exposed in medical medicine and industry. If certain principles and dose limits are not followed, harmful effects of radiation are observed on human health. In order to protect against the harmful effects of radiation, organizations such as the International Atomic Energy Agency (IAEA), the International Commission on Radiological Protection (ICRP) have determined dose limits for radiation workers and the public in emergencies and annually. 
In radiation applications, shielding is important both to protect human health and to prevent the destruction of the devices in which the source is located. Since the penetration and properties of the radiation types are different from each other, the shielding properties also vary.
Geant4 is a Monte Carlo method adapted to the C++ language developed by physicists and software engineers. This toolkit makes use of different physics processes to simulate the passage of particles through material. It can be used in  nuclear physics, medical physics, accelerator design and aerospace engineering studies.
Within the scope of this master's thesis; The gamma and neutron radiation shielding properties of the composites formed in different ratios of polyaniline, formic acid, boric acid and boron nitride were investigated with the help of experimental and Geant4 simulation application for gammas, theoretical calculation and Geant4 simulation application for neutrons.
Four samples examineted were obtained by adding 1%, 2%, 5%, %0 boron nitride (BN) to the sample obtained by mixing 50% ratios of polyaniline (C6H5NH2)n and formic acid (CH₂O₂). The samples were named S1, S2, S3 and S5, respectively. Another sample, named S4, which was mixed with polyaniline and boric acid at 50% ratios, was also examined. The thickness of the samples was measured with a micrometer.
For gamma shielding studies,  137Cs source with an energy of 662 keV and  60Co source with an energy of 1253 keV were used. Neutron cross-section calculations were made for 0,025 eV, 1 keV and 10 MeV energy values. 
In Geant4 applications, for simulations made with gamma source, the source focused in one direction and adjusted to 662 keV and 1253 keV energies respectively. The sample is placed just in front of the collimator as in the experimental setup. In order to determine the linear attenuation coefficient of the samples, the simulation was run at 0.5 cm, 1.0 cm, 1.5 cm and 2.0 cm thicknesses of the samples, and the number of gamma photons that passed through the material was determined. The linear attenuation coefficient was found from the graph equation by plotting the I/I0 ratio against the thickness.
Linear attenuation coefficients from Geant4 simulations; at 662 keV, the maximum value was obtained from the S1 sample with a value of 0.112 cm-1 and the minimum value was from the S4 sample with 0.103 cm-1, at 1253 keV, the maximum value was obtained from the S1 sample with a value of 0.815 cm-1 and the minimum value was from the S4 sample with, 0.0743 cm-1. 
In experimental applications, the samples were placed in front of the source without leaving a distance. For 137Cs source  the detector was 60 cm away from the samples, for  60Co measurements  the distance between the samples and the detector was 12 cm. NaI scintillation detector was used as the detector. For each sample, the experiments were repeated three times and the averages were calculated. As a result of the experiments, linear attenuation coefficients, mass attenuation coefficient, percent attenuation ratios and half value layers of the samples were determined.The obtained results were compared with the simulation results and the percent errors were calculated.
Linear attenuation coefficients from experiment; at 662 keV, the maximum value was obtained from the S2 sample with a value of 0.116 cm-1 and the minimum value was from the S4 sample with 0.00991 cm-1, at 1253 keV, the maximum value was obtained from the S4 sample with a value of 0.127 cm-1 and the minimum value was from the S2 sample with, 0.00865 cm-1. 
In Geant4 neutron simulations, the neutron cross-section was calculated by two different methods;
In the first method, the samples are defined as a semi-infinite box (10 m) in order to determine the path taken by the neutrons in the material. Half value layers were calculated by using the macroscopic cross-sections given as a result of the simulation.
Macroscopic neutron cross-section from first method; at 0.025 eV the maximum value was obtained from the S4 sample with a value of 6.53 cm-1 and the minimum value was from the S5 sample with 1 .66 cm-1, at 1 keV, the maximum value was obtained from the S4 sample with a value of 1.34 cm-1 and the minimum value was from the S3 sample with, 1.22 cm-1. At 10 MeV, the maximum value was obtained from the S4 sample with a value of 0.121 cm-1 and the minimum value was from the S2 sample with, 0.114 cm-1.
In the second method used for neutron simulations is cross-section calculation by finding the number of neutrons with the same energy value as the source energy from the spectrum of neutrons passing through the material.
Macroscopic neutron cross-section from second method; at 0.025 eV the maximum value was obtained from the S4 sample with a value of 7.02 cm-1 and the minimum value was from the S5 sample with 1 .75 cm-1, at 1 keV, the maximum value was obtained from the S4 sample with a value of 1.36 cm-1 and the minimum value was from the S1 sample with, 1.21 cm-1. At 10 MeV, the maximum value was obtained from the S4 sample with a value of 0.123 cm-1 and the minimum value was from the S2 sample with, 0.113 cm-1.
In the calculation of the neutron cross-section, the macroscopic cross-section was calculated for each compound, and the product of the macroscopic cross-section and weight fractions of the compounds were added for the macroscopic cross-section calculations of the composites. The results were compared with Geant4 simulation results. At 0.025 eV the maximum value was obtained from the S4 sample with a value of 6.85 cm-1 and the minimum value was from the S5 sample with 1 .80 cm-1, at 1 keV, the maximum value was obtained from the S4 sample with a value of 1.34 cm-1 and the minimum value was from the S3 sample with, 1.12 cm-1. At 10 MeV, the maximum value was obtained from the S4 sample with a value of 0.123 cm-1 and the minimum value was equal for S1, S2, S3 sample with, 0.116 cm-1. 
In the gamma shielding efficiency studies, it was observed that the experimental results and the results obtained with Geant4 were largely in agreement. The reason for the differences can be explained by the fact that the materials defined in the Geant4 simulation application are mixed completely homogeneously, but in reality the mixture may not be mixed as homogeneously as calculated in the simulation, the presence of impurities in the structure of the compounds, the unpredictable interactions of the chemical compounds due to their nature, and the oscillations in the intensity of the source in each measurement due to the statistical nature of the radiation. Considering that the Compton interaction is dominant in the interaction of photons with the material at medium energy levels, the scattering of the photons with 1773 and 1333 keV energies of the 60Co source can be considered as the reason for the large differences in the experimental results.
As a result of the simulations and theoretical calculations with Geant4, the shielding properties of the sample with a high percentage of boron element for thermal energy neutrons (0.025 eV) were observed to be quite effective.
By making use of the results obtained from this research, new materials with high shielding efficiency can be developed by obtaining different materials for different types of ionizing radiation. The shortcomings mentioned above should be addressed in future research.