Türkiye`de üretilen granitlerin gama ve nötronları zayıflatma özelliklerinin incelenmesi

Abstract

Bu Yüksek Lisans tezinde Türkiye'de üretilen granitlerin foton zayıflatma katsayıları ve toplam makroskopik tesir kesitleri deneysel ve teorik olarak incelenmiştir. Günümüzde nükleer teknolojisinin kullanım alanlarınınn artmasıyla birlikte radyasyondan korunmanın önemi bir kat daha önem kazanmıştır. Zırhlama radyasyondan korunmanın en önemli metotlarından biridir. Bu durumda yapı malzemelerinin radyasyon zayıflatma özelliklerinin araştırılması önem arz etmektedir. Bu çalışmada yapılarda yalıtım ve dekoratif özelliklerinden dolayı yaygın olarak kullanılan granitlerin gama ve nötron radyasyonlarını zayıflatma özellikleri incelenmiştir.Çalışmamızda yerli üretim olan 8 farklı granit numune incelenmiştir. Bunlar, Giresun Vizon, Aksaray Pembe, Bergama Gri, Çanakkale Gri, Kozak, Aksaray Yaylak, Hisar Yaylak ve Balaban Yeşil granit numuneleridir. Çalışmanın deneysel bölümünde, gama ve nötron geçirgenlik tekniği prensibine uygun olarak, dar demet geometrisinde çalışılmıştır. Gama ölçümlerinde radyasyon kaynağı olarak 137Cs ve 60Co radyoizotoplarından yararlanılmıştır. Böylece 662 keV, 1173 keV ve 1332 keV enerji değerlerinde çalışılmıştır. Gama ölçümleri, NaI (Tl) sintilayon dedektörü ve çok kanallı analizörden oluşan dijital gama spektrometre sisteminde gerçekleştirilmiştir. Çalışmamızda nötron ölçümleri için İstanbul Teknik Üniversitesi, Enerji Enstitüsü, TRIGA Mark-II Reaktör holünde bulunan Pu-Be nötron kaynağı kullanılmıştır. Nötron ölçümleri, farklı granit kalınlıkları için Polimaster marka dedektörle alınmıştır.Deney sonuçlarını sınamak amacıyla granit blokların gama radyasyonu için kütle zayıflatma katsayıları WINXCOM bilgisayar programı kullanılarak hesaplanmıştır. Nötronların teorik hesabı Nötron Araştırma Merkezi'nin internet ortamında yayınladığı NCNR uygulaması ile gerçekleştirilmiştir. Nötron ve gama radyasyonu teorik hesapları için gerekli olan kimyasal analiz işlemi XRF tekniği kullanılarak gerçekleştirilmiştir. Deneysel ve hesaplamalı olarak elde edilen sonuçlar mukayeseli olarak değerlendirilmiş ve uyum içinde oldukları görülmüştür.

Radiation protection is an essential topic in human life, as the radiation become to be used in a variety of fields with the development of technology. Besides the applications of radiation benefit in many areas such as medicine, industry, agriculture etc, it damages cells, living tissues which should be protected. There are three major methods for radiation protection namely shielding, time and distance. The first one is the largely used method especially for critical and crucial buildings such as medicine or nuclear facilities. In addition, in areas where people are likely to encounter ionizing radiation, it is often necessary to supply shielding to decrease the exposure. Lead and tungsten which are heavy elements are ideal material for radiation shielding. In contradiction they cannot be used directly in building constructions because of durability and economic problems. In this study different granite samples, which are produced in Turkey were investigated. The granites were gathered from local dealers. These are polished granite tiles ready to use in public and commercial buildings. The shielding properties of these granites were studied for gamma ray and neutrons. In radiation physics, measuring the probability of all possible interactions between photon and medium is essential. Linear attenuation coefficient is the basic way of explaining the interactions between the medium and photon. The linear attenuation coefficient magnitude depends on the atomic number, the density of shielding materials and the incident photon energy. Composite materials may be preferred because of their chemical resistance, physical durability, portability and so on. Linear attenuation coefficient of a material for a given specific gamma energy can be determined both experimentally with using narrow beam geometry and theoretically by using WinXCOM or XCOM. In literature various natural rocks (building construction materials) were studied by many scientists in terms of their gamma shielding properties. In this study different granite samples such as Giresun Vizon, Aksaray Pink, Bergama Grey, Çanakkale Grey, Kozak, Aksaray Yaylak, Hisar Yaylak and Balaban Green, which are produced in Turkey were investigated. These granites which are polished tiles are commonly used in public and commercial buildings. The granite tiles measured 10 cm x 10 cm and their masses varied from 0.44 to 0.64 kg. The thickness of granite tiles are between 1,67 cm and 2,31 cm.Theoretical calculations for photon attenuation coefficients require the chemical compositions of granite samples. In briefly XRF is an analytical method to determine the chemical composition of all kinds of materials. X-rays produced by a source which is generally X-ray tube and irradiate the sample. The elements in the sample will emit fluorescent X-ray radiation with discrete energies that are characteristic values for elements. The size of grinded granite samples which were used in the XRF measurements, are smaller than 2 mm. For XRF measurements, homogeneous mixture of grinded granites were packed in the small zip lock bags. XRF measurements were accomplished at chemical analysis laboratory of Turkish Atomic Energy Authority (CNAEM). The brand of the system for XRF measurements is Olympus and its model is Innov-X system. The measurements were accomplished in soil analysis mode. This mode benefits for heavy elements in light materials. For minimizing the measurements errors, the measurements were repeated for 3 times and the average values were taken into account. Penetrating gamma rays through material is the basic of gamma transmission technique. The detector and the gamma source are placed at the opposite sides of material on the same axis. Gamma ray intensity comes from the source counted by the detector. The distance between the source and detector is 100 mm and the holes of collimators are 7 mm. Granite samples placed between the detector and the source at same distances. Cs-137 and Co-60 gamma radioisotopes were used in the experiments where their half lives are 30.1 and 5.23 years, respectively. Cs-137 radioisotope has single energy peak at 0,662 MeV, Co-60 gamma source emits two major energy peaks at 1.17 and 1.33 MeV. NaI(Tl) scintillation detector and multichannel analyzer were used in the experiments. Pu-Be neutron source was used for calculating total macroscopic cross section. Pu-Be sources emit fast neutrons, where neutrons are produced mainly through the 9Be(,n)12C reaction with a relatively small contribution from the self-multiplication effect due to the neutron induced fission on Pu isotopes and (n,2n) reactions on 9Be and the other nuclides present in the source construction materials. In this study the neutron transmision technique was used to investigate total macroscopic cross sections. The experiment system was established in the hall of ITU TRIGA Mark-II Training and Research Reactor. The experiment system included neutron source (howitzer) and the detector. The neutron howitzer was obtained from Nuclear Chicago Corporation. The activity of Pu-Be neutron source is 5 Ci and the neutron flux is almost 106 n/cm2 s. Average energy of Pu-Be neutron source is about 4.5 MeV. Howitzer includes the paraffin wax for slow down the fast neutrons. The reason behind this is to increase the probability of interactions. In the experiments measurements were taken three times for minimizing the errors. The calibration of multi channel analyzer was done with energy peaks of Co-60 ve Cs-137. The sofware Maestro 32 gives the results as net area. Net area means that the background counts are not included. In addition the software supplies the error values of measurements. All measurements were accoplished with an error which were less than %5. Measurements durations were determined to keep the errors less than %5. Measurement durations were determined as one hour for Co-60 energy peaks, half an hour for Cs-137 energy peaks for an error less than %5. Time intervals were 10 s for measurements of neutron flux. Total durations were 60 s. According to the technical documents of the detector, the distance between the granite and the detector was smaller than 10 cm. The mass attenuation coefficients for gamma rays were achieved theoretically with WinXCOM. NIST (XCOM) database supplies the mass absoption coefficient values for wide range of elements and composite materials. XCOM is a 16 byte dos application on the other hand WinXCOM is 32 byte windows platform. WinXCOM is a software that can supply total or partial mass attenuation coefficients between the energies 1 keV to 100 GeV for any elements, compounds and mixtures. The total macroscopic cross sections were calculated with an application which is called NCNR. NIST Center for Neutron Research provides an application which used for theoritical calculation of neutron macroscopic cross sections on the internet enviroment. This application can be applied to elements and mixtures. In this application resonance absorptions were not included. This application requires 3 basic data. These are chemical compositions and densities of material and the wavelength of the flux of neutron source. According to the results if incident photon energy increases the attenuation coefficients decreases. The gamma ray mass attenuation coefficients and neutron macroscopic cross sections were calculated computationally by using WinXCOM and NCNR programs respectively. According to the experimental results lineer attenuation coefficients vary between 0.173 cm-1 (Aksaray Yaylak) and 0.223 cm-1 (Giresun Vizon) for 0.662 MeV, 0.133 cm-1 (Çanakkale Gri) and 0.168 cm-1 (Giresun Vizon) for 1.17 MeV, 0.129 cm-1 (Çanakkale Gri) and 0.158 cm-1 (Bergama Grey) for 1.33 MeV. At different energy levels, different granites have the highest value of attenuation coefficient, that is, Aksaray Pink is for 1.33 MeV, Canakkale Grey is for 0.662 MeV and Giresun Vizon is for 1.17 MeV. Computational and experimental results show a good agreement.In neutron measurements 3 different thickness values were used in measurements to calculate relative counts. Relative counts vs thickness was plotted by using Origin 8 software. The total macroscopic sross sections vary between 0.041 cm-1 (Bergama Grey) and 0.056 cm-1 (Giresun Vizon). According to the results, computational and experimental consequences indicate a good correlation.