Please use this identifier to cite or link to this item: http://hdl.handle.net/11527/15800
Title: Nötron Aktivasyon Analizinde Kullanılmak Üzere Bir Database Ve Matlab Analiz Programı Tasarımı
Other Titles: Production Of A Matlab Code And A Database For The Calculation Of Mass And Its Uncertainity In Neutron Activation Analysis
Authors: Özben, Cenap Ş.
Turan Erdoğan, Gizem
10115321
Fizik Mühendisliği
Keywords: Nötron Aktivasyon Metodu
Matlab
Radyasyon Fiziği
Radyoaktivite
Radyasyon Detektörü
Gama Işınları
Gama Spektroskopisi
Akı Tayini.
Neutron Activation Method
Matlab
Radiation Physics
Radyoactivity
Radiation Detector
Gamma Rays
Gamma Spectroscopy
Determination Of Flux.
Issue Date: 2016
Publisher: Fen Bilimleri Enstitüsü
Institute of Science and Technology
Abstract: NÖTRON AKTİVASYON ANALİZİNDE KULLANILMAK ÜZERE BİR DATABASE VE MATLAB PROGRAMI TASARIMI Nötron aktivasyon analizi (NAA), elementlerin nicel ve nitel analizinde etkin olarak kullanılan bir yöntemdir. NAA metodunda öncelikle, numune nötron akısına maruz bırakılır. Numuneye ait kararlı elementlerin çekirdekleri (n,γ) nötron yakalama reaksiyonu sonucu radyoaktif hale gelir. Oluşan radyoaktif çekirdek çoğu kez α, β yayınladıktan sonra, söz konusu bozunmanın karakteristik özelliğini taşıyan γ- ışını yayınlayarak bozunur. Işınlama işleminden sonra numune, γ- ışınlarının ölçümü için uygun bir detektörün önüne (yada içine) konarak, fotopiklere ait enerjiler ve alanlar elde edilir. Gama spektroskopisinden elde edilen bu enerji ve sayım sonuçları kullanılarak hem numune içinde hangi izotopların olduğu hem de numune içindeki miktarları bulunabilir. NAA yöntemi arkeoloji, kriminoloji, adli tıp, çevre kirliliği ve tıp gibi bir çok alanda kullanılmaktadır. Antik dönemlere ait çanak ve çömlek parçalarının mineral ayrıntılı NAA analizi ile hangi parçaların ortak bir kökenden geldiği tespit edilebilmektedir. Bu tez çalışmasında, numunenin reaktör ya da nötron kaynağında ışınlandıktan sonra hem gama spektroskopisinden elde edilen fotopiklerin alanlarının hem de ışınlama, soğuma ve sayım sürelerinin girilerek pratik bir şekilde ışınlanan element miktarını veren bir program yazılmıştır. Programın kullandığı ana eleman olan, NAA yöntemine uygun elementlerden ve bu elementlere ait kütle numarası,ilgili elementin ışınlanan numune içindeki ağırlıkça yüzdesi, gama ışın şiddeti, enerji, termal ve epitermal etki tesir kesitleri ile bütün bu parametlere ait hata değerlerinden oluşan bir veri tabanı hazırlanmıştır. Tezin deneysel bölümünde ilk olarak, altın (Au) elementi ışınlanarak nötron akı tayini yapılmıştır. Au numunesi, kadmiyum kılıf içerisinde ve kadmiyum kılıfsız olmak üzere iki şekilde Am-Be izotopik nötron kaynağında ışınlanmıştır. Kadmiyum kılıfsız olarak ışınlanan Au numunesinin aktivasyonuna hem termal hem epitermal nötronlar katkıda bulunurken, kadmiyum kılıflı olarak ışınlanan Au numunesinin aktivasyonuna ise sadece epitermal nötronlardan katkı gelmiştir. Kadmiyum elementinin termal etki kesitinin çok yüksek olmasından kaynaklanan bu durum sayesinde yapılan hesaplamalar sonucu epitermal ve termal akı değerleri bulunmuştur. Çalışmanın amacı NAA analizinde kullanmak üzere bir Matlab programının yazılması ve test edilmesi olduğu için detektör olarak, pratik ve ekonomik kullanımı olan talyum katkılı NaI(Tl) sintilasyon detektörü tercih edilmiştir. Sintilasyon detektörünün enerji ve verim kalibrasyonları, aktiviteleri bilinen Cs^137 ve Co^60 standart referans kaynakları kullanılarak yapılmıştır. Programın testi için indiyum (In) elementinin ağırlığı hassas terazide ölçüldükten sonra, Am-Be izotopik nötron kaynağında 14 gün boyunca ışınlanmıştır. Işınlamadan sonra, NaI(Tl) detektöründe sayımlar alınmıştır. Elde edilen fotopiklerin analizi sonucunda elde edilen fotopik alanlarından Σ^gps değerleri hesaplanmıştır. In numunesine ait tüm parametreler; ışınlama, sayım ve soğutma süreleri ile enerji, alan, alanın hatası, enerji arama toleransı programda girilerek, program aranan enerji değerlerine en uygun olan elementi veri tabanından In olarak tayin edip kütle değerlerini ve hatalarını vermiştir. Bulunan kütle değerlerinin ortalaması alınarak, gerçekte ışınlanmadan önce ölçülen In miktarı le karşılaştırılıp programın yüzde kaç hata ile sonuca ulaşıldığı tayin edilmiştir.
PRODUCTION OF A MATLAB CODE AND A DATABASE FOR THE CALCULATION OF MASS AND ITS UNCERTAINITY IN NEUTRON ACTIVATION ANALYSIS Neutron activation analysis (NAA) has been widely used in archeological, industrial and medical applications since 1950’s. NAA practically starts with the sample preparation, continues with the irradiation of the sample and is completed with the analysis of qualitative and quantitative results. When a target nuclei is bombarded with neutrons, it becomes unstable and it emits characteristic gamma rays to become stable. Characteristic gamma rays are emitted mostly right after alpha and beta decays to get rid of the excess energy of unstable nuclei. Depending on the type of radioisotope, half life of the produced radioisotopes range between ms and years. For that reason, irradiation, decay and data collection times are very important. Additional to irradiation, decay and data collection times there are two more additional important parameters. These are thermal and epithermal neutron cross sections of the target material and flux of the neutron source. Cross section is the interaction probability of the neutrons with the target material. If irradiations take place in a reactor, power of the reactor and the irradiation position are the parameters change the neutron flux. The NAA can determine the weight of the objects down to parts per million (ppm). There are other sensitive methods like atomic absorption spectrometry. However, NAA is one of the most sensitive methods among the all. Additionaly, NAA can determine multi-elemental structure at a single shot. In this thesis work, instead of having relatively long and complicated calculation of mass and uncertainity of mass in NAA, we have decided to calculate the mass and its uncertainity using a computer code. The code uses a database of elements which was constructed by us. The elements taking place in the database were selected to be the ones that can be activated with neutron activation. The database contains mass number (A), isotopic abundance (a), absolute gamma transition rates (I_γ), gamma energies (E), thermal and ephitermal neutron cross sections (σ_(th )and σ_ep) and uncertainities of all these parameters. Only the strongest three gamma transitions were taken into database for each element if there are more than three gamma lines. The flowchart of the code is as follows: A sample is first irradiated by neutrons. The neutrons can source from a reactor or from an Am-Be neutron source. We have used an Am-Be neutron source. After a proper irradiation time (this is roughly estimated by the saturation factor in the mass formula), the sample is taken out and placed in front of a detector. Gamma spektrum is taken and analysed to obtain net fotopeak counts. Using these counts and related parameters mentioned, mass of the interested element is calculated. To give more details about our code; Our code is constructed from four main sections. They are ‘INPUT’, ‘MAIN’, ‘fluxes’ and ‘enerji_alan_tol’. A database of all elements in database.txt file is defined in the INPUT code. As we have mentioned before database.txt is the database of all elements where NAA is applicable. MAIN code is as expected is the main section of the code. Some input values such as irradiation time, cooling time and counting time and element database are loaded into the main code here. This section also includes a subroutine called ‘ekrana yazdir’. This is the code responsible for displaying the results into the screen in proper format. ‘fluxes’ is the input file for MAIN and it contains thermal and epitermal fluxes and their uncertainities. Results obtained from gamma spektroscopy (energies and gamma per second values) are hold in ‘enerji_alan_tol’ file. The last column is for tolerances of the energies used for element match in the database. First, neutron flux of the Am-Be neutron source was determined. A 22 carad gold sample was cut into two pieces and one of them was used for direct irradiation and other one is irradiated inside a 1 mm thick cadmium case. They both placed side by side inside the irradiation port of our J Sheppard model Am-Be neutron source for 6 days 21 hours and 3 minutes. After the irradiation, both samples were counted using NaI(Tl) scintillation detector, a multichannel analyser (MCA). Analysis were performed on these two samples for determining the net peak counts for with and without cadmium case. Electronic equipments used in this work were NHQ 102M ISEG model high voltage power supply, Canberra 2020 spectroscopy amplifier, Nucleus Multichannel Analyser and its related software PCA-II. Before the count rate were registered in the detector, energy and efficiency calibrations were performed for the detector and counting system. As known, gamma spectroscopy is performed using detectors with high resolution like HpGe. However since our main goal was to generate a database and code for calculating mass and its uncertainity, we prefered to use NaI scintillation detector. It is easy to work with it and it does not require liquid nitrogen to cool it down. Energy and efficiency calibration was performed using Co^60 ve Cs^137 standard sources. Energy dependent efficiency calibration function is produced between 662 keV and 1332 keV. Efficiency for gold energy (412 keV) is determined from the extrapolation of this function to 412 keV. This calculation would give a better result if we had a gamma line around 412 keV. However none of the standard sources we had have a single isolated peak lower than 662 keV energy. 59 keV of Am^241 was almost non visible due to its low energy. Efficiency at 412 keV was determined to be 7.193×〖10〗^(-2). Using this value, we calculated the Σ^gps values for two gold samples irradiated with and without cadmium cases using related equation given in the text. From this information we have determined thermal and epithermal neutron fluxes. As known cadmium has large thermal neutron cross section of 22000b. This means allmost all of the thermal neutrons are absorbed in the cadmium case. The gold inside the cadmium case face only epithermal neutrons where the bare one see both thermal and epithermal neutrons. To test the code and accuracy of the mass calculated, we have irradiated pure indium with known mass. An indium sample with known mass was irradiated for 14 days in the same port position where the fluxes were determined. A similar procedure was followed using our Matlab code to determine the mass and its uncertainity using three indium peak energies detected in the spectroscopy. The measured weight of In was 21.6 mg. The NAA determined value was 25.6 mg corresponding to 18% error. This value is acceptable for NAA calculations. Statistical uncertainity from the peak area is dominant. Therefore the calculated uncertainities are little larger than the statistical error from the peak area. This is exactly what we see from the results. Geometry correction does not take place in our code. The calibrations performed in this work used point charges. However, irradiated samples have non-point geometries. The systematic errors due to this was not taken into account. However we expect this is not larger than 10%.
Description: Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Fen Bilimleri Enstitüsü, 2016
Thesis (M.Sc.) -- İstanbul Technical University, Institute of Science and Technology, 2016
URI: http://hdl.handle.net/11527/15800
Appears in Collections:Fizik Mühendisliği Lisansüstü Programı - Yüksek Lisans

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