Nükleer Enerji Mühendisliği Bölümü Tez Koleksiyonuhttp://hdl.handle.net/11655/4642019-12-11T11:50:56Z2019-12-11T11:50:56ZCobra-Tf Yazılımı ile Soğutucu Kaybı Kazasında Akışın Yeniden Sağlanması Fazının Modellenmesi ve Model Belirsizliklerinin İncelenmesiKaragöz , Refikhttp://hdl.handle.net/11655/119092019-11-08T10:57:15Z2019-01-01T00:00:00ZCobra-Tf Yazılımı ile Soğutucu Kaybı Kazasında Akışın Yeniden Sağlanması Fazının Modellenmesi ve Model Belirsizliklerinin İncelenmesi
Karagöz , Refik
Best estimate codes used in the safety analysis, licensing, research and development fields frequently are cost efficient, quick and reliable tools. For these codes which are widely used by the industry, universities and public intitutions, it is very important to validate the physical models against experimantal data which are obtained for wide range of conditions. Therefore, the modifications which improve solution methods, increase the accuracy in code predictions and performance are required by developing new models and adapting them to the codes.
In this dissertation, reflood phase of postulated Loss of Coolant Accident in Pressurized Water Reactors is analyzed and modifications were performed to analyze the maximum clad temperature behaviour, one of the most important safety criterion. After specifiying weaknesses of reflood models in COBRA-TF and underlying compenents causing them and applying corrective and improving models to the code it is aimed to obtain better results with respect to FLECHT-SEASET (Full Length Emergency Core Heat Transfer-System Effects and Separate Effects Tests) and RBHT (Rod Bundle Heat Transfer) experimental data.
In the study, it is observed that COBRA-TF calculates the Minimum Film Boiling Temperature (T_min) inconsistently when it’s compared with the experimantal data. Since T_min model affects the heat transfer behaviour completely, applying a more advanced T_min model, better code results and consistent fuel temperature behaviour were obtained. As the results of the analysis carried out after this modification, it is observed that the film entrainment in annular film flow and quench front entrainment generated on cooling fluid surface models existing in COBRA-TF have weaknesses on estimating some experimental conditions. Therefore, adding different models to available entrainment calculation had been applied to COBRA-TF code, additionally. The results of using new models, FLECHT SEASET and RBHT experimental data were compared, and significantly modified results were obtained in calculating quench front, fuel clad temperature and vapor temperature.
2019-01-01T00:00:00ZKullanılmış Nükleer Yakıtın Yatay Jeolojik Bertarafının Isıl AnalizleriÖzeşme, Gürelhttp://hdl.handle.net/11655/93142019-10-30T08:46:08Z2019-01-01T00:00:00ZKullanılmış Nükleer Yakıtın Yatay Jeolojik Bertarafının Isıl Analizleri
Özeşme, Gürel
Geological disposal is the most accepted method for permanent disposal of spent nuclear fuel and high-level waste. There are various geological disposal concepts under development in many countries and these concepts have differences mainly in the geometry and material of disposal canisters, geological formations of host rock and orientation (vertical and horizontal) of disposal canisters. Thermal behavior of disposal facility components is an important determinant of repository design and waste disposal density (the amount of radioactive waste that can be safely emplaced per unit area of the geological repository). Thermal load of the geological repository strictly depends spent fuel characteristics (amount, isotopic composition, heat generation), disposal canister model and the thermal features of the host rock. In this study, canisters loaded with spent fuel assemblies discharged from VVER – 1200 and ATMEA1 reactors and disposed horizontally in the granitic rock are modeled. Spent fuel characteristics are evaluated by using the MONTEBURNS 2.0 code. The ANSYS finite element code is utilized to generate a thermal model of horizontal repository and determine waste disposal densities through thermal analysis by taking into account the thermal constraints. Thermal analysis is repeated for disposal canisters loaded with a various number of spent fuel assemblies (4, 5, 6 assemblies) and with assemblies precooled for various periods (40, 50, 60 and 70 years) in order to assess the impact to waste disposal densities. Most favorable canister model with regard to disposal density is determined.
2019-01-01T00:00:00ZDijital Çakışma Sayım Yönteminin
Geliştirilmesi ve 4pi-Beta-Gama Sayım
Sisteminde Karmaşık Bozunma Şemalı Belirli
Radyoizotoplara UygulanmasıŞahin, Namık Kemalhttp://hdl.handle.net/11655/92862019-11-06T14:25:34Z2019-01-01T00:00:00ZDijital Çakışma Sayım Yönteminin
Geliştirilmesi ve 4pi-Beta-Gama Sayım
Sisteminde Karmaşık Bozunma Şemalı Belirli
Radyoizotoplara Uygulanması
Şahin, Namık Kemal
The methods of data acquiring and processing in the 4πβ-γ coincidence counting systems for radionuclide standardization have been employed by using analogue electronic modules for decades. Since the supply and maintenance of these modules are costly, the need for the digital data acquiring and processing systems which are faster, more flexible and more reliable have been arised in parallel to the new advancements in radiation measurement systems. Achieving the signal parameters from the detector of a 4πβ-γ coincidence counting system by using a digital card and analyzing these signal parameters offline by using computer programs became an easy and innovative alternative method for the conventional analogue systems.
In this thesis, a unique software has been developed to be used as data analysis method of the 4πβ-γ coincidence counting system setup in the Radionuclide Metrology Laboratories of Turkish Atomic Energy Authority (TAEK) for absolute activity standardization. This real-time coincidence counting software includes coincidence and anticoincidence methods. Activity calculations have been performed using unique algorithms of these methods. The validation of the software were done by determining the activity values of the standart radioactive solutions including 60Co, 133Ba, 152Eu and 166mHo radionuclides separately. Some amount of radioactive solutions were used as the source and then measured in the 4πβ-γ coincidence counting system. The data achieved from the counting system were analyzed in the software. The activity values of 60Co and 133Ba, which have relatively simple decay schemes, and of 152Eu ve 166mHo, which have relatively much more complicated decay schemes, have been determined by both coincidence and anticoincidence methods in a good agreement. The relative bias values between the two methods are found as 0.3% for 60Co, 0.18% for 133Ba, 0.1% for 152Eu and 0.08% for 166mHo.
In addition, the Monte Carlo simulation of the 4πβ-γ coincidence counting system was performed, and the response functions of NaI detector for 60Co, 133Ba, 152Eu and 166mHo are determined as the detector efficiencies. The efficiency values and the experimental count rates of 60Co, 133Ba, 152Eu and 166mHo are used to calculate the activities of these radionuclides.
When the activity values of these radionuclides determined by the coincidence method, anticoincidence method and Monte Carlo method were compared, it is found that the results agree well with each other within at most 2.5% relative bias. The developed digital coincidence software is now ready to be applied to the measurements of all radionuclides having simple or complicated decay schemes in the 4πβ-γ coincidence counting system. This primary radioactivity measurement system, which is used in world’s leading radioactivity measurement laboratories, has become availbale in TAEK Radionuclide Metrology Laboratories as a result of this thesis study.
2019-01-01T00:00:00ZDüzleştirici Filtresiz Tedavi Sistemleri İçin
Kalem Huzme Kerneli GeliştirilmesiErtürk , Mehmet Ertuğrulhttp://hdl.handle.net/11655/65112019-05-29T11:42:44Z2019-01-18T00:00:00ZDüzleştirici Filtresiz Tedavi Sistemleri İçin
Kalem Huzme Kerneli Geliştirilmesi
Ertürk , Mehmet Ertuğrul
Pencil beam algorithm is a fast and accurate method for dose computation at Radiotherapy. This algorithm was developed for filtered x-ray energies. The aim of this study was developing pencil beam kernels by utilizing x-ray characteristics of flattening filter free energy, and build a calculation model working at inhomogeneous medium. It was also purposed to make dose calculation at inhomogeneous medium kernel of each medium. Dose distribution data of flattening filter free x-ray at different mediums were gotten with water phantom measurements and Monte Carlo modeling. Phase-space data provided by Varian were used for Monte Carlo modeling. Optimization method was used to obtain kernel parameters. Gama analysis method was used as penalty function at optimization. Dose distribution was obtained with two dimensional discrete convolutions by using these kernel parameters. Gama values of each measurement points were calculated on dose profile by using calculated dose distribution. Accuracy of dose distributions, which were calculated with evaluated kernels, was controlled with three different methods by using gamma analysis technique. At the first, measured profiles used for kernel evaluation were compared with profiles that were calculated with evaluated kernels. Criteria’s of gamma analysis were 1 mm distance to agreement, 1% dose difference and the threshold was 10%. Multileaf collimator shaped fields were used at the second stage of calculation accuracy control. Validity of evaluated kernel was controlled with shaped field with multileaf collimators (MLC) at the second phase. Dose maps of irradiated MLC shaped fields at different depths were measured with two dimensional detector. Criteria’s of gamma analysis were 3 mm 3% dose difference criteria and with 10% threshold. Passing ratio of each comparison was found greater than 95%. Acquired dose maps by irradiation of intensity modulated radiation therapy (IMRT) field were used at the third phase of the comparison. Criteria were used at MLC shaped fields were also used in this phase, and a passing ratio greater than 99% were achieved. At the second stage, capability of dose to medium calculation ability of pencil beam algorithm was investigated. Thus, dose distributions were calculated with the kernels evaluated from different homogeneous mediums. In order to calculate dose distributions at heterogeneous medium, two different kernels were prepared for calculate dose distribution at depth and lateral directions. Dose distributions evaluated at heterogeneous mediums pencil beam calculation were compared with dose distributions evaluated with Monte Carlo simulation by using gamma analysis method. Criteria used for comparison of lateral dose distribution were 2 mm 2% with 10% threshold. A passing ratio greater than 95% was achieved. At the second part of the second stage a percentage depth dose calculation model with forward scatter and backward scatter kernel was suggested. Percentage depth dose curves evaluated with pencil beam algorithm and Monte Carlo methods for different phantoms. These curves were compared with gamma analysis method. Passing ratio’s greater than 95% were achieved by the criteria of 3 mm 3%. As a result, the kernels can be redefined by considering the beam characteristics of the flattening filter free x-ray energies. With this redefinition, the number of parameters in the kernel can be reduced. Thus, the calculation time can be shortened. In addition, dose to medium approach can be used in the pencil beam algorithm.
Mehmet Ertuğrul Ertürk Doktora Tezi
2019-01-18T00:00:00Z