Bir Türk Öğrencinin Dijital Okuryazarlık ve Uzaktan Eğitim Bağlamında Kanadada Eğitim Deneyimi%253A Bir Anlatı Araştırması

Bu çalışma Kanadada öğrenim gören bir Türk öğrencinin eğitimde dijital okuryazarlık ve uzaktan eğitim hakkındaki deneyimlerini çözümlemek amacıyla gerçekleştirilmiştir. Çalışma nitel araştırma yöntemlerinden anlatı araştırmasına göre hazırlanmış, anlatı araştırmasının yedi aşamasına uygun olarak çalışma yürütülmüştür. Görüşme yöntemiyle veriler toplanmış ve Riessmannın veri analiz sınıflandırmasına göre tematik analizle değerlendirilmiştir. Anlatının analiz edilmesiyle elde edilen temalarla dijitalleşmeyi eğitime entegre etmeyi amaçlayan en eski ülkelerden biri olan Kanadada, eğitimde teknolojinin kullanımının ne seviyede olduğu anlaşılmaya çalışılmıştır. Dijital okuryazarlığın lisede Yeni Medya kapsamında ana dil derslerinden biri olarak değerlendirildiği, üniversitede online laboratuvar derslerinde animasyonlarla etkileşimli dersler yapıldığı, derslerde araç gereç olarak bilgisayarın ve ilgili uygulamalarının temel araç gereç olarak kullanıldığı gibi sonuçlara ulaşılmıştır. Ayrıca Kanadada özellikle lise eğitiminde Eğitim 3.0 dönemi uygulamalarının aktif olarak kullanıldığı sonucuna varılmıştır

Contributions of each isotope in structural material on radiation damage in a hybrid reactor

In this study, the fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. In this study, salt-heavy metal mixtures consisting of 93–85% Li20Sn80 + 5% SFG-PuO2 and 2-10% UO2, 93–85% Li20Sn80 + 5% SFG-PuO2 and 2-10% NpO2, and 93–85% Li20Sn80 + 5% SFG-PuO2 and 2-10% UCO were used as fluids. In this study, the effect on the radiation damage of spent fuel-grade (SFG)-PuO2, UO2, NpO2 and UCO contents was investigated in the structural material of a designed fusion–fission hybrid reactor system. In the designed hybrid reactor system were investigated the effect on the radiation damage of the selected fluid according to each isotopes of structural material in the structural material for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library

Three-dimensional Monte Carlo calculation of some nuclear parameters

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20Sn80 + 1-5% RG-Pu, 99–95% Li20Sn80 + 1-5% RG-PuF4, and 99–95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0

The calculation of neutron flux using Monte Carlo method

In this study, a hybrid reactor system was designed by using 99–95% Li20Sn80 + 1-5% RG-Pu, 99–95% Li20Sn80 + 1-5% RG-PuF4, and 99–95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII.0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code

Three-dimensional Monte Carlo calculation of some nuclear parameters

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20Sn80 + 1-5% RG-Pu, 99–95% Li20Sn80 + 1-5% RG-PuF4, and 99–95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0

1-20 MeV Enerji Aralığında 230,232Th ve 236,238,239,240,241,242,244Pu Çekirdeklerinin Nötronlarla Oluşturulan Fisyon Tesir Kesitlerinin Hesaplanması

Özet: Bu çalışmada, 230,232Th ve 236,238,239,240,241,242,244Pu çekirdeklerinin fisyon tesir kesitleri 1-20 MeV gelme enerjili nötronlar için hesaplandı. Hesaplamalar, tesir kesiti hesaplamalarını farklı modellerle ele alan ALICE-ASH ve CEM95 bilgisayar programları ile yapıldı. CEM95 ile yapılan hesaplamalarda seviye yoğunluğu parametresinin (IFAM) tüm sistematikleri kullanıldı. ALICE-ASH ile yapılan hesaplamalarda RFRM (Rotation-Finite-Range Model) ve RLDM (Rotation-liquid-Drop Model) modelleri kullanıldı. Hesaplanan fisyon tesir kesitleri Deneysel Nükleer Reaksiyon Veri (Experimental Nuclear Reaction Data-EXFOR) kütüphanesinden elde edilen deneysel değerlerle karşılaştırıldı. Anahtar kelimeler: Seviye yoğunluğu parametresi, fisyon tesir kesiti, fisyon engeli The Calculation of the Neutron-Induced Fission Cross Sections of 230,232th and 236,238,239,240,241,242,244Pu Nuclei in the Energy Range 1-20 MeV Abstract: In this study, fission cross section of 230,232Th ve 236,238,239,240,241,242,244Pu nuclei have been carried out in the 1-20 MeV incident energy neutrons. ALICE-ASH and CEM95 computer programs was used for calculations. In CEM95 calculations, all sistematics of level density parameters (IFAM) was used. In ALICE-ASH calculations, RFRM (Rotation-Finite-Range Model) and RLDM (Rotation-liquid- Drop Model) was used. The calculated fission cross section was compared with the experimental data taken from the Experimental Nuclear Reaction Data (EXFOR). Key words: Level density parameter, fission cross section, fission barrie