Akademik Veri Yönetim Sistemi
1st International Conference on Radiations and Applications (ICRA), Algiers, Cezayir, 20 - 23 Kasım 2017, cilt.1994
Neutron moderation for nuclear applications has an undeniable importance. To be able to moderate neutrons, one of the important step is to choose the convenient material. The material's response for different type of nuclear reactions, define its possibility and availability for using as a neutron moderator material. The cross-section data of a reaction, which could be expressed as the probability of a reactions occurrence, could provide many benefits in the cases of the experimental difficulties or unsuitable conditions. For such cases, theoretical calculations obtained via verified methods are highly acceptable. In this study, Y-89, which is a neutron moderator material used in nuclear reactors, has been investigated and neutron production cross-section calculations for some gamma and proton induced reactions on Y-89 have been completed in the energy interval of 6-772 MeV with two most known and accepted calculation codes, TALYS 1.8 and EMPIRE 3.2. The Two Component Exciton and Exciton models have been used within the codes, respectively. Obtained results by using both code's mentioned models and exist experimental data reached from EXFOR database have been compared with each other.