SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data
نویسندگان
چکیده
منابع مشابه
A Data Model Validation Approach for Relational Database Design Courses
This paper presents an instructional method for validating a relational database design. Data model validation is often overlooked in course projects involving relational database design, in part because while most database texts stress the importance of validation, few provide an instructional method for performing validation. Validation is a critical step, especially for students. A flawed da...
متن کاملData Modeling and Relational Database Design in Archaeology
Data from archaeological excavation is suitable for computerization although they bring challenges typical of working in non-scientific subjective areas. We present some issues with regard data modeling in the specific field of archaeology.
متن کاملCharacterization of High Level Liquid Waste Generated from Reprocessing of Power Reactor Spent Fuel
High level Liquid Waste (HLW) is generated during the reprocessing of spent nuclear fuel which is used to recover uranium and plutonium. More than 99% of the radioactivity generated during the burning of nuclear fuel in the reactor is present in HLW. For the efficient management of HLW either by vitrification in the suitable borosilicate glass matrix, or partitioning and transmutation (P&T) of ...
متن کاملDesign and Implementation of a Reactor Physics Laboratory Simulation Software
The basic structure of a reactor physics laboratory environment simulation software, developed using object modeling technique (OMT), and based on the reactor point kinetic equation, is presented. Also, various capabilities of the simulator in teaching the fundamental concepts of reactor physics are discussed. In this virtual laboratory, student can perform seven different experiments, ...
متن کاملshielding and criticality safety analyses for spent fuel transportation cask in tehran research reactor
in this research, shielding and criticality safety calculations carried out for interim storage and transportation cask in the tehran research reactor. such dual purpose cask is being designed to the spent fuel elements of research reactors. the monte carlo mcnp5 code calculation was utilized for the criticality safety analysis and origen2.1code was used for shielding calculation. according to ...
متن کاملذخیره در منابع من
با ذخیره ی این منبع در منابع من، دسترسی به آن را برای استفاده های بعدی آسان تر کنید
ژورنال
عنوان ژورنال: EPJ Web of Conferences
سال: 2017
ISSN: 2100-014X
DOI: 10.1051/epjconf/201714606015