Design of Portable Shield for Neutron Sources Using McNp Computer Code

Arab Journal of Nuclear Sciences and Applications
doi 10.21608/ajnsa.2018.2256.1021
Full Text
Abstract

Available in full text

Date
Authors
Publisher

Egypts Presidential Specialized Council for Education and Scientific Research

Related search

Gamma Shield Design for Primary Coolant Sources Using the Ibm Type 650 Computer

1956English

Portable Neutron Detector Using Ce:LiCaAlF6 Scintillator

Sensors and Materials
Materials ScienceInstrumentation
2015English

Advanced Neutron Sources: Plant Design Requirements

1990English

Inspect: An Interactive Computer Code for Unfolding Neutron Spectra From Activation Measurements.

1972English

The Experience of Using Portable Russian Neutron Generator for Gamma-Neutron Therapy of Domestic Animals With Malignant Tumors

Radiation and Risk
Nuclear MedicineRadiologyNuclearPublic HealthRadiationHigh Energy PhysicsEnvironmentalImagingOccupational Health
2018English

Application of Dose Evaluation of the McNp Code for the Spent Fuel Transport Cask

Progress in Nuclear Science and Technology
2011English

Approach for Calculating Population Doses Using the CIDER Computer Code

1993English

Spallation Neutron Sources

2014English

Pulsed Neutron Sources for Neutron Crystallography: New and Future Capabilities

Acta Crystallographica Section A Foundations of Crystallography
2011English

Amanote Research

Note-taking for researchers

Follow Amanote

© 2024 Amaplex Software S.P.R.L. All rights reserved.

Privacy PolicyRefund Policy