Criticality Verification for Open Pool storage of MTR Spent fuel Elements using MCNP5 Code

Abstract

Abstract:

This study aims to verifying the nuclear criticality criteria in spent fuel repository. The repository is an open pool type that is used temporary to store the spent fuel elements removed from the core of 22 MW power of open pool research reactor. The repository could receive up to 800 of spent fuel elements of material testing reactor (MTR) type. Arrangement configuration of spent fuel elements in the repository must verify the criticality criteria during the storage that mean that the multiplication factor should not exceed 0.85. Two models were developed in this study using MCNP5 code to simulate the spent fuel elements configuration including; the basket model, and the storage pool model (contains the maximum permissible capacity). The result shows that the multiplication factors would be less than the permissible limit for the criticality criteria of the storage.

Keywords: MCNP5 code; criticality; spent fuel storage; spent fuel; fuel burn-up; multiplication factor.

Cite this Article: Amr Abdelhady. Criticality Verification for Open Pool storage of MTR Spent fuel Elements using MCNP5 Code. Journal of Nuclear Engineering & Technology. 2020; 10(2): 26–30p.