Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experiment...

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Bibliographic Details
Published in:Journal of radiological protection 2018-09, Vol.38 (3), p.881-891
Main Authors: Suárez, H Saurí, Becker, F, Klix, A, Pang, B, Döring, T
Format: Article
Language:English
Subjects:
flux spectra
Monte Carlo method
NE-213
neutron generator
shielding
spent nuclear fuel
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Summary:To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.
ISSN:0952-4746
1361-6498
DOI:10.1088/1361-6498/aac574