Abstract
The JSI TRIGA reactor has several irradiation facilities with well characterized neutron fields. The characterization was performed by measurements and by utilizing Monte Carlo particle transport computational methods. Because of this, JSI TRIGA has become a reference center for neutron irradiation of detectors for ATLAS experiment (CERN). Thorough γ characterization of the reactor is however yet to be performed. Current Monte Carlo particle transport code only account for the prompt generation of neutron induced γ rays, which have been characterized, but are neglecting the time dependent delayed part, which may in some cases amount to more then 30% of total γ flux in an operation reactor, and is the only source of γ-rays after reactor shutdown. Several common approaches of modeling delayed -rays , namely D1S and R2S exist.In this paper an in-house developed R2S method code is described, coupling a Monte Carlo particle transport code MCNP6 and neutron activation code FISPACT-II, with intermediate steps performed by custom Python scripts.An example of its capabilities is presented in terms of evaluation of utilization of JSI TRIGA nuclear fuel as a viable γ-ray source. In the model, fresh nuclear fuel is considered and a silicon pipe sample is modeled in. Fuel activities, dose and kerma rates on the sample, as well as emitted γ-ray spectra and isotopic contribution to the contact dose are calculated and presented.
Highlights
The Jožef Stefan Institute (IJS) TRIGA reactor is a 250 kW pool type reactor featuring numerous irradiation facilities with different characteristics in terms of size and neutron field properties [1]
The code is written in Python 2.7 (Python) [11] and couples MCNP [12] particle transport capabilities with FISPACT-II [13] neutron activation and transmutation analysis computer code
An R2S method script was developed for calculation of delayed γ-rays to improve photon characterization of irradiation facilities
Summary
This, along with initial material composition of the voxel is input into the activation analysis computer code, generating isotope inventory as well as γ-ray spectra respectively The latter is used as input into particle transport computer code again, so each voxel is represented in terms of a γ source respectively, calculating delayed photon field flux, spectrum and dose. Such approach allows user to calculate for example dose fields around activated or irradiated samples or in our case around irradiated nuclear fuel. The results are presented in terms of fuel element activities, H∗10 and kerma on the sample, as well as fuel element delayed γ spectra and isotopic contribution to contact dose vs. cooling time after activation completion respectively
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