Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code

• Published in: Nuclear engineering and design 2014-07, Vol.273, p.576-583
• Main Authors: Salomé, J.A.D., Guerra, B.T., Pereira, C., de Menezes, M.Â.B.C., da Silva, C.A.M., Dalle, H.M.
• Format: Article
• Language: English
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2014.03.020

•The TRIGA IPR-R1 was modelled using MCNP.•The thermal neutron flux through the samples in eleven irradiation channels was obtained.•The simulated results were compared to experimental values.•The relative error, the relative trend, the z-score test and uncertainty were analysed. The TRIGA IPR-R1 was modelled using MCNP. The model consists of a cylinder filled with water, fuel elements, radial reflectors, central tube, control rods and neutron source. Around the core is placed the Rotary Specimen Rack (RSR) with adequate groove to insert the samples to irradiation. The values of the thermal neutron flux through the samples in eleven irradiation channels were simulated and compared to the experimental results to validate the model. After that, the values of the thermal neutron flux, in the same channels, were simulated on two horizontal planes at different heights and compared to validate the model. These channels were characterized as representative channels of the neutron flux distribution in the RSR. To evaluate the results, the relative errors, the relative trend, the z-score test and the relevance to a confidence interval of 95% were analysed. Good agreement has been obtained for the most channels when compared with the experimental results.