Calibration of neutron detectors on the Joint European Torus.

The present paper describes the findings of the calibration of the neutron yield monitors on the Joint European Torus (JET) performed in 2013 using a 252Cf source deployed inside the torus by the remote handling system, with particular regard to the calibration of fission chambers which provide the time resolved neutron yield from JET plasmas. The experimental data obtained in toroidal, radial, and vertical scans are presented. These data are first analysed following an analytical approach adopted in the previous neutron calibrations at JET. In this way, a calibration function for the volumetric plasma source is derived which allows us to understand the importance of the different plasma regions and of different spatial profiles of neutron emissivity on fission chamber response. Neutronics analyses have also been performed to calculate the correction factors needed to derive the plasma calibration factors taking into account the different energy spectrum and angular emission distribution of the calibrating (point) 252Cf source, the discrete positions compared to the plasma volumetric source, and the calibration circumstances. All correction factors are presented and discussed. We discuss also the lessons learnt which are the basis for the on-going 14 MeV neutron calibration at JET and for ITER.

Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×103Svh-1 to 6×106Svh-1, silicon dose equivalent from 6×102Gy/hsi to 3×105Gy/hsi, and neutron air kerma from 4.3×103Gyh-1 to 2×105Gyh-1. Ratio of fast (1MeV

Spectrum average cross section measurement of 183W (n, p)183Ta and 184W (n, p)184Ta reaction cross section in 252Cf(sf) neutron field.

Neutron induced nuclear reactions are of prime importance for both fusion and fission nuclear reactor technology. Present work describes the first time measurement of spectrum average cross section of nuclear reactions 183W(n,p)183Ta and 184W(n,p)184Ta using 252Cf spontaneous fission neutron source. Standard neutron activation analysis (NAA) technique was used. The neutron spectra were calculated using Monte Carlo N Particle Code (MCNP). The effects of self-shielding and back scattering were taken into account by optimizing the detector modeling. These effects along with efficiency of detector were corrected for volume sample in the actual source-detector geometry. The measured data were compared with the previously measured data available in Exchange Format (EXFOR) data base and evaluated data using EMPIRE - 3.2.2.

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations.

Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization.

In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization.