Benchmark evaluation of reactor critical parameters and neutron fluxes distributions at zero power for the TRIGA Mark II reactor of the University of Pavia using the Monte Carlo code MCNP

Introduction of Graphene/h-BN Metamaterial as Neutron Radiation Shielding by Implementing Monte Carlo Simulation

In this paper, graphene/h-BN metamaterial was investigated as a new neutron radiation shielding (NRS) material by Monte Carlo N-Particle X version (MCNPX) Transport Code. The graphene/h-BN metamaterial are capable of both thermal and fast neutron moderator and neutron absorber process. The constituent phases in graphene/h-BN metamaterial are chosen to be hexagonal boron nitride (h-BN) and graphene. The introduced target was irradiated by an Am-Be neutron source with an energy spectrum of 100 keV to 15 MeV in a Monte Carlo simulation input file. The resulting current transmission rate (CTR) was investigated by the MCNPX code. Due to concrete's widespread use as a radiation shielding material, the results of this design were also compared with concrete targets. The results show a significant increase in NRS compared to concrete. Therefore, metamaterial with constituent phase's graphene/h-BN can be a suitable alternative to concrete for NRS.

Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor

Democratic Republic of Congo (DRC) has a TRIGA mark II research reactor called TRICO II, its design power is 1.00 MW. The reactor was in extended shutdown state since November 2004. The DRC government has decided to resume its operation using the last uploaded core. One of the safety features to be determined before putting the spent fuel into the reactor core is the calculation of its excess reactivity, radionuclide inventories as well as its discharge burn-up. The spent fuel was modeled and simulated by using Monte Carlo software, MCNPX code. The input data and the horizontal and vertical modeling for the fuel pins, control rods and moderator were done. The model results were validated by calculating the effective delayed neutron fraction (β eff) and the worth of the control rods. The results of the criticality and fuel burn-up were compared with the reference design parameters and with the experimental measurements and it were found in good agreement. The calculations showed that the last uploaded core has 47.00 g of 235U which represent only 2% of fissile materials. The depletion analysis results showed that the highest radio-activities come from 151Sm, 137Cs, 90Y, 90Sr and 85Kr.

Calculation of core neutronic parameters in electron accelerator driven subcritical TRIGA reactor

Investigating the dependence of reactor core neutronic parameters on the accelerator related parameters such as source multiplication coefficient (K s ), beam profile and electron beam energy (E e ), have a significant impact on the design of future ADS reactors and the cost of their construction. Using the MCNPX code, core neutronic parameters were calculated for three eigenvalues levels (k s ) such as axial and radial distributions of neutron flux, effective multiplication coefficient (K eff), photoneutron yield (Y n/e ), net neutron multiplication (M), energy gain (G), energy constant gain (G 0), power peaking factor (P max/P ave), and importance of neutron source (φ*) for the tungsten target in the core of TRIGA reactor which was driven by an electron accelerator. According to the results, when K s increases safety margin and the required accelerator current (I e ) decrease, but the parameters G and φ* increase. Also, using the parabolic spatial distribution for the electron source instead of the uniform spatial distribution, the parameters Y n/e , G and M increase. In addition, by increasing E e from 100 to 500 MeV, Y n/e and G are improved, and I e and accelerating power (p acc) decrease. Therefore, these results suggest that the dependence of reactor core neutronic parameters on accelerator related parameters is necessary to optimize the design cost of an ADS reactor.

Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors

In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

Cryoconite as a temporary sink for anthropogenic species stored in glaciers

Cryoconite, the typical sediment found on the surface of glaciers, is mainly known in relation to its role in glacial microbiology and in altering the glacier albedo. But if these aspects are relatively well addressed, the same cannot be said about the geochemical properties of cryoconite and the possible interactions with glacial and peri-glacial environment. Current glacier retreat is responsible for the secondary emission of species deposited in high-altitude regions in the last decades. The role played by cryoconite in relation to such novel geochemical fluxes is largely unknown. Few and scarce observations suggest that it could interact with these processes, accumulating specific substances, but why, how and to what extent remain open questions. Through a multi-disciplinary approach we tried to shed lights. Results reveal that the peculiar composition of cryoconite is responsible for an extreme accumulation capability of this sediment, in particular for some, specific, anthropogenic substances.

Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR

In the current research, sensitivity analysis of the accelerator and fissionable and non-fissionable spallation targets containing U-238, Pb, LBE and W materials, Yn/p (spallation neutron yield), Ep (proton energy), G (energy gain), ϕ* (importance of neutron source), Keff (effective multiplication coefficient), Ks (source multiplication coefficient) and Ip (accelerator current) for two cases of Ks including: 0.91 and 0.97 in an Accelerator Driven Subcritical Reactor TRIGA (TRIGA-ADSR) were studied. These neutronic factors were computed by MCNPX code. The obtained results of this study show that for the Yn/p and G parameters, there is an optimum energy from 800 to 1000 MeV. Furthermore, according to the results, if this reactor would be operated close to criticality, the effect of reactivity insertion on the core power is raised. Also, the optimum value of Ks = 0.97 was chosen as adequate multiplication coefficient in this research because of appropriate margins. Lastly, the results of these investigations show that analysis of sensitivity and specificity of the accelerator and spallation target factors is required to optimize the neutronic plan of ADSR.