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Radiation Damage Investigation on Certain Alternative Fluids in a Hybrid System by Using MCNPX Monte Carlo Radiation Transport Code

100%

Günay M.

Acta Physica Polonica A

2015

vol. 128

issue 2B

B-113-B-117

In this study, the effect of the radiation damage of spent fuel-grade plutonium content was investigated in the structural material of a designed fusion-fission hybrid reactor system. In this study, the molten salt-heavy metal mixtures 99-95% Li_{20}Sn_{80}-1-5% SFG-Pu, 99-95% Li_{20}Sn_{80}-1-5% SFG-PuF_{4}, and 99-95% Li_{20}Sn_{80}-1-5% SFG-PuO_{2} were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. Proton, deuterium, tritium, He-3 and He-4 gas production rates are the parameters of radiation damage. In this study, damage to the total structural material and each 1.0 cm thickness thereof was measured as a function of radiation energy, using the selected fluid rates, for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.

The Effect on Neutronic Calculations of Certain Alternative Fuels in a Boiling Water Reactor by Using MCNPX Monte Carlo Method

51%

Günay M., Sanchez Espinoza V., Travleev A.

Acta Physica Polonica A

2015

vol. 128

issue 2B

B-110-B-112

We were studying on the modeling of boiling water reactor fuel assemblies at pin-by-pin level by using Monte Carlo method. The designed boiling water reactor system is cylinder, and the radius of the cylinder is 228 cm. The total active core height is 315.79 cm. The reactor core was divided into the square lattice 7 × 7 type with a constant pitch of 30 cm. The core was surrounded with the reflector. The reflector was surrounded by SS316LN ferritic steel with width of 3 cm. The mixtures 0.2-1% PuF₄ and PuO₂ were used as fuel. In this study, the effect on the neutronic calculations of PuF₄ and PuO₂ fuels was investigated in the designed boiling water reactor system. There were calculated k_{eff}, heat deposition and the fission energy in the designed boiling water reactor system. The three-dimensional (3D) modelling of the reactor core and fuel assembly into the designed boiling water reactor system was performed by using MCNPX-2.7.0 Monte Carlo method and the ENDF/B nuclear data library.

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2 Acta Physica Polonica A

2 2015

2 Günay M.

1 Sanchez Espinoza V.

1 Travleev A.

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Radiation Damage Investigation on Certain Alternative Fluids in a Hybrid System by Using MCNPX Monte Carlo Radiation Transport Code

The Effect on Neutronic Calculations of Certain Alternative Fuels in a Boiling Water Reactor by Using MCNPX Monte Carlo Method