Monte Carlo Transport Code MCNP Research Articles

Experiments and calculations of 14MeV neutron streaming through multi-layered slit assembly.

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.

Some features of fast neutron spectra in graphite cube with collimated 14.7MeV neutron source.

An analysis of fast neutron spectra in an integral experiment with a collimated monoenergetic (d, t) neutron beam has been carried out by means of the MCNP Monte Carlo transport code. The calculations were performed on a 95 × 95 × 95 cm3 graphite cube. Results of the transport calculations before smoothing the output spectra with a Gaussian response function of a real neutron detector, reveal a very close resemblance between the calculated neutron spectra on one hand and angular distributions of elastically scattered and first level inelastically scattered neutrons on the other. This conclusion holds especially for neutron spectra obtained on the axis of the collimated neutron beam. Neutron spectra obtained in an integral experiment with a collimated (d, t) neutron beam can therefore be effectively used for an experimental determination of the neutron angular distributions of elastic and first level inelastic scattering at the energy of the collimated neutrons.

Measurement and analysis of reaction rates in a large spherical depleted uranium pile surrounding a 14-MeV neutron source

Measurement was made of the reaction rate distributions for 238U( n, f), 235U( n, f), 238U( n, γ), 27Al( n, α) and 58Ni( n, p) in a large depleted uranium (DU) pile. The pile consisted of DU blocks forming a spherical shell of 45.72 cm radius and 40.64 cm thick. 14-MeV neutrons were generated at the center. Fast neutron leakage spectrum was also measured by an NE-213 spectrometer. In order to assess the 238U neutron cross sections of JENDL-2, the experiment was analyzed using the Monte-Carlo transport code MCNP with continuous energy cross sections. The agreement between the calculation and the experiment is generally unsatisfactory. The ratios of calculation to experiment of low energy reactions decreased with the thickness of the DU layer. The analysis of the Livermore pulsed sphere experiment for the small DU sphere revealed underestimation of leakage neutron spectrum around 10 MeV. The 238U cross sections of JENDL-2 should be improved for 14-MeV neutronics calculation.

Tritium production and foil reaction rates in a large lithium sphere by D-T neutrons

Abstract The benchmark data on tritium production-rate in a 120cm diameter lithium sphere are compared with one dimensional calculations by the Monte Carlo transport code MCNP using the nuclear data files of ENDF/B-IV and JENDL-3PR1, as the reaction-rate ratio data to the 93Nb(n,2n) reaction. Overestimation of the 7Li(n, n'α)t cross section values in the ENDF/B-IV file has been clearly concluded.

A Comparison of Different Concepts of Integral Experiments for Fusion Reactor Blanket Design

Different concepts of integral experiments for fusion blanket neutronics are investigated. The first is with the neutron source (tritium target of a neutron generator) located inside of or in immediate proximity to the stack of blanket materials under consideration. The second is based on irradiation of the stack by means of a collimated and, therefore, monoenergetic T(d,n)UHe neutron beam with a tritium target placed outside the stack. The comparison between the different concepts is carried out by means of the Monte Carlo transport code MCNP with continuous energy treatment. The comparison between the two approaches reveals that the integral experiments with a collimated monoenergetic T(d,n)UHe neutron beam result in a neutron spectrum that is better correlated with the details of elastic and inelastic scattering to the first level of the material's nuclei than the one with a neutron source inside a stack. In the case of a collimated neutron beam, there is a clearer separation between energy regions of different neutron interactions and, therefore, the source of discrepancies between measurement and calculation can be identified more easily and corrected by a proper treatment of the cross sections of the specified nuclear reactions.