Related papers: Subroutines to Simulate Fission Neutrons for Monte…
Employing a recently developed Monte Carlo model, we study the fission of 240Pu induced by neutrons with energies from thermal to just below the threshold for second chance fission. Current measurements of the mean number of prompt neutrons…
Many neutron transport Monte-Carlo codes can randomly sample fission neutron energies from a Watt spectrum. The quality of simulations depends on how well the Watt spectrum represents the true energy spectrum of the fission neutrons, and on…
The $^{235}$U(n,f) reaction cross section was measured relative to neutron-proton elastic scattering for the first time in the energy region from 10 MeV to 440 MeV at the CERN n\_TOF facility, extending the upper limit of the only previous…
Background: Spontaneous fission events emit prompt neutrons correlated with one another in emission angle and energy. Purpose: We explore the relationship in energy and angle between correlated prompt neutrons emitted from 252Cf spontaneous…
Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and $\gamma$-ray~observables. Beyond simple average quantities, the study of distributions…
Quantifying inherent neutron sources in matter, particularly $(\alpha, n)$ reactions and spontaneous fission, is important in nuclear engineering and other fields. The SOURCES code is a common tool for calculating the yield and spectrum of…
Neutrons can induce background events in underground experiments looking for rare processes. Neutrons in a MeV range are produced in radioactive decays via spontaneous fission and (alpha, n) reactions, and by cosmic rays. Neutron fluxes…
The CGMF code implements the Hauser-Feshbach statistical nuclear reaction model to follow the de-excitation of fission fragments by successive emissions of prompt neutrons and $\gamma$ rays. The Monte Carlo technique is used to facilitate…
A Monte Carlo simulation algorithm to investigate the measurement of the average prompt neutron multiplicity as a function of pre-neutron mass, for fragments from the spontaneous fission of 252Cf is presented. The input data consist of…
Prompt fission neutron spectra from $^{239}$Pu($n,f$) were measured for incident neutron energies from $0.7$ to $700\,$MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center. A newly designed…
The Pu-239 fission chambers are widely used to measure fission spectrum neutron flux due to a flat response to fast neutrons. However, in the mean time the resonance and thermal neutrons can cause a significant influence on the measurement…
Coincidence and time-of-flight measurement techniques are employed to tag fission neutrons emitted from a 252Cf source sealed on one side with a very thin layer of Au. The source is positioned within a gaseous 4He scintillator detector.…
Isotopic fission-fragment distributions of $^{240}$Pu have been measured, for the first time, as a function of the initial excitation energy, and the prompt neutron multiplicity has been derived from these data. The $^{240}$Pu fissioning…
Accurate multiplicities of prompt fission neutrons emitted in neutron-induced fission on a large energy range are essential for fundamental and applied nuclear physics. Measuring them to high precision for radioactive fissioning nuclides…
Excited nuclear fragments are emitted during nuclear fission. The de-excitation of these fission fragments takes place as sequential emission of neutrons followed by photons. A correlation between neutron and photon multiplicities…
The mass and kinetic energy distribution of nuclear fragments from thermal neutron induced fission of 235U have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening on the standard deviation of the…
Monte Carlo (MC) neutron transport provides detailed estimates of radiological quantities within fission reactors. This involves tracking individual neutrons through a computational geometry. CPU-based MC codes use multiple polymorphic…
Neutron-induced fission cross-section studies are being pursued with a double Frisch grid fission detector. We are working towards conducting neutron-induced fission cross-section measurements of high priority 241Am(n, f) and 244,245Cm(n,…
Within MCNP6, users can use multiplicity distributions for spontaneous and neutron-induced fission from an evaluation performed by Lestone (LA-UR-05-0288). This evaluation assumes the multiplicity distributions are Gaussian and adjusts the…
This article describes an original approach to analyze simultaneously cross sections and surrogate data measurements using efficient Monte Carlo extended $\mathcal{R}$-matrix theory algorithm based on unique set of nuclear structure…