The authors used the Monte Carlo neutron-photon (MCNP) code to design the scanner and review optimum materials and geometries. The yield-mass curve indicates a value of 5 ± 1 for the average number of neutrons. Both light- and heavy-fission product groups are shifted to higher masses relative to spontaneous fission of Cm 242. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. Fission is observed to be asymmetric with some indication of fine structure in the mass distribution associated with the 82-neutron shell. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. (LANL), Los Alamos, NM (United States) Sponsoring Org.: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP) USDOE Laboratory Directed Research and Development (LDRD) Program OSTI Identifier: 1899129 DOE Contract Number: 89233218CNA000001 Resource Type: Conference Journal Name: Transactions of the American Nuclear Society Additional Journal Information: Journal Volume: 125 Journal Issue: 1 Conference: 2021 ANS Winter Meeting and Technology Expo, Critical and Subcritical Experiments - I, Washington, DC (United States), 30 Nov - Related Information: Journal ID: ISSN 0003-018X Publisher: American Nuclear Society Country of Publication: United States Language: English Subject: 73 NUCLEAR PHYSICS AND RADIATION PHYSICS Nuclear Criticality Safety Program (NCSP) Integral Experiments Special Nuclear Material (SNM) Neutron Multiplicity Counting = Cf source is located at the center of the scanner very near the through-hole for the fuel rods. Publication Date: Research Org.: Los Alamos National Lab. Although demonstrated here for a 4π-detection system, future plans include simulating a more detailed model of the MC-15 detection setup to determine exact correction factors for future experiments. This suggests that further investigation is required to determine exact correction factors when calibrating neutron detectors with a 252Cf. 240Pu, a common spontaneous fission source of interest, ranged from 2% - 3% within this space. Neutrons produced in spontaneous fission are emitted with characteristic energy distributions, or spectra, that de- pend on various factors from one isotope to. It was also observed that configurations comparable to most detection setups, 10 atm and 6 - 8 cm polyethylene, were closer to the upper bound of deviations from a 252Cf. Incident neutron fission spectrum average Nup 3.1231 94-Pu-240. The Maxwellian average cross sections are for a peak neutron energy distribution at 0.0253 eV (a room temperature thermal distribution). Results showed that deviations between the simulated isotopes become more prevalent at higher energies with 238U showing the greatest deviation. Spontaneous Fission Rate - SF (F/sec/kg) Fission Cross Section - sigma. A total of 880 configurations were simulated by varying 3He pressure and polyethylene reflector thickness for 8 different isotopes. A parametric study was simulated in MCNP to investigate the impact of spontaneous fission neutron energy on neutron detector response.
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