Title

Characterization of Prompt Neutron Spectrum of the Missouri University of Science and Technology Reactor

Abstract

An activation-foil method was used to obtain the energy spectrum of the prompt-neutron flux at the Missouri University of Science and Technology Reactor (MSTR). The foils were irradiated at the center of the reactor core (120 W configuration). The neutron spectrum was determined using the unfolding method implemented in the SAND-II code. Multiple foils of dysprosium, vanadium, indium, gold, aluminum, copper, cobalt, silver, nickel, and iron were placed at the source holder position and irradiated for 3 min each at 100 kW. The foil set cover energies from 0.025 eV to 7.2 MeV for a broad spectrum analysis. After irradiation, the sample was cooled using water until the dose rate reached a level that allowed safe handling. For each material, both bare and cadmium covers were used to obtain the thermal and epithermal neutron flux. The activity of each foil was determined by counting for 3 min with a high-purity germanium detector, and the results were used in SAND-II to obtain the neutron flux spectrum for the MSTR. Monte Carlo N-particle (MCNP) model was used to obtain the initial guess of the spectrum at the source holder position, which was used as input for SAND-II. The spectrum is unfolded using two different energy binning structures in the initial guess for SAND-II. As expected, the results provided highly thermal neutron spectra in source-holder environment, where water thermalizes most of the neutrons; 93% of the total flux at the source-holder position is no greater than 0.55 eV for the 120 W reactor-core configuration. The epithermal and fast neutrons were 6% and 1%, respectively at fast/epithermal energy boundary of 100 keV. A similar experiment and analysis in the past determined that 7%, 38% and 55% of the total flux at the source-holder position were respectively thermal, epithermal and fast for the 101 W configuration. Notwithstanding the difference between the two configurations, the current determination is an improvement to the characterization of the MSTR flux. This is because 7% thermal flux is unlikely in an environment primarily surrounded by water. The total neutron flux calculated by MCNP is 4.98 x 1011 ± 1.72 x 1010n/cm2s. The flux determined through SAND-II is 4.79 x 1011 ± 6.10 x 1010n/cm2; a difference of 4%.

Department(s)

Mining and Nuclear Engineering

Keywords and Phrases

Chemical activation; Neutron flux; Neutron spectrometers; Reactor cores; Safe handling; Sand; Experiment and analysis; Flux spectrum; High purity germanium detectors; MCNP; Monte carlo n particles; MSTR; Reactor core configurations; Science and Technology; Spectrum analysis; Multiple foil activation; Neutron flux spectrum; SAND-II

International Standard Serial Number (ISSN)

0029-5493

Document Type

Article - Journal

Document Version

Citation

File Type

text

Language(s)

English

Rights

© 2017 Elsevier, All rights reserved.

Publication Date

01 Aug 2017

Share

 
COinS