Program - Brookhaven National Laboratory

J. P. Lestone, E. F. Shores
Los Alamos National Laboratory
The neutron experiment (NUEX) was a common diagnostic on nuclear device tests conducted at the
Nevada Test Site (NTS). In these experiments neutrons from a device pass up a collimated line of site,
and in the case of a Faraday cup (FC) NUEX, the neutrons pass through a thin CH2 foil. Some of these
neutrons interact with the nuclei in the foil, generating light charged particles (predominantly protons)
which are collected in a Faraday cup. The time dependence of the Faraday cup current is a measure of the
energy spectrum of the neutrons that leak from the device. With good device models and accurate neutrontransport
codes, the leakage spectrum can be converted into a prompt fast-neutron-induced fission-neutron
energy spectrum (PFNS) from ∼1 to 11 MeV. This has been done for one of our last events containing
a plutonium primary, where the NUEX data were of a particularly high quality, and for an earlier event
containing a uranium primary. The fission-neutrons in these devices were produced by fission events
induced by neutrons over a broad range of energies. We have inferred 2 MeV n + 239Pu and n + 235U
fission-neutron spectra for outgoing neutron energies from 1.5 to 10.5 MeV, in 1-MeV steps. These spectra
are in good agreement with the Los Alamos fission model [1].
[1] D. G. Madland and J. R. Nix, Nucl. Sci. and Eng. 81, 213 (1982), and http://t2.lanl.gov/data/fspect.
LA 2 4:00 PM
A New Measurement of the Neutron Capture Cross Section of 235 U Below 5 keV
David Williams, Yaron Danon, Rian Bahran
Gaerttner LINAC Center, Rensselaer Polytechnic Institute, Department of Mechanical, Aerospace and
Nuclear Engineering, 110 8th St, Troy, NY 12180
Devin Barry, Greg Leinweber, Robert Block, Michael Rapp
Bechtel Marine Propulsion Corporation, Knolls Atomic Power, Laboratory, P.O. Box 1072, Schenectady,
New York 12301-1072
The neutron microscopic capture cross section for 235 U is a critical parameter for the design and operation
of nuclear reactors. The evaluated nuclear data libraries of ENDF/B-VII.1 and JENDL-4.0 have nearly
identical values for the neutron capture cross section for neutron energies below 0.5 keV. In the most recent
release of the JENDL library the onset of the unresolved resonance region (URR) was changed from 2.25 keV
to 0.5 keV, and in this region the average neutron capture cross section from ENDF/B-VII.1 is about 10%
higher than that from JENDL-4.0. In an attempt to address the discrepancies between the libraries, a
new measurement of the neutron capture cross section of 235 U was conducted at the Gaerttner LINAC
Center located at Rensselaer Polytechnic Institute. This measurement used a 16 segment γ-multiplicity
NaI(Tl) detector to detect the prompt gammas emitted from neutron interactions with a highly enriched
235 U sample. Using the time-of-flight method, detected events were recorded and grouped based on the
total gamma energy per interaction and observed multiplicity. A method was developed to separate fission
from capture based on total energy deposition and gamma multiplicity. Application of this method in
the thermal and resonance region below 0.5 keV for both the fission and capture cross section produced
cross sections in excellent agreement with both the ENDF/B-VII.1 and JENDL-4.0 evaluations. The
measurements support the conclusion to lower the 235 U neutron capture cross section below the ENDF
values in the energy range 0.5-2.25 keV.
LA 3 4:20 PM
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