Abstract
Liquid-fueled Molten Salt Reactors (MSRs) are an advanced reactor design concept
with material accountancy challenges due to continuously-circulating fuel in bulk
form with online additions of fresh fuel. With over 24 distinct liquid-fueled MSR
designs being pursued by design companies around the world, including within the
U.S., facility operators and the International Atomic Energy Agency need approaches
and technologies to quantify nuclear material within MSR facilities for safeguards and
security purposes. Safeguards technical objectives were defined across a prospective
liquid-fueled MSR and a high-priority objective of quantifying nuclear material in
fresh initial salt (i.e., salt added to the MSR at the beginning of life) and makeup
salt additions to the reactor system (i.e., salt added over time while the MSR is
operational) was selected for further analysis. Computational simulations combining
source terms from SCALE/ORIGEN into MCNP models were run for different
MSR design parameters as well as different gamma- and neutron-based measurement
techniques to determine the impact of each parameter on the feasibility of radiation
monitoring on the outside of fresh fuel salt in piping for material accountancy. MSR
design parameters included a fluoride-based and chloride-based salt, four uranium
enrichments, two pipe materials, multiple pipe outer diameters and thicknesses, and
the optional presence of insulation and aluminum jacketing external to the piping.
NaI and high purity Ge gamma detector systems, total passive neutron counting
in a moderated 3He collar, coincident neutron counting in a moderated 3He collar,
and coincident neutron counting in in a moderated 3He collar with an interrogating
neutron source were assessed for feasibility. While no measurement technique
worked well for both salt types and all design parameters assessed, passive gamma
spectrometry and total passive neutron counting in a moderated neutron detector
collar both seem promising, especially for salts with high assay low-enriched uranium
salt and larger diameters. Active neutron interrogation with a neutron coincidence
collar counting thermal neutrons saw very high relative uncertainties due to high
accidental coincidences. A neutron coincidence collar counting fast neutrons will
likely be a promising alternative for quantifying 235U within unirradiated, uranium-
bearing fuel salts within piping.
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