ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site

neutronic losses to 233Pa.
3-32
clude reduced poisoning effects from in-core fission products and an in-
creased f i ss il e inventory a
Other effects of the %ow power density in-
The reactor core would ~~nsist of a central region containing 20
vol Z fuel salt and a larger surrounding zone containing 13 vsl % salt.
Neutron moderation would be provided by vertical cylindrical unclad graph-
ite Gologs 911 with fuel salt flowing upward thrseaih central passages and
between the moderator elements. me core would be'surrounded first by
salt plenums and @~pansio~ spaces and then by a graphite reflector and
the reactor vessel.
With this core design, a 1-GWe plant would require an initial fis-
sile loading of 3450 kg 235U at 28% enrichment (extractable from about
870 short tons sf U,O,>.
makeup requirement would be about 4478 kg of 20% enriched 235U (from 1125
short tons of ~ ~ 0 , for ) a lifetime U308 demand of 2800 short tons,
ever, at the end of plant life, the fuel salt would contain denatured fis-
sile uranium (233U and 235U) equivalent to at least $00 short tons of
natural U308.
and reenriched, it would substantially reduce the net fuel requirement of
the BMSR.
Over 30 years at 75% capacity factor, the fuel
Row-
If this material could be recovered (eogop by fluorination)
Preliminary calculations of the kinetic and dyrtamic characteristics
of the BbfSB system indicate that it would exhibit high levels of control-
lability and safety. 'Fhe system would also possess inherent dynamic sta-
bility and wouM require only modest amounts of reactivity control cap-
$ iI ity e
A first-round analysis of the thermal-hydraulic Characteristics of
the DMSR core conceptual design indicated that the cylindrical moderator
elements would be adequately cooled by the flowing fuel salt and that
reasonable salt temperature distributions could be achieved with some
orificing of the fuel flow passages. While some uncertainties about the
detailed flow behavior in the salt-graphite system remain which would
have to be resolved by developmental testing, the results would not be
expected to affect the fundamental feasibility of the concept.
Tke primary fuel salt would be a molten mixture of EiF and BeF2 eon-
taining ThFb9 denatured UF4, and some PuF3. Lithim highly enriched in