Fusion solution to dispose of spent nuclear fuel, transuranic elements and highly enriched uranium. Page: 4 of 6
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inventory of the transuranic elements and the
highly enriched uranium by utilizing its energy
content and transmuting its long lived fission
products, which represents a complete and
attractive solution for this problem.
The main objectives of this investigation are to
introduce the fusion solution and to size the
required fusion system for solving the US spent
fuel problem as an example for such application.
In this solution the transuranic elements and the
long-lived fission products are eliminated, the
radioactive waste volume is minimized, the
generated energy from the transuranic elements is
used for generating revenue for the system, and the
need for a geological repository is eliminated. D-T
fusion neutrons are used for achieving these goals
with minimum fusion power.
2. Molten Salt Blanket Concept
The self-cooled molten salt blanket concept is
considered as an example to quantify the
performance of the fusion solution and the results
are used to size the required fusion system to
achieve the elimination goal for solving the US
spent fuel problem. Flibe molten salt (Li2BeF4)
was developed and used for the Molten Salt
Breeder Reactor (MSBR) [1 and 2]. Molten salt
technologies were developed for MSBR in the
1960's, for the fuel cycle of the fast breeder
program in the 1990's, and for decommissioning
the molten fuel salt of the MSBR in the 1990's.
Flibe characterization included the physical
properties, the corrosion issue, the chemical
processing, and the solubility of the fuel
compounds and the fission products in the salt [2
and 3]. UF4, ThF4, and PuF3 are the fuel
compounds selected for the MSBR. For the
disposal of transuranic elements, PuF3 is the
material of interest. PuF3 is a solid with a density
of 9.32 g/cm3 and it has a melting point of 1425*C.
The solubility of PuF3 in Flibe was measured [3]
for composition ranging in BeF2 from 28.7 to 48.3
mole in the temperature range of 450 to 650*C.
A poloidal blanket configuration is considered
where the inlet and the outlet manifolds are located
at the top section of the reactor. The salt coolant isfirst introduced to the front section of the blanket
to remove the surface heat flux from first wall.
Then, the flow direction changes at the bottom of
the reactor to leave the blanket module at the top.
This flow pattern simplifies the reactor manifold
system.
At the start, it is foreseen to operate the blanket
without transuranic elements to confirm and
calibrate the operation of the different systems.
This mode of operation reduces the shielding
capability of the blanket due to the absence of the
neutron absorbers, the transuranic elements.
Therefore, the blanket radial thickness was first
defined with pure Flibe as a function of the blanket
radial thickness, where the Flibe zone thickness
was varied from 0.2 to 0.6 m. The results show
that a 0.5-m Flibe zone thickness is required to
reduce the energy deposition in the shield to -4%
of the total energy deposition. At this blanket
thickness, all the shielding performance
parameters are quite satisfactory.
The second step defined the blanket performance
with PuF3 dissolved in the Flibe. The PuF3 weight
fraction varied parametrically in the range of
0.0025 to 0.0275, which is consistent with the
experimental results. The results show that the
blanket performance parameters change
monotonically in this range. The first two blankets
of Table 1 give the main blanket performance
parameters for the extreme values of the range. As
the PuF3 concentration in the Flibe salt increases,
all the blanket performance parameters improve.
At the highest PuF3 concentration, the plutonium
transmutation rate is 4.4 kg/MW.y of fusion
power. The corresponding tritium-breeding ratio
and the blanket energy multiplication factor are 2.2
and 15.3, respectively. This high tritium-breeding
ratio indicates that the lithium-6 concentration can
be reduced to increase the plutonium transmutation
rate and to reduce the tritium-breeding ratio.
The blanket performance analyzed parametrically
as function of the lithium-6 concentration in the
range of 2.5 to 7.5 (natural) % with constant PuF3
concentration. As the lithium-6 concentration
decreases, the blanket performance parameters
improve monotonically. The plutonium
transmutation rate changes from 3.4 to 48
kg/MW.y as the lithium-6 concentration varies
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Gohar, Y. Fusion solution to dispose of spent nuclear fuel, transuranic elements and highly enriched uranium., article, August 31, 2000; Illinois. (https://digital.library.unt.edu/ark:/67531/metadc723850/m1/4/: accessed April 18, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.