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4
Preferred Technical Approach
Based on present understanding of the condition of the system
and its salts, a number of observations are offered that are relevant to any
decision on remediation options and strategy. As further information is
obtained, these observations may have to be modified. These comments
form the basis for a preferred technical approach developed by the panel.
COMMENTS ON PROCESS STEPS
As discussed earlier, it is possible to conclude the following:
1. Since the charcoal trap is "at the end of the line" (Peretz,
1996c, Figure I.9, p. I-13) and all the migrated uranium salts originated
in the drain tank, it is prudent to assume that all the piping contains
uranium in volatile and nonvolatile forms such as uranium hexafluoride
(UFO; solid and gas), urany! difluoride (UO2F2), uranium pentafluoride
(UFs), uranium tetrafluoride (UF4), and uranium trifluoride (USA.
2. Since the freeboard gas space above the salt is at a lower
temperature than the salt, there are probably significant uranium
compounds deposited on the walls of the tanks and the thimbles and on
the underside of the tank cover.
3. Current plans to resolve issues 1 and 2 above (with collection
of uranium compounds in the piping and the auxiliary charcoal bed
tACB], for use in a mass balance estimate) are important to determine the
1The panel is aware that Oak Ridge National Laboratory (ORNL) documents (e.g.,
ORNL, 1996b) have used various assumptions about technology and process steps for
remediation. It is the panel's understanding that these assumptions were made for
administrative reasons such as budgeting and staffing and do not represent actual
decisions for implementation.
45
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46
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
appropriate strategy for dealing with the uranium in the salts. Indeed, it is
conceivable that there may not be as much uranium left in the salt as
initially anticipated; the uranium may be concentrated in certain regions
exterior to the bulk salt. A directional, high-resolution gamma survey
(e.g., using a welI-shielded, small-bore collimated germanium [Ge]
detector) could help clarify this question.
4. Separation of uranium from the salt and from the system
should be one of the major milestones, because it would eliminate
subsequent criticality concerns and reduce the rate of radiolysis. Doing
this without first moving the salt mixture to another location is attractive
because it avoids the hazards (and costs) of any transfer operation and
avoids leaving a residue of residual uranium and salt that is inherent in
such liquid salt transfers.
5. Except for fluorination in place, all other separation
technologies appear to require extensive new facilities with resulting long
times and high costs. With the present state of knowledge, the pane}
considers that some type of fluorination procedure to extract UFO from
the salt is the most promising technical approach to address comment 4
above.
DEVELOPMENT OF A PREFERRED APPROACH
With consideration given to the present state of knowledge of the
condition of the Molten Salt Reactor Experiment (MSRE) system and
salts and to the above-mentioned list of technical approaches that seem
credible, several information-gathering exercises are recommended to
assist the decision-making process, as discussed further in Chapter 8.
Subject to confirming evidence yet to be gathered, the panel recommends
that the following considerations, outlining a cautious, stepwise strategy,
be taken into account in the development of a preferred approach:
1. Additional chemical, physical, and process data on the
distribution of uranium and the remelting and refluorinating behavior of
the salts, as discussed in Chapters 2 and 3, are essential to validate
assumptions and reduce uncertainties; therefore, the plan must be
developed in stages either as the information becomes available or as it
becomes clear that such information cannot be made available.
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PREFERRED TECHNICAL APPROACH
47
2. Venting of the off-gas system piping and the gas space above
the salt in the drain tanks, together with volatile UFO removal from these
areas, is an essential first step for any alternative, and the resulting
reduction in system pressure will be a significant safety step. The
collection of uranium will also provide the first data point in an ongoing
material balance program.
3. The next step depends on the success of step 2 and an
assessment of system integrity. If possible, it should involve removing
the presently nonvolatile uranium salts from the piping and tank
freeboards and obtaining a material balance.
4. During steps 2 and 3 above, noninvasive methods of locating
significant uranium salt inventory (e.g., high-resolution gamma
spectroscopy scans) should be undertaken as feasible.
5. After "cleaning" of the piping and tank gas space, obtaining a
sample of the fuel salt could be considered. This sample seems useful for
evaluating the methods and possible hazards of salt melting, but such
merits should be weighed against the hazards of taking a sample and of
contamination problems.
Specifically, the gaseous radon hazard (see Appendix C), and
contamination with alpha- and beta-gamma-emitting residues, should be
considered. With an inventory of 5 g of uranium-232 (232U; 130 curies) in
the salt, a I-pound sample (0.01 percent by weighty would contain
approximately 13 millicuries (mCi), not only of224 U. but al2s20o of the
decay products thorium-228 (2 Pith), radon-224 ~ Rn), and Rn, the
latter of which is a highly radioactive, alpha-emitting gas. The 220Rn gas
permeates neoprene gloves and distributes rapidly to the air even in
countercurrent air flows. It then decays rapidly to polonium-216 (typo)
and on to a 3.3-hour half-life daughter, lead-212 `2~2Pb), that finally
decays (with a 36 percent yield) to thallium-208 ~ TI), which emits a
2.6-MeV (million electron volt) gamma ray (see Figure 2.~.
6. The alternatives for salt melting include the following. but
choices cannot be made until more information is available:
a. Melt as is.
melt.
A,
b. Hydrofluorinate below the liquidus temperature and then
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48
ANEVALUATION OF DOE ALTERNATIVES FOR MSRE
c. Use pool melting by either heater or laser, and
hycirofluorinate as melting occurs.
d. Perform step 6b without melting, and then step 6c.
7. At this point, if melting seems feasible, the decision can be
made whether to fluorinate in place or to transfer the molten salt to
another vessel for fluorination.
a. If the decision is to fluorinate in place, choices must be
made concerning fluoridating agents (see Chapter 3 and Appendix B),
melting methods (all at once or stepwise), and whether to use existing
=7 ~ ~ ~
tank heaters or a new heat source.
b. If the decision is to fluorinate elsewhere, the choice must
be made as to where and how.
Hydrofluorination in place, followed by melting (to test the salt
solubility), followed by transfer of the melt to a new vessel of ensured
integrity for fluorination is a strategy with merit if corrosion concerns
persist, because the hydrofluorination step, which may enhance the melt,
is a less corrosion-aggressive process than fluorination. The options of
thoroughly inspecting the vessel walls for localized thinning, and
possibly plugging the thimbles or filling them with HF-resistant material,
could be used to reduce any likelihood of leakage.
8. If melting either does not appear to be possible or does not
appear to be safe, the solid removal alternative would have to be
developed.
In the development of steps 6 through 8, some laboratory and
mockup tests and use of these test results in deciding what large-scale
process to apply to the drain tank salts would seem to be in order. For
example, the flush salts might be a preliminary surrogate, but because of
the (assumed) large difference in constituent concentration and oxidation
state, either melting or fluorination of the flush salts is only a
demonstration of the equipment and technique and does not confirm the
chemistry of the fuel salts. If all else fails and solid salt removal becomes
the proposed operation, it may be necessary to conduct a mockup test to
determine the mechanical problems (such as those introduced by the
cooling thimbles; the mechanical challenge is greater to remove solids
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PREFERRED TECHNICAL APPROACH
49
from regions interior to the tank that are behind cooing tube arrays, as
compared with regions directly under a flange on the tank head). Because
of the uncertainties inherent in a solid removal operation, it would be
prudent to do the work in stages and not to commit to doing the drain
tanks lentil after the mockup and the flush tanks have been done
successfully.
A particularly significant issue in carbon dioxide (CO2) blasting
or other solid removal techniques is gaseous radon contamination.
Whereas fine particles of salt can be filtered from a gas stream, radon
cannot. This consideration alone could well preclude the use of such
means of breaking up the salt if fluorination (with all the caveats
regarding melting) fails.
Also there should be efforts to determine the status of the system
(leak-tightness, pressure worthiness, corrosion damage to date) and
projected damage, as well as complete assurance of the absence of water
from the cells or the systems (since essentially all the credible postulates
of criticality excursions require the presence of water).
Tests on surrogate samples can be useful to address some of the
uncertainties associated with the fuel salts. For example, such tests could
be used to bound the range of redox conditions finder which
hydrofluorination and fluorination would be successful and to probe the
extent of segregation that could be tolerated. Peretz (1996c, p. I-30)
noted that segregation was observed on freezing surrogate salt samples
(uranium concentration varied by a factor of two, from top to bottom).
This effect would likely have been enhanced by the slow (melt refining)
process of cooling the several tons of irradiated salt. Radiation effects
and transport of UFO from the mass may have further increased the
inhomogeneity of the uranium and fission product distribution.
A sample of the actual fuel salt, preferably a core sample down
through the mass, would make it possible to test several remedies on the
actual salt itself before committing to major action that could be
irreversible. However, as noted earlier, the decision to obtain a core
sample should consider the disadvantageous consequences of possible
22sTh 224Ra and 220Rn contamination
Chapter 5 discusses specific technical processes mentioned above
that are finder consideration (Peretz, 1996c) for stripping uranium from
the salt.
Representative terms from entire chapter:
preferred technical
