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OCR for page 37
4
Retrieval and Pretreatment
Retrieval of waste consists of removing the bulk of the waste from
the underground tanks or bins and safely transporting it to the process
facilities. Waste retrieval is treated in this chapter as part of the pre-
treatment process and will also be discussed in Chapter 6 with regard
to tank closure issues. Pretreatment consists of separating the HLW
components from the bulk of the waste and preparing the waste stream
for feed i ng to i mmobi I ization processes. Th is preparation is necessary to
remove constituents that interfere with the HLW immobilization process
and achieve the final objective of reducing the volume of solidified
HLW produced and meeting quality criteria.
Since most defense HLW was neutralized (with the exception of
INEEL waste) to minimize tank corrosion, it contains large amounts of a
wide variety of non-radioactive salts and metal hydroxides. Sidebar 4.1
describes the pretreatment programs for the HLW sites. The pretreat-
ment of defense HLWis a much more complicated problem than the
pretreatment of relatively pure acidic HLW from processing commercial
reactor fuel (as performed in some foreign countries such as France,
United Kingdom, and Japan) because HLWis not neutralized. The pre-
treatment required depends on the chemical and radiochemical compo-
sition of the waste. Because of the complex and varied chemistry of the
elements present in the mixture, it is usually not possible to find a sin-
gle reagent or chemical process efficient enough to accomplish all of
the separations needed. Therefore, multistep processes are normally
required. The unit operations that are performed as part of pretreatment
(Figure 4.1 ~ include the fol lowing:
· retrieval and blending;
solids-liquid separation;
sludge leaching; and
· liquid decontamination.
R e t r i e v a I a n d
P r e t r e a t m e
37
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SIDEBAR 4.1 SITE-BY-SITE PRETREATMENT BASELINES
Pretreatment at the Hanford Site
Retrieval of waste at the Hanford Site has just begun; in fact, to date, none of the tanks has been com-
pletely emptied. After retrieval, the alkaline waste will be sent to a pretreatment facility to separate
hazardous constituents from bulk material.The SLS process is followed by enhanced sludge washing.
Cesium and strontium will be removed by ion exchange from the liquids and sent to the HLW immobi-
lization facility along with the HLW solids.The Hanford Site has selected vitrification in borosilicate
glass for immobilization of the LLW stream, as well as the HLW.
Pretreatment at INEEL
The proposed plan includes the "non-separations alternative early vitrification option,"which encom-
passes (1 ) direct vitrification for the liquid SEW, with use of the current tank farm ending by the end of
2012, and (2) for the calcite, enhancement of characterization, retrieval, and treatment technology,
leading to vitrification with or without separations by 2035.The method for processing and disposal
for calcite will depend on the results of future separations research (Huntoon, 2000). A previous NRC
committee on alternatives for HLW treatment at INEEL concluded that calcite is chemically stable and
safely stored; therefore no immediate action should be taken (NRC, 1 999b).
Pretreatment at SRS
The alkaline salt solutions retrieved from different tanks are mixed in pretreatment tanks to"blend
down"components that might cause problems in the vitrification step.The sludge is separated and
pretreated by caustic leaching to solubilize aluminum from the HLW feed.The SRS is still in the process
of selecting a method to separate cesium from the supernate, dissolved saltcake and sludge-leach
solutions.The previous method (simultaneous large-scale in-tank precipitation with tetraphenylborate
to remove cesium, and sorption with hydrous sodium titanate to remove strontium) was discarded
because of technical difficulties related to rapid release of benzene (NRC, 2000b). Studies to resolve the
benzene problem, as well as to identify alternative processes, are currently under way. In the mean-
time, HLW sludge is being vitrified for future disposal (see Chapter 5). Pretreatment at SRS also
includes adding formic acid to reduce mercuric compounds in the waste to metallic mercury.This mer-
cury is then steam-stripped from the feed directed to the immobilization facility to avoid its presence
in the melter.
Pretreatment at WVDP
After the sludge remaining at the bottom of the tanks was washed, the wash supernate and the origi-
nal supernate liquid were pretreated with zeolite to remove the radioactive cesium, strontium, and
residual transuranic elements.The decontaminated supernate was concentrated by evaporation and
immobilized with cement.The high-level sludge and zeolite were combined and vitrified in borosilicate
glass.The WVDP has retrieved and vitrified more than 95 percent of the initial waste. Zeolite was not
selected for treating supernates at other sites because it would generate an excessive volume of vitri-
fied HLW.
·~e
H ~ G H - L E V E E W A S T E
38
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Water
· Retrieval
Process
Slurry ~ ~ ~
Liquid
1K ~ ~ ~
HEW
Storage
Tank
HEW
Storage
Tank
Retrieved Slurry
Slurry
Feed
Storage
Tank
Slurry
Solids-Liqujd | Slurry
Separation I ·| Leaching
Pretreatment
Processes
Solution
Blending
Tank
Washed
Slurry Solids Ljquid I Solids
· Separation
Solution I A_
Separation
Processes
Cesium
Strontium
Transuranic elements
Technetium
Other Elements
I
LLW l l HL~
I m mobi I ization I m mobi I ization +
Disposal
Disposal
The first step consists of the retrieval of waste slurry from the tanks
and the mixing of various batches of feed materials to achieve a reason-
able volume of feed with uniform properties. SLS is performed to sepa- process.
rate solids from the slurry, leaving a solids-free solution. This separation
may recur at several points in the process. Sludge leaching consists of
R e t r i e v a I a n d P r e t r e a t m e
FIGURE 4. 1 Simplified
scheme of the pretreatment
39
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the selective dissolution of specific constituents from the solids in the
retrieved waste slurry. Liquid decontamination consists of the selective
removal of radioactive and hazardous species from the liquid streams.
The solids from sludge leaching and the species removed during the liq-
uid decontamination step are combined to generate the feed for HEW
immobilization. If key radionuclides have been adequately extracted,
this waste is incidental to reprocessing, or WIR, and it is classified
according to its composition (see p. 1 1~. In addition to small concentra-
tions of radioactive species, WIR may contain significant amounts of
hazardous wastes (such as chromium, mercury, and lead) as well as
large amounts of soluble salts (such as NaNO3, NaOH, and NaAIO2~.
Additional constraints are imposed if WIR contains hazardous material
causing it to be classified as mixed waste, as defined in Sidebar 2.2.
Disposal of WI R. although less expensive than that of H LW, sti l l repre-
sents a high cost because of its large volume.
Pretrealment Issues
Pretreatment operational issues, long-term research needs, as well as
general approaches used at the different sites are described below. The
objective of the long-term basic research recommendations for pre-
treatment is to provide the scientific basis for developing high-efficien-
cy, high-throughput separation methods that could reduce HEW pro-
gram costs over the next several decades.
Retrieval and Blending
To date, waste has been completely retrieved from two tanks, both
located at the SRS. The DOE plans to retrieve the remaining waste from
the storage tanks with methods used in the past at the Hanford Site and
SRS. Most of the waste will be retrieved from the tanks by pumping out
the slurry of supernate and solids. In the case of INEEL waste, calcine
will likely be retrieved from the storage bins by vacuuming. Stabilized
water (water adjusted for pH and REDOX potential to minimize corro-
sion of the steel tank) will be added to the residual waste. Following
mixing to Dissolve soluble components e.g., saltcake containing alkali
metal salts including cesium) and suspend insoluble solids (e.g.,
hydroxides of iron, chromium, nickel, transuranic elements, alkaline
earth sulfates, and phosphates) the slurry will be pumped out as before.
Recovery of residual waste when the tank is nearly empty will be done
by sluicing, which may present problems in the case of leaking tanks,
depending on the extent of recovery required.
. . . .. . . . . . .
. . . . .
H ~ G H - L E V E E W A S T E
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Wastes from multiple tanks can be blended in a waste receipt tank,
as currently done at the SRS, but this may not be done at the Hanford
Site because of the lack of free tank space. Instead, the Hanford waste
will be treated on a tank-by-tank basis (see Chapter 5~. Blending of the
waste feeds can be used to reduce the final volume of immobilized
waste, by diluting those components that (1 ) have low solubility in
borosilicate glass, which limits the achievable waste loading, and
(2) have elevated concentrations in some tanks. The "blended" compo-
nents will therefore be less limiting with regard to waste loading. In
general, the waste will reside in various tanks for an extended period
prior to transfer for pretreatment, which will allow time for mixing,
sampling, and analysis to ensure that it meets the feed specifications for
the pretreatment steps, discussed below.
tong-Term Research Need
After some discussion, the committee decided that blending and
retrieval is not a fertile topic for basic research because it would over-
lap with other research activities undertaken within the EM, such as
those of the Tanks Focus Area (TEA) (DOE-TFA, 2000b).
So~ids-[iquid Separation
Solids-liquid separation is required to physically separate the insolu-
ble components from the supernate and leach solutions. This is a key
operation because most of the hazardous material (e.g., strontium,
transuranic elements, but not cesium) is associated with the solids.
Since the clarified liquid must meet the stringent decontamination
requirements for LLW, it is necessary to remove solids to an unusually
high extent. For example, decontamination factors (DFs) in the order of
1 0,000 to 1 00,000 may be requ i red for some red ionucl ides to meet
specifications for the immobilized LLW product; but solids removal by
such large factors is not commonly achieved in practice in a single
process cycle.
Solids-liquid separations will likely be required at more than one
point during the process. Although simple settling and Recantation will
be adequate for some processes, it is necessary to remove the solids
with an extremely high efficiency, at least once, during the process
cycle to meet the decontamination requirements for all of the con-
stituents i n the clarified I iqu id. I nadequate SLS wi 11 i m pact downstream
operations for example bv causing olu~in~ of ion-exchange columns
, , , , , in, w~ ~
Ratio of concentration of a species in the feed to that in the effluent of a
process (see the glossary in Appendix G).
R e t r i e v a I a n d P r e t r e a t m e
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and by degrading DFs for most processes. This operation is critical to
the operability and performance of pretreatment, and its success is not
assured.
tong-Term Research Needs
Long-term basic research is needed on filter media material and on
filtration methods using appropriate simulated slurries followed by veri-
fication using actual waste. The objective is to improve the rate of filtra-
tion, filter media life, resistance to plugging or fouling, and removal
efficiencies for very small solid particles and colloids (removal factors at
least 104 with a range of particle sizes down to 0.1 micrometer, and
possibly colloidal). Additionally, SLS methods other than filtration (such
as centrifugal separation or flocculation and settling) should be investi-
gated.
Sludge [caching
After being separated from the retrieved slurry, the sludges are
leached by mixing them with a solution of sodium hydroxide to dis-
solve some of the bulk constituents. The purpose of leaching is to
remove compounds that would be detrimental to the immobilization
process or to the quality of the immobilized product. Examples of such
compounds are sodium and aluminum, which are large contributors to
the bulk waste material, and chromium, sulfate, and phosphate, which,
although present in much smaller amounts, are relatively insoluble in
borosilicate glass and interfere with vitrification operations. If not
removed, these non-radioactive materials would increase the final vol-
ume of the HEW and/or lead to a hard-to-process slurry feed or an
unacceptable final product.
Removal of these undesirable species from the sludges is currently
based on the "enhanced sludge leaching" method that is being used at
the SRS and will likely be applied at the Hanford Site. This technique
consists of leach i ng sl udges with strongly al kal i ne sol utions with the
purposes of (1 ) solubilize aluminum and some other elements and (2)
metathesize2 and thereby partially solubilize anions such as phosphate
and sulfate from insoluble salts (e.g., Cased. The main problem with
the enhanced sludge leaching method is that a large fraction of the
waste (hydroxides of iron, manganese, nickel, zirconium, and other
metals) cannot be dissolved in alkaline conditions. An alternative
approach is to leach with acidic solutions to dissolve more of the
solids, thereby leaving very little sludge. This was considered for
Metathesis is a double-decomposition chemical reaction of the type AB + CD
~ AD + CB that is driven by the law of mass action when there is an excess of
one ~on.
H ~ G H - L E V E E W A S T E
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Hanford waste (Swanson, 1993) and is still under consideration for
INEEL waste, because alkaline leaching is not effective for the zirconi-
um-type calcine (NRC, 1 999b). In fact, except for limited cases involv-
ing complexing with concentrated fluoride or oxalate, zirconium can
be maintained in solution only under acidic conditions. Zirconium is
the dominant element that has to be separated only for INEEL waste;
however, in all cases, the leach solutions require a subsequent SLS step.
Operations using precipitation and solids leaching can be affected
negatively by several factors:
· the presence of high salt concentrations and organic complex-
ants that may solubilize strontium, and perhaps transuranic ele-
ments;
· the solubility and speciation of actinides (plutonium, neptunium,
americium, uranium) and strontium in caustic media containing
carbonates and organics (e.g., oxalic acid); and
· secondary reactions in leach solutions that can lead to inadver-
tent reprecipitation and gel formation. These reactions have been
observed in some aged or mixed Hanford leach solutions that
were saturated in many components.
A number of unknown reactions could occur during solids leaching
and washing, with uncertain consequences for the chemical and physi-
cal properties of the waste stream. Slow reactions can occur between
aluminates, silicates, and other materials to form complex solids. Leach
solutions and clarified supernates may be mixed together and/or evapo-
rated, thereby changing their composition. Since leach solutions were
initially saturated in several components, it is not surprising that new
insoluble compounds can be generated. This has two main implications
for the processing: (1 ) SLS processes must be performed several times,
and (2) solids can continue to precipitate, which can disrupt operations
such as filtration, ion exchange, and evaporation.
A large amount of research has been done on solubilities in multi-
component systems, for example by RapLo and Lumetta (2000), and
efforts have been made to model the results for application to the
extremely complex waste systems. At Hanford, for example, a thermo-
dynamic equilibrium program called Environmental Simulation
Programs is used for several purposes, such as estimating the conse-
quences of waste mixing and dilution during transfers and retrieval
(Mahoney et al., 2000; Papp, 1998~. Extension of such models to
include kinetic data and solid phase identification would be beneficial.
At INEEL, there are only very limited data on dissolution of calcine in
acid solutions, and more studies are needed if the waste is to be
processed before i mmobi I ization (N RC, 1 999b).
R e t r i e v a I a n d
P r e t r e a t m e n t
OCR for page 44
tong-Term Research Needs
Sludge leaching offers several long-term basic research opportunities
for successful operation of the pretreatment processes for all of the
sites. Projects should focus on chemical kinetic data (rates of dissolu-
tion) and equilibrium data (residual concentrations and solubility prod-
ucts) in very complex multicomponent systems. Examples of such com-
ponents are: sodium, aluminum, silicon, iron, chromium, zirconium,
sulfate, phosphate, carbonate, hydroxide, nitrite, and nitrate, many of
them dissolved at saturation limits. In spite of the complexity of the
task, research in this direction can yield useful results in the long term.
Research is also needed on solution stability and precipitation mecha-
nisms in highly concentrated solutions. Of particular interest is the sta-
bility of solutions with time, relative to reprecipitation of new phases
and gel formation. To the extent practical, effects of radiolysis should be
i n c I uded i n expert ment with s i m u I ants.
Long-term basic research is needed to develop methods to enhance
removal or to cope with the presence of problematic material and con-
stituents not readily dissolved by alkaline leaching (e.g., decomposition
of organic materials and dissolution of Cr3+ by oxidation to Crop. In
addition, the option of acid leaching of the sludge should be investigat-
ed to dissolve a greater fraction of the solids. Finally, research leading
to the development of a predictive model for solids-liquid behavior in
concentrated salt solutions could result in reduced costs and time.
Liquid Decontamination
After sludge leaching and SLS, the supernate and all leach solutions
are carried through one or more processes designed to remove specific
hazardous or radioactive constituents so that the liquid product can
meet the acceptance criteriafor LLW immobilization and disposal. As
indicated in Figure 4.1, these processes may include removal of
radioactive materials such as cesium, strontium, transuranic elements,
technetium, and other elements, as well as non-radioactive toxic metals
and organics. Several processes are known for each of these separa-
tions, such as zeolite ion exchange to remove cesium and strontium
and hydrous sodium titanate ion exchange to remove strontium and
transuranic elements. Since none of the processes is as simple and
effective as might be desired,3 there is continuing interest in finding
improved separation methods.
3For instance, the monosodium titanate process employed at the SRS may not
be successful for all waste streams since it may require extensive blending
because of low DFs.
H ~ G H - L E V E E W A S T E
44
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There is potential advantage in combining two or more separation
processes into a single unit operation, for example, by combining
processes for both strontium and transuranic elements removal, as well
as for finding a single reagent that effectively will remove more than
one problem constituent. An example of a combined process is the use
of a mixture of chlorinated cobalt dicarbollide, polyethylene glycol and
octyltphenyl)-N, N-d i isobutylcarbamoyl methyl phosph i ne oxide
(CMPO), which has been used in Russia for simultaneous removal of
cesium, strontium, and transuranic elements and has received limited
testing at INEEL (Thompson, 1998~.
Absorption few., using activated carbon), coprecipitation, and ion
. .
. ~ ~ .
exchange provide selective removal of contaminants, either on particles
in columns or on absorbent solids that are added to the solution and
then removed by SLS. The main issues to be addressed with absorption
methods are the maximum achievable decontamination effectiveness
and the amount of sorbent required to achieve such effectiveness. The
loaded sorbent must either be routed to the HLW for disposal or be
eluted and reused for subsequent loading cycles. If reused, the attain-
ment of sufficiently large DFs (in the range up to 105) in subsequent
cycles is usually very difficult. Solvent extraction is another option: a
solvent extraction process using a crown ether is under study for SRS to
remove cesium from the HLW liquid (NRC, 2000b).
Solvent extraction may be particularly advantageous for acidic solu-
tions. Because of the unique composition of the INEEL HLW, the caus-
tic leaching process used at the other sites offers little benefit. Acid
leaching and separation of inactive components from the acidic solu-
tion is an option. The final immobilized HLW volume could be
decreased substantially if acid leaching and acid-side processing
(Swanson, 1 993) were selected for either Hanford Phase 11 or INEEL cal-
cine. The value of this approach depends on the availability of reposito-
ry space for DOE HLW. Current plans call for HLW from Hanford and
SRS to go to a first geological repository. HLW from INEEL will be
directed to a second repository, particularly if processing is not com-
pleted before the closure of the first repository (NRC, 1 999b, page 83~.
However, at present the United States is investigating only one site
(Yucca Mountain) to determine its suitability to host a geological reposi
tory, and a decision to proceed has not yet been made. The possibility
of a second repository is h igh Iy conjectu ral at th is ti me.
In all cases, each separation process generates some quantity of sec-
ondary waste, such as ion-exchange sorbents, organic solvents, or vari-
ous conditioning solutions. Dealing with these secondary waste streams
is sometimes difficult because they can interfere with the separation
processes and increase final immobilized waste volume. Therefore, it is
necessary to develop processes that minimize secondary wastes or gen-
R e t r i e v a I a n d
P r e t r e a t m e n t
OCR for page 46
erate secondary wastes that are relatively innocuous. The strategies for
pretreatment end for both HLWand LLW immobilization and disposal
are mutually interdependent. Because of this connection, the nature of
both immobilized wastes will be determined by the separation and pre-
treatment strategy adopted.
An important research effort in the pretreatment area is already car-
ried out within the Efficient Separations Program-lntegrated Program
(ESPI P), a crosscutti ng project with i n the EM (DOE-ESPI P. 2001 ). Its
mission is to identify, develop, and perfect separation technologies to
separate cesium, strontium, and transuranic elements from radioactive
waste streams. Additionally, the EPA has developed a large database of
technologies that might be applicable for the removal of non-radioac-
tive toxic materials (metals and organics from the HEW and LLW). The
EPA Superfund Innovative Technology Evaluation (SITE) program evalu-
ates all available information on the technology for hazardous waste
remediation and analyzes its overall applicability to other site charac-
teristics, waste types, and waste matrices. The objective of the SITE pro-
gram is to encourage the development and implementation of (1 ) inno-
vative treatment tech nologies and (2) man itori ng and measu remeet. For
further information see EPA-SITE (2001~.
tong-Term Research Needs
A long-term basic research effort is needed to complement the activ-
ities within the ESPIP to identify sorbents and separation methods for
cesium, strontium, technetium, and transuranic elements that are effec-
tive and operable in high-ionic-strength alkaline or acidic solutions and
high radiation fields. Sorbents must have high selectivity and capacity
and must either be capable of elusion and regeneration for a number of
cycles or be easily decomposed, such that disposal of the sorbent has a
minimal impact on the final volume of immobilized waste. For exam-
ple, inorganic sorbents should be investigated if their selectivity is suffi-
ciently h igh (e.g., monosodium titanate for strontium and acti n ides,
ammonium molybdato-phosphate for cesium).
Decontamination methods that combine removal of more than one
target constituent within a single unit operation (e.g., a combination of
sorbents or extractants that can remove cesium, strontium, and
transuranic elements simultaneously from a feed composition) are
needed and warrant further basic research. Research in solvent extrac-
tion separation methods is needed to identify stable, selective, high-
capacity extractants that also are inexpensive and commercially avail-
able. In addition, research on solvent extraction methods would be par-
ticularly valuable for use with acidic waste solutions for which solid
sorbents are often less effective. Solvent extraction from acidic solutions
is the standard for reactor fuel reprocessing, because fuel is soluble
H ~ G H - L E V E E W A S T E
46
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only in acidic solutions. The waste treatment methods now being used
are generally alkaline, mainly because the feed is usually alkaline and
there is reluctance to add large quantities of nitric acid to re-acidify it.
However, investigators should not restrict consideration to alkaline
methods only, even though that is the starting medium. There could be
benefit from treating solutions from acid leaching of alkaline sludge or
INEEL calcine, but the methods would be very different from fuel repro-
cessing.
R e t r i e v a I a n d P r e t r e a t m e
47
Representative terms from entire chapter:
leach solutions
