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OCR for page 13
Overview of HEW Challenges
at DOE Sites
The DOE has approxi mately 400 mi l l ion l iters (1 00 mi l l ion gal ions)
of I iquid H LW stored in underground tanks and approximately 4,000
cubic meters of solid HLW stored in bins. The current DOE estimate of
the cost of converting these liquid and solid wastes into stable forms for
shipment to a geological repository exceeds $50 billion to be spent
over several decades (DOE, 2000~.
The most important HLW sites in the United States are the follow-
ng:
· the Hanford Site, near Richland, Washington,
· the Idaho National Engineering and Environmental Laboratory
(I N EEL), near Idaho Fal Is,
· the Savannah River Site (SRS), near Aiken, South Carolina, and
· the WestValley Demonstration Project (WVDP), near Buffalo,
New York.
All of the HLW sites listed above are briefly described later in this
chapter.
Sidebar 2.1 shows a simplified diagram of one of the processes that
produced HLW during chemical recovery of plutonium and uranium
from spent fuel and target materials for the production of nuclear
weapons. Most of the HLW is a multiphase mixture of solids and liquids
stored in underground tanks. Some has been retrieved and solidified
and is stored in underground bins or surface vaults. Processing and
handling of these HLW forms generate other radioactive wastes man-
aged as transuranic or low-level waste (LLW, see Sidebar 2.2), depend-
ing on their characteristics. The reclassification of HLW residues to LLW
or transuranic takes place according to DOE's procedures to determine
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
OCR for page 14
SIDEBAR 2.1 THE PLUTONIUM AND URANIUM EXTRACTION (PUREX) PROCESS
_m.~
Ed | [d |~ |~] | ~~ x. 0~ ~ ~ ~ ~ ~
~~ ~~ em.
~F ~ ~
~ ~ I:. O~s atop
_ ~ _
. ::
_ ~
_E1~1
~ :
. . my
,...
^..
11. it [~
It|| jiTd{~ ~11 iatlel~.~. 1~ - Nll.iTd{~ mill i.~I.I} nI [I
.~
UO 2+X
:
to
hi 0~32 ~
._
lit
The PUREX process, shown above, purified and separated both uranium and plutonium from defense
spent nuclear fuel. It was first used at the Hanford Site in 1952 and subsequently at the other HEW
sites.The first step consists of removing the cladding of spent fuel (fuel preparation) to expose fuel
rods. In the second step, spent fuel is dissolved in nitric acid.The third and key step is the separation
and recovery of uranium and plutonium from the fission products. It is done in a continuous counter-
current solvent extraction process using tri-N-butyl phosphate in a hydrocarbon diluent (such as
kerosene). Plutonium and uranium are extracted in the organic solvent while fission products and
other impurities remain dissolved in the highly acidic aqueous nitric solution stored in tanks and clas-
sified as HLW.The fourth step consists of separating plutonium from uranium by reducing plutonium
to the organic-insoluble, trivalent state while uranium remains in the organic phase in the oxidized
form. Both plutonium and uranium are then converted to solid oxide or metal. Except at the INEEL, the
H ~ G H - L E V E E W A S T E
14
OCR for page 15
waste was neutralized with sodium hydroxide prior to storage in the tank farms. A major advantage of
the PUREX process over the earlier processes used (bismuth phosphate and REDOX processes, see the
glossary in Appendix G) is that nitric acid is the only salting agent used in the aqueous phase; there-
fore HLW has a lower salt content than waste from other processes. Other advantages of PUREX are
reduction of the waste volume, greater flexibility in process conditions, reduction of fire and other haz-
ards, and reduction of costs. Processes based on PUREX have been adopted for nearly all fuel repro-
cessing throughout the world. SOURCE: Adapted from Benedict et al. (1981 ) and from JEER (1996).
whether the waste can be treated as waste incidental to reprocessing
(WIR) described in Chapter 6, Sidebar 6.1. The committee has not con-
sidered issues with these other wastes in order to focus its attention on
the retrieval and processing of HLW from tanks and bins.
HLW Management Strategy at DOE Sites
The current DOE plans to treat and dispose of HLW face many tech-
nical uncertainties. Many of the planned treatment activities are first-of-
a-kind efforts involving highly radioactive liquid and solid wastes.
Moreover, the waste streams are fundamentally different in character,
both among the HLW sites and within the sites, because waste streams
were generated from processing different kinds of nuclear fuel and were
treated and stored differently. Consequently, each HLW stream presents
different challenges and may require specific processing and immobi-
lization strategies.
However, the overall HLW management strategy applied by DOE is
the same throughout all HLW sites. The generic HLW remediation flow
sheet is shown i n Figu re 2.1 and comprises the "backbone" of th is
report. This flow sheet applies mainly to HLW in the tanks. In the case
of HLW stored at the INEEL in a calcined form, some details are slightly
different, as described later in this chapter. The HLW management strat-
egy consists of different steps, described as follows.
Characterization of waste in the tanks is performed to ensure the
continued safe storage of the waste and to facilitate the following HLW
management operations. At a given point, HLW must be retrieved from
the storage tanks and bins and processed for ultimate disposal in a
repository. The process of removing waste from the tanks and transport-
ing it to the processing facil ities is Cal led retrieval. After retrieval, the
HLW proceeds to the pretreatment phase, to separate constituents that
O v e r v i e w o f H L W C h a I I e n 9 e s
a t D O E S i t e s
OCR for page 16
SIDEBAR 2.2 NUCLEAR AND RADIOACTIVE WASTE TERMS
A. High-level waste: The radioactive by-product generated from processing irradiated fuel to sepa-
rate usable plutonium and other isotopes for weapons, research, and new fuel; and irradiated
fuel itself if destined for direct disposal.
B.
Transuranic waste: Waste that contains alpha-emitting transuranic elements with half-lives
greater than 20 years whose combined activity level is at least 100 nanocuries (3,700 becquerels)
per gram of waste at the time of assay.
C. Low-level waste: All radioactive waste not classified as HLW, transuranic waste, spent nuclear
fuel, or the by-product material (see definition F.3.b below) generated by the processing of ores
for extraction of natural uranium or thorium.
D. Mixed waste: Waste that contains both radioactive material regulated under the Atomic Energy
Act and hazardous chemical compounds regulated under the Resource Conservation and
Recovery Act, such as mercury, polychlorinated biphenyls, or organic solvents.
Spent nuclear fuel: Nuclear fuel that has been withdrawn from a nuclear reactor following irradi-
ation, has undergone at least one year's decay since being used as a source of energy in a power
reactor, and has not been chemically separated into its constituent elements by reprocessing.
Spent nuclear fuel, through fission and neutron activation and decay, contains multiple radioac-
tive elements with varying chemical and radiological properties. Spent fuel includes the special
nuclear material, byproduct material, source material, and other radioactive materials associated
with fuel assemblies.
Nuclear material (Atomic Energy Act material): Nuclear material generally includes (1 ) source
material, natural uranium, thorium, or any other material that, alone or in combination, is deter-
mined to be a potential fuel for a fission reactor; (2) special nuclear material, fissile material such
as uranium enriched in uranium-235, or plutonium that can undergo fission; and (3) byproduct
material, meaning (a) any radioactive material (except special nuclear material) yielded in or
made radioactive by exposure to the radiation incident to the process of producing or utilizing
special nuclear material, and (b) the tailings or wastes produced by the extraction or concentra-
tion of uranium or thorium from any material processed primarily for its source material content.
G. Radioactive material: Radioactive material includes all of the nuclear material defined above.
Radioactive material also includes naturally occurring radioactive elements such as isotopes of
radium as well as material made radioactive by processes other than nuclear fission.
SOURCE: Adapted from DOE (2000).
are radioactive, hazardous, or detrimental tothe immobilization step
from the bulk of the waste. After pretreatment, HLW is ready for immo-
bilization in a borosilicate glass matrix. This process is called "vitrifica-
tion." The HLW glass canisters produced are stored at the sites until
H ~ G H - L E V E E W A S T E
16
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High-
Level
Waste
F
\ 'I Retrieval
Safety | ~:
Characterization I
Characterization
Transport &
Pretreatment _1 Immobilization _ Disposal in
- - —I —Repository
LLW Stream ]~
~ Tank Closure ~ Monitoring
New Science & Technology Process Path
they can be transported to a HLW repository.] Once the waste is
removed from the tanks to the maximum extent that is technically and
economically practical, the next step is tank closure. The retrieval of
HLW must be adequate to ensure that residues can be declared as inci-
dental to reprocessing and can be disposed of as LLW (see Sidebar 6.1~.
Tank closure consists of characterizing the residues, filling the tanks
with cementitious material, and leaving the tanks in place. After closure,
monitoring activities must take place to identify any release of residual
waste into the environment.
HEW Generation and Cleanup
Strategies at DOE Sites
The first HLW was generated at the Hanford Site from reprocessing
reactor fuel for the production of plutonium used in nuclear weapons.
The Yucca Mountain site in Nevada is currently under consideration as an
HLW repository.
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
-
FIGURE 2. 1 Generic waste
remediation flow sheet used
throughout this report.
NOTE: LLW = low-level waste
OCR for page 18
Subsequently, H LW was generated at three other sites, I isted on p. 1 1 .
Most of the HLW was produced during World War 11 and the ensuing
Cold War.2 The generation of HLW and the cleanup strategy for each
site are briefly described in the following sections.
The Hanford Site
The first recovery of plutonium at the Hanford Site began in late
1944 using the bismuth phosphate separation process (Gephart and
Lu ndgren, 1 998~. Th is process recovered pi uton i u m, but not u ran i u m,
from the spent fuel and produced large quantities of waste. Later efforts
to recover uranium from the bismuth phosphate waste further changed
the nature of HLW produced. The higher-activity liquid wastes from bis-
muth phosphate reprocessing were neutralized chemically to reduce
their corrosiveness and stored in carbon steel underground tanks.
The first successful solvent extraction process for plutonium and ura-
nium recovery in continuous plant operation was the REDOX
(REDuction and OXidation) process, which began at Hanford in 1952.
The REDOX process generated a lower volume of waste, which was
chemically neutralized and stored in carbon steel underground tanks. In
early 1956, a new solvent extraction process, PUREX (Plutonium and
Uranium Recovery by Extraction), came into use at Hanford (Sidebar 2.1~.
PUREX used a different organic solvent and nitric acid. PUREX wastes
were highly radioactive, self-boiling, and were also neutralized and
stored in carbon steel underground tanks. Therefore, the tank wastes at
Hanford include many different chemical compositions and physical
characteristics, since different processes for plutonium recovery (as well
as other operations) were developed and applied during the operational
I ife of th is site. Table 2.1 presents a si mpl if fed i nventory of the wastes at
the Hanford Site.3 Only one percent of the mass of waste material is
radioactive, but this is enough to make the tank contents highly danger-
ous.4 Most of the radioactive species can be separated and immobilized
for disposal as solid HLW; the residues remaining can be treated as
other forms of waste.
2The HLW inventory may still grow slightly at the SRS because waste from
deactivation and decommissioning operations is stored in HLW tanks. A small
quantity of HLW is produced also from the reprocessing of part of damaged
research reactor fuel and of the fuel "blanket" from the Experimental Breeder
Reactor 11. For more details on research reactor spent fuel, see (NRC, 1998).
3HLW from the other sites has a different and less complex composition com-
pared to the Hanford waste.
4The radiation field inside the tanks can be as high as 10,000 red per hour
(100 gray per hour).
H ~ G H - L E V E E W A S T E
OCR for page 19
TABLE 2.1 Total Waste inventory Based on Process Know~ecge for Aid Hanforc Tanks. Only some of the radionu-
c~ides present are listed, and the inventory does not take into account the Decay in radioactivity since ~ 996.
Although water is not listed, it represents a significant fraction of HEW (>200,000,000 kilograms). NOTE: ~ curie
3.7 x 103° becquere~s.
=
Total Inventory Total activity Half-Life
Species (103 kilograms) (103 curies) (years)
Na+
Al3+
Fe3+
Cr3+
Bi3+
Zr as ZrO(OH)2
pb2+
Nix+
Ca2+
K+
TOC
NO3-
NO2-
CO32-
po43-
SO42-
Si as SiO32-
F-
49,170
8,372
1,872
994
492
214
279
183
620
521
1,818
50,430
1 4,093
4,836
3,795
3,226
662
575
Cl- 1,099
Sr-90/Y-90 0.44 123,203 29.12/(64 hours)
Tc-99 1.94 33 21 3,000
Cs-137/Ba-137m 0.54 91,966 30/(2.6 min)
Np-237 0.20 0.141 2,140,000
Pu-239 0.77 47 24,1 30
Pu-Total 0.77 225.59
NOTE:TOC = total organic carbon.The half-life of a radionuclide is the time after which half of atoms on average have disinte-
grated. SOURCE: Agnew et al., Appendix E (1997).
In addition to the evolution of the waste characteristics, the designs
of the waste storage tanks evolved as well. The earliest tanks were sin-
gle-shell tanks (SSTs) of carbon steel. Reliance on a single shell has
resulted in leakage of waste from some of the SSTs into the subsurface
soil. Of the 177 tanks at Hanford, 67 SSTs are known or suspected to
have leaked. The total quantity of leaked waste is estimated to be
between 3.4 and 6.7 million liters containing between 0.45 and 1.8
O v e r v i e w o f H LW C h a I I e n 9 e s a t D O E S i t e s
19
OCR for page 20
million curies (1.7 x 10~6 and 6.7 x 10~6 becquerels), although there
are large uncertainties in these estimates (Waite, 1991; ERDA, 1975;
Agnew et al., 1997~. The later double-shell tanks (DSTs) consisted of a
first tank surrounded by a secondary containment tank. Figure 2.2
shows a diagram of a SST and a DST. The external shell of the DST pro-
vides an additional barrier to leakage of waste into the subsurface soil.
There has been some leakage into the annulus space lying between the
shell liners of some DSTs. All the free liquid from SSTs has been trans-
ferred to DSTs, thereby greatly diminishing the potential impact of
future leaks. The risk from leakage in the future is expected to be asso-
ciated almost entirely with retrieval of waste from the SSTs, depending
on the retrieval method used. The vulnerability to leakage into the sub-
surface soil is an important reason for closure of the Hanford tanks. A
summary of the key facts and figures about the tank wastes at the
Hanford Site is presented in Table 2.2.
In its record of decision of 1997, DOE adopted a phased approach
to tank waste management at the Hanford Site (DOE, 1997~. Under
Phase 1, which will last approximately until 2018, 10 percent of the
tank waste and 20 to 25 percent of the radioactivity is slated for
retrieval and immobilization on a pilot scale for demonstration purpos-
es. After val idation of the processes, DOE wi 11 implement Phase 11, the
full-scale production phase, which will last approximately 30 years.
F/GURE2.2 Diagram ofa SST Figure 2.3 shows DOE's baseline plan for HEW management in
(left) and aDST(right) atthe Hanford.
HanfordSite.SOURCE:DOE- So far, the waste has been characterized to the extent feasible for its
Tanks Focus Area. heterogeneous contents. Waste retrieval operations have just begun.
H ~ G H - L E V E E VV A 5 T E
~ A ~
OCR for page 21
TABLE 2.2 Facts anc Figures about the Hanforc Site
Type of information
Data
Surface of the site 1,450 square kilometers (560 square miles)
Number of HLW tanks 177
Volume of HLW 208 million liters (54 million gallons)
Radioactivity About 200 million curies (7.4 x 1048 becquerels)
Characteristic of HLW Alkaline waste (pH approximately 12), saltcake, viscous liquid, sludge, and highly non
homogeneous waste
Types of tanks
Carbon steel, SST and DST
Waste sluicing by pumping liquid to entrain the sludge is the technique
chosen for retrieval of the waste from the tanks. According to the plan,
after retrieval, the waste will be sent to a pretreatment facility to sepa-
rate the solids, handled as HLW, from the liquid phase, which will be
decontaminated further to remove the hazardous constituents.5 The
solids-liquid separation (SLS) process is followed by sludge leaching of
the separated solids to dissolve non-radioactive components. Cesium,
strontium, transuranic elements, and technetium (as necessary) will be
removed from the I iqu id phase and sent to the H LW vitrification faci I ity
5Further details on pretreatment and immobilization processes are given in
Chapters 4 and 5.
-
I _| Solids-Liquid I Solid
it=
| Sluicing
. 1.-
· Radionuclide
Removal
Sl edge l l
Washing ~ Solid
LLW Vitrification
= Liquid
HLW Vitrification
·h~
HLW Interim Storage
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
FIGURE 2.3 Based on infor-
mation gathered from DOE,
this figure represents a sim-
plified view of the baseline
plan for waste management
at the Hanford Site.
Closed Tanks
· ·~
_=
LLW Disposal Vaults
Off-Site Transport
OCR for page 22
along with the HLW solids. The HLW glass canisters will be stored tem-
porarily on site for eventual disposal in a geological repository.
According to the 1994 revision of the Tri-Party Agreement (TPA)
between DOE, the U.S. Environmental Protection Agency (EPA), and
the Washington State Department of Ecology, LLW wil I also be immobi-
lized in glass and stored on site (TPA, 1998~.
The Idaho National Engineering and
Environmental Laboratory
Between 1 953 and 1 992, the I N EEL reprocessed spent nuclear fuel
mainly for recovery of the fissile isotope uranium-235. In the reprocess-
ing operation, spent nuclear fuel and its cladding materials (aluminum,
zirconium, stainless steel, and graphite) were dissolved in highly acidic
solutions (nitric acid, hydrofluoric acid, and sulfuric acid were used)
and Graphite was burned. INEEL's waste type, containing uranic and
O , ,, , O
. . . . . . . .. . . . . . .
transuranic Isotopes generated tram the reprocessing at spent fuel, IS
unique within DOE HLW sites in that it was stored while still in highly
acidic form.
Table 2.3 presents some general information on the wastes at the
IN EEL. The majority of INEEL's waste has been calcined to a granular
solid (ceramic), which is considered an interim storage form by the
State of Idaho. The calcine is stored in partially buried stainless steel
bins (designed to last 500 years) grouped inside concrete vaults.
Calcine waste can be processed further to convert it into a more con-
solidated long-term waste form. In addition, the INEEL has some liquid
sodium-bearing waste (SBW) to be immobilized. This waste is mixed
transuranic waste but, because of its radioactivity and large volume, it
TABLE 2.3 Facts anc Figures about the KNEEL
Type of information
Data
Surface of the site 2,300 square kilometers (890 square miles)
Number of HLW containers 1 1 tanks, 7 calcine vaults
Volume of HLW About 5.3 million liters (1.4 million gallons) of liquid waste and 3.8 million liters (1 million
gallons) of calcine
Radioactivity 520,000 curies of radioactivity (1.9 x 1 on becquerels) in liquid form and 24 million of
curies (8.9 x 1017 becquerels) of radioactivity in calcined form
Characteristic of HLW Very acidic (pH approximately 0), sludges and viscous liquid, calcine solids
Types of tanks Stainless steel, single shell
H I G H - L E V E L VV A 5 T E
~ A ~
OCR for page 23
is stored in underground tanks and managed as H LW (N RC, 1999b).
DOE has recently decided (Huntoon, 2000) to seek direct vitrification
as the preferred alternative for SBW. The preferred alternative for INEEL's
calcine is to continue work on characterization, pretreatment, and vitri-
fication options to be ready for shipment to a disposal facility by the
end of 2035. Immobilized calcine and SEW will be stored temporarily
on site awaiting eventual disposal in a geological repository.
Figure 2.4 shows the proposed baseline plans for the management
of HLW calcine and liquid at INEEL. DOE is still studying this strategy
and in 1999 was advised by the NRC on the different possibilities
(NRC, 1999b).
The Savannah River Site
Plutonium and tritium for nuclear weapons were produced at the
SRS near Aiken, South Carolina, by reprocessing spent fuel and targets.
The PUREX process was used from the time the SRS opened in 1952; as
FIGURE 2.4 Simplified base-
line plan for the manage-
ment of liquid and calcined
HLW at INEEL. Question
marks represent alternative
treatments for calcine. The
final decision is still pending
in Hanford, acidic H LW generated by PU REX was neutral ized for storage
in carbon-steel tanks. SRS also produced plutonium-238 and uranium- (Huntoon,~OOO).Formore
processes other than standard PU REX. The details, see NRC (1999b).
233 and recovered them by r
HLW Tanks
Calcine
Bins
?
?
Evaporator
Liquid Dissolver
Vitrification Facility
Grout Plant LLW Disposal
(On-Site or Off-Site)
HLW Interim Storage
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
Off-Site Transport
OCR for page 24
waste tanks contain HLW in the form of sludge, concentrated super-
nate, and saltcake. The sludge, of thin peanut butter-like consistency,
contains 67 percent of the radioactivity and represents 9 volume per-
cent; the supernate and the saltcake represent 91 volume percent and
3 3 percent of the red inactivity. HLW at the S RS is stored i n 51 tan ks, 2
of which have already been closed. Table 2.4 presents some of the prin-
cipal data on the tank wastes at the SRS. Figure 2.5 shows the baseline
pi an for waste management at th is site.
Radioactive vitrification operations at the SRS began in 1996, and
approximately 1,000 HLW glass logs have been produced to date. The
sludge is retrieved from different tanks and mixed in pretreatment
batches to lower (blend down) the concentration of components that
might add to the complexity of the vitrification step (more details are
given in Chapter 5~. The sludge is also pretreated by caustic extended
sludge washing to remove aluminum from HLW. Aluminum is known to
increase the viscosity of the molten glass, which complicates pouring
the glass melt into the canisters and also considerably increases the
total volume of immobilized HLW. The SRS is still in the process of
selecting a preferred method to separate cesium, strontium, and
actinides from the saltcake and supernate, after the previous process
(large-scale in-tank precipitation) was abandoned because of technical
difficulties (NRC, 2000b). An NRC panel is currently advising DOE on
viable alternatives (NRC, 2001 a). The leached sludge solids are then
combined with cesium, strontium, technetium, and transuranic ele-
ments from the liquid decontamination step, and sent to the vitrification
pi ant, the Defense Waste Processi ng Faci I ity (DWPF). The decontam i-
nated LLW filtrate is then grouted (i.e., immobilized by cementation).
The HLW glass canisters produced are temporarily stored on site wait-
ing for eventual disposal in a geological repository.
TABLE 2.4 Facts anc Figures about the SRS
Type of Information
Data
Surface of the site 800 square kilometers (300 square miles)
Number of HLW tanks 51 (2 tanks already closed)
Volume 125 million liters (33.4 million gallons)
Radioactivity 240 million curies (8.9 x 1048 becquerels) in the sludge and 180 million curies (4.4 x 1049
becquerels) distributed between the supernate and the saltcake
Characteristic of HLW Alkaline (pH approximately 14)
Types of tanks The carbon steel tanks are differentiated into Type I, II, III, and IV tanks, according to their
characteristics. Of the 51, only 27 tanks (type ill) are double shell tanks
H I G H - L E V E L VV A s T E
~ A ~
24
OCR for page 25
Salt
._..
Sludge
l
Cs, Sr, and
TRU
remove I
· | Extended
Sludge | Slurry
Raffinate
Precipitate
Saltstone
Grout Plant
LLW Vaults
HLW Vitrification HLW Interim Storage HLW Glass Logs
1
Failed Melters
The West Valley Demonstration Project
The WVDP is situated near Buffalo, New York. The State of New
York owns this site, not DOE. However, by the terms of the WVDP Act
of 1980, DOE is responsible for retrieving the HLW on site, removing it
for disposal, and decommissioning all parts of the site used for defense-
related tasks. The U.S. Congress appropriates funds to DOE for this
work, and New York State shares the costs. Because the WVDP has
almost completed its HLW cleanup operations, the recommendations in
this report are not as relevant to this site as they are for the Hanford,
Savannah River, and INEEL sites. However, the committee gathered
information on the WVDP because of the "operational experience"
acqu i red at th is site.
From 1966 to 1972, commercial spent nuclear fuel was reprocessed
at the WVDP to recover uranium and plutonium by a licensed, com-
mercial fuel reprocessing plant. The reprocessed reactor fuel included a
substantial quantity of typical light water reactor (LWR) fuel, a single
thorium-based core from an LWR, and a substantial amount of low-bur-
nup defense spent fuel from the Hanford N Reactor, provided as a base
load to support this first commercial reprocessing plant. The N Reactor
fuel and the typical LWR fuel were reprocessed to recover uranium and
plutonium using the PUREX process; the thorium-based core was
reprocessed to recover uranium-233 and uranium-235.
The single tank farm at the WVDP contains two carbon steel tanks
(each with a capacity of 2.4 million liters), one for storage of alkaline
waste and one spare, along with two stainless steel tanks (each with a
capacity of 1 10,000 liters), one for storage of the acid waste from the
Off-Site Transport
FIGURE 2.5 Simplified base-
line plans for management
of the waste at the SRS. The
Defense Waste Processing
Facility (DWPF) is the vitrifi-
cation plant at the site. The
waste supernate is the main
source of cesium, strontium,
and transuranic elements
(Cs, Sr and TRW).
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
OCR for page 26
TABLE 2.5 Facts anc Figures about the WVDP
Type of Information
Data
Surface of the site
Number of HLW tanks
Volume of HLW (initial conditions)
Radioactivity
Characteristic
Types of tanks
0.3 square miles (0.8 square kilometers)
2 carbon steel tanks of 2.4 million liters (660,000 gallons), 2 stainless steel tanks of 1 10,000
liters (30,000 gallons)
2.4 million liters (0.66 million gallons), after blending a small amount of acid waste into
the alkaline waste
0.6 million curies (2.2 x 1 on becquerels)
Alkaline waste (pH approximately 14)
Double shell, carbon steel
thorium-based core and one spare. Table 2.5 presents some of the prin-
cipal data on the tank wastes at the WVDP. All of the waste was blend-
ed together in one of the larger storage tanks. All of the tanks were inte-
grated into a recycling pretreatment system. As a result, there was a sin-
gle waste composition to process.
Figure 2.6 shows a simplified baseline plan for the management of
HLW at the WVDP. The supernatant liquid was removed and pretreated
with zeolite, a granular aluminosilicate ion-exchange material, to
remove the radioactive cesium and strontium, yielding a separate LLW
salt solution. The sludge remaining at the bottom of the tanks was
FIGURE 2.6 Simplified base- washed, and the I iqu ids were treated with zeal ite. The remai n i ng H LW,
line plans for the WVDP consisting of the washed sludge and the zeolite used for pretreatment,
basedoninformationpro- was vitrified in borosilicate glass using a slurry-fed Joule melter of a
videdby DOE design similar to that employed at the SRS. The LLW was further evapo-
Washed sludge ~
Supernate Supernate LLW
fig Program ~:
Zeolite
HLW Glass
HLW Vitrification Interim Storage
Possible On-Site
I Relocation
Off-Site Transport
H ~ G H - L E V E E W A S T E
26
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rated and blended with cement for interim storage on site. To date, the
WVDP has almost completed6 retrieval and vitrification of its HEW. The
255 glass canisters produced are stored temporarily on site, awaiting
eventual disposal in a geological repository. The WVDP is now perform-
ing final flushing of the tanks to prepare for shutdown of the waste
melter.
6More than 99 percent of the alpha emitters and more th
beta and gamma emitters have been retrieved to date.
a
in 97 percent of the
O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
27
Representative terms from entire chapter:
hanford site
