Treatment, Stabilization,
and Transmutation of
High-Level Wastes
“My uncle’s enthusiasm, always a little more
than was required, was now excusable.”
—Henry Lawson in “Journey to the Center of
the Earth” by Jules Verne.
Composition of reprocessing
wastes per 1,000 kg of SNF
(Murray, 2003)
Fission products 28.8 kg
U 4.8
Pu 0.04
Np 0.48
Am 0.14
Cm 0.04
Reprocessing 68.5
chemicals
Reprocessing wastes
The weight of reprocessing waste is about
one-tenth of the weight of SNF.
Sr-90 and Cs-137 are the major problems
during the first few centuries of waste
storage. Can they be eliminated from
HLW? Will be discussed later. For now. . .
by definition, reprocessing wastes are HLW
Reprocessing wastes
Aqueous/nitric acid solutions that contain fission
products such as Cs, Sr, Zr, Ni, La and others.
Derived from SNF from military applications in the
US. Because we do not reprocess SNF, HLW
treatment research has not been a major priority
in the US.
In general, these are high-level liquid wastes that are
stored in underground tanks.
Typical first-cycle raffinate
Treating High-Level
Liquid Wastes
Reduce the volume of waste
Calcination—heat the liquid at a high temperature
to evaporate moisture and volatile constituents
without fusing the residue (a granular form is
easier to handle).
Fixation, immobilization, —adding some agent to the
calcinate to reduce the potential for leaching
(waste stabilization) as another safety factor.
Immobilization of calcinate/HLW
Basic approach is to mix and isolate the
calcinate with
1. Amorphous glass
2. Ceramic-formation minerals
3. Glass-ceramic combinations
4. Basalt glass-ceramic combinations
5. Cements and concretes
Immobilization of calcinate/HLW
Immobilization of calcinate/HLW
Concrete and cement are used for the
immobilization of LLRW such as in the
Netherlands.
Problems with radiolysis of water when used with
HLW:
H2O + alpha, beta, and gamma radiation yields H+,
OH-, H2 (gas), O-, H2O2, H3O, and H2O+
Low-moisture concretes have been proposed, but
application is complicated. Research abandoned.
Immobilization of calcinate/HLW
Best understood and widely used additive is
borosilicate glass.
70 to 81% SiO2
7 to 13% B2O3
4 to 8% Na2O
2 to 7% Al2O3
Immobilization of calcinate/HLW
Borosilicate glass is more resistant to thermal
shock such as from the heat from
radioactive decay than ordinary glass.
Also called “Pyrex glass” as in lab glassware.
Ultimate goal is to immobilize and make
insoluble radionuclides in HLW
Immobilization of calcinate/HLW
Solid calcinate is mixed with ground
borosilicate glass. Sometimes the liquid
waste is fed directly into the glass-melting
furnace.
The mixture is heated to about 2,012º F.
Molten glass is then poured into a steel
container, and allowed to cool.
Commercial treatment
Example: The Advanced Vitrification Method
(France) on pages 120-125.
Calcination and vitrification are coupled (two-
stage process).
Also used in the UK, Japan, and Germany.
Storage of high-level
vitrified waste
Ceramic Wasteforms
Synroc “Synthetic rock”
Invented in 1978 at the Australian National
University
Basic premise: to imitate geologic storage of
radionuclides by using a mixture of
minerals and heat to create a ceramic
wasteform.
Dr. Alfred E. Ringwood
(1930-1993)
Australian Geochemist
Fellow, Australian
Academy of Science,
Fellow, American
Geophysical Union,
Fellow, Royal Society of
London
Synroc
Use titanium oxide minerals that have stored U, Th,
and other rare earth elements (REE) for million
years.
Hollandite (BaAl2Ti6O16)
Rh, Ru, can replace Al and Ti
Zirconolite (CaZrTi2O7)
Sr, REE, and others can replace Ca
Perovskite (CaTiO3)
Sr, Cm, Am, Pu, and others can replace Ca or Ti
Hollandite
Zirconolite
Perovskite
Process
About 10% calcine is added to a mixture of ground
hollandite, perovskite, and zirconolite (and others,
depending on the specific formulation)
Hot press at 1,100 to 1,350° C. Radionuclides
become part of mineral structures.
Allow to cool and recrystalize in sealed containers.
Containers then are to disposed in a geologic
repository.
Synroc
Different formulations
Synroc being tested
The Synroc technology is not being used
commercially at this time. DOE stopped
funding ceramic-alternative research in
1983.
Being used at a large demonstration project
the Sellafield Plant in England
(reprocessing wastes).
Leaching of wasteforms
Whether glass or polyphase ceramic forms are used
then placed in canisters and overpack, what
happens if the canister is comprised?
Groundwater could flow through the disposal area.
Can radionuclides leach from these immobilized
HLW forms?
Leaching Behavior
Everything is soluble at some level
Extent of leaching depends on
*pH and chemical composition of water
*glass/ceramic composition,
*surface chemistry and area,
*flow rate (static vs dynamic conditions),
*time.
Standardized tests used
Leaching Tests
MCC-1. Static (no mixing)
Developed at the Material Characterization
Center at Pacific Northwest Laboratory
(WA).
Monolith sample placed in a volume of
distilled water at 40°, 70° or 90º C for 28
days (or longer).
MCC-1
The volume of water (V) is
V = 10 x the surface area (0.1/cm)
If the surface area of the monolith is 300 cm2, the
volume of water would be 3 L.
Also called ASTM C-1220-98
MCC-2 (150º C)
Leach Rate (LR)
LR = (mi - mf)/(Sa x t)
mi = initial mass
mf = final mass
Sa = surface area
t = time Typical units: g/m2-day
Normalized leach rate
NLRj = m2/(Sa x t x m1)
m2 = mass of component j leached
m1 = mass of component j initially present
Example of MCC-1 data
Other standardized
durability tests
MCC-2. Like MCC-1 but at 110, 150, and 190 C.
MCC-3. Powered sample mixed
with a fixed volume of extracting
solution (“solubility limited”)
MCC-4. Low flow rate
MCC-5. Soxhlet Extraction (most
aggressive)
Other standardized
durability tests
The Product Consistency Test (PCT)
for nuclear mixed waste glasses and
multiphase glass ceramics
ASTM C 1285-02
Method A, 7 day, distilled water leachant at
90° C, disaggregated sample, static in
stainless steel vessels.
Method B, static, variable conditions.
NEW!
Transmutation of
Radioactive Waste
Basic concepts
Transmutation. Transformation of one
isotope into another by neutron absorption.
Products: Next heavier isotope or two or more
fission products.
Fissile: Fissionable by thermal neutrons.
235U is fissile whereas 238U is not.
Energy production results in
transmutation
235U + η → 236U* → fission products + η +
β + γ
The fission products include 90Sr (28.8 years)
and 137Cs (30.1 years). And by neutron
capture
238U + η → 239U* → 239Np + β- →
( 23.5 min) (2.35 d)
239Pu + β-
(24,400 y)
Transmutation as a
curse and cure?
Transmutation creates waste management
issues with respect to either once-through
SNF or in reprocessing SNF.
Can transmutation be applied to SNF to
reduce it’s radiotoxicity by converting
radionuclides with long half-lives to ones
that decay more quickly?
The Roy Process
Some people think so. Several transmutation
processes have been proposed. Take for
example “The Roy Process.”
In 1979, the late Dr. Radha Roy announced he
“had invented a new method to render all
radioactive waste elements, including
plutonium, into non-radioactive
elements.”
The Roy Process
“With the Roy Process, high level nuclear
waste can be neutralized and totally
eliminated at each reactor site, where the
waste is now stored in cooling ponds. When
treated with the Roy Process, these unstable
radioactive isotopes rapidly decay into
stable, non-radioactive elements . . .”
From:
http://members.cox.net/theroyprocess/
Realities of Transmutation as a
Waste-Treatment Technology
Transmutation of persistent fission products:
99Tc + η → 100Tc → 100Ru
(2.12 x 105 y) (16 sec) (Stable)
129I + η → 130mI → 130I → 130Xe
(1.6 x 107 y) (9 min) (12 hours) (Stable)
These are examples of desirable reactions.
Realities of Transmutation
The process of transmutation can also initiate
undesirable side reactions that produce new
radionuclides with long half-lives. For example,
133Cs + η → 135Cs
(stable) (2.3 x 106 y)
241Pu + η → 242Pu
(13.2 y) (389,000)
35Cl + η → 36Cl
(stable) (3.1 x 105 y)
Realities of Transmutation
Some fission and activation products do not
transmute significantly because their cross
section for capturing thermal neutrons is
too small. The term “cross section” is the
probability of a nuclear reaction resulting in
transmutation. Some of these products
include 79Se, 126Sn, 36Cl, and 14C. This also
includes 90Sr (1.34 barns) and 137Cs (0.176
barns).
Realities of Transmutation
Transmutation cannot be applied to solid
SNF. Because SNF contains 235U and 238U,
the addition of thermal or fast neutrons
would produce more Pu which is not the
goal.
Transmutation must be coupled with chemical
separation of the radionuclides into
different wastes streams.
Separation and Transmutation
Under study:
Aqueous chemical separation (PUREX,
UREX, TRUEX, etc.) followed by
transmutation in light water reactors or fast
breeder reactors.
Pyroprocessing separation followed by
transmutation in light water reactors of fast
breeder reactors.
Pyroprocessing
Current research results
“SNF is placed into a cathode basket that is
then immersed in a pool of molten LiCl-
Li2O. When a sufficiently high electrical
potential is applied, oxygen gas bubbles are
evolved at the anode, and actinide oxides
are reduced to metals at the cathode. Rare
earth fission products appear to remain
unreduced in the basket. Alkali and alkaline
earth fission products (Cs, Sr, Rb, and Ba)
partition into the salt, presumably as
chlorides” (Simpson, 2006).
Still have waste issues . . .
“The accumulation of these alkali and
alkaline earth fission products in the salt
will require periodic disposal of the salt into
a waste form that can be safely stored for
approximately 200 years to allow decay of
the 137Cs and 90Sr. Salt can be simply
removed from the process once it reaches a
contamination limit, blended with zeolite,
and formed into a ceramic waste.”
(Simpson, 2006).
Barriers to Separation and
Transmutation
Separation requirements for transmutation:
U and Pu must be separated (PUREX).
Cs and Sr must be separated (under study).
Methods for separating Am, Cm, Np, and
turning them into targets for transmutation
still at the experimental stage.
All extractions need to be optimized to extract
nearly all of each radionuclide.
Barriers to Separation and
Transmutation
Any S-T approach would increase the volume
of LLRW.
What is the best source of neutrons for S-T?
Light-water reactors? Fast reactors
breeder? Coupled with accelerators?
(Accelerator Transmutation of Waste—
ATW)? Generation IV reactors?