Синтез и свойства неорганических соединений
ISSN 0036-0236, Russian Journal of Inorganic Chemistry, 2019,
Vol. 64, No. 13, pp. 1611–1624. © Pleiades Publishing, Ltd.,
2019.
On selection of matrix (wasteform) material for higher activity
nuclear waste immobilisation
©2019 г. C. M. Jantzen1, M. I. Ojovan2,*
1The University of South Carolina, USA, Aiken, South
Carolina
2Imperial College London, UK, London, SW7 2AZ, South Kensington
Campus, Exhibition Road
*e-mail: [email protected]
Selection of wasteform materials for higher activity nuclear
waste containment is considered. Utilisation of materials such as
glasses, ceramics, glass composite materials and cements is
discussed as practiced in different countries. Emphasis is on
multiple parameter approach on selecting the wasteform where the
durability is not solely the most important characteristic.
Ключевые слова: nuclear waste, wasteform, glass, ceramics,
cement
DOI: 10.1134/S0036023619130047
Introduction
There is a generic consensus that the radioactive waste end
points (e.g. storage and final disposal) and conditioning methods
(e.g. immobilization and packaging in containers) depend on the
level of radioactivity and waste radionuclide lifetimes [1, 2].
Safety of storage and disposal is achieved mainly by concentration
and containment. Containment uses many barriers around the
radioactive waste to restrict the release of radionuclides into the
environment e.g. the multi-barrier concept. The restricting
barriers can be either natural or engineered, e.g. obtained via
processing of radioactive waste and using relevant wasteforms. The
generic approach is to use more reliable barriers for more
hazardous waste, including engineered barriers which result from
the radioactive waste treatment and conditioning process. Very low
level waste (VLLW) typically does not require immobilization,
although some forming packaging could be used. Low level waste
(LLW) is typically immobilized using cements, although glasses
provide better retention of contaminants. Intermediate level waste
(ILW), with a higher hazard than LLW, can be immobilized using
glasses, bitumen, cements and geopolymers. LLW and ILW can be
packed in high-integrity metal and concrete containers. The most
hazardous waste, such as high level waste (HLW), requires the most
durable and reliable wasteforms, e.g. ceramics and glasses and
containers such as stainless steel canisters for vitrified HLW and
copper containers for spent nuclear fuel declared as waste
(SNFW).
Immobilization of radioactive wastes in a stable, solid matrix
reduces the potential for migration or dispersion of radionuclides.
The product of incorporating a waste into a suitable matrix is a
wasteform. Immobilization is the conversion of a waste into a
wasteform by solidification, embedding or encapsulation. It
facilitates handling, transportation, storage and disposal of
radioactive wastes. Immobilization of waste is achieved by its
chemical incorporation into the structure of a suitable matrix
(typically glass or ceramic) so it is captured and unable to
escape. Chemical immobilization is typically applied to HLW.
Encapsulation of waste is achieved by physically surrounding it in
materials (typically bitumen or cement) so it is isolated, and
radionuclides are retained. Physical encapsulation is often applied
to ILW but can also be used for HLW especially where chemical
incorporation of radionuclides in the surrounding matrix is also
possible. Choosing the wasteforms and packages in which to
immobilize higher activity radioactive wastes (e.g. ILW and HLW) is
a crucial decision depending largely on the waste streams and
following proven technologies used elsewhere.
Purpose of a Wasteform
The purpose of a wasteform is to immobilize nuclear waste in a
solid matrix that is chemically and physically more durable, and
more heat and radiation resistant than the waste itself [3-5]. The
wasteform is meant to stabilize the waste during packaging, an
interim storage period, transportation, emplacement within a
geological repository, and post closure disposal period. Notable
that deep geologic disposal is considered to be the most suitable
option for disposing higher activity radioactive wastes worldwide
because of the predicted effectiveness of many natural geological
systems in reducing the transport of radionuclides that can be
augmented by robust engineered barriers, thereby enhancing disposal
facility performance.
Within the repository, the wasteform is one part of a multiple
engineered barrier system (EBS). During storage the wasteform is
the primary barrier preventing the release of radionuclides into
the environment while during post closure disposal, the wasteform
will reduce the release of radionuclides from breached and
compromised containers that could result due to corrosion,
earthquake, human intrusion, igneous intrusion (volcano), or other
disruptive phenomena.
The natural barrier system (NBS) primarily establishes the
geochemical environment wherein the engineered barriers reside. In
some repository concepts, chemical buffers are included in the EBS
to reduce the mobility of radionuclides (e.g., using clay
backfill), reduce the solubility of certain radionuclides (e.g., by
adding phosphates), or reduce the wasteform degradation rates
(e.g., by adding reactive materials such as ductile iron). The
wasteform, as part of the EBS, is the source of radionuclides and
controls the release of those nuclides when in contact with water.
Figure 1 shows schematically the various barriers including the
wasteform, waste package, and NBS.
Figure 1.
The performance of a repository is assessed using a performance
assessment (PA) model in which the effect of each of the barriers
on the release and migration of radionuclides is taken into account
and coupled to calculate the radionuclide mass flux that could
potentially reach the accessible environment should the engineered
barriers fail. Typically, interactions such as those among thermal,
hydraulic, mechanical, and chemical (THMC) processes are
sufficiently complex that models fully describing their behaviour
are not fully coupled and the processes are abstracted into an
overall PA model. Each significant variable is assigned a
probability distribution. Dose conversion factors are then applied
to determine the impact on human health and the environment
resulting from radionuclides reaching the environment.
The resistance of the wasteform to aqueous corrosion and release
of radionuclides in the disposal environment – chemical durability
– is a critical parameter.
Immobilization and Matrix Requirements
Wasteform matrix selection and optimization cannot be
dissociated from selection of technology and waste characteristics
and is a compromise between various constraints including
flexibility, ease of processing, waste loading, chemical
characteristics of the waste and required lifetime performance
which will be affected by irradiation and water alteration [6]. The
waste hosting matrix is expected to possess certain features that
are desirable for easy processing, handling, storage and eventual
disposal:
•Fabrication should be accomplished under reasonable conditions,
including thee lowest temperatures to enhance the life of
processing equipment and reduce volatilization of fission products
and ideally in an air atmosphere, using well established methods to
minimize worker dose and capital cost of the facility.
•Adequate viscosity if the wasteform is molten during processing
to facilitate homogenization and pouring.
•Adequate electrical conductivity in the case of molten liquids
while in Joule Heated Ceramic and Cold Crucible Melters.
•Good mechanical strength and shock resistance so that handling
during transportation, storage or disposal is safe.
•Compatibility with the storage container (canister).
•Compatibility with processing equipment (melting temperature,
corrosion, refractory, etc.).
The general requirements for matrices which incorporate HLW in
wasteforms include, but are not limited to [7]:
•High waste loading: the system must be able to accommodate a
significant amount of waste (typically 20-35 %) to minimize the
wasteform volume, saving on storage and transportation costs and
space in a potential repository (waste minimization concept).
•High radiation stability: the wasteform should show a high
tolerance to radiation effects from decay products to reduce phase
changes (which may be detrimental to durability) and volume
swelling (resulting in cracking).
•Chemical flexibility: if required, the system should be able to
accommodate a mixture of radionuclide and other contaminant
species, with minimum formation of secondary phases which may
compromise wasteform performance or processability.
•Durability: the wasteform should be resistant to alteration
under conditions relevant to storage and disposal to minimize
release of radiotoxic species.
Wasteform durability is measured using laboratory and field
tests [5, 8-10] which measure corrosion in aqueous environments,
the better of which try to match the test to the likely conditions
in an underground repository and/or determine parameters for PA
durability rate law models. Laboratory testing of vitreous
wasteform durability, both short- and long-term has been reviewed
in [11-13]. Wasteforms also need to be acceptable for disposal in a
repository and a range of procedures are used by different
countries, so-called Waste Acceptance Criteria (WAC) or Waste
Acceptance Product Specifications (WAPS).
The long-term behaviour of a waste disposal facility is a
function of the entire disposal system, including the wasteform,
engineered barriers, and surrounding environment. In order to
assess the ability of a given disposal concept to meet regulatory
requirements it is necessary to consider the influence of each of
these system components on short-and long-term performance. This is
accomplished through the PA process. For HLW many countries are
proposing long storage life for the glass wasteforms in steel
canisters during geological repository siting and preparation.
During that time a great many of the radionuclides will decay
leaving the long lived radionuclides as the primary sources that
need be considered in a PA. Figure 2 is a schematic of a generic
HLW repository which shows the relative role of the wasteform, the
role of the multiple barriers (canisters, containers, overpacks,
and casks) in the waste disposal system.
Figure 2.
Ultimately the role of the repository or disposal environment is
to isolate the waste from the biosphere until all the barriers have
failed, i.e. a canister is anticipated to remain intact for time
periods such as 100000 years, at which time almost all of the
radionuclides will have decayed. While the wasteform is the source
term and should be as durable as reasonably possible, multiple
barriers must corrode before the wasteform will be exposed to
groundwater.
The final evaluation of any wasteform is done using safety
assessment (SA) and/or PA cases. SA of nuclear waste disposal
evaluates the wasteform and repository performance and its
radiological impact on the environment and humans. SA results in
data demonstrating either compliance with safety standards or
unacceptability of the proposed concept. PA provides data on the
performance of all repository subsystems and the repository itself.
PA results in data which are used to improve the overall safety and
to demonstrate compliance with safety standards and performance
targets. PA comprises scenario development and potential impact
calculations. As a rule, the impact is characterised by the most
probable doses arising from radionuclides which may be released and
interact with the biosphere. Scenarios considered are not limited
to the most probable but also include abnormal events.
Mathematical models are key components of the SA and PA of waste
disposal. A chain of mathematical models is used to describe the
long-term degradation of the engineered barriers and the eventual
release of radionuclides into ground water and their transport to
the biosphere. All studies to date have shown that disposal of
radioactive waste including geological disposal, can be safe;
adequate isolation from the environment can be assured for hundreds
of thousands of years and, thereafter, radionuclide releases are
negligible in comparison to natural radiation exposures.
Types of Matrix
Once radioactive waste is generated its management involves a
series of steps aimed at its eventual permanent disposal. These
include: (1) pre-treatment which may involve collection,
segregation and decontamination, treatment aimed at volume
reduction and where possible radionuclide removal, (2) conditioning
which can involve physical and chemical immobilisation of waste
radionuclides into the structure of a stable solid (such as glass,
glass composite material or ceramic), or physical encapsulation in
a matrix (such as cement or bitumen), followed by packaging often
in a steel container. The packaged waste can then be placed in
temporary storage under control and where retrievability is ensured
until a suitable permanent geological disposal facility is
available [6, 10].
Choosing a suitable wasteform (matrix) to use for nuclear waste
immobilisation is not easy and its durability is not the sole
acceptance criterion. In any immobilisation process where
radioactive materials are used, the process and operational
conditions can become complicated, particularly if operated
remotely and equipment maintenance is required. Therefore, priority
is given to reliable, simple, rugged technologies and equipment,
which may have advantages over complex or sensitive equipment and
processes. A variety of matrix materials and techniques are
available for immobilisation. The choice of the immobilisation
technology depends on the physical and chemical nature of the waste
and the acceptance criteria for the long-term storage and
geological disposal facility (GDF) to which the waste will be
consigned. A host of regulatory, process and product requirements
has led to the investigation and adoption of a variety of matrices
and technologies for waste immobilisation. The main immobilisation
technologies that are available commercially and have been
demonstrated to be viable for higher activity radioactive wastes
including ILW and HLW are vitrification, cementation and
ceramification [2, 5, 10, 14, 15]. Reports comparing the properties
of crystalline ceramics to glass generated considerable controversy
[16, 17] in the 1970s and 1980s as “low leachability” had become
the main criteria for comparison of vitreous and crystalline
wasteforms rather than the multiple criteria presented above.
Although SYNROC (SYNthetic ROCk) ceramics had been reported to have
superior product quality to glass wasteforms, the high temperature
processing, especially for SYNROC-D for defence wastes, formed an
intergranular glassy phase from the alkali-and silica in the
wastes. This intergranular glass limited the ceramic product
stability and durability as often the radionuclides (137Cs and
90Sr) migrated to the glassy phase [18-21]. While ceramics were
credited with having higher chemical durability than glasses, if
the radionuclides were incorporated in the intergranular glassy
phases, they were determined to leach at similar rates as those
from glassy wasteforms [22]. This illustrates the danger of
considering only one of the general requirements for matrices which
incorporate HLW as given above, instead of optimizing the general
requirements against one another for the waste stream being
considered.
Wasteform Options and Requirements
Nearly every reprocessing nation uses glass as the primary HLW
form; most of which use alkali-borosilicate glass; with the
exception of Russia which produces an alkali-aluminophosphate glass
[2, 5, 10]. Glasses were selected in the 1970s primarily for the
following reasons (after References [10, 23]):
· A known and reliable large scale non-nuclear commercial
technology;
· More cost-effective process than ceramics and other HLW
options;
· High solubility of waste components in the glass;
· Tolerance to variation in waste composition;
· Low raw materials costs;
· Sufficiently high durability to assure long-term
performance;
· Continuous, high-throughput, operation of glass melters;
· High resistance to damage from radiation and radioactive
decay;
· Reasonably well understood corrosion/release mechanisms for a
homogeneous wasteform;
· Existence of natural analogues.
While borosilicates are the predominant glass wasteform
phosphate based glasses are used in Russia – a glass similar to
alkali-borosilicate with a network of PO43- tetrahedra instead of
SiO44- tetrahedra: usually aluminophosphate or iron phosphate. The
advantages of these glasses include higher solubility for a number
of waste components such as P, Mo, Tc, Fe, Cr, S, and halides. Two
disadvantages are the relatively high corrosivity of the melt and
higher tendency to crystallize during cooling. Several recent
studies suggest that these short-comings have been overcome by new
glass formulations [24-28]. However, other wasteforms have several
advantages including higher durability in the case of some ceramics
or ease of fabrication for wasteforms made at low-temperature.
Vitrification is a particularly attractive immobilisation route
because of the high chemical durability of the glassy product. HLW
has been incorporated into alkali borosilicate or phosphate
vitreous wasteforms for many years and vitrification is an
established technology in France, Japan, Russia, USA and the UK.
Large streams of Low and Intermediate Level Waste (LILW) are
planned to be vitrified in the USA, South Korea and Russia [10].
The glass structure is able to accommodate most of the Periodic
Table and radioactive species can go into the glass network or
other sites [29, 30]. For example, alkalis and alkaline earths
(137Cs, 90Sr) sit on the large Na sites in a sodium borosilicate
glass. The key issue during production of vitreous wasteforms is to
balance the waste loading (to minimise volume), viscosity (ease of
pouring), volatilisation (of high vapour pressure species) with
wasteform durability and simplicity and economics of the process.
For a variety of reasons some species present in wastes including
tetravalent actinides, 99Tc, 36Cl, 129I and 14C are inherently
difficult to immobilise and dependent on melt pool
REDuction/OXidation (REDOX) behaviour. In particular, halogens have
low solubility in glasses and are volatile at typical glass
processing temperatures of 1150-1200C while the melting
temperatures of components such as Al2O3 in a HLW can create
glasses that are well above these glass processing temperatures,
leading to bubbles or crystals in the glass; the latter becoming de
facto Glass Composite Materials (GCMs) [10, 14]. As illustrated in
Table 1 a range of wasteform matrices can be considered for
immobilizing higher activity waste e.g. HLW and potentially ILW.
The choice of wasteform depends on the physical and chemical nature
of the waste; the criteria for storage, transportation, and
geological disposal; and a host of regulatory, policy, process, and
cost considerations.
Table 1
Recent reviews of the developments in wasteform research have
been provided in [31-41]. The most recent interest has been
associated with the desire to create new nuclear materials as part
of new or advanced nuclear fuel cycles that chemically process the
Spent Nuclear Fuel Waste (SNFW) [42].
At the other end of the spectrum shown in Table 1 the use of
predominantly crystalline ceramic wasteforms (ceramification) has
also been proposed including single-phase ceramics such as zircon
or zirconolite to accommodate a limited range of active species
such as Pu and multiphase systems such as SYNROC to accommodate a
broader range of active species [43]. To date these systems have
not been extensively used to immobilise active waste.
Recently, however, there has been a trend to systems
intermediate between the “completely” glassy or crystalline
materials [7, 10, 14, 29, 44 ]. An illustration of nuclear
wasteforms used and developed for industrial application is given
in Figure 3.
Figure 3.
Typically nondurable crystals are Na2SO4, Na2MoO4, NaF and NaCl
while durable crystals include BaAl2Ti6O7, CaZrTi2O7, CaTiO3,
CaMoO4, TiO2, and spinels such as (Ni,Mn)(Fe,Cr)2O4.
GCMs sit between the fully amorphous glasses and fully
crystalline ceramics and contain both glassy and crystal phases.
GCMs include: (i) glass ceramics where a glassy wasteform is
crystallised in a separate heat treatment (ii) materials in which a
refractory waste such La2Zr2O7 is encapsulated in glass to
immobilise minor actinides and (iii) wasteforms which form glass
and crystals during processing such as those made at La Hague by
cold crucible melting which partly crystallise on cooling [45]
derived from difficult wastes such as the French HLW
U-Mo-containing materials.
GCMs may be used to immobilise long-lived radionuclides (such as
actinide species) by incorporating them into the more durable
crystalline phases, whereas the short-lived radionuclides may be
accommodated in the less durable vitreous phase. Historically,
crystallisation of vitreous wasteforms had been regarded as
undesirable as it has the potential to alter the composition (and
hence durability) of the remaining continuous glass phase which
would (eventually) come into contact with water.
However, there has been a recent trend towards higher
crystallinity in ostensibly vitreous wasteforms so that they are
more correctly termed GCMs. This is particularly apparent in the
development of hosts for more difficult wastes or where acceptable
durability can be demonstrated even where significant quantities of
crystals (arising from higher waste loadings) are present such as
the high sodium Hanford wastes. Acceptable durability will result
if the active species are locked into the crystal phases that are
encapsulated in a durable, low activity glass matrix. The GCM
option is currently being considered in many countries including
Australia, France, Russia, South Korea, UK and USA.
Cement matrices [46, 47] are included in this section because of
the IAEA definition of HLW being anything slated for geological
disposal. As a result much waste, which in the UK would be regarded
as ILW as it is not heat generating, is regarded as HLW. Hydraulic
cements are inorganic materials that have the ability to react with
water under ambient conditions to form a hardened and
water-resistant product. The most common cements are those based on
calcium silicates, such as the Portland cements. Cementation of
radioactive waste has been practised for many years basically for
immobilisation of Low Level Waste (LLW) and Intermediate Level
Waste (ILW) and commonly (in France and the UK and US in
particular) composite systems are used with high proportions of
blast furnace slag or pulverised fuel ash compared to Portland
cement, typically a 9:1 ratio to reduce setting heat generation and
improve wasteform durability. The waste is physically encapsulated
by the (amorphous) calcia-silica-hydrate gel and other phases in
the microstructure, hence its inclusion next to GCMs in Table 1. A
summary of the different wasteforms currently used or under
development in various countries is given in Table 2.
Table 2
Glasses
The most common matrix for immobilizing higher activity wastes
(HLW and ILW) is glass mainly due to its ease of processing,
flexible structure able to accommodate a range of radionuclides,
and high durability [5, 10]. Generically phase separated glasses
(glass-in-glass phase separation) should be avoided. Control of
REDuction/OXidation (REDOX) of the melt during the processing also
helps retain radionuclides. Vitrified HLW is predominantly glass
although it may contain some crystals and bubbles. Several types of
glasses have been proposed for HLW conditioning, however two types
of glasses namely borosilicate and phosphate glass are currently
used for engineering scale HLW vitrification in various countries
(Table 2). Borosilicate glass is widely accepted as a waste matrix
due to the acceptable chemical durability of glass. The formation
temperature is moderate, in the range 950-1100 oC. The modifiers
including alkali oxide and boric oxide lower the formation and
pouring temperatures of the glass but also reduce chemical
durability. To maintain a balance between these two properties
i.e., (i) lower formation and pouring temperatures and (ii)
excellent chemical durability, multivalent modifiers such as MnO,
TiO2, BaO, Fe2O3, ZnO, CaO and MgO can be added if not already
present in the waste being solidified. However, too many divalent
cation oxides such as BaO, CaO, and MgO can cause undesirable
glass-in-glass phase separation in borosilicate glasses [48] which
can negatively impact glass durability.
An important aspect in the use of borosilicate and phosphate
glasses for HLW is the solubility of key species in the glasses.
Table 3 shows the solubility of U, Pu, Am, Cm and Np in a range of
glass compositions (see also [49]).
Table 3
It is important to note that the solubility is also a function
of the glass REDOX as driven by the processing conditions.
Phosphate glass matrices are suitable for wastes rich in
molybdenum, chromium, sulphates, strontium, and uranium. These
elements have lower solubility in borosilicate glasses. Typical
phosphate compositions incorporate about 30% of waste oxides. These
glasses have low melting temperature in the range of 900 oC and low
viscosity at pouring temperatures. However, as the main raw
material is phosphoric acid there are concerns about container and
melter corrosion. Also, these glasses have poorer mechanical
strength and are more prone to devitrification compared with
borosilicate glasses. Some wastes are not amenable to incorporation
into glass structures e.g., because they are volatile (e.g.,
alkalis) or too refractory to melt (e.g., Al2O3, ZrO2) or they do
not dissolve in silicate melts (noble metals). These may require
alternative matrices or if present in small quantities may stay
suspended in the glass and not impact glass behaviour.
Ceramics
Almost fully crystalline matrices have been developed for
immobilizing reprocessing waste but not yet commercially used such
as SYNROC by ANSTO. In contrast there are many single phase ceramic
matrices for stabilization of a variety of radionuclides. A more
detailed description of a variety of ceramic wasteforms can be
found in [43] including a discussion of radiation damage.
Multiphase systems have been developed to immobilize complex
multi-radionuclide waste streams. These include SYNROC, which
consists of the titanates hollandite (BaAl2Ti6O16), zirconolite
(CaZrTi2O7), perovskite (CaTiO3) and TiO2. The hollandite mainly
fixes the Cs and Rb and some process chemicals, whereas actinides
and rare earth fission products (FPs) are bound in zirconolite and
perovskite, and the latter also incorporates Sr. Important that
SYNROC compositions can be altered so they contain phases which
target radionuclides in a particular waste stream. These minerals
can accommodate various actinides and FPs in their crystalline
structures which results in good resistance to chemical alteration
even under hydrothermal conditions. Single-phase ceramics have been
examined targeting difficult species such as Pu. Pyrochlore-based
ceramics such as those based on zirconolite have been developed for
immobilization of Pu and zircon (ZrSiO4) was examined extensively
[43, 50], although its tendency to go amorphous (metamictisation)
under self-irradiation may limit its application. Ceramics tend to
be stable in aqueous environments but the alteration in durability
arising from loss of crystallinity by irradiation may impact this
behavior [51]. Finally, it should be noted that SNFW is a
crystalline ceramic (uranium oxide or mixed plutonium/uranium
oxides in MOX) but one which was not designed with post irradiation
stability in mind.
Glass Composite Materials
Between the almost fully glass and fully crystalline matrices
lie Glass Composite Materials. Early work on GCMs [52] was
performed at Pacific Northwest National Laboratory (PNNL) and the
Pennsylvania State University (PSU) in the 1970s on “supercalcine
ceramics”, regarded as the forerunner of SYNROC, with the aim of
achieving a waste loading of 60% by additions to HLW liquids which
on calcination and sintering at ~1100 oC formed durable crystalline
phases [53]. For high density a hot press was needed along with
silica which formed a liquid, i.e. liquid phase sintering, at the
formation temperature and an intergranular glass on cooling. Often
the intergranular glass contained the radionuclides rather than the
desired ceramic phases. GCMs as alternative matrices are still
being developed e.g. to enable increased waste loading or to
incorporate difficult wastes. GCMs include: (i) Glass ceramics
where a glassy wasteform is crystallized in a separate heat
treatment, (ii) GCMs in which e.g. a refractory waste is
encapsulated in glass such as hot pressed lead silicate glass
matrix encapsulating up to 30 vol% of La2Zr2O7 pyrochlore crystals
to immobilise minor actinides [54], (iii) GCM formed by
pressureless sintering of spent clinoptilolite from aqueous waste
processing [55], (iv) Some difficult wastes such as the French HLW
UMo-containing materials were immobilized in a GCM termed UMo glass
formed by CCIM and which partly crystallises on cooling [29], (v)
alkali Mo-containing wastes are immobilised in Russia in a yellow
phase GCM containing up to 15 vol.% of sulphates, chlorides and
molybdates [56], and (vi) GCMs which immobilise ashes from
incineration of solid radioactive wastes [57]. Note that
alkali-rich wastes at the Hanford site are also immobilized in
glassy wasteforms with high crystal contents which characterize
them as GCMs [58].
GCMs may be used to immobilize long-lived radionuclides (such as
actinide species) by incorporating them into the more durable
crystalline phases, whereas the short-lived radionuclides may be
accommodated in the less durable vitreous phase. Historically,
crystallization of vitreous wasteforms has always been regarded as
undesirable as it has the potential to alter the composition (and
hence durability) of the remaining continuous glass phase which
would (eventually) come into contact with water. That is why it is
also important to determine that the grain boundaries created by
these durable crystals do not create accelerated grain boundary
dissolution which can impact the overall wasteform durability.
However, there has been a recent trend towards higher
crystallinity in ostensibly vitreous wasteforms so that they are
more correctly termed GCMs. This is particularly apparent in the
development of hosts for more difficult wastes or where acceptable
durability can be demonstrated even where significant quantities of
crystals (arising from higher waste loadings) are present such as
in some of the Hanford wasteforms. Acceptable durability will
result if the active species are locked into the crystal phases
that are encapsulated in a durable, low activity glass matrix or
the active species are locked into the glass and the crystalline
species contain no active species. The former occurs in the Mo GCMs
while the latter occurs in the US HLW defense wastes. The GCM
option is currently being considered in many countries including
Australia, France, Russia, South Korea, UK and USA. The processing,
compositions, phase assemblages and microstructures of GCMs may be
tailored to achieve the necessary material properties.
GCMs are under consideration for a variety of partitioned
difficult wastes such as 129I separated from SNFW [59] and 99Tc as
pertechnetate as well as e.g. spent clinoptilolite filters. One
option being examined for the zeolite wastes from the Fukushima
clean-up is use of a mobile vitrification system using a form of
in-can melting. In the Areva process for cleaning up contaminated
sea water a sludge is created and the Crystalline Silico Titanate
Ion Exchange inorganic resin CST IE-911 is also being used. No
decision has yet been made on what to do with the spent resins but
cementation and vitrification are being considered [60].
Novel technologies
Apart from conventional cold isostatic pressing (CIP) and cold
uniaxial pressing (CUP) of ceramic precursors (calcined waste and
additives) followed by sintering, a number of novel thermal methods
are at various stages of development for immobilising difficult
wastes such as those in the legacy ponds and silos at Sellafield,
UK and at the Hanford site in the USA (Figure 4).
Figure 4
Hot isostatic pressing (HIP) and hot uniaxial pressing (HUP) are
mature technologies in the ceramics and metallurgical processing
fields but have had limited application in immobilization of
radioactive wastes. The joint operation of high temperature and
pressure increases the speed of sintering and reduces the fraction
of volatiles lost during processing compared to pressureless
processing. Hot pressing is an ideal technology for production of
ceramic wasteforms which would otherwise be difficult to obtain
with full density and for immobilizing wastes containing highly
volatile ions such as TcO4-, I-, or Cl-. The HIP process is
typically operated up to ~1250 °C and ~500 atm. when producing
multiphase ceramics, but can vary the temperature and pressure
widely to accommodate waste-containing systems.
Novel non-thermal (cement like) wasteforms are discussed in
References [61-65]. These can be used together or alone to treat
some of the problematic legacy wastes [66]. The flexibility of
cement and other non-thermal systems means that a toolbox of
systems made at close to room temperature are being developed
(Figure 5) applicable to the complex array of wastes which they
must encapsulate [67].
Figure 5
In particular, some metals (such as Al, Mg and U) are unsuitable
for immobilisation in the alkali composite cement systems currently
employed as they react with them in an expansive manner forming
gases (e.g. H2). Alternative non-thermal systems for encapsulating
these reactive metals based on calcium sulpho-aluminate cements and
geopolymers [68] have shown promise in early investigations. A
number of non-thermally produced wasteforms with improved
durability and compatibility with specific ILWs are under
development.
Processing technologies
In any immobilization process where radioactive materials are
used, the process and operational conditions are complicated by
operation and maintenance in a remote environment. Therefore
priority is given to reliable, simple, rugged technologies and
equipment, over complex or sensitive equipment. As described above
a variety of wasteform materials are available for immobilization.
The technologies commercially available to process these forms
summarized in Table 4 [2, 5, 6, 15].
Table 4
Conclusions
Section of wasteforms for nuclear waste immobilisation is a
complex task. Choosing a suitable wasteform for nuclear waste
immobilisation is difficult and durability is not the sole
criterion. In any immobilisation process where radioactive
materials are used, the process and operational conditions can
become complicated, particularly if operated remotely and equipment
maintenance is required. Therefore priority is given to reliable,
simple, rugged technologies and equipment, which may have
advantages over complex or sensitive equipment. A variety of matrix
materials and techniques is available for immobilisation. The
choice of the immobilisation technology depends on the physical and
chemical nature of the waste and the acceptance criteria for the
long-term storage and disposal facility to which the waste will be
consigned. A host of regulatory, process and product requirements
has led to the investigation and adoption of a variety of matrices
and technologies for waste immobilization. For higher activity
nuclear waste the main immobilisation technologies industrially
used are cementation, vitrification and ceramication.
Acknowledgements
Authors acknowledge contribution to this overview of W.E. Lee,
A.I. Orlova, S.V. Stefanovsky, F. Takats, and late E.R. Vance.
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Table 1. Classification of higher activity wasteforms
Glasses
Glass Composite Materials (GCM’s)
Cements
Crystalline Ceramics
Borosilicate
Magnox, Blend, UK.
Glass Ceramics: Zirconolite for MA or Pu.
Multiphase
Composite Cements
UK and France, WVDP US
Single Phase: zirconolite for MA or Pu, hollandite for Cs
Borosilicate
DWPF*, WVDP**, USA
Crystal Waste Encapsulated in Glass Matrix, glass bonded
sodalite (PNNL/KAERI and ANL)
Novel cements,
e.g. calcium sulphoaluminates
Multiphase (e.g., SYNROC) compositions tailored for specific
waste streams
Borosilicate
R7/T7 (France), IR111 (India)
Crystal-tolerant Glasses:
U-Mo CCM French wastes, Hanford US wastes
Flyash geopolymeric cements at SRS for low activity fraction of
HLW wastes, US
Spent Nuclear Fuel Waste (SNFW)
Aluminophosphate, Russia
* Defense Waste Processing Facility; **West Valley Demonstration
Project
Table 2. Summary of matrices (wasteforms) used in different
countries
Country
SNFW
Glass
Cement
Ceramics
GCM
China
No
Yes
May be under development
Finland
Yes
No
No
No
No
France
No, closed fuel cycle
Yes for HLW
Yes for ILW
May be under development
Yes (CCIM) UMo wasteforms, zirconolite glass ceramics under
development
Germany
May be under development
India
No, closed fuel cycle
Yes for HLW
Yes for ILW
Japan
No, closed fuel cycle
Yes for HLW
Yes for LLW and ILW
No
Under development for some of the wastes at Fukushima.
Korea, Republic of
Yes
Yes for LILW
Yes for LILW
No
No
Russia
No, closed fuel cycle
Yes for HLW and ILW
Yes at Mayak and Radon (ILW and LLW)
Under development
Under development in CCIM and also full scale plasma melting at
Radon
Sweden
Yes
No
No
No
No
UK
Yes, AGR* and PWR from Sizewell B
Yes for HLW
Yes for ILW
Under development for Pu
Under development for ILW from Sellafield.
USA
Reference wasteform for commercial power reactor fuels
Wasteform for defense HLW and WVDP reprocessing waste.,
Potential wasteform for advanced fuel cycles.
Yes for LLW and LAW at SRS and WVDP.
Under development for some advanced fuel cycle waste streams
Under development for some advanced fuel cycle waste streams and
Hanford WTP
*Advanced Gas-cooled Reactor
Table 3. Incorporation limits of key radioactive species in
various glasses
Glass type
Processing temperature, oC
Specie
Oxide AnO2 solubility, mass %
Borosilicate
1150
U(IV)
9
1150
U(VI)
40
1200-1300
U
25 – 28
1100, air atmosphere
Pu(IV)
0.55
1200, air atmosphere
Pu(IV)
<2
1500, air atmosphere
Pu(IV)
<10
1400 +graphite
Pu(III)
>13, <25
1000
Am
2
1175
Am
5
1000, air atmosphere
Np(IV)
2
1250, air atmosphere
Np(IV)
3
1350
Np(III)/Np(IV)
>5
La-borosilicate
1450, air atmosphere
Pu
>7, <10
1500
Pu
>11.4
1450, Air atmosphere
Am
0.1
Sodium silicate
1250, CO/CO2
U(VI)
19
1250, reducing conditions
U(VI)
52.7
1250
Np(V)
33
Phosphate
1100
Pu(III)
>2, <10
Table 4. Wasteform processing technologies for higher activity
nuclear wastes, after [5].
Processing Technology
Processing Mode
Treatment and Waste Stream Scale
Wasteforms Produced
Advantages
Disadvantages
Joule Heated Melter (JHM)
Continuous
Large
Boro Silicate Glass, Other Glasses (LaBS, FeP, AIP, chalcogenide
and others)
Proven technology; typically operates with a "cold cap" to
minimize volatility of species of concern
Electrode and refractory erosion may be a problem; solubility
control of certain species
Advanced Joule Heater Melter (AJHM)
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AIP, chalcogenide and others)
Increases throughput and melt rate compared to JHM
Operates with minimal or no "cold cap" with associated increases
in volatility of species of concern
Cold Crucible Induction Melter
(CCIM)
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AlP, chalcogenide and others), crystalline
ceramics/simple oxides, Metal Matrix
Allows the processing of corrosive glasses; No refractories; No
metallic or oxide electrodes;water cooled; self-cleaning; high
purity,can be stirred if needed; increases capacity compared to JHM
and AJHM;can operate at higher temperatures than JHM and AJHM;
operates with a "cold cap" to minimize volatility
Higher temperature operation can increase volatilization of
species of concern but "cold cap" coverage minimizes these
impacts
Cold Press and Sinter
Batch
Small
Glass Ceramic Materials, crystalline ceramics/simple oxides,
metal matrix,zeolites, hydroceramic
Higher waste loadings; minimum diposal volume
Usually small scale; may require pre-calcining or pre-treating
waste to an oxide form to avoid shrinkage of form
Hot Isostatic Pressing (HIP)
Batch
Small
Borosilicate glass (lab demonstration only), Glass Ceramic
Materials, crystalline ceramics/simple oxides, metal
matrix,zeolites, hydroceramic
Zero off-gas emissions; higher waste loadings; minimum diposal
volume; mature flexible technology; no major secondary wastes;
mature industrial process
Processes small quantities; can overpressurize if large amounts
of volatiles (e.g. nitrates/hydrates) are present; requires
pre-calcining or pre-treating waste to an oxide form (shrinkage
handled by dumbbell-shaped canisters)
Hot Uniaxial Pressing (HUP)
(largely supplanted by HIP)
Batch
Small
Glass Ceramic Materials, crystalline ceramics/simple oxides,
metal matrix,zeolites, hydroceramic
Higher waste loadings; minimum disposal volume, mature flexible
techology; mature industrial process
Usually small scale; may require pre-calcining or pre-treating
waste to an oxide form for shrinkage control
In-Container Vitrification (ICV); also known as "Bulk
Vitrification"
Batch
Depends on container size (could be medium or large)
Borosilicate glass; Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AlP, chalcogenide and others)
Relatively cheap and simple for low activity wastes or
contaminated soils; not applicable to HLW
Inhomogeneous wasteforms; no temperature control so radionuclide
vaporization is high; little convection in melt causes processing
problems
Self-sustaining Vitrification (SSV)
Batch
Small
Glass Ceramic Materials
Low capital requirements; can be used to process small amounts
of wastes at remote locations
May require some pre-processing, for example, grinding of the
waste and pre-mixing
Fluidized Bed Steam Reforming (FBSR)
Continuous
Large
Crystalline ceramics/simple oxides, zeolites as formed;
hydroceramics, geopolymers (as encapsulated)
Industrially proven technology; acidic or basic tank wastes
processed without neutalization; destroys organics and
nitrates;converts aqueous components in a waste to stable water
insoluble mineral products; immobilizes S, Cl, and F are in a
stable mineral form, with no secondary waste
Product is granular and requires a high integrity container
(HIC) or encapsulation in a binder to make a glass ceramic
material, a geopolymer, or a hydroceramic; radionuclide
partitioning amonst the phases needs to be further studied
Cyclone Furnaces
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AlP, chalcogenide and others), crystalline
ceramics/
simple oxides, metal matrix
Suitable for soils containing low volatility radionuclides;
Secondary Recovery Process to treat off gases may be
necessary
Rotary Kilns
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBs, FeP, AlP, chalcogenide and others), crystalline
ceramics/
simple oxides, metal matrix
Established industrial practice
Secondary recovery process to treat off gases may be necessary;
not proven for nuclear waste processing; used as pre-processing in
French HLW vitrification
Electric Arc Furnaces
Batch
Medium/Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AIP, chalcogenide and others), Mineral Ceramic/Simple
Oxides, Metal Matrix
Established Industrial Practice; Similar technology is used for
ICV
High temperatures (1600°C); requires significant off-gas
treatment; volatilization of constituents of concern
Plasma Furnaces
Batch
Small
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AIP, chalcogenide and others), Mineral Ceramic/Simple
Oxides, Metal Matrix
Plasma generating electrode erosion; Efficient for the
destuction of organics
No large scale practice; high tempeatures, increaced
volatilization
Microwave Heating
Batch or Continuous
Small
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LABS, FeP, AIP, chalcogenide and others), Mineral Ceramic/Simple
Oxides, Metal Matrix
Suitable for mixed wastes; Can be used as a heat source in other
equipment (e.g fluidized bed)
Limited to small scale; Microwaves need to be guided
Processing Technology
Processing Mode
Treatment and Waste Stream Scale
Wasteforms Produced
Advantages
Disadvantages
Joule Heated Melter (JHM)
Continuous
Large
Boro Silicate Glass, Other Glasses (LaBS, FeP, AIP, chalcogenide
and others)
Proven technology; typically operates with a "cold cap" to
minimize volatility of species of concern
Electrode and refractory erosion may be a problem; solubility
control of certain species
Advanced Joule Heater Melter (AJHM)
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AIP, chalcogenide and others)
Increases throughput and melt rate compared to JHM
Operates with minimal or no "cold cap" with associated increases
in volatility of species of concern
Cold Crucible Induction Melter
(CCIM)
Continuous
Large
Boro Silicate Glass, Glass Ceramic Materials, Other Glasses
(LaBS, FeP, AlP, chalcogenide and others), crystalline
ceramics/simple oxides, Metal Matrix
Allows the processing of corrosive glasses; No refractories; No
metallic or oxide electrodes;water cooled; self-cleaning; high
purity,can be stirred if needed; increases capacity compared to JHM
and AJHM;can operate at higher temperatures than JHM and AJHM;
operates with a "cold cap" to minimize volatility
Higher temperature operation can increase volatilization of
species of concern but "cold cap" coverage minimizes these
impacts
ПОДПИСИ К РИСУНКАМ
Fig. 1. Schematic representation a of multiple barrier system.
THMC stands for thermal, hydraulic, mechanical, and chemical.
Fig. 2. A generic vitrified HLW waste disposal system, after
[15]
Fig. 3. Phase composition of nuclear wasteforms, after [10]
Fig. 4. Range of thermal technologies under development for
difficult waste streams. The products of the vitrification
processes for such waste streams are often GCMs.
Fig. 5. Non-thermal technologies under development for wastes
destined for geological disposal which are reactive in Ordinary
Portland Cement (OPC) systems.
On selection of matrix (wasteform) material for higher activity
nuclear waste immobilisation
C. M. Jantzen1, M. I. Ojovan2,*
1Savannah River National Laboratory, USA, Aiken, South Carolina
29803, Wood Valley Dr. 3922
2Imperial College London, UK, London, SW7 2AZ, South Kensington
Campus, Exhibition Road
Figure 1. Schematic representation a of multiple barrier system.
THMC stands for thermal, hydraulic, mechanical, and chemical.
Figure 2. A generic vitrified HLW waste disposal system, after
[15]
Figure 3. Phase composition of nuclear wasteforms, after
[10]
Figure 4. Range of thermal technologies under development for
difficult waste streams. The products of the vitrification
processes for such waste streams are often GCMs.
Figure 5. Non-thermal technologies under development for wastes
destined for geological disposal which are reactive in Ordinary
Portland Cement (OPC) systems.
1
Geologic SystemUnderlying StrataGeologic Formation
Conceptual High LevelWaste Repository -CanisterAssumed to Last
1000 years
OverlyingStrataBackfillGeologicFormationOverpackContainerWaste
Form(glass)
