Inert-matrix fuel for transmutation: Selected mid- and long-term effects on reprocessing, fuel fabrication and inventory sent to final disposal

Progress in Nuclear Energy 145 (2022) 104106

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Progress in Nuclear Energy
journal homepage: www.elsevier.com/locate/pnucene

Inert-matrix fuel for transmutation: Selected mid- and long-term effects on
reprocessing, fuel fabrication and inventory sent to final disposal
Friederike Frieß ∗, Wolfgang Liebert
University of Natural Resources and Life Sciences, Vienna, Department of Water, Atmosphere and Environment, Institute of Safety and Risk
Sciences, Peter-Jordan-Straße 76/I, 1190 Vienna, Austria

ARTICLE

INFO

Dataset link: />S-Transmutation-Fuel-Paper.git
Keywords:
Accelerator-driven-system (ADS)
Transmutation
Minor actinides (MA)
Long-Lived Fission Products (LLFP)
Inert-matrix fuel (IMF)

ABSTRACT
Partitioning and transmutation (P&T) fuel cycles provide a technical approach to ease the problem of
radioactive waste disposal. Some of the partitioned components of the waste stream are irradiated while others
can be used for energy production or are sent to final storage. Minor actinides are planned to be irradiated in
a fast spectrum nuclear facility to transmute them into stable or short-lived isotopes. As minor actinides have
negative effects on reactor dynamics, subcritical, accelerator-driven systems are proposed to increase their

fraction in the fuel. An example is the MYRRHA research reactor to be built in Mol, Belgium.
This reactor is modeled for depletion calculations. The behavior of special fuel elements that mirror fuel
composition as envisioned for large scale transmutation facilities, namely inert-matrix fuels with an increased
minor actinide content, are investigated within this reactor environment. It turns out that gamma dose rates,
activity and residual heat from the spent fuel elements present significant challenges for implementing a P&T
program. Spent inert-matrix fuel element show significantly higher levels than spent fuel elements from fast
reactors. This requires long cooling periods and poses unprecedented challenges to reprocessing technology.
The problem is amplified by the fact that it is generally agreed upon that due to low transmutation efficiencies
several transmutation steps would be necessary. Looking at the radiotoxicity index, the efforts suggested
to reduce the minor actinide content in a final repository are justified. The long-term safety case of deep
geological repositories, however, implies that certain long-lived fission products are more relevant. The buildup of some of these radionuclides is investigated for two hypothetical German P&T scenarios. Naturally, the
amount of fission products increases with continued irradiation. But namely the fraction of Cs-135 increases
over-proportionally when inert-matrix fuel rich on minor actinides is used.

1. Introduction
The Belgian government announced its decision to finance roughly
one third of the MYRRHA1 (Multi-purpose hYbrid Research Reactor for High-tech Applications) project in 2018 (WNN, 2018). According to the European Strategy Report on Research Infrastructure
(ESFRI), MYRRHA is supposed to become operational in 2027 (ESFRI, 2018). The Belgian Nuclear Research Center (SCK⋅CEN), however,
states operating the reactor and full power accelerator-driven system
(ADS) in 2036 (SCK-CEN, 2020). MYRRHA is supposed to be the
first hybrid nuclear research reactor: its design comprises critical and
sub-critical core configurations. Recently, much work was put into
the final adjustments of the cooling system for the MYRRHA reactor
core and accelerator technologies (Van Tichelen et al., 2020; Kennedy

et al., 2020; Gladinez et al., 2020; Moreau et al., 2019). The focus
of the MYRRHA project shifted from research activities to project
implementation recently (SCK-CEN, 2019).
Besides being introduced as the Experimental Technology Pilot
Plant (ETPP) for a lead-cooled fast reactor (LFR) (ESFRI, 2018),

MYRRHA is intended to demonstrate the concept of an acceleratordriven system. ADS are key components for one possible concept
of irradiating high level nuclear waste in a partitioning and transmutation (P&T) fuel cycle. Partitioning means the separation of the
spent fuel into different waste streams such as uranium, plutonium,
minor actinides and fission products. It is foreseen that the separated
minor actinides are incorporated into a special fuel form and are then
irradiated in a (usually fast) neutron spectrum with the objective of

∗ Corresponding author.
E-mail addresses: (F. Frieß), (W. Liebert).
1
ADS: Accelerator-driven system, BWR: Boiling water reactor, EFIT: European Facility for Industrial Sized Transmutation, FE: Fuel element, FP: Fission products,
FPD: Full power days, IM-fuel: Inert-matrix fuel, IPS: In-pile test section, LBE: Lead–bismuth eutectic, LLFP: Long-lived fission products, LWR: Light water reactor,
MA: Minor Actinides, MOX: Mixed oxide fuel, P&T: Partitioning and Transmutation, PWR: Pressurized water reactor, UOX: Uranium oxide fuel.

/>Received 13 May 2021; Received in revised form 9 November 2021; Accepted 17 December 2021
Available online 1 February 2022
0149-1970/© 2022 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license ( />

Progress in Nuclear Energy 145 (2022) 104106

F. Frieß and W. Liebert

their transmutation into stable or short-lived isotopes prior to final
disposal.
P&T concepts bring several challenges with them. To efficiently
transmute minor actinides, it is important to irradiate them in a fast
neutron spectrum.2 Additionally, they have undesired safety-relevant
effects on reactor dynamics such as the delayed neutron fraction and
the Doppler effect. This has to be considered in fuel and reactor design.
The possible amount of minor actinides and plutonium in the fuel for

critical core configurations is thus limited (Palmiotti et al., 2011). For
a medium-sized sodium-cooled fast reactor, a minor actinide content
of approximately 10% seems manageable. For a large-size sodiumcooled fast reactor (3000 MWth), this fraction drops to 2.5%–3% (Fazio
and Boucher, 2008, 7). In contrast, accelerator-driven systems should
allow for minor actinide fractions up to almost 50% (Artioli et al.,
2007). Therefore, many P&T concepts rely on accelerator-driven systems that could ensure an efficient throughput of minor actinides.
Usually, ADS are envisioned for minor actinide transmutation while fast
reactors would use the excess plutonium for power generation (Mueller,
2013; Abderrahim et al., 2013; Doligez, 2017; ESNII, 2020). This is
called a double-strata fuel cycle. In any case, multiple reprocessing and
irradiation steps are necessary.
An ADS consists of a particle accelerator, a spallation target, and
a sub-critical reactor core. The core never reaches criticality during
operation but amplifies the neutrons supplied by an external neutron
source, usually a spallation target. The number of neutrons in the
core is regulated by the variation of the beam current. The amplifying
nature of the sub-critical core is one key aspect in the safety concept
since it prevents exponential criticality excursions in most possible
cases (Sarotto, 2017). Consequently, a significantly increased fraction
of minor actinides should be possible in the fuel composition.
The amount of minor actinides in a deep geological repository could
be reduced with the implementation of P&T programs. But other radionuclides impact radiological safety as well. Some of these are fission
products with very long half-lives well beyond some 100,000s of years.
In the 1990s, the transmutation of at least Tc-99 and I-129 was also considered a relevant contribution of P&T to reduce the burden of nuclear
waste disposal (NRC, 1996; Jameson et al., 1992; DoE, 1999). This has
however been proven to be by far more complicated than anticipated.
Obstacles are for example that single fission products, sometimes even
single isotopes, need to be separated from the spent fuel. If they are
then placed in a reactor for irradiation, they only consume neutrons and
thus effect the neutron economy in the core negatively. Consequently,

research has ceased (NEA/OECD, 2006a; Doligez, 2017; Frießet al.,
2021). The emphasis shifted to the irradiation of minor actinides only.
If transmutation efficiency for certain systems is evaluated, only the net
balance of minor actinides is discussed in most cases (Mueller, 2013;
Mansani et al., 2012; Sarotto et al., 2013; Renn, 2014; Liu et al., 2020).
The need to transmute long-lived fission products as well is only rarely
mentioned, e.g. in Shwageraus and Hejzlar (2009) and Chiba et al.
(2017).
In a first step, this article explores the effects a high minor actinide
content in inert matrix fuel (IM-fuel) has on the dose rate, the activity
and the decay heat of spent fuel elements. Simulations are based on a
computer model of the planned accelerator-driven system MYRRHA.
Nuclide compositions in the spent fuel are derived from depletion
calculations. In a second step, the concentration of selected long-lived
fission products in the spent fuel is extracted to estimate the influence
of a P&T scenario on the inventory of a deep geological repository.
The latter is illustrated using two hypothetical scenarios of a potential P&T implementation in Germany. The scenarios are based on the
highly radioactive waste accumulated by the German nuclear energy
program until 2022. With the end of that year, Germany will have shut
down all its nuclear power reactors.

Fig. 1. Core layout of the generic ADS core for equilibrium sub-cycle. Control rods
are not inserted.

2. Methods
In this chapter the methods used for the analysis are introduced.
It starts with the description of the reactor model and the computer
programs. Then the procedure of evaluating the amount of certain
isotopes in the spent fuel is explained.
2.1. Reactor model

In the 7th Framework Program of EURATOM, a FAst Spectrum
Transmutation Experimental Facility (FASTEF) was designed (Sarotto
et al., 2013; Sarotto, 2017; CORDIS, 2019). The rather detailed FASTEF
design is very similar to the MYRRHA reactor. The large-scale European
Facility for Industrial Sized Transmutation (EFIT) is planned to follow
after the proof-of-concept reactor MYRRHA is in operation (Mansani
et al., 2012; Artioli et al., 2007; Sarotto, 2017).
Since MYRRHA is designed as a hybrid facility, critical and subcritical core configurations have been modeled for validation of the
model. The critical layout could also be used for transmutation, but
within the already mentioned limitations: only a slight increase compared to minor actinide content in current MOX fuel would be possible
due to safety reasons. Thus the focus of this work is set to the subcritical core. The simulations of criticality and neutron flux show
good accordance with average values for a typical fast reactor system (Sarotto et al., 2013; Frieß, 2017). The cross section of our generic
ADS model based on the MYRRHA core design Sarotto et al. (2013) and
Sarotto (2012) is depicted in Fig. 1.
There are six different fuel zones in the core. After each sub-cycle of
90 days, the elements in one zone are replaced with fresh fuel elements.
The other elements are shuffled to the next fuel zone. After one full
cycle, consisting of six sub-cycles, all fuel elements are replaced by fresh
fuel elements. Plutonium enrichment and fuel composition are the same
for all batches (Sarotto et al., 2013). Geometric dimensions are provided in Table 1. The center assembly hosts the spallation target, which

2
The reason is the relation between fission and absorption cross sections.
It is significantly more favorable in a fast neutron spectrum.

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Progress in Nuclear Energy 145 (2022) 104106

F. Frieß and W. Liebert
Table 1
Geometric dimensions of the generic ADS model for the MOX fuel elements and the
P&T IM-fuel elements. The number of fuel assemblies is given at begin of cycle, not at
beginning of life. The center pin in each assembly is a structure pin.
Source: The values are taken from Sarotto (2012).
Parameter

Unit

MOX Fuel

EFIT/IM-Fuel

# of Fuel Elements
# of Fuel Rods
Radius of Fuel Pin
Active Height
Cladding Thickness
Fuel Assembly Pitch
Can Thickness
Can Inner Diameter

–
–
cm
cm
cm
cm
cm

cm

72
126+1
0.271
60.00
0.045
10.45
0.20
9.755

6
60
0.36
60.00
0.06
10.45
0.20
9.755

The depletion calculation is split into steps of 30 days each. After
each sub-cycle of 90 days, a 30 day decay step is included. This accounts for the reshuffling of the fuel elements. Additionally, every third
sub-cycle, there is a longer maintenance interval of 90 days (Sarotto
et al., 2013). The power of the ADS is 400 MWth. The six IM-fuel
elements placed in the IPS elements are irradiated 1080 full power days
(FPD). This is the irradiation time planned for the EFIT reactor. With
cladding and structure materials currently available, this irradiation
time is not feasible due to high material stress. Those burn-ups are
considered here nevertheless, since only then efficient transmutation
rates can be achieved.

Only equilibrium sub-cycles where the full number of fuel elements
is placed in the core are considered. The depletion calculations are
started in equilibrium configuration with fresh fuel elements. The first
cycles of the evaluation were skipped before evaluation to allow 𝑘𝑒𝑓 𝑓 to
reach decline. During burn-up, 𝑘𝑒𝑓 𝑓 ranges between 0.971 and 0.954.
More details on the burn-up calculation, including mass balanced, can
be found in Frieß (2017).

Table 2
Composition of MOX fuel used in the basic MYRRHA design and IM-fuel under research
for transmutation purposes in weight percent. Plutonium comprises 45.7% of the
transuranium content in IM-fuel.
Isotope

MOX Fuel
in wt%

EFIT/IM-Fuel
in wt%

U-235
U-238
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Am-242 m

Am-243
Cm-243
Cm-244
Cm-245
Cm-246
Cm-247

0.5
69.5
–
0.5
17.4
6.8
3.6
1.2
–
4.9
–
–
–
–
–
–

–
–
2.1
1.7
21.2
15.6

1.8
5.4
41.0
0.14
8.71
0.04
1.63
0.62
0.05
<0.01

2.2. Simulation code
The general-purpose Monte Carlo radiation transport code MCNPX
(Monte Carlo N-Particle eXtended) Version 2.7 is used for criticality,
neutron and gamma flux distributions (Pelowitz, 2011). This MCNPX
version also provides the neutron flux distribution for the depletion
calculations. Even the newest version, however, MCNP 6, does not
provide the option to calculate various criticality coefficients of a
system with an external neutron source. Thus, two different kinds
of simulations were run: either a criticality calculation for general
assessment of the system in the state of sub-criticality with different
fuel compositions or a simulation of the spallation source within the
surrounding reactor core for depletion calculations (Frieß, 2017). The
energy distribution of the neutron flux in different distances from
the spallation target shows that the characteristic high energy tail of
spallation neutrons only appears in the vicinity of the central element.
Therefore, the reactions caused by the spallation neutrons are limited
to only a small fraction of the core (Malambu and Aoust, 2005). This
is one reason for the unusually high number of fuel zones in the core.
Frequent shuffling of fuel elements is mandatory to gain a flattened

power profile and a more uniform burn-up in the different elements.
The program code VESTA that couples MCNPX with its build-in
depletion module PHOENIX is used for derivation of time-dependent
material compositions (Haeck, 2011). Criticality calculations with MCNPX 2.7 were used since VESTA only processes this kind of MCNP
input. During the depletion step, the number of neutrons in the core
is scaled according to the power level input given by the user. The
spallation neutrons only undergo a relevant amount of reactions after
they have lost most of their energy. Their negligence does not influence
the results of the depletion calculations significantly. Consequently, for
those calculations only neutrons were tallied.
From the results of the depletion calculation, values for decay heat
and activity can be derived almost directly. The gamma dose rates are
calculated using MCNP 6 and the dose conversion factors published by
the International Commission on Radiological Protection (ICRP, 1996,
2012).

has been modeled for tally and source efficiency calculations (Frieß,
2017). Spallation target and coolant are assumed to be lead–bismuth
eutectic (LBE).
MYRRHA is designed to use mixed-oxide (MOX) fuel with natural
uranium. The plutonium content is set to 30wt% (Sarotto et al., 2013).
The fuel composition is derived from averaged PWR fuel at a burnup of 45 MWd/kgHM (see Table 2, MOX fuel). The cooling period is
assumed to be 15 years for the plutonium and 30 years for the minor
actinides (Artioli et al., 2007; Sobolev et al., 2011). The density of the
fuel is 10.27 g/cm3 (Sarotto et al., 2013; Eriksson et al., 2005).
The MYRRHA core comprises six in-pile test sections (IPS) for
various experiments. In the simulation, those are filled with EFITlike IM-fuel elements. These elements do not contain uranium but
Magnesium-Oxide as an inert matrix to avoid further plutonium breeding. As proposed for EFIT, the matrix accounts for 42%wt of the
fuel. The fraction of plutonium out of all transuranium elements is
0.457 (Artioli et al., 2007; Mansani et al., 2012) (Table 2, IM-fuel).

The density is set to 6.27 g/cm3 which equals a bulk density of 95%
of the theoretical density (Eriksson et al., 2005; Maschek et al., 2008).
Unlike the fuel itself, the geometry of the EFIT-like IM-fuel elements
had to be adapted for irradiation in the in-pile test sections: To fit the
smaller assembly size in the model based on the MYRRHA geometry,
the number of rods per fuel element is reduced (cf. Sarotto et al.
(2013)). The geometric measures that are used can be found in Table 1.
The core is surrounded by dummy elements filled with lead–bismuth
eutectic and reflector elements consisting of yttrium-stabilized zirconium (YZrO). In the sub-critical core configuration, the control rods
function as six absorbing devices to ensure a sufficiently high level of
negative reactivity during refueling (Sarotto, 2017). During operation,
these elements are filled with coolant.

2.3. Long-lived fission products in a hypothetical german P&T scenario
Energy and particles emitted by certain radionuclides determine its
effects on the environment. Using dose conversion coefficients that mirror those differences and a given amount and composition of material,
the dose rates caused by ingestion can be calculated. The resulting
radiotoxicity is often used when the effects of partitioning and transmutation on a possible final repository are discussed (e.g. in (Abderrahim
et al., 2013; NEA/OECD, 2006b; Romero and Abderrahim, 2007).
3


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F. Frieß and W. Liebert

Table 3
Expected German spent fuel inventories of important isotope (groups)
in 2022 (Schwenk-Ferrero, 2013). Entries marked with an asterix are
calculated using Eq. (1).

Fig. 2. The radiotoxicity of the spent fuel after discharge in Sievert per ton heavy
metal. The values are calculated based on ingestion dose rates. For reference, the value
of an uranium ore is plotted.
Source: Figure adapted from (NEA/OECD, 2006b).

Spent Nuclear Fuel in 𝑡

Total
Transuranium
Minor Actinides
Plutonium
Fission Products
Long-lived fission products
Se-79*
Sn-126*
Zr-93*
Tc-99
I-129
Cs-135

10,300
152.5
21.5
131.0
415.7
0.02
0.38
8.31
9.28

2.27
5.43

Estimated fractions for Tc-99, I-129, and Cs-135 in the German spent
fuel are derived from (McGinnes, 2002). With those, the variables
in Eq. (1) can be derived to be 𝛼 = 0.662, 𝛽 = 0.058, and 𝛾 = 0.105.
Relevant spent fuel inventories for the German case are listed in
Table 3. Already vitrified waste is not considered for transmutation,
since this material would be very complicated to reprocess again. It
will most likely be sent to storage as it is even in a hypothetical P&T
scenario (CEA, 2008; NNL, 2014; Frießet al., 2021, p. 153).
Two future scenarios were evaluated to illustrate the change in inventories due to possible P&T implementation compared to the baseline
scenario of nuclear phase-out in 2022. They are similar to scenarios
used by others (Salvatores et al., 2008; Renn, 2014; Kirchner et al.,
2015; Frießet al., 2021).
In those hypothetical scenarios, it is of only minor concern that
isotopes undergo different reactions in the reactor core. Besides the
(desired) fission, the most likely reaction is neutron absorption. If a
minor actinide undergoes neutron absorption, the reaction product will
be another minor actinide, which might undergo further fission or
absorption reactions. As long as the isotope did not fission, it cannot be
considered to be transmuted (neglecting very unlikely spontaneous and
induced decays). Consequently, each transmuted radionuclide results in
roughly its mass of fission products that need to be taken care of.
The two hypothetical scenarios can be described as follows:

Fig. 2 shows the radiotoxicity resulting from PWR spent fuel with
a burn-up of 48 GWd/tHM assumed to be sent to a deep geological
repository. The cumulative radiotoxicity is shown for different isotope
groups in the spent fuel separately. Assuming that the plutonium can

be extracted and used for energy production in suitable reactors, the
total radiotoxicity is mainly caused by the minor actinides in the spent
fuel. If those could be removed in a P&T fuel cycle, the time interval
until the radiotoxicity falls below the threshold of the uranium ore is
significantly reduced.
The long-term safety of a deep geological repository, however, is
more strongly affected by other radionuclides. Certain fission products
e.g. dissolve more easily than the minor actinides once the canister is
breached (Nagra, 2002, p. 145). As a result, it can be shown that in
most cases the dose rate from long term waste storage is dominated
by a few long-lived fission products (Brasser et al., 2008; NEA/OECD,
2006a; IAEA, 2004). Exemplary calculations for the final waste disposal
site SAFIR-2 in Mol, Belgium, reveal that for the case of a clay dome
repository Se-79, I-129, Sn-126, and Tc-99 are the most influential
nuclides in the long term (Schmidt et al., 2013). Other studies point
out the importance of long-lived fission products such as Tc-99, Cs135 and Zr-93 (Nagra, 2002, p. 208). One exception was the planned
deep geological repository in Yucca Mountain, NV, USA (Krall and
Macfarlane, 2018).
Therefore, we analyzed the production of some long-lived fission
products in a transmutation fuel cycle. The amounts were compared
to the spent fuel inventory already produced by the German nuclear
power generation program. Keeping in mind that the common approach
of providing concentrations in Becquerel per ton heavy metal does not
make sense with IM-fuel, the weight percentage of single radionuclides
in all fission products is considered instead. This approach also bears
the advantage that it is not burn-up dependent.
An estimate of the accumulated fission products and the long-lived
fission products I-129, Cs-135 and Tc-99 in Germany until the phase-out
year 2022 is given in Schwenk-Ferrero (2013). The expected amount of
the remaining three long-lived fission products (Se-79, Zr-93, Sn-126)

was estimated from existing inventories in the following way Frieß
(2017): Detailed compositions for Switzerland’s inventory of spent
BWR-UOX, PWR-UOX, and PWR-MOX fuel can be found in McGinnes
(2002). It is assumed that the German nuclear waste foreseen for direct
disposal so far can be modeled by combining these three vectors, since
mainly these three types of spent fuel were accumulated. The fraction
𝐹 of a certain isotope 𝑖 in the German spent fuel can be estimated by
𝑖
𝑖
𝑖
𝐹𝐺𝐸𝑅
= 𝛼𝐹𝑃𝑖 𝑊 𝑅 + 𝛽𝐹𝐵𝑊
𝑅 + 𝛾𝐹𝑀𝑂𝑋 .

Constituent

Regional Scenario : Germany cooperates with international partners
that operate a fast reactor fleet for power generation. These
partners are willing to take the German plutonium to fuel their
nuclear power reactors. Germany would only be responsible
for the waste resulting from the transmutation of its minor
actinide inventory. In this case, it is desirable if only minor
actinides were fissioned and the amount of plutonium in the core
remained constant.
For the current design of the EFIT transmutation facility, simulations show that 40.16 kg minor actinides and 1.74 kg plutonium
isotopes are fissioned per generated TWhr (Artioli et al., 2007).
Consequently, it is assumed that minor actinide transmutation
alone is not possible and that there is always some plutonium
fission. This results in the increased amount of 22.5 tons (instead
of 21.5 tons) of fission products from the transmutation facility

that would need to be sent to final disposal.
National Scenario : Germany implements a P&T fuel cycle to transmute its total inventory of plutonium and minor actinides. The
sole goal of this approach is the transmutation of the transuranium elements. This would lead to an additional 152.5 t (resulting from all minor actinides and plutonium as given in Table 3)
of fission products sent to final storage. This scenario is quite
optimistic, since, among others, it would need facilities that are
able to transmute small amounts of minor actinides. In reality,
there would always be some amount of transuranium elements
that could not be transmuted.

(1)
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F. Frieß and W. Liebert

significantly longer time periods before it falls for another order of
magnitude.
The longer the fuel element is irradiated in the core, the more the
concentration of short-lived fission products approaches an equilibrium
state. For the fuel with higher burn-up, the contribution of the shortlived fission products to the dose rate is thus smaller than for fuel with
low burn-up. After removal from the core, the isotopes continue to
decay but the effect on the derived dose rates is stronger for the case
where the contribution of the short-lived fission products is higher.
The reference values for a Westinghouse BWR and PWR spent fuel
element shown in the plot are derived from Lloyd et al. (1994). It
should be kept in mind that an IM-fuel element contains only 47.7 kg
of fuel while the heavy metal content of the pressurized water reactor
fuel elements is almost ten times as high.

Fig. 3. Gamma dose rates after different cooling periods for P&T/IM-fuel with high
and low burn-up. Reference values for Westinghouse BWR and PWR fuel elements
taken (Lloyd et al., 1994) are also shown.

3.1.2. Activity
The values for the total activity and the 𝛼-activity of spent IM-fuel
elements (irradiated for 1080 full power days) after various cooling
periods resulting are given in Table 4. The transuranium elements in
general account for more than 99% of the total 𝛼-activity of spent fuel
in the short term (Fanghänel et al., 2010, p. 2982). Consequently, 𝛼activity is approximated by summing the activity of all transuranium
elements except Pu-241 and Am-242. For these two isotopes, 𝛼-decay
is not the most likely decay mode.
The total activity of the spent IM-fuel after discharge is about 5 ⋅
1013 Bq∕cm3 . It falls almost one order of magnitude in the first year.
After ten years it is still about 2.7% of the original value. If the elements
are irradiated for only 270 full power days, the trend is more or less
the same, but the absolute values are lower.
The contribution of the transuranium elements to the total activity
increases with the duration of cooling period. Directly after discharge,
several short-lived fission and activation products are present in the
spent fuel. With their disappearance, the contribution of the transuranium elements becomes more important. This is illustrated by the
𝛼-activity: directly after discharge, it is responsible for only about
17% of the total activity while after ten years it accounts already
for around 60%. After one century, 83% of the disintegrations are 𝛼decays. This fraction of 𝛼-activity and the absolute activity values are
several orders of magnitude higher than in common spent LWR fuel
(compare e.g. McGinnes (2002) and Schwenk-Ferrero (2013)).
Neutrons are almost exclusively produced in the spent fuel by
spontaneous fission of Pu-238, Pu-240, Pu-242, Cm-242, and Cm-244.
Preliminary calculations show that for an appropriate estimation it
is sufficient to consider these five isotopes for the neutron activity

of the spent IM-fuel elements. The last column of Table 4 lists the
calculated spontaneous fission rates. It is 1.16⋅106 s−1 cm−3 directly
after discharge. After a ten year cooling period the spontaneous fission
rate is still around 40% of its original value. The spontaneous fission
rate of fast reactor spent fuel after a one year cooling period can be
approximated3 to be around 3 ⋅ 103 s−1 cm−3 . It is roughly 250 times
less than for the P&T IM-fuel. In general, neutron activity from fast
reactor spent fuel declines more rapidly than the rate from the IM-fuel
elements. One reason is the high amount of Cm-242 in IM-fuel which
is created during reactor operation via 𝛽 − -decay of Am-242.

These scenarios suffice the purpose of this paper to get a sense of the
added material in a final repository if a P&T strategy is chosen. They do
not consider the multiple recycling needed to reach the goal of minor
actinide (and plutonium) transmutation and thus neglect, among other
aspects, losses due to insufficient partitioning technology.
3. Results
The following section starts with examining characteristics of spent
IM-fuel element which influence cooling periods, reprocessing procedures, and fuel fabrication. It then looks at the build-up of long-lived
fission products in the spent fuel that influences the safety of a final
disposal site.
3.1. Characteristics of spent inert-matrix fuel elements from transmutation
facilities
In order to establish a P&T fuel cycle, various reprocessing steps
of the spent radioactive fuel are necessary before new fuel can be
fabricated. It is often claimed that these procedures are basically available since they have been proven on laboratory scale. On an industrial
level, however, this is only true for the hydrochemical separation of
plutonium and uranium from LWR spent fuel.
Possible reprocessing procedures that allow for high separation
efficiencies on a large scale must still be developed. Such partitioning

options and fuel development have to deal with novel spent fuel
characteristics, in particular high gamma dose rates, specific activities
and the decay heat. Those are discussed in the following.
3.1.1. Gamma dose rate
High radiation of spent fuel elements makes handling only possible
using specialized equipment and facilities. Even with heavy shielding,
the radiation emitted by the spent fuel elements must fall below a
certain level before they can be chopped up and dissolved for further
reprocessing steps. In the following, the ambient gamma dose rate in
one meter distance emitted by the spent fuel elements is evaluated.
The results for irradiated elements from the outer fuel zone with a
burn-up equaling 270 full power days and 1080 full power days of irradiation, respectively, are shown in Fig. 3. These elements contain more
fissile material than elements from the inner and intermediate core
zone. About one month after discharge from the core (35.6 days), the
dose rates are extraordinarily high with levels up to almost 3000 Sv/hr
(2676 Sv/hr for 270 FPD and 2817 Sv/hr for 1080 FPD). Due to the fast
decay of short-lived fission products into long-lived or stable isotopes,
the dose rates decline quickly. After approximately four years, the does
rate is reduced by a factor of ten and approaches a plateau. It takes

3.1.3. Decay heat
The evolution of the decay heat from spent IM-fuel elements after
one full irradiation cycle of 1080 full power days is depicted in Fig. 4.
The Cm-242 contribution is plotted separately to highlight its relevance for the total heat. In the first year, Cm-242 is the main contributor. About one month after removal from the core, the IM-fuel

3
This value is derived from the figure for the Russian BN-600 given
in Orlov et al. (1974) using a MOX density of 10 g∕cm3 and the average
number of neutrons per fission as three.

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Progress in Nuclear Energy 145 (2022) 104106

F. Frieß and W. Liebert
Table 4
The total activity for one IM-fuel element irradiated for 1080 full power days in the
MYRRHA reactor after different cooling periods. The 𝛼-activity and the spontaneous
fission rate per volume are given separately. The spontaneous fission (SF) rate is
calculated only considering Pu-238, Pu-240, Pu-242, Cm-242, and Cm-244.
Cooling
period

Activity
Bq/cm3

𝛼-Activity
Bq/cm3

SF Rate
1/(s cm3 )

0.0 a
0.1 a
0.6 a
1.0 a
5.0 a
10 a
40 a

100 a

5.11 ⋅ 1013
1.68 ⋅ 1013
7.54 ⋅ 1012
5.17 ⋅ 1012
1.70 ⋅ 1012
1.38 ⋅ 1012
7.58 ⋅ 1011
4.10 ⋅ 1011

9.05 ⋅ 1012
7.93 ⋅ 1012
4.19 ⋅ 1012
2.71 ⋅ 1012
9.18 ⋅ 1011
8.32 ⋅ 1011
5.37 ⋅ 1011
3.40 ⋅ 1011

1.16 ⋅ 106
1.09 ⋅ 106
8.48 ⋅ 105
7.48 ⋅ 105
5.52 ⋅ 105
4.56 ⋅ 105
1.46 ⋅ 105
1.63 ⋅ 104

Fig. 5. The concentration of certain long-lived fission products in IM-fuel, PWRMOX and PWR-UOX fuel after a 40 year cooling period. The fractions for IM-fuel

are derived from depletion simulations while PWR fractions are estimated using data
from (McGinnes, 2002; Schwenk-Ferrero, 2013).
Table 5
Simulated inventories in tons for two hypothetical scenarios. The baseline scenario
refers to the German inventories as of 2022, which is the endpoint of the German
nuclear energy program (Schwenk-Ferrero, 2013). Material already vitrified is not
considered. For the hypothetical P&T scenarios, the additionally produced long-lived
fission products (LLFPs) with relevance for the long-term safety analysis of a deep
geological repository are listed. All values are given in tons.

All fission products
Important LLFPs
Se-79
Zr-93
Tc-99
Sn-126
I-129
Cs-135
Sum

Fig. 4. Residual heat derived from simulations after different cooling periods for IMfuel with a burn-up of 1080 full power days. The contribution of Cm-242 is plotted
separately. The BN-600 fuel elements emit about one order of magnitude less heat
than the IM-fuel elements (Orlov et al., 1974). The values for LWR fuel are even
smaller (McGinnes, 2002).

elements emit about 14 W/cm3 of heat. This figure drops by almost
90% within two years. It takes a decade before the heat from the
IM-fuel elements is comparable to the heat from MOX elements one
month after discharge from the core. Values shown are for fuel elements
that have been exposed to the highest neutron flux. For elements from

other positions in the core, namely the outer region, values are slightly
smaller.
For comparison, the decay heat of the spent fuel elements from the
fast BN-600 reactor is depicted in Fig. 4. Values are taken from Orlov
et al. (1974). Even though also originating from a fast neutron spectrum
in a metal cooled reactor, the heat per volume one year after discharge
is almost twenty times lower than the heat originating from IM-fuel
elements. One reason is that the BN-600 is a sodium-cooled reactor
while MYRRHA is lead–bismuth cooled. Using lead(-bismuth) coolant
leads to an even harder neutron spectrum than sodium (Cinotti et al.,
2010, p. 28). Further, the IM-fuel contains more transuranium elements
that have high specific powers. It takes the IM-fuel elements a century
to reach decay heat values comparable to the fast reactor fuel after a
one year cooling period.
The effect of the increased transuranium content in MOX fuel can
also be seen in Fig. 4: the heat for the MOX fuel does not drop as rapidly
as for the uranium oxide fuel (UOX). Still, spent MOX fuel emits less
than one tenth of heat per volume over the first century after discharge
compared to IM-fuel elements.

Baseline
scenario

Regional
scenario

National
scenario

415

+22.5

+152.5

0.02
8.31
9.28
0.38
2.27
5.43
25.7

<0.001
+0.33
+0.51
+0.03
+0.15
+0.98
+2.00

+0.01
+2.20
+3.42
+0.17
+0.98
+6.63
+13.41

main contributors to the long-term cumulative dose caused by the final

repository.
The first question is in how far the distribution of fission products
changes when introducing IM-fuel with an increased content of minor
actinides into the fuel cycle. The concentrations that are derived from
the simulation are compared to the fractions in average spent PWR
(MOX and UOX) fuel. An irradiation period of 1080 full power days
is assumed. The concentrations of the five most important long-lived
fission products as a fraction of all fission products in the spent fuel
are plotted in Fig. 5. A cooling period of 40 years is assumed to allow
for the decay of short-lived radionuclides. Se-79, also an important
contributor because of its solubility and its dose conversion coefficient,
is not shown because the concentrations are so low that they would
appear as zero in the plot.
The result is diverse: in comparison to PWR fuel, for Zr-93 and Tc99 the fraction in the spent fuel is declining. For the Sn-126 and I-129
the fraction is more or less the same. Most significant is the rise of Cs135 in the IM-fuel. Its fraction is around three times as high as in UOX
fuel. The factor is still nearly two times when comparing to MOX fuel.
This is a strong indicator that the rise in Cs-135 concentration can be
attributed to the transuranic elements (minor actinides and plutonium)
in the fuel. Further investigations of different fuels with a high minor
actinide content should be conducted to clarify the influence of certain
isotopes in the initial input on the composition of the spent fuel.
These relative concentrations can be used to derive the total inventory for the two hypothetical scenarios and the baseline introduced in
the previous section (see Table 5).
For the regional scenario, the inventory of all fission products
increases by 5.4% and for the national scenario by 36.7%. From all

3.2. Accumulation of long-lived fission products in radioactive waste
Implementing a transmutation fuel cycle affects the inventory of the
final storage. The following section aims at a deeper analysis how the
inventory of certain long-lived fission products might be influenced by

a transmutation fuel cycle. These long-lived fission products are the
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Progress in Nuclear Energy 145 (2022) 104106

F. Frieß and W. Liebert

fuel during irradiation leads to increased swelling (NEA/OECD, 2005,
p. 122). The stress posed on fuel structure materials is one reason why
with current materials irradiation time would need to be reduced to
meet failure-safe mechanical criteria (NEA/OECD, 2017, p. 130). This,
in turn, reduces the transmutation efficiency per fuel cycle. Those problems are even more complicated to tackle since there exist only little
experience with fuel containing transuranium elements (NEA/OECD,
2017, p. 67) and the current lack of appropriate irradiation facilities
(which is supposedly filled by MYRRHA (Christian et al., 2020).
Spent IM-fuel elements emit more than order of magnitude higher
residual heat than spent fuel elements from other reactor types. For
IM-fuel, the decay heat per volume is dominated by Cm-242 (half-life
0.45 years) in the beginning. The strong contribution of Curium to the
heat load is also seen in MOX fuels (Nafee et al., 2012). When the
contribution of Cm-242 ceases, the total heat of the IM-fuel almost
reaches a plateau value (compare Fig. 4). It would take decades from
there for the heat to decrease notably. Such long cooling periods are
not possible in a P&T fuel cycle which would require multiple recycling
steps to reduce the minor actinide inventory to the target value (e.g. 1%
of the original inventory). This is probably the reason why a cooling
periods of five years is assumed by Salvatores et al. (2008). However,
still at this time, the heat is unprecedented. Without a breakthrough
in reprocessing technology, continuous cooling within the reprocessing

and fuel fabrication process would be needed (NEA/OECD, 2005, p.
51). The problems posed by fuel rich in minor actinides on reprocessing were already mentioned in Heidet et al. (2017). One option
to ease the burden on reprocessing of the spent fuel elements might,
however, could be the separate transmutation of Curium in special
targets (Matveev et al., 1999; Kooyman et al., 2018). Cm-242 has a
half-life of only 0.45 years and is strong neutron emitter (Fanghänel
et al., 2010).
Assuming the above mentioned problems will be solved and a
transmutation scenario could be implemented in the future, it is still
not clear in how far the burden on a deep geological repository would
be eased.

long-lived fission products, Cs-135 is the outstanding isotope: while for
the others, the total amount increases almost linearly with the total
increase of fission products and for Zr-93 even a slightly lower fraction
is produced, the Cs-135 generation is disproportionately high. For the
regional scenario it increases by 18% and it is more than twice as high
as in the national scenario (+120%). In a national scenario, there would
be about 12 tons of Cs-135 to be stored instead of only 5.43 t in the
base-line phase-out scenario. The total amount of the selected longlived fission products is increased by more than 50% in a hypothetical
national scenario.
In a regional scenario, there is only a slight increase (around 10%)
of long-lived fission products — at least in Germany. This analysis
somehow neglects that the burden of taking care of some of the
generated fission products is transferred to partnering countries: even
though they can use the plutonium for energy production these partners
will have to cope with the left-overs, in particular with the additional
amount of long-lived fission products. The additional long-lived radiological burden can be expected to be comparable to amounts calculated
for the national scenario.
4. Discussion

The transmutation of certain radionuclides, namely minor actinides
and plutonium, might be a strategy to cope with already existing radioactive waste. In many concepts it is foreseen to use the plutonium for
nuclear energy production while minor actinides are, as far as possible,
transmuted into stable or long-lived isotopes. Advanced reprocessing,
specialized fuel and fuel fabrication facilities as well as fast reactor
systems are mandatory. Most of these procedures and facilities are yet
to be demonstrated on an industrial level.
4.1. Characteristics of spent IM-fuel
IM-fuel does not contain uranium. Thus, plutonium breeding during transmutation cycles is prevented. For transmutation purposes, it
could be used in the accelerator-driven systems in a double-strata fuel
cycle. During fuel fabrication and reprocessing, spent IM-fuel poses
unparalleled requirements during all process steps: Simulations show
unprecedented levels of gamma dose rates, the (alpha and neutron)
activity and the decay heat from those elements. This holds true for
low and high burn-ups.
Despite the notably smaller size of the IM-fuel elements, their
gamma dose rates are significantly higher than for modern LWR fuel
elements. These fuel elements must be shielded appropriately during
storage and transport to reprocessing facilities.
Unlike the dose rates, activity and decay heat are analyzed looking
at values per volume. This mirrors the situation in a reprocessing
facility when the spent fuel elements are already disassembled and
dissolved. The activity of spent IM-fuel is several orders of magnitude
higher than the activity of spent fuel from present reactor types, including fast reactor spent fuel. This affects the required shielding during
storage, transport and reprocessing. The surrounding materials are
exposed to high material stress. Solvents and agents used in the process
must function under these radiation levels. This is a big challenge for
aqueous processes (NEA/OECD, 2018).
Comparing activities of different fuel types, values for IM-fuel are
not only higher directly after discharge. Due to the high contribution

of transuranium elements, the values decline slower than in conventional UOX and MOX fuel. Some of these transuranium elements such
as Californium contribute significantly to the neutron activity, even
though present only in small amounts (NEA/OECD, 2005, p. 51). In
2006, the Nuclear Energy Agency stated that the activity of the spent
fuel high on minor actinides is higher ‘‘than the potentiality of current
reprocessing methods’’ (NEA/OECD, 2006a, p. 34). Besides challenges
to reprocessing, the extremely high 𝛼-activity must be considered during fuel design and fabrication: the increased helium content in the

4.2. Accumulation of long-lived fission products
The radiotoxicity index based on ingestion dose rates implies that
the transmutation of minor actinides and the use of plutonium for
further nuclear energy production could reduce the time for which the
integrity of the final repository must be ensured. For the long term
safety case of a deep geological repository, however, other radionuclides are more important: e.g. the long-lived fission products Zr-93,
Cs-135, I-129, and Sn-126.
Based on the estimated German spent fuel inventory in 2022
(Schwenk-Ferrero, 2013), the additional inventory of the long-lived fission products in case of two different hypothetical P&T implementation
strategies is assessed. In these scenarios, accelerator-driven systems are
used — either for burning all transuranium elements or for incinerating
the minor actinides while using the plutonium in a (foreign) fast reactor
fleet for energy production.
Naturally, the production of fission products continues with ongoing
use of nuclear reactors. However, the adapted IM-fuel, namely its
increased fraction of minor actinides, leads to changes in the spent fuel
composition. As a consequence, in a national scenario, the total amount
of fission products that need to be disposed increases by 36.7% (from
415 t in the baseline scenario) to 567.5 t. At the same time, the amount
of long-lived fission products increases by roughly 50% and the amount
of Cs-135 more than doubles (about 13 tons compared to 5.4 tons in
the base line scenario).

In a regional scenario, minor actinides are irradiated for the sake of
destroying them while the plutonium is burned for energy production
abroad. The fission products must be sent to final storage either way.
In a regional scenario, only the radioactive waste originating from the
7


Progress in Nuclear Energy 145 (2022) 104106

F. Frieß and W. Liebert

waste treatment: P&T programs require a long-term commitments to
the use of nuclear technology.
Based on the results in this paper, it could be further doubted
in how far the implementation of a P&T fuel cycle would ease the
requirements that are placed on a deep geological repository. The
amount of radionuclides that are relevant for the long-term safety
analysis increases.
Accelerator-driven systems are optimized for the transmutation of
minor actinides. Consequently, the deployment is envisioned in doublestrata fuel cycles where the plutonium can be used in critical reactors
for energy production. Due to the nuclear energy phase-out, this could
only be possible in the German case if the plutonium would be transferred to foreign partners. This case is analyzed by the hypothetical
regional scenario. It assumes willing partners that use the plutonium
for energy production and does not consider the additional waste
streams produced in doing so. Consequently, only its effect on the
German inventory that must be sent to final disposal looks comparably
favorable.
In a national scenario, in which all transuranium elements are to be
transmuted in Germany, the inventory of safety relevant radionuclides
increases significantly. Additionally, both scenarios would require a

long-time commitment to operation and use of nuclear facilities.

transmutation of minor actinides must be taken care of by the original
country (in this case: Germany).
As a trade-off for the minor actinide reduction in the deep geological
repository, the inventory of long-lived fission products is increased. But
those are the most relevant isotopes for the long-term safety analysis of
a deep geological repository.
4.3. Conclusion
The challenging characteristics of spent inert matrix fuel (IM-fuel)
and the build-up of long-lived fission products are two facets that must
be considered in a P&T fuel cycle which has to deal with higher minor
actinides content at begin of cycle. They were analyzed in more detail
in this paper.
The irradiation of IM-fuels in an accelerator-driven system was
modeled to analyze their activity, their decay heat and the gamma
dose rate. The results show that the values are sometimes even orders
of magnitude higher than the values obtained from the spent fuel
elements of commercial, thermal nuclear power plants. It is widely
accepted that in a P&T fuel cycle several irradiation cycles are needed
— with intermediate cooling, reprocessing and fuel fabrication steps.
Currently available technologies for reprocessing and fuel fabrication
would not be able to handle the spent IM-fuel elements. To avoid
extraordinary long cooling periods, pyroprocessing technologies for
the separation process would be needed. Those processes have only
been demonstrated on a laboratory level yet (OECD/NEA, 2018). The
possible benefit of a P&T fuel cycle depends highly on the achievable
separation efficiencies for the different transuranium elements. Lower
efficiencies inevitably lead to a higher transuranium inventory in the
final repository. Additionally, fuel fabrication facilities would have

to be cooled continuously independent of the cooling period (Pillon,
2012).
Besides those significant technological challenges there is a question
to the usefulness of transmutation of minor actinides. One common
criteria used while arguing in favor of minor actinide transmutation
is the radiotoxicity index. This index does not consider the mobility of
radionuclides in a repositories environment and the biosphere as well
as other relevant parameters. Therefore, it cannot provide a sufficient
criterion for the long term safety case of a deep geological repository.
If one looks at the dose rate to the public emerging from such a
repository, the main contribution originates from certain long-lived
fission products. The amount of these fission products is inevitably increasing if minor actinides are transmuted through fission. Simulations
show that especially Cesium-135 is over-proportionally produced in a
transmutation fuel cycle. Thus, additional amounts of long-lived fission
products – beyond those already accumulated in the nuclear power
program – would have to disposed of in a final repository.
It is generally agreed upon that five to ten transmutation steps are
needed. Each of those steps changes the isotopic composition of the
fuel. For a detailed analysis of a possible P&T implementation, the
whole period should be simulated and analyzed. This would include
at least several decades, but more likely even centuries (Lyman and
Feiveson, 1998; Kirchner et al., 2015; Frießet al., 2021).
Consequently, those long simulation periods also increase the errors
introduced e.g. by the simulation method (Skarbeli et al., 2020), the
uncertainties in capture and fission cross sections of relevant isotopes (Takeda et al., 2017; Stanisz et al., 2019), and the presence of
unusual isotopes like Cf-252 that reach equilibrium only after long
time (Kooyman et al., 2018; Wu and Wang, 2020). These challenges
must be faced when assessing the actual feasibility and practicability
of a P&T scenario.
In any P&T scenario a deep geological repository is needed. Additionally, there are good reasons why ADS are considered to be

deployed in double-strata fuel cycles. Consequently, it could be questioned whether this strategy is labeled correctly by the term radioactive

CRediT authorship contribution statement
Friederike Frieß: Conceptualization, Methodology, Software.
Declaration of competing interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to
influence the work reported in this paper.
Data availability
Input files related to this article can be found at />juleylene/ADS-Transmutation-Fuel-Paper.git.
Funding
This research did not receive any specific grant from funding agencies in the public, commercial, or not-for-profit sectors.
The authors would like to thank the Viennese Ombudsoffice for
Environmental Protection for its financial support in writing this paper.
Open access funding is provided by University of Natural Resources
and Life Sciences Vienna (BOKU).
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