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

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.

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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 (http://creativecommons.org/licenses/by/4.0/)

Contents lists available atScienceDirect 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

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

A R T I C L E I N F O

Dataset link: https://github.com/juleylene/AD

S-Transmutation-Fuel-Paper.git

Keywords:

Accelerator-driven-system (ADS)

Transmutation

Minor actinides (MA)

Long-Lived Fission Products (LLFP)

Inert-matrix fuel (IMF)

A B S T R A C T 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

build-up 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

Reac-tor for High-tech Applications) project in 2018 (WNN, 2018)

Ac-cording to the European Strategy Report on Research Infrastructure

(ESFRI), MYRRHA is supposed to become operational in 2027 (

ES-FRI,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

∗ Corresponding author

E-mail addresses: friederike.friess@boku.ac.at(F Frieß),liebert@boku.ac.at(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

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 accelerator-driven system ADS are key components for one possible concept

of irradiating high level nuclear waste in a partitioning and trans-mutation (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

https://doi.org/10.1016/j.pnucene.2021.104106

Received 13 May 2021; Received in revised form 9 November 2021; Accepted 17 December 2021

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

sodium-cooled 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

sys-tems 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

ra-dionuclides 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

con-sidered 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

poten-tial 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

2 The reason is the relation between fission and absorption cross sections

It is significantly more favorable in a fast neutron spectrum

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 sub-critical 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 com-pared 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 sub-critical core The simulations of sub-criticality and neutron flux show good accordance with average values for a typical fast reactor sys-tem (Sarotto et al.,2013;Frieß,2017) The cross section of our generic ADS model based on the MYRRHA core designSarotto et al.(2013) and Sarotto(2012) is depicted inFig 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 pro-vided inTable 1 The center assembly hosts the spallation target, which

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 ).

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.

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

burn-up 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

EFIT-like IM-fuel elements These elements do not contain uranium but

Magnesium-Oxide as an inert matrix to avoid further plutonium

breed-ing 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 inTable 1

The core is surrounded by dummy elements filled with lead–bismuth

eutectic and reflector elements consisting of yttrium-stabilized

zirco-nium (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

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 ac-counts 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 𝑘 𝑒𝑓 𝑓to

reach decline During burn-up, 𝑘 𝑒𝑓 𝑓 ranges between 0.971 and 0.954 More details on the burn-up calculation, including mass balanced, can

be found inFrieß(2017)

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 MC-NPX 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)

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 mir-ror 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 transmu-tation on a possible final repository are discussed (e.g in (Abderrahim

et al.,2013;NEA/OECD,2006b;Romero and Abderrahim,2007)

Progress in Nuclear Energy 145 (2022) 104106

F Frieß and W Liebert

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 ).

Fig 2shows 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,

Cs-135 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 inSchwenk-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 inMcGinnes

(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

𝐹 𝑖

𝐺𝐸𝑅 = 𝛼𝐹 𝑖

𝑃 𝑊 𝑅 + 𝛽𝐹 𝑖

𝐵𝑊 𝑅 + 𝛾𝐹 𝑖

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)

Constituent Spent Nuclear Fuel in 𝑡

Long-lived fission products

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 in-ventories 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:

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, simula-tions 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

trans-mute its total inventory of plutonium and minor actinides The sole goal of this approach is the transmutation of the transura-nium elements This would lead to an additional 152.5 t (result-ing from all minor actinides and plutonium as given inTable 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

Progress in Nuclear Energy 145 (2022) 104106

F Frieß and W Liebert

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.

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

proce-dures, 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

avail-able 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

irra-diation, respectively, are shown inFig 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

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 short-lived 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

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 inTable 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,

𝛼-activity is approximated by summing the 𝛼-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 ⋅

1013Bq∕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 transura-nium 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 𝛼-decays 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) andSchwenk-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−1cm−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

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 inFig 4 The Cm-242 contribution is plotted separately to highlight its rel-evance for the total heat In the first year, Cm-242 is the main con-tributor 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

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.

0.0 a 5.11⋅ 10 13 9.05⋅ 10 12 1.16⋅ 10 6

0.1 a 1.68⋅ 10 13 7.93⋅ 10 12 1.09⋅ 10 6

0.6 a 7.54⋅ 10 12 4.19⋅ 10 12 8.48⋅ 10 5

1.0 a 5.17⋅ 10 12 2.71⋅ 10 12 7.48⋅ 10 5

5.0 a 1.70⋅ 10 12 9.18⋅ 10 11 5.52⋅ 10 5

10 a 1.38⋅ 10 12 8.32⋅ 10 11 4.56⋅ 10 5

40 a 7.58⋅ 10 11 5.37⋅ 10 11 1.46⋅ 10 5

100 a 4.10⋅ 10 11 3.40⋅ 10 11 1.63⋅ 10 4

Fig 4 Residual heat derived from simulations after different cooling periods for

IM-fuel 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 inFig 4 Values are taken fromOrlov

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 inFig 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

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

Fig 5 The concentration of certain long-lived fission products in IM-fuel,

PWR-MOX 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.

Baseline Regional National scenario scenario scenario

Important LLFPs

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 inFig 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

Tc-99 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

Cs-135 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 inven-tory for the two hypothetical scenarios and the baseline introduced in the previous section (seeTable 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

Progress in Nuclear Energy 145 (2022) 104106

F Frieß and W Liebert

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

long-lived 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

radio-logical 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

ra-dioactive 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

dur-ing 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,

includ-ing fast reactor spent fuel This affects the required shieldinclud-ing durinclud-ing

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

conven-tional 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

dur-ing fuel design and fabrication: the increased helium content in the

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 prob-lems 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 (compareFig 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 bySalvatores 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 repro-cessing 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

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 radionu-clides 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 fis-sion 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

Progress in Nuclear Energy 145 (2022) 104106

F Frieß and W Liebert

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

in-creasing 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

iso-topes (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

Ad-ditionally, there are good reasons why ADS are considered to be

deployed in double-strata fuel cycles Consequently, it could be

ques-tioned whether this strategy is labeled correctly by the term radioactive

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 double-strata 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 trans-ferred 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

CRediT authorship contribution statement Friederike Frieß: Conceptualization, Methodology, Software Declaration of competing interest

The authors declare that they have no known competing finan-cial 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 athttps://github.com/ juleylene/ADS-Transmutation-Fuel-Paper.git

Funding

This research did not receive any specific grant from funding agen-cies 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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