Preliminary analysis of gaseous radiocarbon behavior in a geological repository hosted in salt rock

Abstract

A preliminary evaluation of gaseous radiocarbon ( ¹⁴ C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of ¹⁴ C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous ¹⁴ C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (K d ) in materials used to seal the disposal underground facility. Results show the possibility to correlate the K d values, the volume and the location of the sealing materials to the amount of ¹⁴ C migrating toward the surface.

Full text

PRELIMINARY ANALYSIS OF GASEOUS RADIOCARBON BEHAVIOR IN A
GEOLOGICAL REPOSITORY HOSTED IN SALT ROCK
Riccardo Levizzari
1*
•Barbara Ferrucci
2
•Alfredo Luce
1
1
ENEA Saluggia Research Centre, Strada per Crescentino 41, 13040 Saluggia (VC), Italy.
2
ENEA Bologna Research Centre, Via Martiri di Monte Sole 4, 40129 Bologna (BO), Italy.
ABSTRACT.A preliminary evaluation of gaseous radiocarbon (
14
C) behavior under geological repository conditions
for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although
in Italy there is still no defined project for a geological disposal facility, current work may support future safety assess-
ment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the per-
centage of
14
C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most
important radiological source. Data about the radiological inventory has been collected to simulate production and
migration of gaseous
14
C in a hypothetical geological repository. Three different conceptual models have been devel-
oped and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the
whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite.
A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution
coefficients (K
d
) in materials used to seal the disposal underground facility. Results show the possibility to correlate the
K
d
values, the volume and the location of the sealing materials to the amount of
14
C migrating toward the surface.
KEYWORDS: geological repository, radioactive waste, radiocarbon, salt rock.
INTRODUCTION
The present work has been carried out within the EC CAST project (CArbon-14 Source Term),
which aims to develop understanding of the generation and release of radiocarbon (
14
C) from
radioactive waste materials under conditions relevant to waste packaging and disposal
in geological repositories. In the framework of different analyses supporting the long-term radi-
ological safety of underground repositories, the migration of
14
C into the environment is a key
issue. Radiocarbon has a relatively long half-life (5730 yr) and, depending on its speciation, a high
mobility in the environment and a high efficiency of incorporation into the human body via the
food-chain. During the post-closure phase in the life-cycle of a geological repository, significant
quantities of
14
C labeled gases (i.e. methane and carbon dioxide) might be produced by corrosion
of metals, irradiated graphite and by microbial degradation of organic waste; these chemical
species might then be released from the disposal system and reach the biosphere in various ways.
The understanding of the key mechanisms influencing
14
C transport from the repository to the
surface is thus critical in this context and within the safety assessment of geological waste disposal.
Therefore, it is important to gain new scientific understanding of different mechanisms dealing
with the dynamics of radionuclides inside and outside future repositories.
In Italy, according to international practices, radioactive high level waste-long lived (HLW-LL) and
intermediate level waste (ILW) are intended to be disposed of in an underground repository. In the
meantime, until a site for underground disposal is selected, these waste products will be placed in a
near surface repository for interim storage along with the final disposal of low level waste (LLW).
The site for the interim storage has not been yet selected. The model simulations discussed here
present a preliminary safety analysis of a hypothetical Italian geological repository hosted in salt
*Corresponding author. Email: riccardo.levizzari@enea.it.
Radiocarbon, 2018, p. 1–14 DOI:10.1017/RDC.2018.133
Selected Papers from the European Commission CAST (CArbon-14 Source Term) Project–A Summary of
the Main Results from the Final Symposium
© 2018 by the Arizona Board of Regents on behalf of the University of Arizona. This is an Open Access
article, distributed under the terms of the Creative Commons Attribution licence (http://creativecommons.
org/licenses/by/4.0/), which permits unrestricted reuse, distribution, and reproduction in any medium,
provided the original work is properly cited.
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rock. In preparation for future assessment studies of post closure safety, the fate of gaseous
14
Cin
geological disposal conditions and its radiological relevance are analyzed. The first phase of
this work has consisted in defining the radiological inventory of
14
C bearing radioactive waste
(HLW-LL and ILW); in a second phase, the evaluation of the radiocarbon release and its dynamic
in gaseous phase inside the repository has been evaluated. Three 3D models have been designed and
evaluated; their main differences consist of the layout of the underground repository, of the type,
features and volume of sealing materials and of the type and volume of waste disposed of. In
particular, whereas in the former case, all the waste in the Italian inventory has been considered; in
the latter two cases only the irradiated graphite waste (ILW) has been considered. Irradiated gra-
phite generated from nuclear power plant dismantling has been a major issue in past nuclear
research, due to various products of activation, including
14
C and tritium. These two radionuclides
are critical in the context of the safety analysis for both underground and surface repositories,
because they are a possible source of radioactive gaseous species. In fact, even today there are
uncertainties related to many aspects of irradiated graphite management and disposal, as its inter-
action with the geochemical environment in which it is disposed, the distribution of the
14
Cwithin
the waste and the form in which the
14
C is released (i.e. organic, inorganic, liquid or gaseous). Some
uncertainty also remains for the possible degradation of radioactive waste due to microbial activity
in harsh conditions, caused by the absence of water, low porosity and high alkalinity induced by the
cementitious near-field of a repository (Grant et al. 1997; Sorokin 2005). The behavior of
14
Cforthe
final disposal of HLW and ILW has been analyzed in various geological formations, as for example
for crystalline rocks by Poskas et al. (2016). They have considered the disposal of irradiated graphite
in a deep geological repository, evaluating the
14
C transfer into the geosphere via the groundwater
pathway. The speciation of the released
14
C into organic and inorganic species was taken into
account. The role of backfill and sealing material in a geological repository is fundamental to retard
the migration of radionuclides. Indeed, the hydraulic conductivity of the backfill material for
organic
14
C, and its distribution coefficient K
d
(m
3
/kg) for inorganic
14
C, that quantify the mass
partitioning between the solid and the gaseous phase, represent the key parameters influencing the
gaseous carbon flux into the geosphere.
MATERIALS AND METHODS
This work has been performed taking into account data of previous Italian studies on the
management and disposal of radioactive waste (Luce et al. 2009) and data of the Italian
radioactive waste inventory (Bove et al. 2009; Capone et al. 2011). The current inventory of
Italian radioactive HLW-LL and ILW containing
14
C to be placed in a geological disposal is
reported in Table 1. The inventory does not include the residual waste from the reprocessing of
irradiated fuel sent abroad (UK and France), to be returned to Italy as vitrified waste.
The conceptual models of the underground disposal facility have been designed by analogy with
other existing geological repository projects (e.g. the U.S. Waste Isolation Pilot Plant, WIPP)
(Nuclear Waste Partnership LLC 2016; Sevougian et al. 2016). The hypothetical underground
facility was assumed to be located to a depth of about 800 meters, hosted in a salt body, 200 m
thick, below clay rock about 700 m thick (Figure 1). A geological barrier is guaranteed by the
salt rock, isolated from the surface by two different clay formations; the salt formation is
considered quite homogeneous, although some discontinuities, as clay inclusions, cannot be
excluded. Bentonite, concrete, asphalt and crushed salt have been used as sealing materials
inside the underground facility.
The conceptual models of the waste-repository system have been elaborated considering the
source of the
14
C. The predominant percentage of
14
C in the inventory is related to irradiated
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graphite, which represents the most important radiological source in the context of Italian
14
C bearing radioactive waste. In nuclear power plants, where graphite is used as moderator of
nuclear reactions, the production of
14
C is related to the activation of
14
N,
13
C, and
17
O, which
are the precursor species. The volume of graphite used in this study is referred to as untreated
waste and is related to the decommissioning of the shutdown Italian MAGNOX nuclear reactor
(Latina NPP).
Figure 1 Geological stratigraphy of the host rock.
Table 1 Inventory of Italian HLW-LL and ILW containing
14
C.
Origin Materials Volume (m
3
)
Inventory
of
14
C (GBq)
GCR-Magnox reactor
(Latina NPP)
Graphite 3.30E03 2.83E04
Nuclear power plants Resins, metals n.a. 3.30E03
Medical, industrial,
research
Conditioned
sources
172.60 (863 drums 200 dm
3
) 0.42
Not treated
sources
6.78 (113 metallic drums 60 dm
3
) 92.80
Cemented
sources
113.60 (284 metallic drums 400 dm
3
) 106.28
Solid treated
sources
42.00 (105 drums) 11.00
Not treated
liquid waste
4.19 (192 plastic drums 20 dm
3
;
2 metallic drums 120 dm
3
;
1 metallic drum 110 dm
3
)
4.05
Not treated
solid waste
15.06 (251 metallic drums 60 dm
3
) 2.36
Simulation of
14
C Dynamic in Geological Repository 3
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The analysis about the release and migration of
14
C in near-field and far-field of a hypothetical
geological repository has been carried out by means of numerical modeling, using
the TOUGH2/EOS7R multi-phase code, with PetraSim software as the graphical interface
(Oldenburg and Pruess 1995). In the transport modeling, the sorption of
14
C in the solid phase is
taken into consideration through the distribution coefficient K
d
. The objective of the modeling
is to assess both the role of the layout and of the volume of different materials constituting the
engineered barriers, and the importance of the K
d
coefficient in delaying the migration of
14
C,
released in the gaseous form from the waste. Although the K
d
parameter is a factor related to
the partitioning of a radionuclide for the aqueous phase only, and there is no explicit dis-
tribution coefficient for the gas phase, in the TOUGH2 volatilization model, the retardation
occurs for volatile species in the gas phase, just as it does in the aqueous phase, through the
assumption that the gas phase consists primarily of air (Oldenburg and Pruess 1995).
To evaluate the dynamics of the gaseous radiocarbon in the repository, and to quantify its
radiological relevance, three conceptual models have been developed, associated with two
different 3D layouts. The first one, referred to a 3D multi-room model (MR), represents a whole
system containing all the waste inventory listed in Table 1; three different cases have been
simulated (MR1, MR2, MR3) using different volumes of bentonite to seal the underground
repository. The latter two conceptual models describe systems containing only the irradiated
graphite and are referred to as a 3D single-room model with conservative approach (SRC) and
with realistic approach (SRR). In the SRC model four different cases have been simulated
(SRC1, SRC2, SRC3, SRC4) using different K
d
values for sealing materials; in the SRR model
only one case has been simulated. All the models include a single shaft that connects the
underground repository to the ground surface. No assumption has been made about the
14
C
speciation. For what concern the emission from graphite, an important fraction of the
14
C is not
releasable, because it is strongly bounded to the graphite matrix. Moreover, the total fraction
and the fraction of each gaseous species (mainly CO
2
and CH
4
) that migrates outside the waste,
depend on the history of graphite (i.e. irradiation history, operational conditions, etc.) (Toul-
hoat et al. 2018); but due to the lack of information about the Italian graphite inventory, these
data are not currently available and conservative assumptions will be used in the simulations.
Since at the moment some aspects of the complex processes simulated in disposal rooms are not
known, the elaboration of the conceptual models has been performed with an arbitrary set of
conservative assumptions. This approach to conservative parametrization supports aspects of
sealing performance studies. Salt formations are favorable rocks to host a geological disposal
because they maintain unsaturated conditions; however, salt formations might not be con-
sidered completely dry because of the presence of water/humidity related to repository opera-
tions and especially because of the presence of brine, that might move from the hosting rock to
the excavated zone due to pressure gradient, thermal gradient or rock stress gradient (Rübel
et al. 2013; Sandia National Laboratories 2013). This condition, combined with the possible
assumption of complete saturation with brine of the disturbed rock zone close to the under-
ground excavation, results in a possible water flow into the repository (Sandia National
Laboratories 2006). In this work, a conservative approach has been adopted and the most
unfavorable repository conditions have been assumed, considering a constant brine flow into
the storage rooms, which can cause the corrosion of carbon steel waste packages and the
leaching of
14
C from waste. It is assumed that no overpack is provided to contain drums and, in
each room, waste is considered as compacted volume. No convergence phenomenon of the host
rock has been considered. In all three models, the evolution of described arbitrary scenario has
been modeled with two main time steps: the first, from 0 to 300 yr after repository closure, when
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no release of radionuclides occurs and waste package integrity is maintained (Sullivan 2004);
the second, from 300 yr to 300,000 yr after the repository closure, when steady-state conditions
are established, waste packages corrosion and
14
C release take place, due to the presence of
brine and water inside the repository.
The main purpose of the MR model has been to test the performance of the barrier system in
delaying the migration of gaseous
14
C, by considering different layouts and volumes of the
sealing materials at the repository level. The layout has been designed on the basis of the WIPP
project (WIPP 2009). It consists of 16 rooms arranged in two panels, separated by a central
pillar of intact salt and connected to the ground surface by a single shaft (Figure 2).
The engineered barrier system includes four materials that completely fill and seal the shaft,
rooms and drifts. No disturbed rock zone has been included in this conceptual model; the base
of the shaft, after repository closure, consists of a monolith of concrete, 50 m high, 10 m long,
and 10 m thick. In each storage rooms, the total amount of the estimated
14
C activity is assumed
equally distributed within the volume of the waste. The simulations start from a steady-state
initial condition calculated assuming an initial pressure of 15 MPa (lithostatic pressure) at the
repository level (Kristopher 2014). This is a conservative assumption that implies, immediately
after the repository closure, a lithostatic pressure in the storage room. During all the compu-
tation time, this conditon, assuming a constant gas production within the storage rooms, gen-
erates a pressure profile that provides a gas flow from the underground repository to the top of
the shaft. In this scenario only migration by diffusion has been modeled. The source term is
modeled as a gaseous emission of
14
C with a release rate of 10% per year of the total activity; this
highly conservative and unrealistic release rate has been intended to force the model to produce
results that bring out the effects of different features of the sealing materials on the
14
C
migration. In order to make a preliminary sensitivity analysis, the volume of bentonite has been
selected as the focus of the sensitivity analysis, because of its very low permeability and its
Figure 2 Multi-room conceptual model.
Simulation of
14
C Dynamic in Geological Repository 5
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capacity to retard most radionuclides by sorption (Daeman and Ran 1997). Three case studies
have been developed; in MR1 case, no bentonite is used to seal the underground facility; in
MR2 case, a bentonite volume of 3657 m
3
is placed around the lower part of the monolith
(underground shaft closure); in MR3 case, a bentonite volume of 9508 m
3
is distributed along
the entrance of every disposal room and around the base of the monolith (Figure 3).
In these cases, generic features have been used for the sealing materials. According to Yim and
Caron (2006), a K
d
value of 1.0E-5 m
3
/kg has been assumed for asphalt, waste form and
concrete monolith, while K
d
=0 for the crushed salt and intact salt; because of its sorption
capacity, a high K
d
value (1 m
3
/kg) for bentonite has been assumed (Daeman and Ran 1997).
The layout of shaft sealing materials has been maintained constant in all cases (Figure 4).
In the SRC model, a new barrier system configuration has been taken into account, based on
Freeze et al. (2013), and only irradiated graphite is considered. Only one storage room has been
considered, containing 196 hypothetical waste packages, each with a volume of 1 m
3
. All the
activity of
14
C related to the irradiated graphite is assumed equally distributed throughout these
packages. Crushed salt, waste form and shaft seal materials constitute the repository barrier
system. A disturbed rock zone, 10 m thick, has been modeled as volume of rock, surrounding
the excavated zone, that experiences durable changes in hydraulic parameters, as porosity and
permeability (Freeze et al. 2013). The source term is modeled as a gaseous emission of
14
C,
characterized by a constant release rate of 0.0067% per year of the total
14
C activity contained
in graphite (Fugaru 2018). Where not specified in the following description, input data assumed
for the simulations are the same assumed for the MR model. Figure 5 illustrates the config-
uration of materials featuring the shaft sealing, as defined for a generic vertical shaft in US
WIPP, that consist of concrete, asphalt, salt and bentonite, to enhance physical and chemical
barrier to radionuclide migration. The values of their main parameters are reported in Table 2.
A small inverse Henry’s constant has been assigned, such that gaseous
14
C mostly volatilizes
leading to a worst case scenario (Oldenburg and Pruess 1995). Four main cases have been
simulated (SRC1, SRC2, SRC3, SRC4), starting from the most conservative assumption
(SRC1) in which, for all the sealing materials, the K
d
value is 0. In the other three cases, a K
d
value of 1.0E-5 (m
3
/kg) has been assumed for the waste form, the bentonite, the concrete and the
asphalt (Table 3).
The SRR model has been developed in analogy with the previous one. To evaluate the radi-
ological impact of
14
C in a more realistic way, a less conservative approach has been assumed by
using a K
d
>0 for all sealing materials (Table 3). The main features of the previous conceptual
Figure 3 Sealing materials configuration in the MR model.
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model have been preserved. In this simulation, the release of
14
C from the irradiated graphite
has been analysed with a more detailed approach. According to Doulgeris et al. (2015), two
main dynamics featuring the behavior of
14
C have been considered; first, the CO
2
and CH
4
released from graphite are partitioned between the solution and the gas phase, according to the
Henry’s law. The second is related to the contribution of cementitious materials in attenuating
14
C migration, because of CO
2
carbonation; in fact, as highlighted by Poskas et al. (2016),
encapsulation of graphite inside a container with cementitious material would decrease the
maximum flux of inorganic
14
C into the geosphere, by approximately one order of magnitude.
Methane is the predominant form of gaseous
14
C considered for this case, with no adsorption
capacity into the salt rock formation (K
d
=0m
3
/kg). A congruent release of gaseous
14
C starts
300 yr after the repository closure, with a release rate of 0.0067% per year, as gaseous
14
CH
4
(Fugaru 2018). Migration is assumed to take place only by diffusion.
RESULTS AND DISCUSSION
The
14
C amount was calculated for two specific blocks common to the three models: near the
compacted hearten fill at the top of the shaft, and near the concrete monolith at the base of the shaft.
Figure 4 Sealing materials for the shaft in
the MR model.
Simulation of
14
C Dynamic in Geological Repository 7
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The results of the multi-room and single-room models are not comparable, because of the different
discretization and conceptualization of the models; therefore, the data are reported in different
graphs. The simulation results of the MR model, cases MR1 and MR2, show that, at the base of the
shaft in the monolith, the presence of a bentonite volume of about 3600 m
3
leads to a decrease of 3%
of the cumulative amount of
14
C. Increasing the bentonite volume of about three times, from case
MR2 to case MR3, the amount of
14
C decreases of 23.5%, from 2.3E10 Bq to 5.3E09 Bq (Figure 6).
In all three cases, due to the large volume of bentonite (about 4.0E05 m
3
) and due to the other
sealing materials in the shaft, the maximum cumulative amount of
14
C at the top of the shaft is
Figure 5 Sealing materials for the shaft in the
SR model.
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Table 2 Values of main parameters of the sealing materials used in the simulations.
Gaseous
14
C parameters
Diffusion coefficient in gas phase (m
2
/s) 1.3E-4 (Quintessa Ltd. & Geofirma Engineering Ltd. 2011)
Diffusion coefficient in liquid phase (m
2
/s) 4.9E-11 (Quintessa Ltd. & Geofirma Engineering Ltd. 2011)
Inverse Henry’s Constant (1/Pa) 3.0E-10 (Quintessa Ltd. & Geofirma Engineering Ltd. 2011)
Hydraulic parameters
Permeability (m
2
) Porosity (-)
Concrete/monolith 1.0E-19 (Enssle et al. 2014) 0.250 (Enssle et al. 2014)
Bentonite 1.0E-19 (WIPP 2009) 0.290 (Enssle et al. 2014)
Crushed salt 1.0E-18 (Mariner et al. 2015) 0.113 (Mariner et al. 2015)
Asphalt 1.0E-20 (WIPP 2009) 0.010 (WIPP 2009)
Intact salt 1.0E-23 (WIPP 2009) 0.018 (WIPP 2009)
Disturbed rock zone 1.1E-16 (Sandia National Laboratories 2013) 0.013 (Mariner et al. 2015)
Waste 1.0E-13 (Mariner et al. 2015) 0.300 (Mariner et al. 2015)
Earthen 1.0E-14 (WIPP 2009) 0.300 (WIPP 2009)
Simulation of
14
C Dynamic in Geological Repository 9
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Table 3 K
d
(m
3
/kg) values of the sealing materials used in the simulations.
K
d_14C
(MR 1, 2, 3)
K
d_14C
(SRC1)
K
d_14C
(SRC2)
K
d_14C
(SRC3)
K
d_14C
(SRC4)
K
d_14C
(SRR)
Concrete monolith 1.0E-05
(Yim & C. 2006)
0 0 1.0E-05
(Yim & C. 2006)
0 0.5
(Enssle et al. 2014)
Bentonite 1.0
(Daeman & R. 1997)
0 1.0E-05
(Yim & C. 2006)
0 0 1.9E-02
(Wersin et al. 2014)
Crushed salt 0 0 0 0 0 0
Asphalt 1.0E-05
(Yim & C. 2006)
0 0 0 1.0E-05
(Yim & C. 2006)
1.0E-05
(Yim & C. 2006)
Intact salt 0 0 0 0 0 0
Disturbed rock zone 0 0 0 0 0 0
Waste 1.0E-05
(Yim & C. 2006)
0 0 0 0 1.0E-05
(Yim & C. 2006)
Earthen 1.0E-05
(Yim & C. 2006)
0 0 0 0 1.0E-05
(Yim & C. 2006)
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orders of magnitude lower than in the monolith at the base of the shaft (about 2.3E-1 Bq in the
case MR1, about 1.4E-2 Bq in the case MR2 and about 2.0E-3 Bq in the case MR3) (Figure 7).
The simulation results of the SRC model (cases SRC1, SRC2, SRC3, SRC4) show that sealing
materials with values of K
d
lower than 1.0E-5 m
3
/kg do not affect the cumulative amount of the
gaseous
14
C, with quite similar results in all the four cases. The peak of
14
C amount at the top of
the shaft occurs at about 4000 yr after the repository closure, with a maximum value of 2.7E03
Bq. After 300,000 yr, the
14
C amount at the top of the shaft becomes negligible in all the
simulated cases, decreasing to a minimum value of about 1.0E-15 Bq. The peak of the
14
C
Figure 7 Cumulative amount of
14
C at the top of the shaft (MR model).
Figure 6 Cumulative amount of
14
C in the monolith at the base of the shaft (MR
model).
Simulation of
14
C Dynamic in Geological Repository 11
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gaseous emission in the monolith at the base of the shaft occurs about 400 yr after the repository
closure with value of about 1.0E04 Bq (Figure 8).
In the SRR model, the peak of the
14
C gaseous emission at the top of the shaft, occurs about
17,000 yr after the repository closure with about 1.4E02 Bq. Also in this case, the peak of the
14
C gaseous emission in the monolith occurs about 400 yr after the repository closure with value
of 1.0E04 Bq (Figure 9).
Figure 8 Cumulative amount of
14
C (SRC1 case, representative of the four cases
simulated).
Figure 9 Cumulative amount of
14
C (SRR model).
12 R Levizzari et al.
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CONCLUSION
The paper reports a preliminary analysis of the radiological impact of gaseous
14
C from HLW-
LL and ILW disposed in a hypothetical Italian geological repository hosted in salt rock. Before
this work, no safety assessment or generic safety studies had been developed in the Italian
context for geological disposal. The three conceptualizations of the underground facility, and
the following simulations, have improved the knowledge about the future safety issues that will
have to be discussed for final disposal of Italian HLW-LL and ILW. The results of the MR
model confirm that the capacity of materials to retard radionuclides migration depends on their
K
d
value and their volume in the repository. High values of K
d
significantly affect the barrier
system performances. From detailed analysis of the simulation results of the MR model, taking
into account the cases 2 and 3, it is evident that, for high values of K
d
, the performance of the
sealing materials is proportional to their volume and density. Increasing the total volume of the
bentonite by about three times, moving from the case 2 to the case 3, the cumulative amount of
14
C in the monolith is reduced of about one order of magnitude (from 2.3E10 Bq to 5.3E09 Bq).
At the top of the shaft, in all three cases, the significant decrease of the cumulative amount of
14
C, with respect to the monolith, depends mainly on the large volume of bentonite and of the
other sealing materials in the shaft. In the SRC and SRR conceptual models, a more realistic
scenario than the MR model has been simulated. In particular, in SRR model a release rate of
0.0067% per year and a K
d
>0 for all the sealing materials have been assumed. The maximum
value of the cumulative amount of
14
C calculated in the SRR model at the top of the shaft
(1.4E02 Bq) is about one order of magnitude lower than the cumulative amount calculated in
the cases of the SRC model (2.7E03 Bq).
The results of this study show that, if disposed in a hypothetical geological repository hosted in
salt rock, a low radiological impact of
14
C from the Italian graphite waste and other HLW-LL
and ILW can be expected. The mitigation of
14
C migration towards the surface, and the fol-
lowing reduction of the radiological impact, are related to the combined actions of the low
conductivity of salt host rock, of the adsorbing capacity of sealing materials and of the radio-
active decay. Moreover, the conservative assumptions used in the various simulations (i.e.
saturation with brine for all the calculation time, relatively high permeability of the disturbed
rock zone and backfill around the waste packages, etc.) might have produced overstated values
of
14
C at the top of the shaft.
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