Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system

Nuclear Engineering and Design 330 (2018) 241–249

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Nuclear Engineering and Design
journal homepage: www.elsevier.com/locate/nucengdes

Transmutation considerations of LWR and RBMK spent nuclear fuel by the
fusion–fission hybrid system

T

R. Plukienėa,⁎, A. Plukisa, L. Juodisa, V. Remeikisa, O. Šalkauskasa, D. Ridikasb, W. Gudowskic
a

Center for Physical Sciences and Technology, Savanoriu 231, LT-02300 Vilnius, Lithuania
International Atomic Energy Agency, Vienna International Centre, PO Box 100, 1400 Vienna, Austria
c
KTH (Royal Institute of Technology), AlbaNova University Centre, 106 91 Stockholm, Sweden
b

A R T I C L E I N F O

A B S T R A C T

Keywords:
Fusion–fission hybrid system
Incineration of trans-uranium elements
LWR and RBMK spent nuclear fuel

The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma

based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR
(Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT. The IAEA
nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of
heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE 6 code package has been used for the
determination of the RBMK-1500 spent nuclear fuel composition. The total amount of trans-uranium elements
has been estimated in both countries by 2015. Major parameters of the hybrid system performance (e.g., kscr, keff,
Φn(E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation
cases. Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material
and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system.
About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the 3 GWth
fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition. The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation
time has been estimated.

1. Introduction
When trans-uranium elements (TRU) are removed from the discharged fuel destined for disposal, the radiotoxic nature of the remaining materials drops below that of natural uranium ore within a
period of 500 years (Ewing, 1999). Hence the possibilities of partitioning and transmutation of the long-lived radioactive waste into
stable or short-lived isotopes, which could then be surface-stored with
the little/no proliferation value, are now under investigation. In addition, TRU elements could serve as fuel for transmutation systems. The
transmutation steps include the reprocessing process, fuel fabrication,
management of secondary wastes, etc. and it is likely to be the most
challenging issue to be solved in the sustainable nuclear fuel cycle.
These problems are comparable for all types of transmutation systems
since they are related to the specific TRU content of the transmutation
dedicated fuels (Salvatores, 2009; Yurov and Prikhod’ko, 2014). The
idea of the “closed” fuel cycle of plutonium/higher actinides and TRU
recycling could be implemented in the future fast neutron systems

(Salvatores, 2002). The fast reactor technology is one of the most
promising when compared with others at present. Some commercially
available examples such as BN-600, BN-800 in Russia or FBTR (Fast

Breeder Test Reactor) in India could be mentioned. Other FBRs (mostly
research) have been built and operated in the United States, the United
Kingdom, France, the former USSR, India and Japan (Waltar and
Reynolds, 1981). It is true that light-water reactors dominate nuclear
power today due to nowadays relatively low uranium prices and
availability, lower capital costs (by 25%) compared with fast neutron
reactors, but still in Generation IV reactor initiative three from six GEN
IV reactors are of the Fast Reactor (FR) type: Gas-cooled Fast Reactor
(GFR), Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor
(LFR) and Supercritical Water-cooled Reactor (SCWR) (GEN IV, 2016).
The latest system has an option for the actinide management based on
the particular core design with the fast neutron spectrum (Oka, 2010).
The MSFR concept is also recognized as one of Gen IV options because
of the compact size and relatively low costs to both build and operate
(less metal is needed to fabricate/maintain and no initial fuel

Abbreviations: ADS, Accelerator Driven System; Flibe, LiF-BeF2-(HN)F4 molten-salt blanket; FP, Fission Products; LWR, Light Water Reactor; MA, Minor Actinides; SNF, Spent Nuclear
Fuel; RBMK, High Power Channel-type Reactor (in Russian abbreviation); TBR, Tritium Breeding Rate; TRU, Trans-Uranium Elements
⁎
Corresponding author.
E-mail address: (R. Plukienė).
/>Received 25 September 2017; Received in revised form 20 January 2018; Accepted 27 January 2018
Available online 20 February 2018
0029-5493/ © 2018 The Authors. Published by Elsevier B.V. This is an open access article under the CC BY-NC-ND license ( />

Nuclear Engineering and Design 330 (2018) 241–249

R. Plukienė et al.

power, a 250 MW power fusion device operating with a deuterium–tritium fuel cycle is sufficient to provide an external 14 MeV neutron

source of 9 × 1019 n/s.
In (Ridikas et al., 2006) we have shown that in spite of very promising results on the efficient TRU destruction a significant quantity of
curium isotopes is accumulated, while for the rest of actinides fully
equilibrated concentrations are reached. In the parallel paper (Plukiene
et al., 2006) an optimization procedure has been developed in order to
determine the optimal fission-blanket composition corresponding to the
fast incineration rate of all actinides. For this purpose a different fissionblanket composition (based on Be + F-salt: 14.29% – 6Li + 7Li,
57.14% – Be, 28.57% – F (Be salt)) corresponding to the fast incineration rate of actinides including minor actinides (also Cm isotopes) was
investigated. In both F-salt and Be + F-salt cases the same incineration
rate (1.1 tons/year) of minor actinides was obtained, but Be + F-salt
showed better hybrid-system performance characteristics and a smaller
actinide mass in the blanket: kscr at 0.82 ± 0.002 and lower fusion
power of 260 MW at equilibrium, which is reached after about 700 d.
However, in the present study we have analyzed feasibility of different
spent nuclear fuel transmutation options in the fusion–fission hybrid
system with conventional LiF-BeF2-(HN)F4: 28.57% – 6Li + 7Li,
14.29% – Be, 57.13% – F (F salt), which is also usually employed in ADS
systems analysis (Henderson, 2011). All calculations of the flibe-based
actinide transmutation blanket were made employing the Monteburns
code system (Trellue and Poston, 1999). MCNP (Briesmeister, 2000)
was used to obtain neutron multiplication factors keff and kscr (comprising the external source neutron input to the total neutron multiplication factor of the system) of the TRU blanket and also to estimate
the neutron flux.kscr is defined as:

fabrication, handling, durability, shuffling, transport, reprocessing, or
fuel refabrication and radioactive waste management costs), and the
lesser environmental impact (mine tailings, etc.) (Siemer, 2015). One of
the reasons, among others, why GEN IV reactors are considered is the
fact that they are expected to “close” the fuel cycle by significantly
reducing high level waste. The same applies to Accelerator Driven
Systems (ADS) and fusion–fission hybrid systems, which are considered

by many research organizations. ADS for radioactive waste transmutation and energy production for the first time were proposed by Rubbia
(1997)) and Bowman (1992). For implementation of ADS, the spallation target (neutron production) and proton beam optimization are
needed together with the development of new spent fuel partitioning
technique and materials technology (Henderson, 2011).
The plasma-based fusion device could provide as intense neutron
source as a high power accelerator up to 1019 n/s (Cheng, 1998), and
this does not require very high fusion based reactors (∼200–300 MW
thermal would be sufficient for our purposes). One of the attractive
criteria and advantages of the fusion–fission hybrid system is that, the
same as ADS, it is designed to be always sub-critical. It is still too early
to talk about the self-sustained fusion–fission hybrid system feasibility
from the industrial point of view, but we believe that as soon as the
fusion International Thermonuclear Experimental Reactor (ITER) project is completed, the demonstration of a molten-salt blanket as a
medium for trans-uranium actinides (TRU) for nuclear waste transmutation could be possible as one of the applications of the fusion
system – the next step for closed the fuel cycle accomplishment. The
progress in the plasma-based fusion device technologies is tangible: the
plasma fusion driver can be designed based on latest progress in the
construction and operation of the (ITER) (Merola et al., 2014) and EAST
(Experimental Advanced Superconducting Tokamak) (Wu et al., 2011).
There is a growing interest within the fusion community in US revisiting the concept of the fusion–fission hybrid reactor (Freidberg and
Kadak, 2009; Kotschenreuthera et al., 2009). Ambitious plans have
been set up to generate more than 200 GWe of nuclear power in China
in 2050 and a fusion-driven subcritical system concept based on viable
technologies has been proposed (Wu et al., 2009, 2011). This system is
expected to recycle nuclear waste making the energy production more
environmentally friendly.
The main parameters defining the transmutation process are the
neutron energy spectrum and neutron fluxes of the system. In principle,
any intensive neutron source can be used for waste transmutation
(Slessarev and Bolov, 2003). The more intensive neutron flux determines a shorter lifetime of a certain nucleus in the system flux. On

the other hand, transmutation depends on the neutron capture and the
neutron-induced fission cross sections. A unique solution is impossible
because the capture and fission cross sections vary considerably from
one isotope to another. Looking for a new efficient and economically
viable transmutation system the main consideration is given to the
neutron flux and spectrum characteristics of a particular system
(Salvatores, 2002) as well as the interaction of this neutron flux with
the given composition of transmutable material.
For the fusion–fission hybrid system both the materials and technology development is still needed. However, the neutronic characteristics in the transmutation blanket have been studied in (Cheng
and Wong, 2000; Cheng, 2001; Ridikas et al., 2006). An inertial confinement fusion (ICF) device (based on the D + T → 4He+n nuclear
reaction) could provide a powerful neutron source. The 1 MW fusion
power corresponds to ∼4 × 1017 n/s. A molten-salt blanket (LiF-BeF2(HN)F4 – “flibe”), surrounding this neutron source, then could serve as
a medium for trans-uranium actinides (TRU) to be burned. Flibe also
has a function of both the coolant and the carrier of tritium breeding
material (6Li in this case). A well known advantage of the molten salt is
its possibility of both refueling of burned TRU and extraction of fission
products (FP) on-line. The averaged neutron flux is very high (of the
order of ∼1.5·1015 n s−1 m−2) and it corresponds to the flux typical
only of high-flux reactors. In order to produce 3000 MWth fission

kscr = (Mn−1)/(Mn−1/ ν )

(1)

where ν – is the average number of neutrons per fission and Mn is a total
neutron multiplication factor of the system.
In the scope of the EURATOM project BRILLIANT, which has been
initiated to identify the barriers for developing nuclear power in the
Baltic region countries and to prepare the basis for overcoming them,
the alternative of closed fuel cycle technologies instead of the open

cycle with the spent nuclear fuel (SNF) disposal has been analyzed. One
of the project objectives is to identify the milestones for the development of the sustainable nuclear fuel cycle using the transmutation
technology. The main aim of the work is considerations of the feasibility study of waste transmutation in the epithermal-fast neutron flux
of this hybrid system by comparing transmutation scenarios of actinides
extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear
fuel in the scope of the EURATOM project BRILLIANT. This research
was done from the point of view of the reactor core physics and of the
isotopic composition of the fuels which would ensure at least the theoretical part of the transmutation feasibility. In particular, the transmutation scenario of TRU elements has been simulated by using the
model of the fusion–fission hybrid system for different spent nuclear
fuels accumulated in the Baltic countries: RBMK-1500 (Lithuania), BWR
and PWR (Sweden) reactors. In the framework of the EURATOM BRILLIANT project the results of this study provide valuable data for
analysis and optimization of nuclear fuel cycle options in the Baltic
region in case of the further development of nuclear power.
In order to investigate the transmutation efficiency of the fusion–fission hybrid system for different spent nuclear fuel TRU, the
actinides extracted from the light water reactor (LWR) (averaged
composition of SNF TRU) and RBMK spent fuel (different cases) were
analyzed. Major performance parameters of the transmutation system
(e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) for different TRU
cases have been investigated. Detailed burn-up calculations with continuous feeding of TRU material to refresh the burned one and partial
treatment of fission products have been performed in the modeling.

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Table 1
Existing PWR and BWR type nuclear reactors in Sweden and RBMK-type reactors in Lithuania. The average parameters of LWR and RBMK reactors are marked in bold.

Name

Type

Fuel enrich-ment

Power (MW)
(PRIS)

Loading factor (%)

Operation time, y. (until 2015 y.)

Partial input to SNF

Burn-up MWd/kg
(U.S DOE/EIS, 2008)

Forsmark 1
Forsmark 2
Forsmark 3
Oskarshamn 1
Oskarshamn 2
Oskarshamn 3
Ringhals 1
Ringhals 2
Ringhals 3
Ringhals 4
Barsebäck 1
Barsebäck 2

Sweden
Ignalina-1
Ignalina-2
Lithuania

BWR
BWR
BWR
BWR
BWR
BWR
BWR
PWR
PWR
PWR
BWR
BWR
LWR
RBMK
RBMK
RBMK

2.5
2.1
2.8
2.3
2.5
2.6
2.8
3.2

2.8
4.0
2.9
3.2
2.1–4
2–2.6
2–2.8
2–2.8

984
1120
1167
473
638
1400
881
807
1063
1118
600
600
1000
1500
1500
1500

82.2
80.2
84.8
60.4

73.3
77.4
67.2
67.4
76.1
80.1
74.5
74.9
76
53.9
64.4
59.3

31.91
30.9
27.4
28.33
31.98
26.01
30.75
30.75
29.72
21.82
19.93
22.96
28
21
22
21.5

0.094
0.104
0.096
0.046
0.071
0.114
0.098
0.09
0.101
0.1
0.039
0.046
1
0.488
0.512
1

33.75
34.5
34.25
33.75
29
34.25
29
33.5
39.5
39.5
26.25
29
34

18*
20*
19

*

(Shevaldin et al., 1998; Krivoshein, 2002).

2. Initial conditions of the hybrid system and tru composition in
the blanket

Table 2
Initial TRU compositions (%) in the molten salt blanket for different SNF cases. The main
TRU elements are marked in bold.

For the investigation of the possibilities to incinerate the actinides
separated from SNF of different power plants in the fusion–fission hybrid system we performed an evaluation of an inventory of the spent
nuclear fuel generated in Sweden and Lithuania nuclear reactors (PWR,
BWR and RBMK) up to 2015. Collected data on parameters of existing
PWR, BWR and RBMK-type nuclear reactors in Lithuania and Sweden
are presented in Table 1. Averaged data of the discharged fuel burn-up
depending on the fuel enrichment and the discharge year in BWR and
PWR reactors are taken from (U.S DOE/EIS, 2008); power and the
loading factor of the reactor are taken from (PRIS website (2016)). We
have used the IAEA nuclear fuel cycle simulation system (NFCSS
(2012)) and a generalized reactor model assumption for the country:
those different types of reactors of the same power and the same initial
fuel flow will produce approximately the same amount of SNF. That
means that the assumption of generalized reactor model for Sweden
takes into account the averaged power, the average operation time, the

averaged fuel enrichment and the averaged fuel burn-up of all existing
BWR and PWR nuclear reactors in the country as listed in Table 1.
According to (Plan 2013, 2014) there was 7520 tons of SNF in Sweden
in 2013. According to this model the total amount of heavy metals of
SNF is approximately 7630 tons including 77.5 tons of TRU at the end of
2015. During operation of the Ignalina NPP (Lithuania) about 22,000
assemblies of SNF (UO2 fuel with 2.0, 2.4, 2.6 and 2.8% initial 235U
enrichment (Barkauskaset al., 2016; Shevaldin et al., 1998; Krivoshein,
2002) from the two RBMK-1500 reactors were accumulated. The approximate amount of TRU is 24 tons.
The TRU composition in the LWR SNF significantly depends on the
initial fuel enrichment and the fuel burn-up in the reactor; these
parameters have changed a lot from 1974 to 2015. There exists a wide
variety of fuel assembly designs used in BWR (8 × 8, 9 × 9, SVEA-64,
SVEA-100) and in PWR (15 × 15, 17 × 17) with different fuel enrichments (2.1%–4% of 235U), which also influences the TRU composition
(Favalli et al., 2016). We have used the average spent nuclear fuel
composition of the LWR reactor (3% enrichment 235U) after the 5 years
cooling time. The isotopic composition of the average LWR reactor SNF
composition was calculated by using ORIGEN-ARP (SCALE module)
with specially prepared CE 14 × 14 (Westinghouse) ORIGEN-S multiburn-up libraries, which have been validated (Bowman et al., 2011,
Bowman and Leal, 2000). The considered TRU isotopic compositions in
the molten salt are presented in Table 2.
The RBMK-1500 SNF composition strongly depends on the initial
RBMK reactor fuel enrichment and the burn-up rate (Kimtys et al.,

Isotope

LWR

RBMK 1

RBMK 2

RBMK 3

237

3.78
1.81
51.64
24.61
8.22
4.61
90.89
4.21
0.83
5.04
0.19

2.07
0.19
61.18
27.56
0.64
1.92
91.49
6.32
0.12
6.44
0.005

3.32
0.51
49.62
31.09
0.94
4.70
86.86
9.30
0.51
9.81
0.01

3.82
0.67
46.08
31.86
0.98
6.06
85.64
9.74
0.78
10.52
0.02

Np,%
Pu,%
239
Pu,%
240
Pu,%

241
Pu,%
242
Pu,%
Total Pu,%
241
Am,%
243
Am,%
Total Am,%
244
Cm,%
238

2001, Plukienė et al., 2009). Three representative RBMK-1500 SNF of
the Ignalina NPP RBMK TRU vectors after the 50 years cooling time
have been evaluated: “RBMK 1” – 2% 235U enrichment and 14 MWd/kg
burn-up; “RBMK 2” – 2.4% 235U enrichment, 22 MWd/kg; “RBMK 3” –
2.6% 235U enrichment, 26 MWd/kg. These RBMK TRU cases have been
chosen taking into account the SNF composition of the Ignalina NPP
(Lithuania). The different RBMK SNF cases have been modelled using
ORIGEN-ARP with prepared one-group burn-up dependent cross-section libraries, which are comprehensively explained in publication
(Barkauskas et al., 2017). Depletion calculations for generation of the
cross-section libraries were performed using the SCALE 6.1 code
package with the TRITON control module, which employs a NEWT
deterministic 2D transport code with the 238-group energy library
based on ENDF-B VII library and the ORIGEN-S nuclide composition
calculation code. The calculated composition of actinides has been
validated by comparing the evaluation against experimental data. The
similar isotopic composition has been obtained in benchmark calculations of the RBMK spent nuclear fuel isotopic composition using MCNP

and ORIGEN codes where modeling results of 2% enrichment 235U fuel
were compared with the same experimentally measured data of the
RBMK-1000 fuel isotopic composition (Burlakov et al., 2003). One
should note that if the RBMK-1500 SNF cooling time is 50 years the
larger part of 241Pu decays to 241Am. And, if the cooling time is shorter,
the 241Pu amount in RBMK-1500 SNF would be closer to the LWR 241Pu
amount. Historically, uranium fuel with the 2% enrichment was used
from the very beginning of the Ignalina NPP operation. It is the same
type of fuel loaded in RBMK-1000 reactors in Chernobyl (Ukraine). We
have chosen two SNF burn-up cases with this enrichment: 1) 14 MWd/
kg – assemblies of low burn-up and 2) 20 MWd/kg – assemblies of high
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continuous TRU and 6Li feeding as well as continuous FP removal have
been used in all simulated cases during the irradiation. A mechanism to
remove the fission products is needed in the molten salt transmutation
system not only due to the better neutron balance of the system – if not
removed solidification of fission products will occur in the molten salt
quite soon after the operation. The hybrid system transmutation blanket
parameters, kscr, Φn(E), equilibrium conditions and the tritium breeding
ratio, have been calculated for each analyzed SNF TRU case. Further,
the comparative analysis of the LWR and RBMK SNF TRU incineration
rate has been performed.
3. Modeling results of SNF TRU transmutation and main
performance parameters of the hybrid system

Here we present the results of the modeling of LWR and RBMK SNF
TRU transmutation process by the previously described hybrid fusion–fission system. From Table 2 one can notice that 240Pu part is
considerably larger in all RBMK TRU composition cases compared with
the LWR SNF case (especially for RBMK 2 and RBMK 3 cases), besides
the concentration of 241Am is higher and that of 244Cm is lower. This is
due to the lower initial RBMK fuel enrichment and higher Pu transmutation in the RBMK type reactor. The strong neutron absorption by
240
Pu and a relatively smaller amount of the fissile element concentration determine the smaller kscr values in RBMK TRU as compared
with the LWR TRU case (Fig. 2(a)). Therefore, a larger fusion power is
needed at the beginning of irradiation as it is presented in Fig. 2(b) to
sustain the 3 GW thermal power of the system. The fusion power is
proportional to kscr: Pfus ∼ (1-kscr)/kscr.
By comparing LWR and RBMK TRU transmutation cases in terms of
the main performance parameters of the hybrid system, a number of
important advantages have been obtained in favor of the LWR isotopic
composition: the values of kscr (0.83) and Pfus (180 MW) are almost
stable comparing the beginning and equilibrium stages, the equilibrium
is reached after ∼3 years of the system operation, and the tritium
breeding rate (TBR) is sufficient to supply tritium for the fusion device
(TBR = 1.25). In the case of RBMK 1, the equilibrium is reached after
∼3.3 years, kscr (0.77) is lesser at the beginning but later it approaches
that of the LWR TRU case. In equilibrium kscr = 0.814, Pfus = 190 MW,
TBR = 1.25. In case of RBMK 2 and RBMK 3 transmutation in the
molten salt kscr at the beginning of irradiation is 0.69 and 0.67, respectively, and the 370 MW and 440 MW fusion power is needed for
corresponding cases, and the TBR is not sufficient for such fusion power
(TBR < 1, 0.73 and 0.67, respectively). When the equilibrium for main
isotope concentrations is reached, in both cases the situation is

Fig. 1. A simplified geometry model of the fusion–fission hybrid. FD stands for a Fusion
Device.

burn-up, both representatives of a typical burn-up fuel before 1996.
From 1996 up to 2004 the 2.4% 235U enrichment fuel with 0.41%
burnable poison (erbium) was the most frequently used nuclear fuel in
RBMK-1500; here the average fuel burn-up was 22 MWd/kg. The last
RBMK 3 case was selected to test the SNF originated from later operational campaigns (from 2002) with the 2.6% 235U enrichment and
0.5% of burnable poison – erbium, which was first tested in the Ignalina
NPP in ∼2001. This initial fuel load characteristic was used until the
final shutdown of the Ignalina NPP (Unit I in 2004 and Unit II in 2009).
The simplified spherical geometry setup described in Ridikas et al.,
2006 was used in our calculations (see Fig. 1 for details). The diameter
of the cavity with the fusion device is 400 cm, surrounded by the 1 cm
thick liquid flibe (6Li 0.1% in Li) wall, the 0.3 cm – metallic wall
(corrosion resistant SS316 50%), and the 1 cm thick graphite. A 60 cm
thick transmutation blanket (divided into 4 regions for calculation)
with TRU inside a flibe is placed between the graphite and the metallic
wall. All the structure was surrounded by a 20 cm thick graphite reflector and a 5 cm thick stainless-steel shell (Cheng, 2001; Ridikas et al.,
2006). Initial conditions of the hybrid system were the same for all TRU
isotopic vectors considered: Ptherm 3 GW, initial TRU mass – 3.04 t. The

Fig. 2. (a) kscr as a function of irradiation time in the molten salt blanket for different TRU cases, (kscr calculated with <1 σ ≥ 0.004). (b) Corresponding Pfus is needed to compensate the
neutron balance and sustain the 3 GW thermal power for different TRU cases.

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Fig. 3. (a) The macroscopic fission cross-sections for 239Pu, 241Pu, all MA and all TRU in the molten salt blanket for different TRU composition cases; (b) the macroscopic neutron capture
cross-sections for TRU, 240Pu, 239Pu and all MA (TRU cross-sections are on Y-axis on the right).

improved: kscr – 0.795 and 0.797, Pfus – 220 MW and 230 MW, TBR is
1.11 and 1.04 for corresponding cases. In the latter cases a large part of
TRU are composed of 240Pu i.e., the 241Pu isotope is produced more
intensively, and therefore the kscr rise gradient is larger than in the LWR
case.
The neutron spectrum in the molten salt blanket is independent of
the TRU isotopic composition and is stable during the irradiation time
as it was shown in the paper (Ridikas et al., 2006). Most of the neutrons
are epithermal (about 80%) and when the equilibrium is reached the
values of actinide macroscopic cross-sections also equilibrate. The value
of kscr of the hybrid system with the molten salt blanket depends on the
neutron capture and fission macroscopic cross-sections of corresponding TRU. The fission and neutron capture macroscopic crosssections of Pu isotopes, all TRU isotopes and MA (Minor Actinides) are
presented in Fig. 3(a) and (b), respectively. One can observe a sharp
increase of macroscopic fission cross-sections of TRU in all RBMK cases,
and it results in a sharp rise of kscr at the beginning of the system
performance. The fertile nuclide part is the largest in RBMK 2 and
RBMK 3 TRU cases (see Fig. 4), and this determines the larger macroscopic neutron capture cross-sections for TRU in these cases.
The mass evolution of fissile and fertile isotopes in the system for
different TRU transmutation cases is presented in Fig. 4. The

equilibrium of fissile TRU in LWR, RBMK 2 and RBMK 3 cases is reached
faster as compared with the RBMK 1 case. This is determined by a
higher concentration of 239Pu at the beginning and also by the feeding
concentration of the same TRU isotopic composition. Despite the higher
fissile isotope concentration in RBMK1 compared with the LWR case,
the kscr begins at a lower value and approaches LWR kscr value at
equilibrium. This can be explained by the macroscopic neutron capture

and fission cross-sections in Fig. 3: it is clearly seen that a higher 241Pu
macroscopic fission cross-section determines a higher kscr at the beginning in the LWR case, while the macroscopic neutron capture crosssection of 240Pu equilibrates the neutron balance for the RBMK 1 case,
and, as a result the neutron absorption is similar in both LWR and
RBMK 1 cases.
The kscr behavior depends on the fission product accumulation in
the molten salt blanket as it was discussed in (Ridikas et al., 2006), but
it is also influenced by accumulating volatile materials such as 3H and
3
He. 3H (T-tritium) is produced from 6L, which is continuously replenished with TRU fuel, the other isotope – 3He – is produced by β−
reaction from 3H. 3He is a strong neutron absorber (neutron capture
∼1000 b). The presence of 3H and 3He deteriorates the neutron balance
in the system as it is seen in Fig. 5. Here kscr decreases by 6% after
3000 days due to 3H and 3He in the molten salt. In a real situation both

Fig. 4. The fissile and fertile isotope mass evolution for different TRU transmutation
cases.

Fig. 5. kscr behavior in the molten salt blanket of a hybrid system in the RBMK 1 TRU
transmutation case performing 3H and 3He separation and without 3H and 3He separation.

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R. Plukienė et al.

Fig. 6. The variation of Pu (a) and MA (b) isotope mass against time in the molten salt blanket/RBMK 2 case.

elements should be separated from the molten salt blanket and this

should be taken into account when performing the modeling calculation.
For the establishment of LWR and RBMK TRU equilibrium conditions in the molten salt blanket the analogous procedure as described in
(Ridikas et al., 2006) has been performed. It has been determined that
the system reaches equilibrium when the variation of the Pu isotope
mass is less than 4% (this variation have been chosen to include the
statistical uncertainties of calculations) in respect of the actual plutonium mass in the blanket. The physical effect of saturation for each
nuclide for a certain nuclide composition is achieved in the molten salt
in equilibrium. The variation of the Pu and MA isotope mass against
time in the molten salt blanket in the RBMK 2 case is presented in Fig. 6
a and b). The time needed to reach the equilibrium state of the system is
determined by the fissile and fertile actinide concentrations and their
macroscopic cross-sections of interaction with neutrons for different
TRU compositions. The equilibrium for Pu transmutation is reached
after 1000 days for LWR, RBMK 2 and RBMK 3 TRU cases and after
1200 days for the RBMK 1 case. Meanwhile the equilibrium for MA is
reached in ∼2500 days for all considered cases. The latest equilibrium
stage is reached for 246Cm, 245Cm and 244Cm isotopes.
The actinide composition is different for considered SNF cases at the
equilibrium stage, however, as compared with the initial composition,
the fissile and fertile isotope concentration ratio becomes comparable
for all cases (at the beginning this ratio is 0.9 ÷ 1.62, in equilibrium –
0.6 ÷ 0.77).
In all the analyzed TRU cases the hybrid system incinerates about
1.1 tons per year of TRU with the 3 GW thermal power output. The
initial TRU vector influences the composition of actinides at the equilibrium stage. The actinide isotopic composition at the equilibrium
stage is presented in Table 3. 237Np, 239Pu, and 241Am are incinerated
most effectively: – the transmuted part is 30–51% for 237Np, 28–50%
for 239Pu and ∼26–54% for 243Am for RBMK2-LWR cases, while the
lowest transmutation rate was obtained for RBMK3 case: 17% for
237

Np, 14% for 239Pu and 11% for 243Am. The smallest amount of Cm is
accumulated in the RBMK1 case (smaller amount of 244Cm is replenished). The minor actinides are accumulating in the molten salt
(minimum ∼200 kg in the RBMK 1 case, maximum ∼300 kg in the
RBMK 3 case).
The critical parameter for the hybrid system is the neutron fluence

Table 3
Actinide concentrations (%) at the equilibrium stages for different TRU cases. The main
TRU elements are marked in bold.
LWR

RBMK 1

RBMK 2

RBMK 3

Np,%
Pu,%
Pu,%
240
Pu,%
241
Pu,%
242
Pu,%

1.6
6.5
22.6

30.9
15.2
9.8

0.8
4.5
25.1
34.9
14.9
7.7

1.3
6.6
20.1
32.9
14.3
9.4

1.4
7.1
18.5
31.8
13.9
10.4

Total Pu,%

85.0

87.1

83.3

81.7

241

Am,%
Am,%
Am,%

1.5
0.3
1.7

1.9
0.4
1.4

2.6
0.5
1.6

2.6
0.5
1.8

Total Am,%

3.5

3.7

4.7

4.9

242

Cm,%
Cm,%
Cm,%
245
Cm,%
246
Cm,%

0.7
0.2
6.0
2.6
0.4

0.9
0.3
4.7
2.1
0.4

1.2

0.5
5.9
2.6
0.5

1.2
0.5
6.7
3.0
0.6

Total Cm,%

9.9

8.4

10.7

12.0

237
238
239

242m
243

243
244

Table 4
Neutron fluence (MW·y/m2) to the first wall of the fusion–fission hybrid system for different TRU cases.
Stage

LWR

RBMK 1

RBMK 2

RBMK 3

kscr peak
Equilibrium
3000 days

1.3
6.5
21.1

2.0
7.7
21.0

2.9
6.7
21.3

4.2

6.7
21.3

to the first molten salt wall (see Table 4). Evaluation shows that, independently of the initial TRU composition, the first molten salt wall is
crossed by ∼2.1 × 1027 neutrons during 8 years (∼21 MW·y/m2). The
application of a different molten salt composition (Be salt) as in
(Plukiene et al., 2006) could improve these fusion–fission system performance parameters.

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R. Plukienė et al.

Fig. 7. Inhalation radiotoxicity of Pu and Np from RBMK 2 (a) and LWR (b) SNF at the beginning and at the end of transmutation in fusion–fission hybrid system.

4. Radiotoxicity analysis of the final TRU waste

decay. 241Pu accumulation from 240Pu is larger in the RBMK 2 case as
compared with the LWR case, but it decreases to natural U radiotoxicity
after 200 years.
At the end of transmutation the isotopic composition in both cases
becomes more or less similar and the same happens with the radiotoxicity. In both considered cases the radiotoxicity of 239Pu, 240Pu,
237
Np and 241Am is considerably decreased. In the molten salt transmutation blanket a larger amount of Cm isotopes is accumulated, but
the contribution of this isotope to the total waste radiotoxicity is minor.
The total transmutation time in the system is 8.2 years. During this
operation time about 9 tons of TRU can be incinerated and the radiotoxicity of the remaining final waste decreases approximately by 4–6
times in the period from 1 thousand to 100 thousand years.

The efficiency of transmutation has been investigated in detail for
LWR and RBMK 2 cases by evaluating the inhalation radiotoxicity of
TRU before and after transmutation in the hybrid system (after 10 years
cooling time).
The inhalation pathway of the actinide intake is much more dangerous than the ingestion and it was used as a more conservative reference (Plukiene et al., 2004, 2014). The inhalation radiotoxicity (InhTox) was calculated as follows: InhTox = A × 3,7 × 1010 × EInh , where
A is the nuclide activity (Bq), EInh are committed effective dose coefficients (SvBq−1) for each nuclide taken from International Commission
on Radiological Protection (ICRP) publication 119 (ICRP, 2012).
The radiotoxicity of RBMK 2 at the beginning was considerably
smaller compared with that of LWR. This was determined by the
smaller quantity of 241Pu (T1/2 = 14.2 y) and 238Pu (T1/2 = 87.7 y) due
to the longer RBMK SNF cooling time (50 years) in the RBMK 2 case.
The inhalation radiotoxicity of Pu and MA at the beginning and at
the end of transmutation is demonstrated in Fig. 7(a and b) and in
Fig. 8(a and b) for RBMK 2 and LWR, respectively. The radiotoxicity
increases by a factor of 2 in the first 10 years after transmutation (due to
238
Pu) in the RBMK 2 case, but it decreases afterwards due to 238Pu

5. Conclusions
The closed fuel cycle scenario of LWR and RBMK SNF has been
analyzed in the framework of the EURATOM project BRILLIANT. The
transmutation of TRU elements has been simulated by using the model
of the fusion–fission hybrid system for different spent nuclear fuels
accumulated in the Baltic countries: RBMK-1500 (Lithuania) and LWR
(Sweden). According to the IAEA nuclear fuel cycle simulation system

Fig. 8. Inhalation radiotoxicity of MA from RBMK 2 (a) and LWR (b) SNF at the beginning and at the end of transmutation in the fusion–fission system.

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R. Plukienė et al.

Lithuanian state budget and partially accomplished within the frame of
the European Commission project “Brilliant” (EURATOM program, No.
662167).

(NFCSS) and the generalized reactor model assumption for the country
about 71% of SNF is produced in BWR and 29% in PWR type reactors in
Sweden as well as the total amount of heavy metals of SNF is approximately 7630 tons, including 77.5 tons of TRU at the end of 2015.
During operation of the Ignalina NPP about 22,000 assemblies of SNF
(UO2 fuel with 2.0, 2.4, 2.6 and 2.8% initial 235U enrichment) from the
two RBMK-1500 reactors were accumulated, and the approximate
amount of TRU is 24 tons. The total amount of TRU in both countries is
about 101 tons by 2015.
In order to investigate the transmutation efficiency of the fusion–fission hybrid system for different spent nuclear fuel TRU, the
actinides extracted from the light water reactor (LWR) (averaged
composition of SNF TRU) and RBMK spent fuel (different cases) were
analyzed. The TRU transmutation scenarios in the fusion–fission hybrid
system with different spent nuclear compositions from LWR (PWR) and
from RBMK (RBMK 1 14MWd/kg burn-up of 2% UO2 fuel, RBMK 2
20MWd/kg burn-up of 2,4% UO2 fuel RBMK 3 22MWd/kg burn-up of
2.6% UO2 fuel) have been investigated. We have shown that a better
hybrid-system performance in terms of the ksrc behavior, a lower requested fusion power, a shorter equilibration period and a smaller actinide mass in the blanket was obtained in TRU of LWR SNF compared
with TRU from RBMK SNF:

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