pr 03 - fusion–fission hybrid system for nuclear waste transmutation

7/23/2019 Pr 03 - FusionFission Hybrid System for Nuclear Waste Transmutation

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Fusionfission hybrid system for nuclear waste transmutation (I):

Characterization of the system and burn-up calculations

D. Ridikas a,*, R. Plukiene b, A. Plukis b, E.T. Cheng c

aDSM/DAPNIA/SPhN, CEA Saclay, F-91191 Gif-sur-Yvette, France

b Institute of Physics, Savanoriu pr. 231, LT-2053 Vilnius, Lithuaniac TSI Research, 312 S. Cedros Avenu, Solana Beach, CA 92075, USA

Abstract

Generally, a very intensive neutron source is needed for the efficient transmutation process of nuclear waste and minor actinides

in particular. A plasma based fusion device is naturally a candidate neutron source. In this paper, we analyze a molten salt (flibe)

blanket, driven by a plasma based fusion device, dedicated to burn actinides extracted from the LWR spent fuel. Performance

parameters such as kscr, burn-up, fluxes and equilibrium conditions were calculated to assess the characteristics of a flibe based

transmutation plant. It seems that nearly 1.1metric ton of spent fuel TRU could be burned annually with an output of 3 GWthfission

power. Detailed burn-up calculations show that equilibrium conditions of the fuel concentration in the flibe transmutation blanket

could be achieved, provided that fission products were removed at least in part and fresh spent fuel TRU was added during burn-up

on-line.

q 2005 Elsevier Ltd. All rights reserved.

Keywords: Hybrid system; Nuclear waste transmutation

1. Introduction

The main goal of our present work is to investigate the performance of the dedicated molten salt (flibe) blanket for

nuclear waste transmutation as a function of different treatment of fission products as well as different refueling of

already destroyed actinides. This work is a continuation of the related studies (Cheng and Wong Clement, 2000;

Cheng, 2001;Ridikas et al., 2001a) on a flibe based actinide transmutation hybrid system driven by a fusion device.

Below, we present only a brief summary of the concept sufficient for understanding this follow-up study. A detailed

description of the fusionfission hybrid system including some burn-up calculations can be found in Refs. (Cheng and

Wong Clement, 2000;Cheng, 2001).

In order to transmute efficiently the nuclear waste, a very high intensity neutron source is needed. It seems that an

inertial confinement fusion (ICF) device could provide with a neutron source as intense as a high power accelerator,

i.e. of the order of 10181019 n/s (Cheng and Wong Clement, 2000). A molten salt (LiFBeF2(HN)F4) blanket,

surrounding this neutron source, then serves as a medium for transuranic actinides (TRU) to be burned. Flibe also has

a function of both coolant and carrier of tritium breeding material (6Li in this case). A well-known advantage of

Progress in Nuclear Energy 48 (2006) 235246

www.elsevier.com/locate/pnucene

0149-1970/$ - see front matter q 2005 Elsevier Ltd. All rights reserved.

doi:10.1016/j.pnucene.2005.09.004

*Corresponding author. Tel.:C33 1 690 878 47; fax: C33 1 690 875 84.

E-mail address:[email protected] (D. Ridikas).
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the molten salt is a possibility of both refueling of burned TRU and extraction of fission products (FP) on-line.

Namely, these two parameters we are going to investigate in more detail in this work.

The main characteristics of the fusionfission hybrid system are presented in Table 1. In order to produce

3000 MWthfission power, a 250 MW power fusion device operating with a deuteriumtritium fuel cycle is sufficient

to provide with an external 14 MeV neutron source ofw9!1019 n/s. The initial load of the fuel composition and

subsequent feeding material is a typical TRU vector defined by the LWR spent fuel to be given below. A simplified

fusionfission hybrid system was modelled with the main geometry and material parameters given inTable 2.Fig. 1

presents a geometry setup used in our calculations.As presented inTable 2,flibe (a) contains 0.1% of 6Li compared to the total amount of Li. Flibe (b) density is 2 g/

cm3 and corresponding TRU density is 0.074 g/cm3 of flibe or 0.5 M% in flibe. 0.6% of 6Li enrichment for flibe (b)

was taken as a reference case (Cheng and Wong Clement, 2000). In both cases, 6Li serves as a breeding material for

tritium production. The initial loading and subsequent feeding TRU compositions were those from LWR spent fuel

when fission products and uranium are separated (to be provided in detail later in this paper in Section 3).

2. Modelling tools and procedure

All calculations on flibe based actinide transmutation blanket were made employing the Monteburns code system

(Trellue and Poston, 1999)an automatic-cyclic coupling of the MCNP (Briesmeister, 1993) and ORIGEN2

(RSICC, 1991) codes. Monteburns was already successfully benchmarked by simulating the uranium based fuel cycleof the high flux reactor at ILL Grenoble (Ridikas et al., 2000a) and gas turbine-modular helium cooled reactor

(GT-MHR) in the case of plutonium (uranium free) fuel cycle (Ridikas et al., 2000b). MCNP was also used to obtain

keffin the mode of a criticality eigenvalue problem (kcode card in MCNP input), and also to estimate neutron flux as

well as kscrin its external source mode (sdef card in the MCNP input). In this case,kscris defined as:

kscr Z MnK1=MnK1=n (1)

wherenis an average number of neutrons per fission andMnis a total neutron multiplication factor of the system.

The effect of 14 MeV neutrons is included in the kscr calculationsthe multiplication factor will be different for

Table 1

Main parameters of the fusionfission hybrid system used in this study

Fusion power 250 MW

Fission power 3000 MWthkeffat equilibrium 0.743

MEat equilibrium 19

TBR at equilibrium 1.4Initial TRU composition LWR spent fuel

Initial TRU inventory (kg) 3040

MEstands for the energy multiplication of the system, while TBR for the tritium breeding ratio (Cheng, 2001).

Table 2

Neutronics model: zone dimension and material compositions for a molten salt transmutation blanket as from (Cheng and Wong, 2000)

Zone name RadiusRiKRiC1 (cm) Thickness (cm) Material composition

Cavity 0200 200 Void

Liquid wall 200201 1 Flibe (a)

Metallic wall 201201.3 0.3 SS316, 50%

Graphite 201.3202.3 1 Graphite

Blanket

Region 1 202.3217.3 15 Flibe (b)

Region 2 217.3232.3 15 Flibe (b)

Region 3 232.3247.3 15 Flibe (b)

Region 4 247.3262.3 15 Flibe (b)

Reflector 262.3282.3 20 Graphite

Envelope 282.3287.3 5 SS316, 50%

Also seeFig. 1.

D. Ridikas et al. / Progress in Nuclear Energy 48 (2006) 235246236


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14 MeV and for fission source neutrons. The average value ofn in a problem is calculated by dividing the fission

neutrons gained per fission by the fission neutrons lost to fission.

All burn-up calculations were done employing this external source mode. The advantage of the Monteburns code is

that it can use different nuclear data libraries as long as they are compatible with MCNP format (see for example Ref.

Ridikas et al., 2001b). For all structure materials and actinides, we have chosen ENDF data files (ENDF/B-VI, 1995)

being most frequently employed, while for the fission products JENDL data library (JENDL-3.2,, 1994) was taken

due to the biggest number of fission products available (w200).

The main goal of the modelling was to find the equilibrium conditions of the fusionfission hybrid system and to

investigate how these conditions could be influenced due to the different modes of refueling (continuous or discrete

feeding) and also due to the different treatment of fission products (partial/full removal and/or continuous/discrete

removal). For this purpose, we had to:

stabilize an effective neutron multiplication coefficient and the total TRU mass in the system,

stabilize the neutron spectrum Fn(E) as a function of burn-up,

define the actual amount of fresh TRU material (DMTRU) to be fueled on-line.

In all cases, the fission power is kept constant (3000 MWth), what corresponds to a variable fusion power of the

fusion device (Section 3.2). In our simulations, this implies a renormalization of the absolute neutron flux if an

effective neutron multiplication coefficient of the system is changed.

3. Results

3.1. Neutron flux and energy spectra

Table 3presents the main neutron flux characteristics in different regions of the molten salt transmutation blanket

(Fig. 1). Corresponding neutron spectra are presented in Fig. 2. In the 239-energy group, neutron spectrum the

resonance absorption by transuranium (239Pu, 240Pu and 242Pu) and molten salt components (19F, 6Li) is clearly

visible.

We note that the averaged neutron flux is very high and corresponds to the flux typical only for high flux reactors.

In addition, the amount of thermal neutrons is very small (!1%). Most of the neutrons are epithermal or fast, what, in

principle, could favor the transmutation of actinides in the resonance region. Indeed, the fast neutron flux (EnO

0.1 MeV) is about 1022 n/cm2 per full power year.

Table 4gives a variation of averaged neutron spectrum as a function of burn-up. The neutron spectrum in the

blanket has two components. One is that from the slowing down of 14 MeV source neutrons. The other one is

Fig. 1. A simplified geometry model of the fusionfission hybrid used in the calculations. Also seeTable 2. FD stands for a fusion device.

D. Ridikas et al. / Progress in Nuclear Energy 48 (2006) 235246 237


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the spectrum established by the fission neutrons. We note that the final spectrum due to 14 MeV source neutrons in the

blanket is almost constant. Equally, the flux level is constant too if the fusion power is not changed. The spectrum due

to fission neutrons is also very stable, except the flux level, which depends on fission reaction rates. This is because the

majority of the fission neutrons are characterized by the same fission neutron spectrum and only then moderated in the

blanket. Indeed, most of the fission reactions still occur in 239Pu and 241Pu, even though at a long burn-up stage

significant amount of Am and Cm accumulates. FromTable 4, it is clearly seen that the neutron spectrum remains

rather stable as a function of burn-up. In addition, this is also true if either continuous or discrete feeding of the fresh

fuel is simulated. This might justify a less time consuming approach for the fuel burn-up calculations, i.e. as a good

approximation one could calculate the neutron flux (and consequently flux weighted neutron cross-sections) only

once and perform burn-up calculations with a constant neutron spectrum.

Finally,Table 5gives one group flux weighted neutron capture (sc) and fission (sf) cross-sections including their

ratiosa (aZsc=sf) for transuranium actinides and comparison with other transmutation systems being either critical

or accelerator driven as reported bySalvatores (2002). It is clear that epithermal flux is more favored for the most of

Table 3

Neutron flux characteristics in different regions of molten salt transmutation blanket at the beginning of the irradiation

Zone name Fn(total), n/s cm2 Fn(EnO0.1 MeV), %

Region 1 2.47!1015 26.2

Region 2 1.84!1015 22.4

Region 3 1.22!1015 20.4

Region 4 0.75!1015 20.6

Regions 14 1.48!1015 22.9

Also seeTable 1.

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

101

1010

1011

1012

1013

1014

region 1 region 2

region 3 region 4 all regions

6Li

19F

240Pu

242Pu

E, MeV

239Pu

Neutronflux,n/(c

m2*s*lethargy)

Fig. 2. Neutron energy spectra in different flibe regions. A solid line indicates the average neutron flux over all regions with fuel.

Table 4

Averaged neutron spectrum over all flibe regions, presented in different energy groups (%) as a function of burn-up (time)

Time,

days

Fn,% (Discrete) Fn,% (Continuous)

[0.,1.]

(eV)

[1.,100.]

(eV)

[.1,100.]

(keV)

[.1,1.]

(MeV)

[1.,20]

(MeV)

[0.,1.]

(eV)

[1.,100.]

(eV)

[.1,100.]

(keV)

[.1,1.]

(MeV)

[1.,20]

(MeV)

0 0.16 14.58 62.49 12.28 10.45 0.04 14.58 62.49 12.28 10.45

400 0.16 15.37 62.07 12.02 10.33 0.03 14.84 62.55 12.24 10.19

800 0.16 15.52 61.87 12.07 10.35 0.04 15.09 62.33 12.18 10.22

1200 0.14 15.43 61.87 12.11 10.42 0.04 15.12 62.12 12.2 10.38

1600 0.15 15.30 61.98 12.03 10.51 0.04 15.25 62.03 12.08 10.47

2000 0.16 15.33 61.72 12.14 10.62 0.03 14.93 62.22 12.22 10.47



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actinides compared with thermal flux (only 239Pu and 237Np is transmuted more efficiently in this case). However,Table 5also shows that a values are the smallest in fast neutron spectradirect incineration by fission being maximal

in this case.

3.2. Equilibrium conditions

A typical evolution of the TRU isotopic composition is presented inFig. 3.This is the case when the system is

refueled continuously. In addition, the fission products are also continuously removed on-line at 100% rate.

In order to define the equilibrium, we employ the following procedure. Since not all actinides are stabilized even

after 2000 days of operation (Cm isotopes in particular), we assume that the system reaches equilibrium (or system is

approaching equilibrium) when the variation of Pu isotope mass is less than 4% with respect to the actual plutonium

mass in the blanket. This is represented in detail in Fig. 4, where the time dependence of the ratio DMi/Mi of Pu

isotope masses is illustrated. According to this analysis, we conclude that the equilibrium is reached after about

1000G50 days of irradiation.

A corresponding behavior ofkscrand required fusion power Pfusionat a constant fission power of 3000 MWthare

presented inFig. 5. The excursion ofkscroccurs mainly due to the mass conversion of 240Pu into 241Pu during early

stage of irradiation (Figs. 3 and 4 for detailed mass dependence of 241Pu irradiation). The values of kscr were

calculated in the case when all fission products are continuously removed from the blanket. The variation of fusion

power is between 220 and 242 MW. We remind that 1 MW fusion power corresponds to the external neutron source

intensity ofw4!1017 n/s of 14 MeV neutrons.

Table 5

Average fission, capture cross-sections, and their ratios for a number of selected actinides

Isotope sc ; b sf; b a(fthis work) a(f thermal)

(Salvatores, 2002)

a(ffast)

(Salvatores, 2002)

237Np 18.20 0.26 70.0 63 5.3238Pu 6.79 1.90 3.6 12 0.53239Pu 7.70 10.80 0.7 0.58 0.3240Pu 14.60 0.31 47.1 396.6 1.6241Pu 6.48 20.70 0.3 0.4 0.19242Pu 14.20 0.22 64.5 65.5 1.8241Am 20.60 0.36 57.2 100 7.4242mAm 7.20 44.00 0.2 0.23 0.18243Am 27.50 0.21 57.2 111 8.6243Cm 3.25 21.10 0.2 0.16 0.14244Cm 16.30 0.65 25.1 16 1.4245Cm 3.25 22.70 0.1 0.15 0.18

Average cross-section (barn) is defined as sZsE4EdE=

4EdE.

0 500 1000 1500 200010

2

101

100

101

102

103

237Np

238Pu

239Pu

240Pu

241Pu

242Pu

241Am

242Am

243Am

244Cm

245Cm246Cm

M,

kg

Time, d

Fig. 3. TRU mass dependence on irradiation time: continuous feeding with 3.1 kg/d of TRU and continuous removal of all fission products.



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Finally, we note that w3.1 kg of TRU is burned per day when the equilibrium situation is reached, what

corresponds to nearly 1.1 metric ton of TRU burned annually.Table 6gives detailed isotopic composition of the

initial TRU vector and also at equilibrium, i.e. after w1000 days of irradiation. Different options DD, DC, CD, and

CC correspond to different refueling conditions and also to different treatment of the fission products (see the caption

ofTable 6). For comparison, we also present the results reported in Ref. (Cheng and Wong Clement (2000)). We note

that in all cases equilibrium is reached and a reasonable agreement between Pu, Np and Cm isotopes is obtained when

compared to the calculations reported in Ref.Cheng and Wong Clement (2000).On the other hand, in our case we

obtained much less 242Am and 243Am. This difference is mainly due to the different data libraries used.

3.3. Effect of different feeding options

In general, it is preferred that an effective neutron multiplication factor is as stable as possible for any nuclear

reactor being either critical or sub-critical. Fig. 6, illustrates the effect of different feeding options on kscr. As

expected, the discrete feeding results in sharp variations ofkscr, consequently, possible sharp variations of the reactor

power. Of course, this effect could be diminished with additional control rods. On the other hand, the situation is

improved considerably if continuous feeding option is chosen (see the sameFig. 6).

We also note separately that in both cases as above a slightly different isotopic composition at equilibrium stage is

obtained (Table 6). As it can be seen fromFig. 7, the difference appears mainly in the first step of burn-up calculation,

0 500 1000 1500 2000

50

40

30

20

10

0

10

20

238

Pu

239

Pu

240

Pu

241

Pu

242

Pu

Mi

/Mi,

%

Time, d

Fig. 4. Variation in mass (%) of Pu isotopes as a function of the irradiation time in the case of continuous feeding of TRU material and total removal

of FP; i stands for a different isotope of Pu.

0 500 1000 1500 20000.74

0.76

0.78

0.80

0.82

0.84

200

220

240

260

280

300

320

340

360

380

400

kscrkscr

Time, d

Pfusion,

MW

Pfusion

~(1-kscr

)/kscr

Fig. 5.kscr and fusion power variations: continuous TRU feeding and continuous removal of FP.



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after which the behavior of the curves is rather similar. TRU mass at equilibrium with discrete feeding is 2770 kg and

with continuous feeding isw3100 kg. The reason for such a difference is that in the discrete feeding case a fresh TRU

material was added at the beginning of each step (100 days long), while in continuous feeding caseit was

replenished during each step. Therefore, in the first case less initial TRU material remains in the molten salt from the

very beginning. Smaller equilibrium mass of TRU material could be important for system safety requirements. On theother hand, discrete feeding results in sharp fluctuations ofkscr, as already discussed above, what would be difficult to

tolerate in the realistic system.

One could also ask if different feeding options could change the equilibrium conditions. Fig. 8, clearly shows

that in both cases (with discrete and continuous feeding), 239Pu mass equilibrates with the same speed. In other

words, less than 4% variation is achieved in both cases at the same time. The same is also true for other isotopes of

plutonium.

Table 6

TRU composition (%) in the molten salt at initial and equilibrium stages

Stage Initial Equilibrium

Cases DD DC CD CC (Cheng et al.,

2000)

Time, days 0 1000 1000 1000 1000 w980237Np (%) 3.78 1.50 1.48 1.63 1.62 1.51

Total Np (%) 3.78 1.51 1.50 1.64 1.64 1.51238Pu (%) 1.81 6.92 6.98 6.47 6.57 4.83239Pu (%) 51.64 21.23 21.05 22.95 22.75 20.90240Pu (%) 24.61 29.49 29.16 30.42 30.58 30.10241Pu (%) 8.22 16.09 16.18 15.35 15.31 15.70242Pu (%) 4.61 10.14 9.96 9.90 9.71 10.70

Total Pu (%) 90.89 83.88 83.35 85.10 84.92 82.20241Am (%) 4.21 1.34 1.32 1.49 1.48 1.49242m Am (%) 0.00 0.29 0.28 0.29 0.28 0.59243Am (%) 0.83 1.82 1.88 1.60 1.67 4.80

Total Am (%) 5.04 3.45 3.48 3.38 3.43 6.89244Cm (%) 0.19 6.74 6.87 5.83 6.05 6.31

245Cm (%) 0.00 2.91 2.99 2.45 2.52 2.68246Cm (%) 0.00 0.52 0.54 0.41 0.43 0.49

Total Cm (%) 0.19 10.17 10.40 8.70 9.01 9.49

Others (%) 1.01 1.31 1.21 1.03

DD, discrete feeding and discrete FP removal; DC, discrete feeding and continuous FP removal; CD, continuous feeding and discrete FP removal;

CC, continuous feeding and continuous FP removal; in all cases 100% of FP were removed.

0 500 1000 1500 20000.65

0.70

0.75

0.80

0.85

0.90

discrete feedingcontinuous feeding

100% discrete removal of FP

kscr

Time, d

Fig. 6.kscras a function of discrete and continuous TRU feeding and with discrete removal of all (100%) fission products at the end of burn-up steps.



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3.4. Effect of fission products

The importance of fission products in the molten salt blanket for fusionfission hybrid system was analyzed

partially in Ref. (Ridikas et al., 2000a). It was concluded that on-line removal and disposal of some neutron poisonous

fission products may be needed in order to reach deep burn-up level and, at the same time, to keep the effective

neutron multiplication coefficient of the system (equally blanket fission power) as close to the initial value as possible.

In the same study, it was found that fission products are responsible at least forK3150 pcm decrease in the effective

neutron multiplication coefficient at 15% burn-up. It was also pointed out that the tolerance of the fission products in

molten salt may allow a batch process, and the allowable burn-up tolerance level could be about 10% or slightly more.

InTable 7,we identify the most important fission products contributing more than 0.1% to the total macroscopic

neutron absorption of the entire blanket, being 6.3!10K3 (1/cm) for 0% burn-up at the beginning of the irradiation.

Typically, these fission products are long lived or stable nuclei with an averaged (flux weighted) neutron capture

cross-sections of the order of 580 barns. All these fission products, due to their constant accumulation and also

because of their important neutron capturing, will degrade the performance of the entire system as we already

mentioned above. The only exception is 135Xe, know as the most poisonous fission product, which is short lived

(T1/2Z9.14 h only) but is equally important in a long run due to its very high neutron capture cross-section (about

1200 b in this particular neutron flux). As it can be seen from the same Table 7, the concentration of 135Xe gets

saturated and it will not perturb the system more as it has done by now (so called Xe-effect in reactor physics). One

0 500 1000 1500 20000

200

400

600

800

1000

1200

1400

1600 237

Np

238

Pu

239

Pu

240Pu

241Pu

242

Pu

241

Am

243

Am

244

Cm

245

Cm

solid symbol line:

continuous TRU feeding

open simbol line:

discrete TRU feeding

M,

kg

Time, d

Fig. 7. TRU isotopic composition profile in molten salt as a function of irradiation in the cases of discrete and continuous feeding.

5000 1000 1500 2000

0

4

8

12

16

20

24

M

i/ M

i,%

Time, d

239Pu continuous feeding

239Pu discrete feeding

Fig. 8. Variation in mass (%) of239Pu as a function of the irradiation time in the case of continuous and discrete feeding and with 100% removal of

FP.



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has to note that the actual effect of non-soluble or gaseous fission products for the performance of the entire system

will be somewhat lower than presented in this work; simply because these fission products will have a chance to leave

the blanket before neutron interaction takes place.

Fig. 9presents the behavior ofkscrwhen all fission products are removed in two different modes, i.e. continuously

and discretely. It seems that the way fission products are removed (as long as they are removed 100%) has little

influence on the performance of the system. On the other hand, the percentage of removed FP relative to the total mass

of fission products present in the blanket becomes crucial as shown inFig. 10.

Four different scenarios were considered with the following FP removal options: 100, 50, 20 and 0%, respectively.

0% corresponds to the case when all fission products are left in the molten salt flibe, while in the case of 100%all

fission products are removed.

One can ask why there is such a big difference between removal of 50, 20 and 0% of fission products. This can be

explained by analyzing the creation/destruction and removal ratios of the fission products. We will take, for example,

a stable fission product 132Xe, which has a small neutron capture cross-section. InFig. 11, we present its mass as a

function of burn-up when either 100% or part of the fission products is left. When all fission products are left, the

amount of 132Xe increases linearly (triangles).

Table 7

Fission products contributing more than 0.1% to the total macroscopic neutron absorption in a self-cooled molten salt transmutation blanket

Fission products sc ; b 6% burn-up, atoms/b-cm 10% burn-up, atoms/b-cm 15% burn-up, atoms/b-cm

99Tc 10 4.18!10K7 9.89!10K7 1.38!10K6

101Ru 6 6.21!10K7 1.06!10K6 1.48!10K6

103Rh 9 2.54!10K7 6.17!10K7 1.02!10K6

105Pd 5 5.44!10K7 9.46!10K7 1.33!10K6

107Pd 7 3.80!10K7 6.52!10K7 9.12!10K7

108Pd 13 2.68!10K7 4.66!10K7 6.62!10K7

109Ag 34 1.61!10K7 2.69!10K7 3.98!10K7

131Xe 30 2.53!10K7 4.22!10K7 5.67!10K7

135Xe 1200 6.89!10K9 6.67!10K9 6.44!10K9

133Cs 12 5.77!10K7 1.04!10K6 1.46!10K6

145Nd 8 3.02!10K7 5.11!10K7 7.06!10K7

147Pm 58 1.17!10K7 1.90!10K7 2.32!10K7

149Sm 86 9.88!10K8 1.55!10K7 1.98!10K7

151Sm 67 6.41!10K8 9.92!10K8 1.27!10K7

152Sm 71 6.36!10K8 1.07!10K7 1.45!10K7

153Eu 43 4.88!10K8 9.73!10K8 1.50!10K7

Note that the macroscopic neutron absorption due to a particular fission product in (1/cm) could be easily obtained by multiplying scin (barns) with

a corresponding atomic density in (atoms/barn cm) as a function of burn-up.

0 500 1000 1500 20000.65

0.70

0.75

0.80

0.85

0.90

kscr

Time, d

Continuous removal of FP 100%

Discrete removal of FP 100%

Fig. 9. Comparison ofkscrin the case of continuous feeding but different removal of fission products.



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When 20% of FP are taken away, the 132Xe mass reaches the equilibrium after w1200 days and later its mass

remains stable; in the case of 50% removal the equilibrium is reached already afterw400 days. It should be pointedout separately that as long as fission products reach equilibrium,kscr also varies very little (compare the curves in

Fig. 10keeping in mind the corresponding curves in Fig. 11). Rather similar situation will be with other (stable or

long-lived and with similar capture cross-sections) fission productsthey will reach equilibrium at different times

and at different quantities. For the rest of fission products (short lived or with high capture cross-sections), the

equilibrium may be reached much faster and is some cases even without removal.

Finally, we add that a mechanism to remove the fission products is definitely needed in the molten salt

transmutation system. This has to be done not only due to the better neutron balance of the system as discussed above

but also due the following reason. The burn-up of the TRU materials is about 1.1 metric ton per year, equivalent to

about 30% of the TRU inventory in the blanket. Solidification of fission products will occur in the molten salt quite

soon after the operation, if some fission products are not removed. The next study of this area would be to look into

strategy and reality how these fission products can be removed.

3.5. Effect of the 3

H/3He

The neutron multiplication coefficientkscrin the blanket depends not only on accumulation of fission products as it

was described in the previous chapter, but also on accumulation of volatile materials3H and 3He. 3H is produced

mainly from 6Li (6Li(n,a)3H), which should be continuously replenished with new TRU in the molten salt blanket in

order to provide the continuous supply of 3H for fusion device. At the same time, due to the 3H b-decay the 3He is

0.60

0.65

0.70

0.75

0.80

0.85

0.90

kscr

Time, d

FP removal:

100%

50%

20%

0%

0 500 1000 1500 2000

Fig. 10. kscras a function of different percentage of fission products removed from the flibe.

0 500 1000 1500 20000

50

100

150

200

250

132Xemass,

kg

Time, d

Continuous FP removal:

50%

20%0%

Fig. 11. Mass variation of the stable fission product 132Xe during the irradiation time in the case of continuous feeding of TRU and different removal

percentage of fission products.



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produced. The microscopic neutron capture cross-section of 3He(n,p)3H is about 1000 b. The efficient neutron

absorption in

3

He disturbs neutron balance in the hybrid system blanket as it is shown inFig. 12. Due to the presenceof3H and 3He in the molten salt thekscrof the system decreases all the time starting from 400 days and in such case the

equilibrium stage in the system cannot be achieved. After 2000 days 5% difference in kscr is obtained for the two

cases: blanket with 3H and 3He and blanket without these nuclides. In more realistic scenario volatile 3H and 3He

nuclides either will be removed (separated for fusion process as in the case of 3H) or will escape by leakage from the

molten salt blanket. Indeed, the solubility of hydrogen in the flibe molten salt is extremely low such that it is very

often the total inventory of hydrogen in the flibe blanket is only a few grams, compared to about 80 metric ton of flibe.

On the other hand, the certain part of 3H and 3He will be present in the blanket during the system operation and this

should be taken into account when making modelling calculations for fusionfission hybrid system.

4. Conclusions

The performance a fusionfission hybrid system with the molten salt (flibe) blanket, driven by a plasma based

fusion device and dedicated to burn actinides extracted from the LWR spent fuel TRU, was analyzed. Detailed burn-

up calculations with different modes of replenishing the destroyed TRU material and also with different treatment of

fission products were done using Monteburns code system. Major system performance parameters of the system (e.g.

kscr,Fn(En), equilibrium conditions, etc.) were investigated. We conclude that equilibrium concentration for most of

the TRU materials in the blanket can be achieved after about 3 years of exposure. Such a plant could annually burn

about 1.1 metric ton of spent fuel actinides when the fission power output is 3000 MWth. Different replenishing of the

destroyed TRU material was tested (continuous and discrete feeding). A stable neutron energy spectrum was obtained

in both cases. The behavior ofkscrin the case of a discrete feeding has sharp fluctuations if compared with continuous

feeding. On the other hand, a different TRU mass in the blanket at equilibrium was obtained (3100 and 2770 kg with

continuous and discrete feeding, respectively). Although a smaller TRU mass at equilibrium is preferred due to safety

reasons, a more stable performance of the system obtained in the case of a continuous feeding is recommended.

Accumulation of fission products in the blanket was investigated including their influence on the performance of the

system. We found thatkscris rather stable (0.810.79) during the calculation time when all or at least 50% of fission

products are taken away. The removal of a smaller percentage of fission products is not recommended since,kscrwould fall down considerably (e.g. 0.81(0.76 and 0.81(0.65 with only 20 or 0% FP removal, respectively, within

2000 days of burn-up). The most poisonous fission products were identified in terms of their macroscopic capture

cross-sections. We found that fission products are responsible at least forK3150 pcm decrease inkscrat 15% burn-up.

The presence of volatile 3H and 3He has significant influence on hybrid systemkscrstability and should be taken into

account when performing hybrid system model calculations. In spite of very promising results on the efficient TRU

destruction, we found that at equilibrium, significant amount of curium isotopes is accumulated, while for the rest of

actinides fully equilibrated concentrations are reached. In this respect an optimization procedure should be developed

in order to determine the optimal fission-blanket composition corresponding to the fast incineration rate of actinides

0 200 400 600 800 1000 1200 1400 1600 1800 20000.74

0.75

0.76

0.77

0.78

0.79

0.80

0.81

0.82

0.83

0.84

k

scrwithout 3H and 3He

kscr

with 3H and 3He

kscr

Time, d

Fig. 12.kscr behaviour in the hybrid system blanket with 3H and 3He and in the system with 3H and 3He removal.



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including all minor actinides. Another important issue is directly related to the benchmark of different sets of nuclear

data libraries in the case of minor actinide incineration scenarios. These problems are addressed in our companion

paper byPlukiene et al. (in press).

Acknowledgements

This work was accomplished in the frame of the FrenchLithuanian cooperation program Gilibert (Ref. No

06452PG) and partly was supported by a Lithuanian State Science and Studies Foundation within C-03049 project.

References

Briesmeister J. for Group X-6 (1993), MCNP-A, A General Monte Carlo Code for Neutron and Photon Transport, Version 4A, LANL, preprint LA-

12625-M.

Cheng E.T. and Wong Clement P.C. (2000), Transmutation of Actinide in Fusion Reactors and Related Systems Analysis, Proc. of 10th Int. Conf.

on Emerging Nuclear Energy Systems, ICENES2000, Petten, The Netherlands, 2428 September (and Refs. therein).

Cheng, E.T., Cheng, E.T., 2001. Characteristics of Promising Transmutation Blanket Concepts Proc. of the International Workshop on Blanket and

Fusion Concepts for the Transmutation of Actinides, General Atomics, San Diego, California, USA 2001.

ENDF/B-VI.The US Evaluated Neutrons Nuclear Data Library for Neutron Reactions, IAEA-Vienna, IAEA-NDS-100 (1995), Rev 6.

JENDL-3.2, Japanese Evaluated Nuclear Data Library, IAEA-Vienna, IAEA-NDS-110, (1994), Rev 5.Plukiene R., Plukis A., Ridikas D., Cheng E.T., Fusion - Fission Hybrid System for Nuclear Waste Transmutation (II): From the Burn-up

Optimization to the Tests of Different Data, accepted for publication in Progress in Nuclear Energy (June 2005).

Ridikas D., Fioni G., Goberis P., Marie F., A (2000), On the Fuel Cycle and Neutron Fluxes of the High Flux reactor at ILL Grenoble, Proc. of the

5th Int. Specialists Meeting SATIF-5, OECD/NEA Paris, France, 1721 July.

Ridikas D., Bletzacker L., Goberis P., Fioni G., Marie F., B (2000), Critical and Sub-critical GT-MHRs for Waste Disposal and Proliferation-

Resistant Plutonium Fuel Cycles, paper 7 in the Proc. of the 6th OECD/NEA Information Exchange Meeting on Actinide and Fission Product

Partitioning and Transmutation, p. 71. Madrid, Spain, 1113 December.

Ridikas, D., Fioni, G., Cheng, E.T.A., 2001a. A Importance of Fission Products in the Molten Salt Blanket for Fusion - Fission Hybrid Systems

Proc. of the International Workshop on Blanket and Fusion Concepts for the Transmutation of Actinides, General Atomics, San Diego,

California, USA 2001.

Ridikas, D., Bletzacker, L., Goberis, P., Fioni, G., Marie, F., 2001b. Comparative Analysis of ENDF, JEF, and JENDL Data Libraries by Modelling

High Temperature Reactors (GT-MHR) and Plutonium Based Fuel Cycles Proc. of the Int. Conference on Nuclear Data for Science and

Technology, ND2001, Tsukuba, Ibaraki, Japan 2001.

RSICC Computer Code Collection (1991), ORIGEN 2.1 -Isotope Generation and Depletion Code Matrix Exponential Method, Radiation shielding

Information Center, report CCC-371.

Salvatores, M., 2002. The physics of transmutation in critical or sub-critical reactors. C.R. Physique 3, 9991012.

Trellue H.R. and Poston D.I. (1999), Users manual, Version 2.0 for Monteburns, Version5B, LANL, preprint LA-UR-99-4999.


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