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Nuclear Engineering and Design 240 (2010) 2988–2993

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journa l homepage: www.e lsev ier .com/ locate /nucengdes

n the proliferation issues of a fusion fission fuel factory using a molten salt

atthias Vanderhaegena,∗, Greet Janssens-Maenhouta,c, Paolo Peeranic, André Poucetb,c

Department of Electrical Energy, Systems and Automation, Faculty of Engineering, University of Ghent, BelgiumCentre for Nuclear Engineering, Department of Mechanical Engineering, Faculty of Engineering, Catholic University of Leuven, BelgiumInstitute for the Protection and Security of the Citizen, Joint Research Centre, Ispra, Italy

r t i c l e i n f o

rticle history:eceived 22 January 2010eceived in revised form 17 June 2010ccepted 30 June 2010

a b s t r a c t

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is inthis case taken similar to the source of the Power Plant Conceptual Study – Water Cooled Lithium Lead(PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanketthe first wall of the ITER design is attached to a molten salt zone, in which ThF4 and UF4 solute salts are

transported by a LiF–BeF2 solvent salt. For this blanket design, the fissile material is assessed in quantityand quality for both the Th-U and the U-Pu fuel cycle.

The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket aresimulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated withthe burn-up code ORIGEN-S.

For the Th-U cycle the bred material output remains below 10 g/h with a 232U impurity level of 30 ppm,uper

while for the U-Pu cycle s

. Introduction

Each D-T fusion reaction results in 17.6 MeV of energy, an ˛article and a neutron. This neutron carries away 14.1 MeV of thereated energy, and is not necessary to sustain the fusion reaction.ach fission results in an average energy release of 200 MeV, two orhree fission products and two or three neutrons. But the producedeutrons in a fission reactor are necessary to sustain the fissioneaction. Therefore the energy density per neutron in a fission reac-ion is the highest, while there are more neutrons per unit of energyn a fusion reactor. A symbioses between both processes is knowns a fusion fission hybrid reactor. In the past several studies wereone on this type of hybrid devices, a summary of several differentoncepts can be found in Moir (1981). Though it is not necessary toave the high power fission density in the same reactor, this con-ept is known as the fission fusion fuel factory (FFFF) of H. Bethend W. Manheimer (Manheiemr, 2006; Bethe, 1979). Therefore aow ks value, around 0.2 as found in this study, can be tolerated.

In this paper such a subcritical FFFF reactor using a breedinglanket with a fluorine molten salt is discussed. For this study a very

onservative blanket geometry is assumed which might be appli-able for the ITER machine. The fluorine molten salt is a mixtureith very variable density and stopping power, which functions ascarrier of the breeding materials. These salts also allow easy and

∗ Corresponding author.E-mail address: Matthias.Vanderhaegen@gmail.com (M. Vanderhaegen).

029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2010.06.043

grade material is produced at a rate up to 100 g/h.© 2010 Elsevier B.V. All rights reserved.

continuous extraction of the fissile material bred from thorium ornatural uranium with the fluorine volatility process. The reactionproduct of this process, respectively UF6 or PuF6, is fully compatiblewith the common used fuel cycle’s front-end technology. This highaccessibility for recycling is also a cause of concern for prolifera-tion. To address the proliferation sensitivity, this paper evaluatesthe isotopic evolution of the fissile nuclear material produced in theblanket for a once through tokamak cycle with a perfect removalefficiency. In particular, the impurity level of the bred materialis determined, because this determines the radioactivity and cantake the role of intrinsic material barrier. The efficaciousness ofthis barrier is assessed with the possible dose received during thefabrication of a nuclear device from the bred material.

When using Th as breeding material the 232U impurity levelis important because this U-isotope has the high energy � emit-ter 208Tl in its decay chain, which determines almost entirely theradiotoxicity of the mixture. When breeding Pu, the Pu-isotopes238Pu, 240Pu and 242Pu can be formed as impurities beside the 239Pu.These Pu-isotopes with even atomic number are all characterizedby a high spontaneous fission rate, which can initiate a prematurechain reaction if the material would be used in a nuclear explosivedevice, and so a fizzle.

2. Breeding blanket

The geometry used for the breeding blanket is given in Fig. 1.The tokamak’s toroidal shape is simplified to an infinite slab geom-etry, with average width, to be compatible with XSRDNPM input.

M. Vanderhaegen et al. / Nuclear Engineering and Design 240 (2010) 2988–2993 2989

FH

TsmweslimTisc

adro6ms9

sttTmtfd

TT

Fig. 2. MCNP reaction rate simulations in function of the molten salt zone width

error in this energy range. In order to obtain reliable results in thisthermal energy range without excessively increasing the runtime,the variance reduction technique ‘energy splitting’ is applied.

ig. 1. The MCNP model of the breeding blanket, with the He cooling lines in theastelloy-N cladding voided.

he first wall (FW) module consist out of FW armor, a heat sink andtructural material with water cooling lines similar to the ITER FWodule. This ITER FW module is attached to a molten salt module,hich is build up of a molten salt zone surrounded with a cladding,

mbedding forced cooling. The cladding is made out of a moltenalt compatible material (such as Hastelloy-N), with He coolingines. These lines create a small frozen layer at the salt-claddingnterface which offsets corrosion caused by the interaction of the

olten salt flow and the tokamak magnetic field (Homeyer andhermal, 1965). This modular type of design is considered becauset offers easy maintenance and/or replacement while it providesufficient barriers to prevent a molten salt leak towards the vacuumhamber.

A D-T fusion device needs to be fueled with tritium, thereforeLiF–BeF2 solvent salt is considered to have some tritium pro-

uction. The fertile material solute is ThF4 and nat-UF4, breedingespectively U and Pu fissile material. The LiF–BeF2–HNF4 (HN = Thr U) compositions used in this study are 47.5–47.5–5 mol% and8.2-26.6-5.1 mol%, for which the density is calculated with partialolar volumes (Cantor, 1968). Also the use of Li enriched in 6Li is

tudied by considering natural Li (7.59 at% 6Li) and Li enriched to0 at% 6Li. These material properties are summarized in Table 1.

The molten salt zone size for the LiF–BeF2–ThF4 mixtures, wasimulated in steps of 25 cm with MCNP4c2. These simulations showhat the (n,�) and (n,2n′) reaction rates and the T production starto saturate at a molten salt zone width of 75 cm as shown in Fig. 2.herefore the molten salt zone size is chosen at 75 cm for all the

olten salt mixtures. The cladding width is chosen at 49 mm, with

he ø20 mm cooling lines placed closed to the salt-cladding inter-ace. The details of the FW module are taken from the ITER technicalatabase (ITER (2008), PDD2-3 p7).

able 1he mixtures considered in this study. The density is at a temperature of ≈700 ◦C.

at% 6Li � (g/cm3)

LiF–BeF2–ThF4 mol%47.5–47.5–5 90 2.44747.5–47.5–5 7.59 2.46768.2–26.6–5.1 90 2.49268.2–26.6–5.1 7.59 2.523

LiF–BeF2–UF4 mol%47.5-47.5–5 90 2.46247.5–47.5–5 7.59 2.48168.2–26.6–5.1 90 2.50968.2–26.6–5.1 7.59 2.540

(in cm) for the different LiF–BeF2 solvent mixtures, using ThF4 as solute. The resultsare given in number of reactions per source neutron. Top: T production, middle:(n,�), and bottom: (n,2n′).

3. MCNP simulations

3.1. Neutron flux

MCNP is used to simulate the transport of the fusion neutronsinto the breeding blanket. The external D-T fusion source is definedas a Gaussian energy distribution with a mean energy of 14.1 MeVas done in the PPCS-A study (Pampin-Garcia and Loughlin, 2002). Inthe PPCS-A study a Gaussian distribution with a spread appropriateto a 58 keV central ion temperature is applied, in combination witha spatial spread. Because a simplification to a slab geometry is made,also the neutron source is simplified to a surface source without aspatial spread. To compensate for this, the effective temperature of52.6 keV is used for the spread of the Gaussian energy distribution.Neutron losses, for example to the divertor, are taken into accountby a correction factor in post-processing.

The backscattering of the breeding blanket is included becauseit accounts for approximately 68% of the initial neutrons and has asofter spectrum. To model this, the surface source is placed betweenreflective surfaces.

With the high energy fusion spectrum of the initial particles,only a small fraction of particle runs contribute to the results in thethermal energy range. This normally leads to a very high relative

Fig. 3. MCNP simulation flux averaged over molten salt zone for the differentLiF–BeF2 solvent mixtures, using ThF4 as solute.

2990 M. Vanderhaegen et al. / Nuclear Engineerin

Table 2MCNP reaction rate simulations for the different LiF–BeF2 solvent mixtures, usingThF4 as solute. The reaction rates are given as number of reactions per sourceneutron.

LiF–BeF2–ThF4 mol% T prod 232Th (n,�) 233Th (n,2n′)

47.5–47.5–5, 90 at% 6Li 0.424 1.13 × 10−5 2.38 × 10−6

sttsr

3

tw2r

gtwgtrafrtfs

4

sSctatr

(wrf

•

•

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with �MCNP(E) from Fig. 3, the energy E in MeV and �fast is the fastflux value. The hybrid weighing yields different results than the fis-sion weighing. As a consequence, not all production pathways forimpurities are equally well modelled. These impurities produced

Table 3Comparison of both methods for LiF–BeF2–ThF4, 47.5-47.5-5 mol%, 90 at% 6Li.Results are given for 0.1 h of irradiation time.

47.5–47.5–5, nat-Li 0.269 6.18 × 10−5 2.45 × 10−6

68.2–26.6–5.1, 90 at% 6Li 0.445 8.16 × 10−6 2.60 × 10−6

68.2–26.6–5.1, nat-Li 0.299 5.26 × 10−5 2.73 × 10−6

The result of the total simulations are given in Fig. 3. The figurehows clearly the Gaussian 14.1 MeV fusion peak, the fluorine andhorium resonances. It can be concluded from the simulations thathe 6Li content in the mixture reduces the thermal and epithermalide of the neutron spectrum. But the fast side of the spectrumemains nearly unchanged for all the considered mixtures.

.2. Reaction rates

MCNP calculates the reaction rates normalized per source neu-ron as in Table 2. Geometrical leakage of the initial fusion neutronsas estimated to decrease the calculated values of Table 2 up to

0%, because already the leakage towards the divertor opening iseportedly (Chen et al., 2003) 13.4%.

The tritium breeding ratio (TBR), defined as the ratio of tritiumenerated in the blanket to the tritium burnt in the fusion reac-or, is determined with the T-production rate of Table 1 correctedith a factor 0.8. This T-production rate was calculated with (i) the

eneral (n,T) reactions, (ii) the 7Li (n,˛n’)T reactions, and (iii) theernary fission reactions. The data to compute these ternary fissioneaction T yield was taken from IAEA-TECDOC-1168 (Compilationnd evaluation of fission yield nuclear data, 2000). The TBR variesor the different molten salt cases between 0.215 and 0.356. Theseesults show that this design is not tritium self-sufficient. The con-ribution of the ternary fission is minor (≈10−5% for Th and ≈10−4%or nat-U). The effect of using Li enriched in 6Li is small, pointing toelf-shielding effects by this highly absorbing nuclide.

. Calculating isotopic evolution with ORIGEN-S

The evolution of the nuclear material that is bred in the moltenalt blanket is calculated with the code ORIGEN-S. ORIGEN-S is theCALE module which solves the Bateman equation for the isotopeoncentration starting from initial nuclide concentrations and reac-ion rates. Special attention was given to the isotopic compositionnd impurity level of the bred U and Pu because of the aforemen-ioned intrinsic barriers. The main issue is to input the reactionates.

The spectral parameters used by ORIGEN-S are coupled to the 6Lin,˛) and fertile material (n,2n′) and (n,�) reaction rates calculatedith MCNP, by solving the following set of equations which rep-

esents the method that ORIGEN-S uses to calculate reaction ratesrom the ORIGEN-S MSBR library:

(n,˛) for 6Li:

�eff = THERM · �0 + RES · I + FAST · �fiss−avg

(n,�) reaction for the fertile nuclide:

� = THERM · � + RES · I

eff 0

(n,2n′) reaction for the fertile nuclide:

�eff = FAST · �fiss−avg

g and Design 240 (2010) 2988–2993

Where the solution of this set of equations gives THERM, RESand FAST, which are spectral parameters. �eff is the effective cross-section that can easily be derived from the values of Table 2, bymultiplying by:

AMCNP

AReal

0.8n0

N˚thermal

with N the nuclide density and ˚thermal the thermal neutron flux.The factor AMCNP / AReal is deduced from the area used in MCNP(AMCNP) and the total area in the PPCS-A studies (derived from theratio of the total first wall power and the average first wall heatflux: AReal = PFW,tot / PFW,av ≈ 1488.2 × 104 cm2). The neutron sourceintensity n0 mentioned in the PPCS-A study is 1.88 × 1021 n/s.

The information on �0, I and �fiss−avg, which are typical thermalfission reactor cross-section properties, is taken from JEF Report14 (Nuclear Energy Agency, 1994). This report mentions differentintegration boundaries for the resonance integral I and the fissionaverage cross-section �fiss−avg than used by the ORIGEN-S MSBRlibrary, though according to the ORIGEN-S manual this is acceptablefor estimations of the order of magnitude (Hermann and Westfall,2000).

The method using ORIGEN-S with spectral parameters was con-fronted with an alternative method to calculate the reaction ratesusing the XSDRNPM SCALE module. For the latter a 238 groupsource deduced from the MCNP flux is coupled into XSRDNPM.XSRDNPM then calculates the problem specific reaction rates whichcan be automatically coupled in ORIGEN-S.

To summarize, Fig. 4 represents a simple flowsheet of the afore-mentioned methods.

The solution of the aforementioned set of equations rendersthe most accurate reaction rates for the primary reactions towardsthe impurity isotope. But because there is rather good agreementbetween the XSRDNPM recreated and the MCNP calculated groupflux spectrum over the thermal and epithermal range, an indica-tion of the accuracy is given by comparing both methods in Table 3for the LiF–BeF2–ThF4, 47.5–47.5–5 mol% using 90 at% 6Li. As couldbe expected the fissile material production is similar, while theimpurity level is an order of magnitude less when using XSRD-NPM. For this reason, the ORIGEN-S spectral parameter methodwas used for the results presented in the next paragraph. Remain-ing shortcomings are noticed when comparing the hybrid weightedcross-section (�hybrid−avg) and the fission weighted cross-section(�fiss−avg):

�fiss−avg =

∫�(E) · �fission(E) · dE

1.49 · �fastwith

�fission(E) = 0.77 · √E · exp(−0.775 · E)

�hybrid−avg =

∫�(E) · �MCNP(E) · dE · 0.8 · ntot · (AMCNP/AReal)

1.49 · �fast

XSRDNPM ORIGEN-SSpectr. Par.

TBR 0.203 0.369U-output (g/cm3) 2.90 × 10−17 1.98 × 10−17

232U (ppm) 2.30 × 10−5 3.63 × 10−3

M. Vanderhaegen et al. / Nuclear Engineering and Design 240 (2010) 2988–2993 2991

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The fissile material production can go up to 4.72 g/h withan impurity level of ≈30 ppm after 100 h of irradiation forLiF–BeF2–ThF4, 47.5-47.5-5 mol%, nat-Li, but with a reduced TBRof 0.215. So there will be a trade off between T production and

Fig. 5. The U output for the different solvent salts with Th solute in function ofirradiation time.

ig. 4. Flowsheet of the methods used for isotopic evolution. Top: the general ORIGEpectral parameters to calculate the reaction rates.

y a reaction with a certain energy threshold, such as 232U are notompletely covered and so underestimated. In particular, the impu-ity 232U is preceded by 232Pa, which can be produced starting from32Th via the next two pathways:

232Th + n → 231Th + 2n′

231Th → 231Pa + e− + �̄e

231Pa + n → 232Pa + �

232Th + n → 233Th + �

233Th → 233Pa + e− + �̄e

233Pa + n → 232Pa + 2n′

Both pathways include (n,2n′) reactions which are character-zed by an energy threshold. The weighted cross-section for thesen,2n′) reactions are given in Table 4. For 232Th(n, 2n′)231Th in therst pathway, this difference is adjusted with the FAST spectralarameter. For the 233Pa, the difference can not be compensatednd therefore this reaction path is underestimated.

The fissile material production is calculated in grams per cm3 ofolvent salt for a once through tokamak cycle. To give an estimatef the time and the quantity of production, this value is multipliedith the mass flow to get a result in grams per hour. This mass flow

s calculated as:

˙ = AReal · LBlanket

TIrradiation(in cm3/h)

ith blanket surface AReal = 1488.2 × 104 cm2, the blanket widthBlanket = 75 cm and the irradiation time TIrradiation in h. The results

or the Th-U fuel cycle, which remain below 10 g/h U after 100 hrradiation are given in Figs. 5 and 6, and discussed in Section 5.1. Forhe U-Pu fuel cycle the results mount up to an outcome of 100 g/h Pufter 100 h irradiation, as shown in Fig. 10 and discussed in Section.2.

able 4veraged cross-sections in units of barns for LiF–BeF2–ThF4, 47.5-47.5-5 mol%,0 at% 6Li.

(n,2n′) cross-sections 232Th 233Pa

�fiss−avg (Nuclear Energy Agency, 1994) 0.2059 × 10−1 0.5623 × 10−2

�hybrid−avg 0.3419 0.3453

owsheet, middle: calculating the reaction rates using XSRDNPM, and bottom: using

5. Discussion of the resulting breeding of fissile material

5.1. The Th-U fuel cycle

The results for the Th-U fuel cycle are given in Figs. 5 and 6.The produced uranium, starting from 100% 232Th, consists mainly(≈99.9%) out of 233U directly after irradiation, but the total outputof fissile material decreases with 6Li content. This 6Li reduces theflux in the thermal and epithermal region, which in turn reducesthe neutron-capture in the fertile nuclides. In addition, the 232Uimpurity level shows to be decreased with the 6Li content while the232Th (n,2n′) reactions in Table 2 show to be nearly independent ofthis concentration. This is due to the intermediate step via 231Pawhich is transmuted in a thermal and epithermal flux.

Fig. 6. The 232U impurity level for the different solvent salts with Th solute in func-tion of irradiation time.

2992 M. Vanderhaegen et al. / Nuclear Engineering and Design 240 (2010) 2988–2993

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ig. 7. The U output for LiF–BeF2–ThF4, 47.5–47.5–5, 90 at% 6Li in function of decayime and irradiation time.

ssile material production. But this is only for direct fissile mate-ial extraction. When considering decay of the Pa in the mixturehe output quantity can increase several orders of magnitude as ishown in Fig. 7, while Fig. 8 shows that the impurity level in the Uixture decreases. The latter is due to the difference in half-life of

32Pa (T1/2 = 1.31 days) and 233Pa(T1/2 = 27.0days).The dose equivalent rate due to � exposure to this U mixture is

iven in Fig. 9 per kg 233U, per ppm 232U. For 8 kg of 233U, the quan-ity which is estimated to be sufficient to produce a nuclear weaponIAEA Safeguards Glossary, 2002), the dose equivalent rate becomes8 mSv/h/ppm as long as the material is fresh. But with aging of theaterial, this value goes up to 800 mSv/h/ppm. For a dose equiva-

ent of 10 Sv, fresh material can be handled for 12500–12.5 h ppmor aged material. This dose is already received for almost 1 year

232

ged material with 8 ppm U within 7 days, which complicatesny handling and use, such as e.g. for a nuclear device.

ig. 8. The 232U impurity level for the LiF–BeF2–ThF4, 47.5–47.5–5, 90 at% 6Li inunction of decay time and irradiation time.

ig. 9. Dose equivalent rate at 1 m due to � exposure, per kg 233U, per ppm 232U.

Fig. 10. The Pu output for the different solvent salts with nat-U solute in functionof irradiation time.

5.2. The U-Pu fuel cycle

The result for the U-Pu fuel cycle is given in Fig. 10. Theplutonium production starting from natural U (99.29% 238U)mounts to 78.2 g/h after 100 h of irradiation for LiF–BeF2–UF4,47.5–47.5–5 mol%, nat-Li, but with a reduced TBR of 0.220. Whenusing nat-U as solute the reduction in TBR is slightly smaller thanwhen using Th as solute. This can be explained by the small 235Ucontent in the nat-U mixture, which increases the neutron mul-tiplication by fission. As with the Th solute, the total output offissile material decreases with 6Li content. The bred Pu is always ofsuper grade quality (>97%239Pu), and therefore poses a significantproliferation concern.

When considering decay of 140 days, the output quantityincreases only with a factor of 2 while the quality remainsunchanged. This is the result of the smaller half-life of 239Np(T1/2 = 2.355 days) compared to 233Pa, but also the similar half-lifefor 238Np (T1/2 = 2.117 days) and 239Np.

5.3. The potential capacity of fissile material breeding

To assess the potential breeding capacity, the amount of bredmaterial in 1 year is compared to the amount which is necessary fora yearly PWR core reload. The amount of bred material is calculatedby taking the maximum production rate and from this deduce theproduction in 1 year. To calculate the potential breeding capacityestimate, it is assumed that the flux and power level in the reac-tor core should remain similar. The energy release per fission isassumed constant for all fissile nuclides. For an average 1000 MWePWR reactor, the amount of fuel replaced is approximately cor-responding to 20 ton of uranium. Though this uranium is on theaverage enriched to 3.8 wt% of 235U, which corresponds to 760 kgof fissile 235U.

For 233U mixed with depleted uranium (with an enrichment of0.3 wt%), a mass of 818 kg is necessary. This is much more thancan be produced. Though if decay would be taken into account,this could be sufficient for several core reloads. In the meanwhilethe fuel is proliferation resistant due to the isotopic dilution of thefissile nuclides.

For 239Pu mixed with depleted uranium, a mass of 594 kg isnecessary. This is comparable to the Pu produced. So only onecore reload can be covered. Taking into account decay, up to 2reloads could be covered. But unlike for 233U, no isotopic dilution isobtained and the fuel remains a significant proliferation concern.

5.4. Discussion of the sensitivity of the simulations

The molten salt solvent is a medium in which high energy ˛ par-ticles can be converted to neutrons by 9Be, 6Li and 19F. Togetherwith a high ˛ particle production this produces an extra neu-

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M. Vanderhaegen et al. / Nuclear Engi

ron source. The magnitude of this neutron source is estimated byOURCES-4C. Though only the total ˛ particle production is giveny ORIGEN-S, all of these were given an energy of 6 MeV. It is verynlikely that an ˛ particle with higher energy will be created sincehe ˛ is in most cases the heaviest particle, therefore carrying theeast energy after a reaction. It was found that 3 × 10−5 neutrons,

ith an average energy of 4 MeV, are created per ˛ for the solid saltixture. When multiplying with the speed and ˛ intensity the esti-ate of the neutron flux in the LiF–BeF2–ThF4, 47.5–47.5–5 mol%,

0 at% 6Li mixture is 107 n/s cm2. This is 6 orders of magnitudeower than the total flux (1013 n/s cm2) and therefore representsnly a small perturbation on the results.

Because of the low irradiation times, a linear extrapolation ofhe presented results can be made. The quality and quantity (innits of g/cm3) is proportional to the fusion source intensity dividedy the total first wall area. The output quantity in units of g/h isnly proportional to the source intensity, but reduced by a fac-or AIrradiated / AReal if only a fraction of the first wall has a blanket.o for the ITER design which has a neutron source intensity of anrder of magnitude lower then for the PPCS-A model and a first wallrea which is approximately half of that in the PPCS-A model, thempurity level is a factor of 5 lower. This means that the maximumchievable output quantity for a similar test blanket in the ITEResign, assuming that ITER can operate in a continuous regime, isufficiently low not to be of high proliferation concern.

. Tritium consumption

The PPCS-A model has a source intensity of 1.88 × 1021 n/sPampin-Garcia and Loughlin, 2002), which corresponds ideallyo a T consumption of approximately 300 kg/y. Since the TBR inhis study varies from 0.25–0.45, a share of 55–75% will need toe supplied externally. This corresponds to 165–225 kg annually.herefore this type of device is only suitable for a fuel cycle whichas an excess of T, for which it’s more valuable to convert the T intossile material instead of leaving it to decay to 3He.

Since an ITER-like fusion device has an order of magnitude lower-consumption, due to the lower source intensity, the fuel con-umption of such a hybrid is much better.

. Conclusion

The molten salt LiF–BeF2 for the blanket of a hybrid reactor

eems not to be the most suitable material for a fusion breedingycle when considering the salt’s tritium breeding capacity belowelf-sufficiency in combination with a non-optimized first wall.owever it offers the potential to manage fissile material produc-

ion in a flexible way.

g and Design 240 (2010) 2988–2993 2993

This study shows that it is theoretically possible to use theintense neutron source of a fusion device to breed fissile materialin the subcritical blanket. Two types of fuel breeding in the moltensalt are investigated: (i) with ThF4 solutes and (ii) with UF4 solutes(composed of natural U). When adding ThF4 solutes, U-233 startsbeing produced after 100h irradiation with only a maximum rateof 4.72 g/h U and with a considerable 232U impurity. This howeveris not sufficient for an average 1000 MWe PWR core reload. Thegamma-exposure risk corresponding to this impurity level is notsufficient to take the role of intrinsic barrier, as it takes time tobuild up. However taking into account that a fusion device such asITER has one order of magnitude lower neutron source intensityand a smaller first wall area then the PPCS-A model, this impuritylevel is reduced with a factor 5. In that case, assuming continu-ous operation mode of the ITER device and using the entire FWfor breeding modules less than 2 years are needed to produce onesignificant quantity of 8 kg 233U. This is far less effective than in athermal converter and therefore much less proliferation sensitive.

The case of UF4 solutes allows to breed after 100 h irradiation upto 78.2 g/h supergrade Pu and therefore poses a higher proliferationconcern. On the other hand a complete core reload could be coveredby it. In the case of an ITER-like fusion device, assuming continuousoperation and use of the entire FW, one significant quantity of 8 kgPu could be bred in less than 43 days with a T consumption ofapproximately 2.65 kg. Therefore the addition of UF4 solutes makesthe hybrid reactor as proliferation sensitive as a thermal converter.

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