TMSR

Proceedings of GLOBAL 2005

Tsukuba, Japan, Oct 9-13, 2005

Paper N�

428

Proposal for a Simplified Thorium Molten Salt Reactor

L. MATHIEU1, D. HEUER1, A. BILLEBAUD1, R. BRISSOT1,

C. GARZENNE2, C. LE BRUN1, D. LECARPENTIER2, E. LIATARD1,

J.-M. LOISEAUX1, O. MEPLAN1, E. MERLE-LUCOTTE1 and A. NUTTIN1

1Laboratoire de Physique Subatomique et de Cosmologie , 53, avenue des Martyrs, F-38026 Grenoble Cedex, France

Tel : +33 4 76 28 40 00, Fax : +33 4 76 28 40 042EDF-R&D, Département SINETICS, 1 av du Général De Gaulle, 92140 Clamart, France

ABSTRACT: Although they have been included in the Generation-IV International Forum selection of reactor systems, Molten

Salt Reactors have often been considered only as versions of the old MSBR concept. This concept suffers from several

problems that have not yet been solved. We have worked on these issues with simulation tools, using a coupling between a

neutron transport code and an evolution code. We have simulated numerous reactor configurations in order to obtain a

sustainable concept, the Thorium Molten Salt Reactor. This paper focuses on the advantages of a fast spectrum MSR in terms of

safety, breeding capability and materials life span. Several options may be considered to optimize the reactor, in terms either of

feasibility or of fissile inventory.

KEYWORDS: TMSR, fast spectrum, reactivity feedback coefficients, breeding ratio, fissile inventory, reprocessing,

bubbling system

I. INTRODUCTION

The Molten Salt Breeder Reactor (MSBR) is a well known con-cept [1]. This reactor had to be coupled to an efficient and, asa consequence, constraining [2] reprocessing unit in order tosecure the desired breeding capability. Later studies have de-termined that positive reactivity feedback coefficients can oc-cur [3, 4], leading to a possibly unstable reactor. MSBR devel-opments were discontinued in 1976.

Our goal is to bring a sustainable solution to some of theMSBR’s shortcomings. In this report we emphasize new devel-opments made in order to simplify the global operating schemeof a MSR that ensures appropriate neutronic behavior or sim-plifies the chemical constraints. In this view, we simulated alarge number of reactor configurations, including some basedon a fast neutron spectrum. These reactor types constitute anew concept, the Thorium Molten Salt Reactor (TMSR).

In Section II, we present the motivations for a fast neutronMSR. We show that such a reactor can operate with satisfac-tory properties, in terms of breeding ratio, feedback coeffi-cients, Transuranian production and materials steadiness to ir-radiation. In Section III, we consider simpler configurationswhich for example do not require a reprocessing unit. SectionIV is devoted to studies addressing the fissile inventory issue.Two approaches are presented, one playing on the reactor spe-cific power, the other on the chemical composition of the salt.

This work is based on the coupling of a neutron transportcode (MCNP [5]) with a materials evolution code. The formercalculates the neutron flux and the reaction rates in every cellwhile the latter solves the Bateman equations for the evolutionof the materials composition in the cells. These calculationstake into account the input parameters (power released, crit-icality level, chemistry), by adjusting the neutron flux or thematerials composition of the core on a regular basis. Our cal-culations are based on a precise description of the geometryand consider several hundreds of nuclei with their interactionsand radioactive decay; they allow detailed interpretation of theresults.

II. FAST NEUTRON MOLTEN SALT REACTOR

1. Description of the General TMSR

The general concept of the Thorium Molten Salt Reactor(TMSR) is based on a 2500 MWth (1 GWe) graphite moder-ated reactor. Its operating temperature is 630

�C and its ther-

modynamic efficiency is 40 %. The graphite matrix comprisesa lattice of hexagonal elements with 15 cm sides. The density ofthis nuclear grade graphite is set to 1.86. The salt runs throughthe middle of each of the elements. One third of the 20 m3 offuel salt circulates in external circuits and, as a consequence,outside of the neutron flux. The salt used is a binary salt, LiF- (HN)F4, whose (HN)F4 proportion is set at 22 % (eutectic

1

point), corresponding to a melting temperature of 565�C. The

233U proportion in HN is about 3 %. The salt density at 630�C

is set at 4.3 with a dilatation coefficient of 10� 3/

�C [6].

A graphite radial blanket containing a fertile salt surroundsthe core so as to improve the system’s regeneration capability.The properties of the blanket are such that it stops approxi-mately 80 % of the neutrons, thus protecting external structuresfrom irradiation while improving regeneration. We assumethat helium bubbling in the salt circuit is able to extract thegaseous Fission Products (FP) and the noble metals within 30seconds. For the first study presented, we consider a delayedreprocessing of the total salt volume over a 6 month periodwith external storage of the Pa and complete extraction of theFPs and the TRansUranians (TRU). We assume that the 233Uproduced in the blanket is also extracted within a 6 monthperiod.

The moderation ratio can be modified by changing the chan-nel radius. This alters the neutron spectrum of the core, placingit anywhere between a very thermalized neutron spectrum anda relatively fast spectrum. The core size is adjusted to keep thewhole salt volume constant. Figure 1 shows the influence of thechannel radius on the neutronic behavior . In order to evaluatethe performance of a reactor configuration, we check numer-ous parameters: total reactivity feedback coefficient, breedingratio, graphite life span and initial fissile inventory.

1 2 3 4 5 6 7 8 9 10 11 12 13channel radius (cm)

-6

-4

-2

0

2

t

otal

feed

back

coef

ficie

nt (

pcm

/°C

)

0.950

1.000

1.050

1.100

bree

ding

rat

io

1 2 3 4 5 6 7 8 9 10 11 12 13channel radius (cm)

0123456

233 U

inve

ntor

y (

met

ric to

ns)

05

1015

2025

ne

utro

n flu

x(x

1e1

4 n/

cm²/

s)

0

2

4

6

8

10

grap

hite

life

spa

n

(

year

s)

Figure 1: Influence of the channel radius on neutronic behavior.

A wide variety of neutronic behaviors is available by chang-ing the moderation ratio. We define three configuration types:thermal, epithermal and fast spectrum. Each one has advan-tages and drawbacks: a thermal spectrum leads to a low fissile

inventory but positive feedback coefficient, while a fast spec-trum implies a high breeding ratio but large fissile inventory.Studies have been carried out on these three fields of researchto improve these configurations and to find appropriate solu-tions [7, 8]. In this paper we will focus only on the fast neutronspectrum since it appears to be the most promising and simplestconfiguration.

2. Description of the fast neutron MSR

When the hexagons are fully filled with the salt (which corre-spond to an equivalent radius of 13.6 cm), there is no graphite inthe center of the core. The reactor is then composed of a singlebig salt channel. In such a configuration there is no graphiteirradiation problem, since there is no graphite inside the highneutron flux area 1.

Figure 2: Horizontal section of the simulated single salt chan-nel.

Figure 3: Vertical section of an upper quarter of the core

The horizontal section and schematic of this configurationare shown in Figures 2 and 3. The salt channel, of 1.25 m ra-dius and 2.60 m height, is surrounded by a thorium blanket and

1 The graphite part of the blanket structure is much less irradiated and thushas a longer life span.

2

two axial reflectors. These reflectors are made of ZrO2 in orderto avoid the use of a moderator material. As previously stated,it uses 20 m3 of salt (one third in the external salt circuit andout of the neutron flux). An efficient bubbling system extractsgaseous FPs and noble metals within 30 s, and a reprocess-ing unit slowly removes other FPs and TRUs of the fuel salt in6 months.

10-2

10-1

100

101

102

103

104

105

106

107

energy (eV)

10-10

10-9

10-8

10-7

10-6

10-5

neut

ron

flux

(n/

cm2 /n

-sou

rce/

dlnE

)

single channel MSR

BN800

Figure 4: Neutron spectrum of the single salt channel configu-ration compared to a typical fast spectrum.

0 20 40 60 80 100

time (years)

0

2

4

6

8

Ura

nium

inve

ntor

y (m

etri

c to

ns)

U

233U

234U

235U

236U

Figure 5: Inventory evolution of total uranium and its isotopesfor the single channel configuration.

Figure 4 shows the neutron spectrum of this configuration,compared with that of a sodium Fast Neutron Reactor (BN800).The neutron spectrum is not as fast as for BN800 because of theFluor (in)elastic scattering. Figure 5 shows the evolution of thetotal uranium inventory (and its isotopes) during 100 years ofoperation. The formation of 234U and heavier uranium isotopesis clearly visible.

From Figure 1, one can see that this configuration has a to-tal feedback coefficient of -5.37

�0.04 pcm/

�C 2. This coef-

ficient can be separated in two parts, one concerning the saltdilatation, and the other concerning only the temperature in-crease. Both are negative, respectively -2.02

�0.04 and -

3.14�

0.04 pcm/�C, thus ensuring a very good safety level.

Thanks to the thorium blanket and the 6 months reprocessing

2 The uncertainty given concerns the statistical error only.

of the FPs, the breeding ratio reaches 1.12. The averaged neu-tron flux in the core is 2 � 0 � 1015 n/cm2/s, for a specific powerof 250 W/cm3 in the salt 3. There is no graphite in the centerof the reactor, and the graphite blanket structure is much lessirradiated. It can withstand irradiation for about 10 years. Theinventories after 100 years of operation are reported in Table1. The FP inventory is stabilized at a low value thanks to thereprocessing unit. The TRU inventories are given for open cy-cle (when extracted) and closed cycle (when sent back in core).Even after this long period, the equilibrium state is not reached,especially for Curium, Berkelium and Californium.

Open cycle Closed cycleLi 5 t Np 9.7 kg Np 160 kgF 29 t Pu 600 g Pu 330 kg

Th 43.7 t Am 750 µg Am 8.8 kgPa 86 kg Cm 22 µg Cm 7.6 kgU 8.3 t Bk - Bk 4.8 gPF 210 kg Cf - Cf 30 g

Table 1: Inventory after 100 years of operation for the singlechannel configuration (in open and closed cycle).

III. SOME POSSIBLE ADDITIONAL SIMPLIFICA-TIONS

1. Suppression of the reprocessing

The breeding ratio of the single salt channel configuration ishigh enough to allow a reduction of the reprocessing efficiency.In this section, we analyse the behavior of this reactor with noreprocessing, but including the bubbling system as well as theblanket 233U recovery. First of all, we have to check the breed-ing capability of such a configuration. As no equilibrium stateis reached, the breeding ratio decreases continuously. In thiscase we examine the 233U stockpile, which represents the 233Ustocked next to the reactor (if positive) or needed to sustain crit-icity (if negative). Figure 6 shows this stockpile for the singlechannel configuration.

As shown, this configuration can withstand FP accumulationduring several decades. The stockpile reaches a maximum after15 years (+100 kg of 233U) and then slowly decreases (+0 kgof 233U after 20 years and -400 kg after 30 years). This con-figuration can thus operate a number of years while keeping anacceptable breeding capability. This performance is stronglyrelated to the thorium blanket. This conclusion applies onlyto the neutronic behavior of the FP accumulation, not to thechemical disturbances they may trigger. Problems of FP solu-bility may appear. They pile up at the rate of 0.35 %m 4 peryear , and 0.17 %m.y

� 1 for the lanthanides alone. This pointcan limit the extent of reactor operation to 20 years or so. Then,the salt has to be cleaned up and a fresh salt must be used.

3 These values are averaged on the 10 m3 of salt located inside the core, anddo not take into account the external salt circuit.

4 It represents the molar ratio of (FP)F2 over the sum of (HN)F4 andLiF. 0.35 %m.y � 1 is equal to 420 kg.y � 1.

3

0 5 10 15 20 25 30 35 40

time (years)

-2

-1.5

-1

-0.5

0

0.523

3 U s

tock

pile

(m

etri

c to

ns)

Figure 6: 233U stockpile for the single salt channel configura-tion without FP reprocessing.

FP accumulation slightly changes the other characteristicsof the reactor. The hardening of the neutron spectrum slowlyalters the feedback coefficients. Nevertheless, they are suffi-ciently negative to withstand this degradation. After 20 yearsof operation, the total feedback coefficient value is � 4 � 65

�

0 � 11 pcm/�C (compared to � 5 � 37 pcm/

�C with FP reprocess-

ing).

2. Dilution process

The aim of this process consists in diluting FPs andTRansUranic elements in a larger amount of salt, as shown inFigure 7. n is the ratio of the tank volume to the core volume.This process can be used as a substitute for the reprocessingunit, and added to the configuration studied above.

Corevolume = V

Tankvolume = n.V

fluorination

U

Th,Pa,TRU,PF

Th,Pa,TRU,PF

Figure 7: Dilution process scheme.

The volume V of the fuel salt is extracted in a time T, asfor a standard reprocessing. The uranium is separated and sentback into the core via a fluorination step. The salt containingFPs, TRUs, Th and Pa, is stocked in a tank of volume n.V.This tank is previously filled with fresh salt, which is injectedcontinuously in the core for a period of time of n.T. This outputflow balances the fuel salt input. During operation, the tank saltcomposition evolves as FPs and TRUs are accumulated in thecore. As there is no reprocessing, the FP inventory grows at thesame rate as previously shown, but this inventory is now dilutedin a larger amount of salt (n � 1 times larger). The poisoningdue to these elements is then reduced by a factor n � 1.

Figure 8 presents the 233U stockpile of single salt channelconfigurations for various dilution factors. When n � 0, there

0 20 40 60 80 100

time (years)

-4

-3

-2

-1

0

1

2

3

4

233 U

sto

ckpi

le (

met

ric

tons

)

with dilution process (n = 2)

without dilution process (n = 0)




Figure 8: 233U stockpile for single salt channel configurationswithout FP reprocessing and for various dilution factors.

is no dilution process and the stockpile is the same as that dis-played on Figure 5. Better breeding capabilities are obtainedthanks to the dilution process. The stockpile remains positiveduring 60 years as the salt volume is doubled (n � 1) and morethan one hundred years for larger tanks. As before, FP sat-uration in the fuel salt may occur. But this chemical limit isdelayed, as the FPs are diluted. Indeed, the operating time isextended by a factor n � 1 since the rise of the FP concentra-tion in the core salt is slowed down. The limit, set arbitrarilyat 20 years, is then extended to 40 years (n � 1, stockpile = 1ton), 60 years (n � 2, stockpile = 2.5 tons) or more.

Moreover, the TRUs are diluted too and that reduces theircapture rate. This has no impact on the breeding capabilitybut significantly reduces the formation of heavy nuclei (theequilibrium state is obtained later). After several decades,few TRUs have captured neutrons and formed Curium orCalifornium.

The dilution process can be implemented at low cost. It onlyrequires a fluorination step and a larger amount of salt and Tho-rium. No additional fissile matter is needed since Uranium isnot stocked in the tank. Although it leads to an increase of thefuel price, the dilution process seems to have major advantages.

3. Low Bubbling System Efficiency

All results presented before are obtained for configurations as-suming a very efficient bubbling system where gaseous FPs andnoble metals are extracted in about 30 seconds. Despite experi-ence gained with the MSRE, the bubbling system is not yet val-idated, specially for noble metals extraction. It is worthwhileto analyze the influence of a less efficient bubbling system onthe neutronic behavior.

We consider the single salt channel configuration studied inSection III.1. We assume that the bubbling system extracts FPsin a period of 1 month instead of 30 seconds. This less ambi-tious reprocessing is easier to implement than in the standardcase. With no bubbling system, gaseous FPs are supposed tomigrate towards surfaces, and noble metals to agglomerate onthe walls. For these reasons, extraction of these FPs in 1 monthwith a low efficiency bubbling system makes sense.

4

0 20 40 60 80 100

time (years)

-8

-6

-4

-2

0

223

3 U s

tock

pile

(m

etri

c to

ns)

with dilutionprocess (n = 1)

without dilution process (n = 0)

bubbling :low efficiency

bubbling :high efficiency

Figure 9: 233U stockpile for single salt channel configurationswithout FP reprocessing and for various bubbling system effi-ciencies and dilution factors.

Figure 9 shows the impact of a low efficiency bubbling sys-tem on the 233U stockpile, for configurations with and withoutthe dilution process. As shown, the breeding capability of thereactor is altered by the reduction of the bubbling efficiency.But this degradation is not as troublesome as one could imag-ine. Thus, the reactor can operate 10 years without needing alarge amount of 233U. With a small dilution process (n = 1), itcan last for 40 years.

With a low efficiency bubbling system, FPs accumulatemore quickly since many of them decay in not-extractedFPs. The accumulation rate reaches 0.53 %.y

� 1 instead of0.35 %.y

� 1. Concerning lanthanides, their formation rate re-mains unchanged. As the reactor can operate a shorter time be-cause of breeding degradation, the faster accumulation of FPsis not a constraining limit.

IV. LOWER FISSILE INVENTORY

The configurations presented before are very simple whilekeeping acceptable breeding capability. Nevertheless, one mayprefer to reduce the fissile inventory rather than the reprocess-ing efficiency. In this section we will study reactors with a re-processing unit and a bubbling system, as presented in SectionII.2. Since this reactor needs 5.5 tons of 233U to start, it couldbe interesting to reduce that to a lower value.

1. Specific power increase

The first solution consists in a specific power increase, i.ehigher than 250 W/cm3. This parameter can be modified intwo ways: by changing the fuel volume at fixed power or bychanging the total power at fixed salt volume. We have simu-lated a smaller reactor, using 10 m3 of salt instead of 20 m3.In order to keep a constant salt flow in the reprocessing unit,the whole core volume is reprocessed in 3 months. The char-acteristics of this configuration are compared with the standardconfiguration in Table 2.

The reactivity feedback coefficient is slightly improvedthanks to the neutron leakages, which are more likely in smallerreactors. The breeding ratio is affected both by neutron escapes

Salt volume 20 m3 10 m3

Feedbackcoeff. (pcm/

�C)

-5.37 -6.33

Breeding ratio 1.12 1.05Neutron flux

(x1015 n/cm2/s)2.0 4.1

233U inventory(m. tons)

5.5 2.7

Table 2: Influence of the salt volume on neutronic behavior.

and by the higher specific power. The FP inventory is stabilizedat a lower level thanks to the faster reprocessing, but the neutronflux increase compensates exactly for this inventory evolutionand leads to the same capture rate as before. On the contrary,the Pa inventory is mostly limited by radioactive decay and isalmost unchanged. A higher neutron flux leads to a higher cap-ture rate on Pa and degrades the neutronic balance. Finally, thefissile inventory is twice as low since the salt volume, of aboutthe same chemical composition, is reduced.

One last important aspect is the thermal hydraulic constraint.The thermal power is evacuated by the fuel which thus has tocirculate in the exchangers. Heat evacuation becomes more dif-ficult as the specific power increases. Small sized or high powerreactors are at a disadvantage in this respect, barring significantprogress in heat exchanger technology.

2. Chemical composition modification

The second way to reduce the fissile inventory is to change thechemical composition of the fuel salt. If the fuel contains alower proportion of Heavy Nuclei, the total amount of uraniumwill be reduced. This can be done by adding a third componentto the salt, for example beryllium, or simply by lowering theHN proportion in the LiF - (HN)F4 salt.

Figure 10: LiF - ThF4 phase diagram

The first solution consists in using the same salt as plannedfor the MSBR, and will not be discussed here. The second oneleads to a higher melting temperature, as one can see on Fig-

5

ure 10. Common structure materials cannot withstand such atemperature increase. However, new promising solutions basedon carbon (carbon-carbon, carbon fiber, carbides) could helpsolve this problem [9]. If this technology is not implemented,then the HN proportion parameter cannot be modified in thisway and this solution must be ignored.

We have simulated several configurations operating at1030

�C with 20 m3 of fuel salt. The proportion of (HN)F4

varies from 22 % to 2 %. At this temperature, the thermody-namic efficiency is assumed to increase from 40 % to 60 %and it has an incidence on the thermal power of the reac-tor: 1666 MWth instead of 2500 MWth are needed to pro-duce 1000 MWe. Similarly, the salt density decreases from4.3 to 3.89 because of the temperature related expansion effect.Moreover, we suppose that the HN flow 5 in the reprocessingunit must be kept constant, in order to keep the same reprocess-ing difficulty. It leads to a reduction of the reprocessing timefor low proportions of HN.

(HN)F4 proportion 22 % 10 % 5 % 2 %

Feedbackcoeff. (pcm/

�C

-5.12 -8.19 -10.87 -13.15

Breeding ratio 210 kg 1.024 0.941 0.850233U inventory

(m. tons)5.1 3.0 1.9 1.2

Table 3: Influence of the (HN)F4 proportion on the neutronicbehavior of a single salt channel configuration at 1030

�C

The neutronic behavior of these configurations is shown inTable 3. Results given for 22 % of (HN)F4 are slightly differentfrom those usually presented (in Table 2 for example). Thiscomes from the different total thermal power and salt densitydue to the temperature increase.

When the HN proportion is decreased, the neutrons are scat-tered for a longer time before they encounter a fissile or fertileelement (those that dominate neutron absorptions). This leadsdirectly to a more thermalized neutron spectrum and greatly in-creases neutron leakages and absorptions by the light elementsin the salt. The total feedback coefficient is improved partiallyby the larger neutron leakages, and by the neutron spectrumevolution. In the same way, the breeding ratio suffers deeplyfrom neutron escapes and neutronic absorptions on 19F. As an-ticipated, the fissile inventory is strongly decreased, with only1 ton for 2 % of HN.

The configuration with 10 % of (HN)F4 is interesting, since itkeeps a good breeding capability and a very negative feedbackcoefficient while needing a quite small fissile inventory.

V. CONCLUSION

We analyzed the impact on the behavior of the TMSR of suchparameters as the reprocessing, the moderation ratio, the core

5 In the previous section, we had referred only to the salt flow since the HNproportion was constant.

size, and the proportion of heavy nuclei in the salt. We haveidentified numerous interesting reactor configurations, using athermal, epithermal or fast spectrum. The solution that removesthe moderating block seems especially attractive: it has verygood neutronic properties, both for breeding ratio and for thereactivity feedback coefficients, and has no graphite life spanproblem due to irradiation. On the other hand, it needs a largeamount of fissile matter.

There are numerous solutions to improve this reference con-figuration. Very simple reactors may be designed by suppress-ing the FP reprocessing. Even in this case, reactors can operatesome ten years before reaching excessive FP poisoning. Thereactor may also be strongly improved in terms of breeding ca-pability and chemical behavior by a dilution process. Otherpaths of investigation are possible in order to reduce the fissileinventory, such as increasing the specific power or modifyingthe salt composition.

Our studies show that the MSR, and particularly a fast spec-trum configuration, is an interesting and sustainable conceptwhich can be used for Generation-IV deployment scenarios.

References

[1] M.W.Rosenthal et al., “Molten Salt Reactors - History, Sta-tus, and Potential”, Nuclear Applications and Technology,vol. 8 (1970)

[2] E.Walle, J.Finne, G.Picard, S.Sanchez, O.Conocar,J.Lacquement, “Molten Salt Reactors: Chemistry of FuelSalt and Fuel Salt Cleanup”, Global, New Orleans, USA(2003)

[3] D.Lecarpentier, “Le concept AMSTER, aspects physiqueset sûreté”, PhD thesis, Conservatoire National des Arts etMétiers, Paris (2001)

[4] A.Nuttin et al., “Potential of Thorium Molten Salt Reac-tors: Detailed Calculations and Concept Evolution With aView to Large Scale Energy Production”, Progress in Nu-clear Energy, Vol. 46, No.1, pp. 77-79 (2005)

[5] J.F.Briesmeister, “MCNP4B-A General Monte Carlo NParticle Transport Code”, Los Alamos Laboratory reportLA-12625-M (1997)

[6] I.Victor, E.Walle et al., “Density of Molten Salt ReactorFuel Salts”, Nureth, Avignon, France (2005)

[7] L.Mathieu et al., “The Thorium Molten Salt Re-actor: Moving On from the MSBR”, base arXiv,http://arxiv.org/abs/nucl-ex/0506004 (2005)

[8] L.Mathieu, “Cycle Thorium et Réacteurs à Sels Fondus:Exploration du Champ des Paramètres et des Contraintesdéfinissant le Thorium Molten Salt Reactor”, PhD the-sis, Institut National Polytechnique de Grenoble, Grenoble(2005)

[9] B.Tahon, Sgl Carbon Group, private communication(2004)

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