J. Plasma Fusion Res. SERIES, Vol. 10 (2013)

Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University
Satoshi FUKADA

Department of Advanced Energy Engineering Science, Kyushu University(Received: 7 May 2012 / Accepted: 6 August 2012)

There is increasing interest in using ionic molten-salt Flibe not only as self-cooled tritium(T)-breeding material in afusion reactor blanket but also as fuel solvent of molten-salt fission reactors. Application of Flibe to T-breeding fluid for astellarator-type fusion reactor operated at a high magnetic field brings large simplification of its blanket structure,
allowing continuous operation under high-beta plasma conditions. Using mixed Flibe-ThF4+UF4 fuel in molten salt
fission reactors permits stable long-term operation without fuel exchange. When Flibe or Flinak is irradiated by neutrons,
however, acid and corrosive TF is generated, and some T permeates through structural walls. In order to solve these
problems, chemical conditions of Flibe are changed using the redox-control reaction, Be+2TF=BeF2+T2. In addition,
permeation of hydrogen isotopes is lowered by enhancing T recovery rates. Part of Flibe-tritium researches are
performed at Idaho National Laboratory (INL) under the Japan-US collaboration work of JUPITER-II. Our own
contributions to the topics are shortly introduced in this paper.
Keywords: Flibe, tritium, fusion reactor, molten salt fission reactor, permeation
TBR>1 [20], which is required for a fusion blanket. Flinak
also gives a higher TBR value with use of Be [21].
Since Flibe heated to above m.p. can dissolve ThF4 andUF4 in any composition, it is used as fuel solvent in molten
salt fission reactors. 232Th is a fertile material which can beconverted into fissile 233U by a sequence ofneutron-absorption reactions and the two beta decays, i.e.
232
Th + n → 233Th → 233Pa → 233U. The earliest molten salt
reactor was set up in Oak Ridge National Laboratory(ORNL) in the 1960’s. China is recently interested in its
application to fuel solvent for the Thorium Molten Salt
Reactor (TMSR), because of their large possible reserves
and less generation of radioactive waste. TMSR can be
operated for extended time without fuel discharge. The
experimental fission reactor will be constructed in Shanghai
Institute of Applied Physics, China Academy of Engineering
Physics (SINAP, CAEP) in the near future. In this context, a

1. Introduction

Flibe is a stable mixed molten salt composed of
2LiF+BeF2 in molar ratio (2:1) with melting point of 459oC.
Flinak (LiF+NaF+KF) is also a stable molten salt with
0.465:0.115:0.42 in molar ratio and a similar m.p. of 454oC.
The Flibe and Flinak ionic salts are planned to be used not
only for a tritium(T)-breeding blanket in fusion reactors
such as FFHR-2 at National Institute for Fusion Sciences(NIFS) in Japan [11,13,15,17] but also as fuel solvent insmall molten salt fission reactors such as FUJI [21]. The
former is intensively investigated under the direction of Prof.
Sagara of NIFS [11,13,15,17]. Flibe is chemically andphysically stable even at high temperature. Its low electric
conductivity makes more suitable when utilizing instellarator-type fusion reactors, which can be operated underconditions of a high beta magnetic field. Flibe can achieve
the T self-sufficient condition of tritium breeding ratio

Fig. 1 Photos of solid Flibe at room temperature and transparent molten salt when its melts
author’s e-mail: sfukada@nucl.kyushu-u.ac.jp22©2013 by The Japan Society of Plasma
Science and Nuclear Fusion Research

S. Fukada, Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University

2. Redox Control

collaboration work on tritium release and measurement in
the TMSR was recently initiated between researchers of
SINAP, CAEP and our Kyushu University.
Generally, molten ionic salts of Flibe and Flinak arestable even under air at high-temperature. They become
transparent when they are melted as shown in Fig. 1. When
Flibe or Flinak is used for fusion or fission reactors under ahigh-intensity neutron flux or high-temperature conditions,various advantages and disadvantages are observed. The
advantages of Flibe or Flinak can be summarized asfollows: (i) comparatively low reactivity with O2 or H2O, (ii)low electric conductivity, (iii) low T solubility, and (iv) easy
T recovery. On the other hand, the disadvantages whenusing Flibe or Flinak are (i) generation of chemicallyreactive TF under neutron irradiation, (ii) comparatively
high viscosity (8.6 mPas at 600oC of Flibe compared with
1.0 mPas at 20oC of H2O), which leads to high pumping
power, (iii) the experimental increase of viscosity withincrease of BeF2 content, and (iv) high T permeability
through some metals.
Many experiments have been carried out under theJapan-US collaboration work called JUPITER-II in
2001-2006 and our contributions within the frame of thedomestic collaboration have continued since then. Part of
the tritium research was performed in the Safety and Tritium
Applied Research (STAR) facility of Idaho National
Laboratory (INL). Many achievements are summarized in
the references of the present paper [1-18]. Our T-related
experimental or design-based research on applications of themolten salts Flibe and Flinak to fusion or fission reactors isbriefly introduced in the present paper. The introduction ismainly devoted to our own studies due to the limit of paper
length.

When mixed molten salt of Flibe or Flinak is exposed
to neutron flux in fission or fusion reactors, tritium fluoride(TF) is generated by the reaction of LiF with neutrons asfollows:
6
LiF + n = TF + 4He
7
LiF + n = TF + 4He + n’
The reaction cross-section of 6Li with thermal neutrons
is very large. On the other hand, that with 7Li is small andhas the threshold energy of 2.47 MeV. Therefore,
enrichment of 6Li is often necessary to achieve high TBR,
when applied in a fusion reactor blanket. Since the reaction
product TF is a strong acid, it corrodes metallic walls.
Consequently, TF is converted into molecular T2 or HT,
which is chemically compatible with metallic walls and thefluorine is allowed to react with metallic Be. The
reduction/oxidation (redox) control reaction in the liquid
phase of Be (dissolved in Flibe) + 2TF (HF) → BeF2 + T2
(H2) is used for the purpose. The experimental results are
summarized in Fig. 2. It is assumed in the calculation thatthe Be dissolution rate is constant, the above reaction issecond-order in HF concentration and contributions of thereverse reaction are negligibly small. As seen in the figure,close agreement is obtained between experiment andcalculation regardless of the HF concentration and the Be
dissolution concentration. A reaction rate constant isobtained from fitting between experimental results andcalculations as shown in Fig. 2.
Since the solubility of T2 generated by the reaction in
Flibe is very low, T easily permeates through structural
materials of fusion or fission reactors. Therefore, highly
efficient procedures for T recovery from the Flibe flow infission or fusion reactors and an effective T permeation
protection are necessary in order to operate safely fusion or

Fig. 2 Comparison between experiment and calculation of reaction of 2HF+Be→BeF2+H2 in liquid Flibe

23

S. Fukada, Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University

fission reactors. Our previous work on T recovery andpermeation through Flibe covers the following five topics,(i) diffusivity and solubility of hydrogen isotopes in Flibe or
Flinak, (ii) Flibe purification and redox control by Be, (iii)chemical analysis of T in Flibe, (iv) design of equipment torecover T from Flibe, (v) design of Flibe blanket, etc. Manyof these results are described in the references [1-18]. In thenext section we focus on the solubility and permeability ofhydrogen isotopes in Flibe or Flinak.

3. Solubility and Permeability of H Isotopes in Flibeor Flinak

Fusion or fission communities are much interested in
two liquid breeders Flibe and Flinak, and many of theirvarious properties have been investigated experimentally.
Solubility and permeability of H isotopes in Flibe or Flinak
molten salts are very important factors for the design offusion reactors and to estimate the T-breeding performance.
Solubility is related to the T inventory and permeability to T
leakage through boundaries. Fig. 3 correlates solubility
constants for various gases in Flibe and Flinak. Dissolutionof gases in Flibe or Flinak follows Henry’s law described byc k = K H ,k pk , where ck and pk are the concentration of gas kdissolved in the molten salt and the partial pressure of gas k,respectively. KH,k (k=H2, He, Ar, Xe) is the Henry constant ofgas k. These gases are dissolved in atomic or molecularform in Flibe and Flinak, and the major force acting
between molten salt and dissolved inert gas is surface
tension. Therefore, the solubility constant KH,k is given by
the following equation regardless of Flibe and Flinak:⎛ E + 4 πN r 2γ ⎞
A b,k
⎟⎟
(1),
K H,k = exp⎜⎜− 0
R
T
g
⎝
⎠

Fig. 3 Solubilities of H2, He, Ar, Xe in Flibe or Flinak
The product of diffusivity DH2 and solubility KH,H2 isequal to the permeability when no retardation onmaterial-gas interface occurs. The steady-state H2
permeation rate denoted by jH2 is proportional to the partialpressure difference between the upstream and downstreamsides denoted by pH2,up and pH2,down, respectively, and isexpressed as follows:
K H, H 2 D H 2
j H2 =
p H 2 ,up − p H 2 ,down(2),δ
where δ is the thickness of the molten salt layer and DH2 isthe diffusivity of H2 in Flibe or Flinak.

()

4. Recovery of Tritium from Flibe or Flinak

The equilibrium partial pressure of T dissolved in Flibeor Flinak and vapor pressure of Flibe are shown in Fig. 4
together with other liquid blanket candidates. As seen there,
the T equilibrium pressure over Flibe is the highest, being
about 105 Pa for 1ppm T dissolved in Flibe. Therefore,
simple He gas purge is promising to recover T in Flibe if
direct contact between gas and liquid is possible. If directcontact is unsuitable, the problem is avoided by means of awhere rb,k is the atomic or molecular radius of the gaseous
species k, NA the Avogadro number, Rg the gas law constant,
T temperature and γ the surface tension of Flibe or Flinak.
Although E0 is a constant for inert gases, in case of H2
another attractive force between this molecule and themolten salt plays a role. Therefore the solubility of H2 showsdifferent tendency from that of inert gases (see Fig. 3).

Fig. 4 Comparison of T equilibrium pressure and vapor pressure among Li, Li17Pb83 and Flibe

24

S. Fukada, Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University

permeation window. If the redox control is not successful,the chemical form of tritium remains to be TF. The
equilibrium pressure of TF is, however, about 1/200 of T2.
Therefore the efficiency of gas purge is reduced.
Fig. 5 shows a schematic view of a He-Flibecounter-current extraction tower. Tritium dissolved in Flibeat the tower inlet diffuses from the Flibe bulk to the
Flibe-He interface. When it arrives at the surfaces, T desorbs
to the He side. Then T2 gas diffuses through the He-sideboundary layer. T2 in He is purged out to the gas outlet. In
order to keep good contact between He and Flibe, inert
metal packing is placed inside of the tower. Rashig ring,
Dixon ring or equivalents can be used as usually packed incommercial gas-liquid extraction towers such as distillation
process. Analytical study on T recovery from Flibe using thecounter-current tower is being performed in Japan and
China, but no experimental result has been presented.
Here, we compare Li chemical activity in the liquidblanket candidates among Li, Flibe, Li17Pb83 and Li25Sn75
eutectic alloys. Li has the highest vapor pressure among
them. Since the chemical activity of Li is high, its use infusion reactors may be limited for safety reasons eventhough its TBR is the largest among the four candidates. The
vapor pressure of Li17Pb83 is lower than that of Li. The
chemical activity of Li in the Li17Pb83 eutectic alloy
decreases with increasing temperature as one can see whencomparing the difference in vapor pressure between Li and
Li17Pb83, which slightly spreads with increasing temperature.
Judging from the viewpoint of vapor pressure, Flibe and
Li25Sn75 are the lower group. The chemical activity of Li islow. Since the vapor pressure of Flibe is about 10-2 Pa at 800
K, its vaporization is negligibly small even when its
accidental release happens (see Fig. 4). Therefore, Flibe is
the safest materials among the four candidates. Li25Sn75
cannot constitute any blanket design of TBR>1.

Flibe in

Flibe out
Fig. 5 Overview of He-Flibe counter-current extraction
tower
When a Flibe molten salt mixed with UF4 and TF4 isused as fuel of a molten salt fission reactor, the largest
disadvantage is the strong gamma ray originated from 232U
or its daughter nuclei 208Tl. The gamma ray energy from208
Tl is 2.8MeV. Therefore, the designed molten salt reactor
should be buried under the ground in order to decrease
non-necessary exposure to gamma rays (see Fig. 6). The
figure shows an example for the molten salt reactor cited in
the reference [19]. One generation path of 232U from 232Th
fuel is considered to be 232Th(n,2n)231Th, 231Th→231Pa,
231
Pa(n,γ)232Pa, and finally 232Pa→232U. The first (n,2n)reaction occurs only with fast neutrons. The thresholdenergy is 6.5MeV. Therefore, neutron slow-down using aresonable moderator is effective in order to decrease thegeneration of 232U. Another path for the generation of 231Th
is the neutron capture reaction of 230Th. This isotope ispresent as a natural impurity in 232Th. 230Th is a daughter

5. Thorium as Fertile Material for Molten Salt Reactor

Flibe out

Fig. 6 Conceptual design of underground molten salt fission reactor. Figure is cited from the address [19].

25S. Fukada, Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University

product of the alfa decay of 238U. When 231Th is generated,as described above, the same path works to generate 208Tl.
Therefore, it is effective to separate natural U from naturalTh (almost 100% 232Th) beforehand in order to minimize thegeneration of high intensity gamma ray. The chemical
separation between U and Th will be performed by a solvent
extraction technique using Tributyl phosphate (TBP) as
solvent.

tritium leak from fusion reactor system”, Fusion Science andTechnology, 48 (2005) 666-669.
[8] S. Fukada, A. Morisaki, “Hydrogen permeability through amixed molten salt of LiF, NaF and KF (Flinak) as aheat-transfer fluid”, Journal of Nuclear Materials, 358 (2006)
235-242.
[9] Y. Edao, S. Fukada, H. Noguchi, A. Sagara, “Tritium releasefrom neutron-irradiated Flibe purged out by Ar-H2 or Ar atelevated temperature”, Fusion Science and Technology, 55(2009) 140-151.
(iii) Design study of tritium recovery from Flibe blankets:[10] S. Fukada, M. Nishikawa, A. Sagara, “Calculation of recoveryof tritium from a Flibe blanket”, Fusion Technology, 39(2001) 1073-1077.
[11] A. Sagara, H. Yamanishi, T. Uda, et al., “Studies on Flibeblanket designs in helical reactor FFHR”, Fusion Technology,
39 (2001) 753-757.
[12] S. Fukada, M. Nishikawa, A. Sagara, T. Terai, “Mass-transport
properties to estimate rates of tritium recovery from Flibeblanket”, Fusion Science and Technology, 41 (2002)
1054-1058.
[13] A. Sagara, T. Tanaka, T. Muroga, et al., “Innovative liquidblanket design activities in Japan”, Fusion Science andTechnology, 48 (2005) 524-529.
[14] S. Fukada, A. Morisaki, A. Sagara, T. Terai, “Control of tritium
in FFHR-2 self-cooled Flibe blanket”, Fusion Engineeringand Design, 81 (2006) 477-483.
[15] A. Sagara, O. Mitarai, S. Imagawa, et al., “Conceptual design
activities and key issues on LHD-type reactor FFHR“, Fusion
Engineering and Design, 81 (2006) 2703-2719.
[16] S. Fukada, K. Katayama, T. Terai, A. Sagara, “Recovery oftritium from Flibe blanket in fusion reactor“, Fusion Scienceand Technology, 52 (2007) 677-681.
[17] A. Sagara, O. Mitarai, T. Tanaka, et al., “Optimization
activities on design studies of LHD-type reactor FFHR”,
Fusion Engineering and Design, 83 (2008) 1690-1695.
[18] S. Fukada, Y. Edao, A. Sagara, “Effects of simultaneoustransfer of heat and tritium through Li-Pb or Flibe blanket”,
Fusion Engineering and Design, 85 (2010) 1314-1319.
[19] http://energyfromthorium.com/2010/12/07/not-so-fast-withthorium/moirteller/[20] H. Yamanishi et aI., “Design studies on nuclear properties on
the Flibe blanket for helical-type fusion reactor FFHR”,
Fusion Engineering and Design, 41 (1998) 583-588.
[21] H.Yamanishi, A.Sagara, Annual Report of National Institutefor Fusion Science, April 1999–March 2000 (2000) 177.

6. Conclusions

Applications of molten salt Flibe or Flinak to fusion or
fission reactors and our previous research on these topics
were presented. Since Flibe and Flinak show not only
proper T breeding ability for fusion reactors but also lessreactivity with O2 or H2O regardless of slightly larger
viscosity than other breeding materials, they can be used asa blanket fluid of a FFHR-2 fusion reactor. The advantagesof molten salts were discussed here. On the other hand,corrosive TF acid is generated when exposed to neutrons. In
order to solve this problem, the Be redox control reaction
was shown to be effective. Since the solubility of T2 (H2) iscomparatively small, although it is larger than that of inert
gases, the T2 recovery is readily carried out with a He-Flibecounter-current extraction tower. When Flibe mixed withfertile Th is exposed to neutrons, strong gamma rays are
emitted from the Th fuel. The problem of strong gamma
radiation is mitigated by placing the whole Th fission
reactor under the ground. Impurity removal in natural Th
can decrease the strong gamma ray originated from 208Tl.

7. References(i) Redox control and Flibe purification study:[1] S. Fukada, R.A. Anderl, Y. Hatano, et al., “Initial studies oftritium behavior in Flibe and Flibe-facing material”, Fusion
Engineering and Design, 61-62 (2002) 783-788.
[2] R.A. Anderl, S. Fukada, G.R. Smolik, et al., “Deuterium/tritium
behavior in Flibe and Flibe-facing materials”, Journal of
Nuclear Materials, 329-333 (2004) 1327-1331.
[3] M.F. Simpson, G.R. Smolik, J.P. Sharpe, et al., “Quantitative
measurement of beryllium-controlled redox of hydrogen
fluoride in molten salt”, Fusion Engineering and Design, 81(2006) 541-547.
[4] D.A. Petti, G.R. Smolik, M.F. Simpson, et al., “JUPITER-II
molten salt Flibe research: An update on tritium mobilization
and redox chemistry experiments”, Fusion Engineering and
Design, 81 (2006) 1439-1449.
[5] S. Fukada, M.F. Simpson, R.A. Anderl, et al., “Reaction rate ofberyllium with fluorine ion for Flibe redox control”, Journalof Nuclear Materials, 367-370 (2007) 1190-1196.
(ii) Solubility, diffusivity, and permeability of hydrogen isotopes
through Flibe and Flinak:[6] S. Fukada, R.A. Anderl, R.J. Pawelko, et al., “Flibe-D2
permeation experiment and analysis“, Fusion Science andTechnology, 44 (2003) 410-414.
[7] S. Fukada, R.A. Anderl, T. Terai, et al., “Diffusion coefficient oftritium through a molten salt Flibe blanket and evaluation of26