Conference PaperPDF Available
Transactions of the American Nuclear Society, Vol. 116, San Francisco, California, June 11–15, 2017
Recent Advancements in Liquid and Solid Molten Salt Reactors—I
Feasibility of a Breed-and-Burn Molten Salt Reactor
Michael Martin, Manuele Aufiero, Ehud Greenspan, Massimiliano Fratoni
University of California, Berkeley, Department of Nuclear Engineering, Berkeley, CA 94720-1730 USA,,,
This work addresses the feasibility and characteristics of
the breed and burn (B&B) cycle for Molten Salt Reactors.
In the general B&B cycle, natural Uranium or Thorium is
used as fuel for a reactor which breeds a sucient amount
to maintain criticality at equilibrium without reprocessing.
Interest in applying this scheme to an MSR has developed
recently [
] For the MSR B&B scheme considered here, fresh
salt containing fertile material is continually fed into the core
while salt with the composition of the homogeneous fuel is
expelled without reprocessing. Through an analysis of the
equilibrium conditions and startup requirements, the feasibility
of such a cycle operating with various salts is assessed.
For a given salt, fertile material, and power level, the
equilibrium core composition will be determined by the salt
feed/removal time constant. To find this equilibrium compo-
sition, long burnup calculations were run using a modified
version of Serpent [
] that removes all isotopes from the fuel
with a given time constant. These dumped elements were re-
placed with an adequate amount of fresh fertile salt. The total
actinide molar share of the fuel was kept constant by altering
the proportion of fertile heavy metal in the feed stream. Each
fission product replaced one of the constituents of the origi-
nal salt to keep a constant total atomic density. In all cases,
noble metals and gasses were removed with a time constant
corresponding to an in-core halflife of 30min.
Infinite Geometry
To determine the feasibility of the MSR B&B cycle for
various salts, the equilibrium burnup calculations were con-
ducted using infinite media. The salts considered are listed in
table I. The densities of the salts, from top to bottom, were
taken from [
] and [
]. For each salt, equilibrium cal-
culations were run using both a natural Uranium feed and
pure Thorium feed employing various time constants. The
lithium in the lithium-fluoride salt was enriched to 99.99%
while the chlorine in the chloride salt was enriched to
. The power density of all salts was chosen to be
. The
kin f
for each salt as a function of burnup is
shown in figure 1. Neither of the cycles using fluoride salts
were feasible breeders because at equilibrium they present a
kin f
lower than unity for all time constants explored. How-
TABLE I: Salts considered for analysis; x =1 for all FP nuclides.
Salt Molar Proportions Density (g/cc)
(NaF+[FP]Fx)KF [Actinides]F443-24-33 4.263
(LiF+[FP]Fx)[Actinides]F477.5-22.5 4.418
(NaCl+[FP]Clx)[Actinides]Cl367-33 3.107
ever, the chloride salt showed breakeven potential employing
the Uranium-Plutonium cycle which presented peak excess
reactivity at equilibrium around a burnup of 0
. This
burnup was achieved with a residence time (inverse of removal
time constant) around 9
. In terms of equilibrium
kin f
, the
Uranium-Plutonium cycle outperforms the Thorium-Uranium
cycle in all cases.
Fig. 1: Equilibrium
kin f
as a function of burnup. From top to bot-
(NaF+[FP]Fx)KF [Actinides]F4
Transactions of the American Nuclear Society, Vol. 116, San Francisco, California, June 11–15, 2017 Transactions of the American Nuclear Society, Vol. 116, San Francisco, California, June 11–15, 2017
Recent Advancements in Liquid and Solid Molten Salt Reactors—I
Eect of Actinide Density
Several heavy metal chloride molar proportions were cho-
sen to investigate the eect of actinide density on the chloride
Uranium-Plutonium cycle equilibrium
kin f
. The results are
shown in figure 2. The increased actinide molar share in-
creases the excess reactivity at equilibrium for a given time
constant. The peaks of the curves shift to slightly higher resi-
dence times with increasing actinide density.
Eect of power level
The equilibrium
kin f
curve for the chloride Uranium-
Plutonium cycle at the arbitrarily chosen power density of
is compared with the same salt/cycle at a power
level of 100
in figure 3. The peak of the 100
curve occurs at a residence time approximately three times
longer than the 300
curve. The eect of changing the
power level is thus essentially a shift in the residence time
required to achieve the same
kin f
. The peak of the 100
curve is slightly higher than the 300
curve (1.10240
vs. 1.09910) which suggests better breeding performance at
lower power levels possibly due to lower insoluble and gaseous
fission product absorption.
Fig. 2: Eect of
molar proportion. The equilibrium
kin f
stretches upward with increasing actinide density.
Fig. 3: Eect of power level. A decrease in power density slightly improves
the equilibrium kin f for a given burnup level.
Finite Geometry
Equilibrium calculations were also run using finite re-
flected cores for the chloride salt with the Uranium-Plutonium
cycle at a power density of 300
and various heavy metal
densities. The core geometry was chosen to be a cylinder of
equal height and diameter. Both lead and steel reflectors were
employed. All reflectors were 1
thick on all sides, providing
an upper bound estimate. Each equilibrium calculation was
iteratively run with dierent core radii until a peak
kef f
300pcm of unity was achieved. This was chosen here to be the
minimum critical dimension. The minimum critical radii, ini-
tial Uranium content, initial Uranium enrichment, and burnup
for the core at equilibrium for each reflector and heavy metal
content combination is presented in table II.
TABLE II: Specifications for cores critical at equilibrium employing various
reflectors and heavy metal densities. A 1:1 diameter to height ratio was used.
[Actinide]Cl_3% Radius
Initial load
Initial Enrichment
33 2.30 240 11.20 0.403
40 1.95 157 11.20 0.397
50 1.70 112 11.20 0.407
33 2.80 432 11.50 0.404
40 2.45 312 11.58 0.409
50 2.25 258 11.47 0.432
The start-up of a chloride salt B&B reactor employing the
Uranium-Plutonium cycle as described here will require the
initial feed salt to contain fissile material. Choosing enriched
Uranium as the feed material keeps the MSR B&B cycle
reprocessing-facility free. To investigate the feasibility of such
a startup process, the same modified Serpent was employed
as for the equilibrium calculations. The lower bound of the
feed enrichment was chosen to be natural Uranium. The upper
bound on the enrichment was chosen to be that of the initial
core load.
The time constant and geometry for the minimum criti-
cal dimension core with 33% [
for both types of
reflectors was used. The results are shown in figure 4. There
is no reactivity drop in the cores fed with salt of the initial
enrichment level. In the steel reflected core fed with natural
Uranium, sucient breeding has occurred at 2.08 years such
that all reactivity has been recovered. The lead reflected core
recovered from the initial reactivity drop in only 1.69 years.
These results indicate that the enrichment levels required for
the startup feed are manageable (no greater than the initial
requirement) and not long lived.
The lifetime enrichment requirements of the MSR B&B
cycle were compared with those of an AP1000 [
]. The MSR
case taken as reference was the 33% heavy metal content, steel
reflected core operating at 300
. The power normalized
SWU requirements as a function of lifetime length are shown
in figure 5. The cumulative normalized separative work re-
quired for the operation of the AP1000 remains constant at 539
Transactions of the American Nuclear Society, Vol. 116, San Francisco, California, June 11–15, 2017
Recent Advancements in Liquid and Solid Molten Salt Reactors—I
SWU/GWt yr
while that for the MSR begins high around
5SWU/GWt yr
and decrease rapidly. The MSR was
assumed to have been fed using the initial enrichment level
(11.5%) for the amount of time it took the natural Uranium
fed core (figure 4) to recover its reactivity (2.08yrs). The inter-
section of the curves occurs at 5.91 years. This represents the
minimum amount of time the MSR must be run before SWU
savings occur.
Fig. 4: Startup
kef f
curves for steel (top) and lead (bottom) reflected cores
with natural and initial enrichment Uranium feeds.
Fig. 5: Comparison of the cumulative separative work required per unit energy
by the MSR B&B cycle and AP1000. The MSR presents SWU savings after
5.91 years.
The MSR B&B scheme studied here was shown to be
feasible when chloride salts and natural Uranium fuel were
employed. The equilibrium conditions showed adequate breed-
ing performance that resulted in cores of reasonable critical
dimensions with start-up feed requirements within practical
bounds. Future studies will address the feasibility of attaining
passive safety and of practically handling the radiation damage
to structures adjacent to the core. The waste characteristics,
proliferation resistance, and optimal method of approaching
equilibrium for the MSR B&B cycle are also to be addressed.
J. LATKOWSKI, “Emerging Commercial Design Con-
cepts: TerraPower, in “Presented at the Workshop on
Molten Salt Reactor Technologiesâ˘
TCommemorating the
50th Anniversary of the Startup of the MSRE: From the
MSRE to a New Emerging Class of Reactors 50 Years
Later.”, Oak Ridge, TN. (October 14-15 2015).
L. LUZZI, and M. RICOTTI, “An extended version of
the SERPENT-2 code to investigate fuel burn-up and core
material evolution of the Molten Salt Fast Reactor,Journal
of Nuclear Materials,441,1, 473–486 (2013).
3. G. J. JANZ, Molten salts handbook, Elsevier (2013).
O. BENES and R. KONINGS, “Molten salt reactor fuel
and coolant,” Comprehensive Nuclear Materials,
, 374
CHERVINSKII, “Density, surface tension, and viscosity of
uranium trichloride-sodium chloride melts,” Soviet Atomic
Energy,39,1, 650 (1975).
NRC, “Final Safety Evaluation Report Related to Certifica-
tion of the AP1000 Standard Design (NUREG-1793, Initial
Report),” (2004).
... Breed-and-burn cycle is a special type of open fuel cycle, where substantial fuel utilization can be achieved, as outlined in the dedicated breed-and-burn chapter (Qvist, 2020). Both, homogeneous (Hombourger et al., 2015(Hombourger et al., , 2017(Hombourger et al., , 2019Latkowski, 2015;Martin et al., 2017) and heterogeneous (Kasam andShwageraus, 2016, 2017;Kasam et al., 2020;Cuoc et al., 2020). MCFR have a potential to be operated in breed-and-burn cycle. ...
... The k inf curve for liquid fuel reactor in Fig. 6 and left Fig. 7 is rather symmetric with burnup. Similar curve as in (Hombourger et al., 2019) can be found also in (Martin et al., 2017) or (Raffuzzi and Krepel, 2020). ...
... This power density is lower than the 300 W/cm 3 assumed in the calculations the results of which are reported in Figs. 6 and 7. However, Martin et al. (2017) are reporting that k inf slightly increases as the assumed power density is lower implying that it might be possible to slightly reduce the Fig. 7 reported core volumes. The actinides inventory in the respective cores is slightly less than 100 tons and 120 tons. ...
MCFRs belong to the family of liquid-fuel molten salt reactors and utilize chloride salts containing actinides as a liquid fuel. The neutronic properties of chlorine and especially of its isotope ³⁷Cl support a very hard neutron spectrum and very low parasitic neutron absorption. The resulting high neutron economy can be used for legacy waste utilization or for new fissile fuel breeding in the closed U-Pu and Th-U fuel cycles or in an open U-Pu breed-and-burn cycle. At the same time, the small scattering and capture reaction probabilities result in core transparent for neutrons with increased neutron leakage. Typical MCFR must thus rely on an efficient reflector and/or blanket and large cores. It also has implications for achievable doubling times for both breed-and-burn and closed fuel cycles. MCFRs can be generally subdivided into homogeneous and heterogeneous concepts. In the homogeneous MCFRs the fuel salt is acting simultaneously as a coolant. The core is thus filled solely by the fuel salt. The absence of dedicated coolant and of structural materials for its separation from the fuel results in minimal parasitic capture in the core. At the same time, the fuel, as a primary heat transfer medium, must be extensively pumped to an external heat exchanger. This results in increased fissile fuel inventory and in the loss of delayed neutrons in the primary circuit. The heterogeneous MCFR core consists of the fuel salt, dedicated coolant and structural material for their separation. The presence of non-fuel material in the core reduces the neutron economy and introduces challenging radiation damage problems. Since, the fuel salt does not have the heat transport function in the heterogeneous MCFR, its circulation is much milder and driven mainly by the reprocessing needs. The molten salt is an ionic liquid and it is not subjected to neutron induced radiation damage. Typical MCFR might achieve a very high uranium utilization of ~ 35% when operating in the open breed-and-burn cycle and, as other breeder reactors, above 90% when operating in a closed fuel cycle. The history and basic design characterization of the two major MCFR types are briefly described in the first half of this chapter. It also includes the rather exotic third option of directly cooled MCFRs. The second half of this article describes selected example of proposed MCFR designs.
... All these have been already discussed in detail in Section 8. • Fuel salt processing, on the other hand, poses a wide range of concerns depending on the fuel salt, processing technique, fissile material concentration, radiation intensity, and presence of other toxic compounds The most distinctive property of fuel salt preparation is that if moisture, sulfur, or free oxygen ions are present, the fuel salt becomes much more corrosive. 222 As a result, water, sulphur, and oxygen are removed from commercial-grade salts and precursor compounds during the MSR salt synthesis. ...
Molten salts as thermal energy storage (TES) materials are gaining the attention of researchers worldwide due to their attributes like low vapor pressure, non-toxic nature, low cost and flexibility, high thermal stability, wide range of applications etc. This review presents potential applications of molten salts in solar and nuclear TES and the factors influencing their performance. Ternary salts (Hitec salt, Hitec XL) are found to be best suited for concentrated solar plants due to their lower melting point and higher efficiency. Two-tank direct energy storage system is found to be more economical due to the inexpensive salts (KCl-MgCl 2), while thermoclines are found to be more thermally efficient due to the power cycles involved and the high volumetric heat capacity of the salts involved (LiF-NaF-KF). Heat storage density has been given special focus in this review and methods to increase the same in terms of salt composition changes are discussed in the paper. Methods of concatenating energy storage systems with nuclear power plants are also discussed with different types of nuclear reactors like MHTGR, PAHTR, VHTR, etc. Nanomodifications of molten salts are done to improve heat transfer properties and efficiency of the transfer. The best dopants for such modifications were found to be TiO 2 , SiO 2 , MWCNTs, etc. Future challenges for large scale deployment of molten salts viz., high volume expansion ratio, low thermal conductivity, incongruent melting , corrosion, etc., are listed and discussed. Corrosion of molten salts and its
... Previous studies, like [3] and [4], have demonstrated the feasibility of the aforementioned MSR concept with continuous refueling. In this study batch-wise refueling, which is more practical from an application perspective, could be implemented through the BBP routine. ...
Full-text available
The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size.
... This work builds on preliminary findings on the feasibility of BNB in MSRs [8][9][10]. Additionally, [11] investigated the feasibility of implementing a BNB cycle in an internally-cooled MSR in which the fuel is contained in separate tubes and cooled by another salt, based on Moltex energy's stable salt reactor concept [12]. ...
Full-text available
The operation of a reactor on an open but self-sustainable cycle without actinide separation is known as breed-and-burn. It has mostly been envisioned for use in solid-fueled fast-spectrum reactors such as sodium-cooled fast reactors. In this paper the applicability of breed-and-burn to molten salt reactors is investigated first on a cell level using a modified neutron excess method. Several candidate fuel salts are selected and their performance in a conceptual three-dimensional reactor is investigated. Chloride-fueled single-fluid breed-and-burn molten salt reactors using enriched chlorine are shown to be feasible from a neutronics and fuel cycle point of view at the cost of large fuel inventories.
A breed-and-burn molten salt reactor (BBMSR) concept is proposed to achieve high uranium utilisation in a once-through fuel cycle. By using separate fuel and coolant molten salts, the BBMSR may overcome key materials limitations of traditional breed-and-burn (B&B) and molten salt reactor designs. A central challenge in design of the BBMSR fuel is balancing the neutronic requirements for B&B operation with thermal-hydraulic requirements for safe and economically competitive reactor operation. Fuel configurations that satisfy both neutronic and thermal-hydraulic objectives were identified for 5% enriched and 20% enriched uranium feed fuel. A neutron balance method and thermal-hydraulic design algorithm were used to evaluate uranium utilisation and maximum allowable power density, respectively, for a range of configurations. B&B operation is achievable in the 5% enriched version with orders of magnitude greater uranium utilisation compared to light water reactors, but with moderately lower power density. Using 20% enriched feed fuel relaxes neutronic constraints so a wider range of fuel configurations can be considered, but there is a strong inverse correlation between power density and uranium utilisation. The fuel design study indicates the flexibility of the BBMSR concept to operate along a spectrum of modes ranging from high fuel utilisation at moderate power density using 5% enriched uranium feed fuel, to high power density and moderate utilisation using 20% uranium enrichment.
This chapter is devoted to the fuels and coolants of the molten salt reactor (MSR), which belongs to one of the six nuclear reactor concepts considered in the Generation IV initiative. The chapter is divided into ten different sections and starts with an introduction (Section 3.13.1), in which the general characteristics of the MSR are described. This is followed by a report (Section 3.13.2) on the historical developments of this reactor concept, with the main emphasis on the work performed at the Oak Ridge National Laboratory in the 1960s and 1970s. Section 3.13.3 briefly explains the current MSR concepts and their applications as either breeder designs or actinide burner concepts in the management of the long-lived actinides. Section 3.13.4 summarizes the physicochemical properties of typical fuels and coolants, based on the critical evaluation of available experimental data of the fluoride systems. This section is divided into several subsections based on the properties discussed, giving first a brief description of the structure of the fluoride liquids, followed by the description of various phase diagrams. Furthermore, the solubility of actinide fluorides in the fuel matrix is given, as well as information about the density, viscosity, heat capacity, thermal conductivity, and the vapor pressure. One of the key issues is the understanding of the fuel behavior under irradiating conditions. This is briefly discussed in Section 3.13.7. Section 3.13.8 is related to the fission products and their behavior in the fluoride fuel matrix with respect to their solubility, and Section 3.13.9 addresses the corrosion effects of the fuel salt on the structural materials.
In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.
Emerging Commercial Design Concepts: TerraPower, " in " Presented at the Workshop on Molten Salt Reactor Technologiesâ ˘ A ˇ TCommemorating the 50th Anniversary of the Startup of the MSRE: From the MSRE to a New Emerging Class of Reactors 50 Years Later
  • J Latkowski
J. LATKOWSKI, " Emerging Commercial Design Concepts: TerraPower, " in " Presented at the Workshop on Molten Salt Reactor Technologiesâ ˘ A ˇ TCommemorating the 50th Anniversary of the Startup of the MSRE: From the MSRE to a New Emerging Class of Reactors 50 Years Later. ", Oak Ridge, TN. (October 14-15 2015).
Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design (NUREG-1793
NRC, "Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design (NUREG-1793, Initial Report)," (2004).
Presented at the Workshop on Molten Salt Reactor TechnologiesâȂŤCommemorating the 50th Anniversary of the Startup of the MSRE: From the MSRE to a New Emerging Class of Reactors 50 Years Later
  • J Latkowski
J. LATKOWSKI, "Emerging Commercial Design Concepts: TerraPower," in "Presented at the Workshop on Molten Salt Reactor TechnologiesâȂŤCommemorating the 50th Anniversary of the Startup of the MSRE: From the MSRE to a New Emerging Class of Reactors 50 Years Later.", Oak Ridge, TN. (October 14-15 2015).