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1
GIF Reactor System
Development Status:
Molten Salt Reactor
Presented by Victor Ignatiev
NRC “Kurchatov Institute”, Moscow, RF
10th GIF - INPRO IAEA Interface Meeting, Vienna, Austria, 11 April, 2016
30-03-2016
There were two people at the [Manhattan Project]
metallurgical laboratory, Harold Urey, the isotope chemist, and
Eugene Wigner, the designer of Hanford, both Nobel Prize
winners who always argued that we ought to investigate
whether chain reactors, engineering devices that produced
energy from the chain reaction, ought to be basically
mechanical engineering devices or chemical engineering
devices. And Wigner and Urey insisted that we ought to be
looking at chemical devices—that means devices in which fuel
elements were replaced by liquids.
2
Mechanical engineering device presumes that the fuel (solid) has tobe used in a max condensed form that excludes reprocessing and hasadvantage of technical simplicity while reactor operating.
Chemical engineering device has not only possibilities of generalbenefits such as unlimited burn-up, easy and relatively low cost ofpurifying and reconstituting the fuel (fluid), but also there are somemore specific potential gains.
Design MSBR DMSR
Reactor thermal power, MW 2250 2250
Overall plant efficiency, % 44 44
Fuel salt inlet/outlet, °C 566 / 704 566 / 704
Coolant salt inlet/outlet, °C 454 / 621 454 / 621
Steam conditions, MPa/ °C 24.3/ 538 24.3/ 538
Core height/diameter, m 4.0 / 4.3 8,3 / 8,3
Salt volume fraction in core, % 13 /37 20
Average core power density, MW /m3 39 5
Estimated core graphite life, years 4 30
Total fuel salt volume, m3 48.7 104
Thorium / Fissile inventory, t 68 / 1.47 140 / 2.37
Breeding ratio 1.06 0,85 (0,8)
Molten salts fluorides
were developed
originally at US ORNL
for MSR in 1970s to
reflect Gen II, but not
Gen IV objectives.
Gen II Th-U MSRs had
mainly graphite
moderated cores
4
Different reactor concepts using molten salt are discussed an GIF
MSR pSSC meetings– Molten Salt Fuelled Reactors (the circulating salt is the fuel +
coolant)
» MSR MOU Signatories France EU and Switzerland work on Th-U MSFR (Molten Salt Fast Reactor). Switzerland joined MOUin 2015.
» Russian Federation works on MOSART (Molten Salt ActinideRecycler & Transmuter) with and without Th-U support. RFjoined the MOU in 2013
» China, Japan and South Korea work on Th-U TMSR withgraphite moderator
– Molten Salt Cooled Reactors (solid fuelled )
» USA and China work on FHR (fluoride-salt-cooled high-temperature reactor) concepts and are Observers to the PSSC
» Australia works with China on materials development forMSR and FHR Australia is joining the MOU in 2016
5
Fluoride-salt-cooled high-temperaturereactors combine three technologies
• Fuel: high-temperature coated-particle fuel developed for high-temperature gas-cooled reactors (HTGRs) with failure temperatures >1650C• Coolant: high-temperature, low-pressure liquid-salt coolant (7Li2BeF4) with freezing point of 460°C and boiling point >1400C (transparent• Power Cycle: nuclear air-Brayton combined power cycle with GE 7FB compressor
Candidate FHR Demonstration (ORNL) Mk1 PB-FHR flow schematic (UCB)
6
• Purpose for CRADA is to Accelerate Development of FHRs
• CRADA supports and is funded by SINAP’s thorium MSR
program
• CRADA is limited to solid fueled MSRs
– Nearly all technology developed will be applicable to
MSRs
– CAS is providing the entirety of CRADA funding, with an
estimated $5 million a year.
– The collaborations under the new agreement are
authorized for 10 yrs.
• University lead integrated research projects ($5 M each) focused on
addressing technical issues for FHRs initiated from 2015 till to 2018
– MIT, UC-Berkeley, U-Wisconsin, and U-New Mexico form one team
– Georgia Tech, Texas A&M, and Ohio State form other team
• US-Czech collaboration on F7LiBe reactivity worth measurement is under
development
U.S. and China Have Begun Cooperating R&D on FHR (CRADA)
DOE’s Focused Investment in FHRsis Through University Research
The near-term Goal of TMSRs project :
�2MW Molten Salt Reactor with liquid fuel (∼2022)TMSR Reactor Site
Source: Zimin Dai (SINAP) 2015
CAS has initiated a TMSR development program with
similar to prior US graphite moderated cores and has
provided resources for R&D, design and construction of
MSR test reactor in China. This initial test reactor will have
a power of 10MWt. A second, 100MWt test reactor is also
planned. Both test reactors will use low-enrichment U fuel.
MOSART
MSFR
Fuel circuit MOSART (RF) MSFR (EU)
Fuel salt,
mole %
LiF-BeF2+1TRUF3
LiF-BeF2+5ThF4+1UF4
78.6LiF-12.9ThF4—3.5UF4--5TRUF3
77.5LiF-6.6ThF4-12.3UF4-3.6TRUF3
Temperature, оС 620 - 720 650 - 750
Core radius /
height, m
1.4 / 2.8 1.13 / 2.26
Core specific
power, W/cm3
130 270
Container material
in fuel circuit
Ni-Mo alloy
HN80MTY
Ni-W alloy
ЕМ 721
Removal time for
soluble FPs, yrs
1 - 3 1 - 3
• strong negative feedback coefficients• good breeding ratio• no problem of graphite life-span
• relatively high initial loading
Europe focused on liquid fuel and no solid moderator inside the core ⇒ possibility to
reach specific power higher than in a solid fuel
Fast
Spectrum
Configuration
In the Li,Be/F MOSART core without U-Th support
it is possible to burn TRUs from used LWR fuel with
MA/TRU ratio from 0.1 up to 0.45 within solubility limit
� Single fluid 2.4GWt core with the rare earth removal time 1 yr containing as
initial loading 2 mole% of ThF4 and 1.2 mole% of TRUF3, after 12 yrs can operate
without TRUF3 make up basing only on Th support
� At equilibrium molar fraction of fertile material in the fuel salt is near 6 mole %
and it is enough to support the system with CR=1
10
Component Cycle
times
Removal
operation
Kr, Xe 50 sec He Sparging
Zn,Ga,Ge,As,Se,Nb, Mo,Cd,InSn,Sb,Te,Ru, Rh,Tc
2.4 hr Plating out on surfaces +To off gas ystem
Zr
1-3 yrs
Reductiveextraction,
Oxideprecipitation,
Electrodeposition
Ni, Fe, Cr
Np, Pu, Am, Cm
Y,La,Ce,Pr,Nd,Pm,Gd,Tb,Dy,Ho,Er,Sm,Eu
Sr, Ba, Rb, Cs >30 yr
Li, Be, Na Salt discard
Reductive extraction of An’s from molten salt into liquid
bismuth with their subsequent re-extraction into purified salt
flow is the most acceptable way of An recycling
MOSART Fuel Clean up:
The grand objective of SAMOFAR is:– prove the innovative safety concepts of MSFR,– deliver breakthrough in nuclear safety and waste management– create a consortium of stakeholders to demonstrate MSFR beyond SAMOFAR
Main results will be:–experimental proof of concept–safety assessment of the MSFR–update of the conceptual MSFR– design roadmap and momentum among stakeholders
SAMOFAR Project (Started 08/2015: 4 years, Euro 5M)
“A paradigm Shift in Nuclear Reactor Safety with Molten Salt Reactor”
EU Partners: TU-Delft, CNRS, JRC, CIRTEN, IRSN, AREVA, CEA, EDF, KIT, PSI, CINVESTAV
Non EU partners: SINAP (China), Univ. of New Mexico (USA) and KI (Russia)
• Integral safety assessment
• Safety related data
• Experimental validation
• Numerical assessment
• Materials compatibility
• Salt chemistry control
• Fuel salt processing
Technical work-packages:
Collaborations: Europe
Company Spectrum Feed Processing Notes
Terrestrial
Energy
Thermal LEU Gas stripping and
mechanical filtering
Canadian company
DMSR - Replace vessel with
salt every seven years
ThorCon Power Thermal LEU Gas stripping and
mechanical filtering
DMSR - Replace vessel with
salt every seven years
Transatomic Thermal &
Epithermal
LWR TRU or Th ZrH moderator would require
significant advances in
cladding. Not apparent that
version from white paper can
maintain criticality.
FLiBe Energy Thermal Th Two Fluid MSBR Close analogy to historic MSR
program
Terra Power Fast No
enrichment
after startup
Polishing only Chloride salt
Hatch Thermal Canadian company
Waterfall design
Moltex Fast Polishing only UK company; Chloride salt
MSRs Are Currently Being Developed Under Both
Commercial Private and Government Sponsorship
RIAR Radiochemical Division
MCFR Commercial Development Roadmap Has Three Phases
Early validation
• Completed by 2019
• Supported jointly by U.S. Government and Southern Nuclear Services led consortium
Critical test reactor
• Mid 2020s
Commercial prototype
• By 2035
Contract is still under negotiation.
The MCFR core is composed of the
reactor vessel, fuel salt, neutron
reflectors, and primary heat
exchangers.
Image courtesy of TerraPower
Heat exchanger
Neutron reflectors
Vessel
Main fuel
salt inventory
14
MSRs Have Several Remaining Technology Challenges
• Ni-based alloys embrittle under high neutron fluxes at high temperature
– Refractory alloys and structural ceramic composites remain at a low
technology readiness levels
• High power density reactors challenge heat exchanger material mechanical
performance and reflector/shield material temperatures
– Minimizing ex-core fuel volume necessitates high performance heat
exchangers
– Strengthening alloy microstructures dissipate over time at temperature
• Proper chemistry control is imperative
– Alkali halide salts can be highly corrosive
– Ratio of U4+/ U3+ is key to maintaining low corrosivity
• Molten salts can generate substantial amounts of tritium
– Especially lithium bearing salts
• Fast spectrum MSR’s operate near solubility limits for actinide trifluorides to
maintain criticality
15
MSR Commercial Deployment Depends
Upon Resolving Multiple Materials IssuesMax temperature of fuel salt in
the primary circuit made of
Alloy N is mainly limited by Te
IGC under strain depending on
salt Redox potential
Element Hasteloy N
US
Hasteloy NM
US
HN80М-VI
Russia
HN80МTY
Russia
HN80МTW
Russia
MONICR
Czech Rep
EM-721
France
Ni base base 82 82 77 base 68.8
Cr 7,52 7,3 7,61 6,81 7 6,85 5.7
Mo 16,28 13,6 12,2 13,2 10 15,8 0.07
Ti 0,26 0,5─2,0 0,001 0,93 1.7 0,026 0.13
Fe 3,97 < 0,1 0,28 0,15 2,27 0.05
Mn 0,52 0,14 0,22 0,013 0,037 0.086
Nb - - 1,48 0,01 < 0,01 -
Si 0,5 < 0,01 0,040 0,040 0,13 0.065
Al 0,26 - 0,038 1,12 0,02 0.08
W 0,06 - 0,21 0,072 6 0,16 25.2
Metallic Materials for Fuel Circuit
Li,Be,Th,U/F HN80МT-VI HN80МTY
[U(IV)]/[U(III)]
500
without
loading at
735oC
K =3360pc×µm/cm; l =166µm K=1660pc×µm/cm; l=68µm
[U(IV)]/[U(III)]
500
Loading
25MPa
750oC
K =8300pc×µm/cm; l =180µm K = 1850pc×µm/cm ; l=80µm
[U(IV)]/[U(III)]
100
Loading
25MPa
750oC
no no
Alloy N Compatibility With Fuel Salts Strongly Depends on Redox Potential
Source: Ignatiev (KI) 2013
U(IV)/(UIII)
Alloy N
enlargement ×160
HN80МTY
enlargement ×160
30
without
loading at
760oC
no no
60
without
loading at
760oC
K = 3500pc×μm/cm; l = 69μm no
90
without
loading at
800oC
K = 4490pc×μm/cm; l = 148μm K = 530pc×μm/cm; l = 26μm
Te Corrosion in LiF-BeF2-UF4
Source: Ignatiev (KI) 2013
19
• Trapping tritium at the primary to intermediate heat exchangerpreserves separation of nuclear and non-nuclear portions of plant
• At MSR temperatures tritium diffuses through structural alloys
– Primary heat exchanger is a significant escape path
– Tritium release potential features prominently in the WASH-1222 report “An Evaluation of the MSBR”, USAEC, 1972
Tritium Control is Necessary for MSR Acceptability
� Refinement of geometric configuration of the intermediate heat exchangers,
minimizing tritium flux, including double wall designs
� Additional development of permeation-resistant coatings, e.g. W-Si, aluminades, etc.
� Ultrasonic degassing to facilitate removal of tritium, reducing required total bubble
volume for gas sparging
� Discovery of reusable solvents for direct tritium removal from molten salt
� The chemistry of sodium fluoroborate and the tritium trapping process
� Tritium uptake on graphite
Main strategies for mitigation include: advanced materials for the piping and heat exchangers,
inert gas sparging, additional coolant lines and metal hydride addition or chemical removal.
-1
-0,5
0
0,5
1
1,5
2
0,90 0,95 1,00 1,05 1,10 1,15 1,20 1,25 1,30
lgS
, м
ол
. %
103/Т, K-1
1
2
3 4
5
y = 0,0206x - 10,2R² = 0,9957
1
1,25
1,5
1,75
2
2,25
2,5
2,75
3
3,25
3,5
3,75
4
4,25
4,5
4,75
5
525 550 575 600 625 650 675 700 725
Раств
ор
им
ость
Pu
F3, м
ол
ьн
. %
Температура,0С
y = 0,0264x - 13,25
R2 = 0,9889
1
1,5
2
2,5
3
3,5
4
4,5
5
5,5
6
525 550 575 600 625 650 675 700 725
Температура,0С
Раств
ор
им
ость
Am
F 3, м
ол
ьн
. %
73LiF-27BeF2+AmF3
73LiF-27BeF2+PuF3
А Р Г О Н
9
2
3
4
6
78
1 1 1 2
1 0
1
5
(1) 45LiF-12NaF-43KF
(2) 78LiF-22ThF4
(3) 75LiF-5BeF2-20ThF4
(4) 58NaF-17LiF-25BeF2
(5) 66LiF-34BeF2
PuF3
local γ-spectrometry, isothermal saturation and reflectance spectroscopy
Min temperature of the fuel salt is determining not
only its melting point, but also the solubility for AnF3
in the solvent for this particular temperature
Temperature, K 72,5LiF-7ThF4-20,5UF4 78LiF-7ThF4-15UF4
PuF3 CeF3 PuF3 CeF3
873 0,35±0,02 1,5±0,1 1,45±0,7 2,6±0,1
923 4,5±0,2 2,5±0,1 5,6±0,3 3,6±0,2
973 8,4±0,4 3,7±0,2 9,5±0,5 4,8±0,3
1023 9,4±0,5 3,9±0,2 10,5±0,6 5,0±0,3
Near the melting point for 78LiF-7ThF4-15UF4 and 72.5LiF-
7ThF4-20.5UF4 salts, the CeF3 significantly displace PuF3
An and Ln Trifluorides Solubility
Temperature, K Individual Solubility, mol.% Joint Solubility, mol. %
PuF3 UF4 PuF3 UF4
823 6.1±0.6 15.3±0.8 1.16±0.06 1.75±0.09
873 11.1±1.1 24.6±1.2 2.9±0.1 3.5±0.2
923 21.3±2.1 34.8±1.7 13.2±0.6 11.0±0.6
973 32.8±3.3 44.7±2.2 19.1±1.0 17.3±0.9
1023 - - 21.0±1.1 19.0±1.0
1073 - - 22.5±1.2 20.0±1.1
LiF-NaF-KF
Up to 873K joint solubility PuF3+UF4 in Li-NaF-KF eutectics
is much less compared to individual ones for PuF3 and UF4
22
Proliferation Resistance Has Become A
Dominant Concern For All Fuel Cycles
• MSRs can be highly proliferation resistant or vulnerable
depending on the plant design
– MSR designs until the mid-1970s did not consider proliferation issues
– Several current MSR design variants do not include separation of actinide
materials
• Liquid fuel changes the barriers to materials diversion
– Lack of discrete fuel elements prevents simple accounting
– Homogenized fuel results in an undesirable isotopic ratio a few months
following initial startup (no short cycling)
– Extreme radiation environment near fuel makes changes to plant
configuration necessary for fuel diversion very difficult
– High salt melting temperature makes ad hoc salt removal technically
difficult
– Low excess reactivity prevents covert fuel diversion
Summary• MSR has flexible fuel cycle and can operate in different modes:
- MSCRs build upon prior MSR heritage
- MSFRs avoid requirement for future uranium enrichment
- TRU fuel utilizes amount of existing long lived TRU’s
• Liquid fuel inherently intimately interconnects the fuel cycle with
the reactor
• MSR fuel cycles can be highly proliferation resistant or have
substantial proliferation vulnerabilities
• Basic elements of MSR fuel cycles have been identified and
demonstrated with varying degrees of sophistication
• Significant research, development, and demonstration remains to
enable any MSR
• Historic MSR program and successful MSRE operation provides
foundational technology and proof-of-concept for future MSRs
24
Our problem is not that
our idea is a poor one –
rather it is different from
the main line, and has
too chemical a flavor to
be fully appreciated by
non-chemists.
-- Alvin Weinberg
Another perspective….