1

GIF Reactor System

Development Status:

Molten Salt Reactor

Presented by Victor Ignatiev

NRC “Kurchatov Institute”, Moscow, RF

Ignatev_VV@nrcki.ru

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.

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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)

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• 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

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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

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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

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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

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• 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

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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….