1

Overview of the GIF MSR

System Activities

Presented by Victor Ignatiev

NRC “Kurchatov Institute”, Moscow

On behalf of MSR pSSC

Ignatev_VV@nrcki.ru

12th INPRO Dialog Forum, Vienna, Austria, 14 April, 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.

• The use of fluid fuel does not mean that FPs are

more movable in matrix. Indeed, some FPs are

soluble and thus are kept in the melt. Others, that

are volatile, could been removed from fuel during

operation.

• The numbers of barriers is not a magic number.

What is of real importance is a reliability of barriers.

And if one wants to keep a number of barriers, why

shouldn't we to put an additional guard vessel for

fluid fuel reactor?

3

Sometimes the critics of the MSR say

that the use of liquid fuel means the loss

of safety barrier (fuel cladding)

4

Molten Salt Reactor Technology Has

60+-yr Development History in the U.S.

• Originally proposed by Bettis and Briant of ORNL in late 1940’s

• Aircraft Nuclear Propulsion Program (1946 – 1961)

– Aircraft Reactor Experiment (1953 – 1954)

– Aircraft Reactor Test (1954 – 1957)

• Experimental Molten Salt Fuel Power Reactor (1960)

• Molten Salt Reactor Experiment (1960 – 1969)

• Collaboration with India on MSR (1968)

• Molten Salt Breeder Experiment (1970 – 1976)

• Molten Salt Breeder Reactor (1970 – 1976)

• Denatured Molten Salt Reactor (1976-1980)

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 for MSR in

1970s to reflect Gen II,

but not Gen IV

objectives

Gen II Th-U MSRs had

mainly graphite

moderated cores

Reactor Fuel Cycle MWe yr / t (Unat+ Th)

LWR UO2 open, 3yr 3.5 closed, 3 yr 6.3

Th-met. open, 3 yr 3 closed, 3 yr 12.6

MSCR UF4 open, 30 yr 17.2 closed, 30 yr 31.2

)(/)()( tGtQt =α

LWR LWR LWR MSR MSR

Requirement No Recycle

(Once through)

Recycle U

(Re-enrichment)

Recycle U & Pu

(Full Reprocessing)

DMSR MSCR

U3O8, tons 5,395 4,342 3,411 796 565

MSR's resource extension is supported by NRC "Kurchatov Institute“ investigators of

MSCR without chemical processing, when a LWR required 5 times the resources

that a MSCR did, and a Th fueled LWR 2.5 times the resources

Q(T) – energy produced for the period t

G(t) – integral consumption of natural U and Th

Reactor

GraphiteModerator

PurifiedSalt

ChemicalProcessing

Plant

FreezePlug

Critically Safe, PassivelyCooled Dump Tanks(Emergency Cooling andShutdown)

Low accident source term withcontinuous removal of mobilefission products

Low pressure (molten saltboiling point ~1400 C)o

Low chemical reactivity

Passive cooling by dumpingfuel to cooled tanks

FuelSalt

566 Co

704 Co

Reactor

GraphiteModerator

PurifiedSalt

ChemicalProcessing

Plant

FreezePlug

Critically Safe, PassivelyCooled Dump Tanks(Emergency Cooling andShutdown)

Low accident source term withcontinuous removal of mobilefission products

Low pressure (molten saltboiling point ~1400 C)o

Low chemical reactivity

Passive cooling by dumpingfuel to cooled tanks

FuelSalt

566 Co

704 Co

MSCR without chemical processing has a high efficiency of natural U and Th resources consumption and

different safety approach that allows passive safety in large unitSource: Novikov (KI) 1990

MSR Projects & Facilities in RussiaIn Russia, the MSR program was started in the second half of 70th. These studies of the MSR

technology were mainly directed on the development of the Th-U graphite moderated concepts.

Existing facilities in Russia provide strong support to molten salt R&D

1998-2014: ISTC#1606 and ISTC#3749, Feasibility of fast MS burner of long-lived radwastes.

2007-2008: Rosatom. Analysis of perspective options for SNF management.

2009-2015: Bilateral Rosatom – Euratom projects, MARS - EVOL, PYROSMANI - SACSESS

8

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

9

Fluoride-salt-cooled reactors 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)

10

• 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 FHRs is 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 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.

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

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

Liquid fuel and no solid moderator inside the core ⇒ possibility to reach specific power much higher than in a solid fuel

Fast

Spectrum

Configuration

Heat source, W/m3

T max wall: 947 K;

T max fluid : 1192 KVelocity, m/s

r = 1,65 m

Li,Be,TRU/F MOSART Core

Core satisfies two most important

requirements:

•maximum temperature of solid

reflectors is low enough to allow it

use for suitable time

•regions of reverse or stagnant

flow are avoided

15

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:

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

MOSART safety basis

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

1.1

0 50 100 150 200 250 300

Time [sec]

rel.

P

ow

er

[fr

]

Power_th

Flux_N

Flow_Cool

550

600

650

700

750

800

850

900

0 50 100 150 200 250 300

Time [sec]

Te

mp

era

ture

[C

]

Fuel_av

Salt_out

Salt_in

Graph_av

ULOF

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

1.1

0 50 100 150 200 250 300

Time [sec]

rel.

P

ow

er

[fr

]

Power_th

Flux_N

Flow_Cool

550

600

650

700

750

800

0 50 100 150 200 250 300

Time [sec]

Te

mp

era

ture

[C

]

Fuel_av

Salt_out

Salt_in

Graph_av

ULOH

0.0

0.5

1.0

1.5

2.0

2.5

3.0

0 50 100 150 200 250 300

Time [sec]

rel.

P

ow

er

[fr

]

Power_th

Flux_N

Flow_Cool

500

550

600

650

700

750

800

850

900

0 50 100 150 200 250 300

Time [sec]T

em

pe

ratu

re

[C]

Fuel_av

Salt_out

Salt_in

Graph_av

Over-Cooling

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

0 2 4 6 8 10

Time [sec]

rel.

P

ow

er

[fr

]

Power_th

Flux_N

Flow_Cool

550

600

650

700

750

800

850

0 2 4 6 8 10

Time [sec]

Te

mp

era

ture

[C

]

Fuel_av

Salt_out

Salt_in

Graph_av

UTOP+200pcm

The MOSART reactor is expected not to be seriously challenged

by the major, unprotected transients such as ULOF, ULOH,

overcooling, or even UTOP

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

Commercial Private and Government Sponsorship

RIAR Radiochemical Division

MCFR Commercial Development Roadmap Has 3 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

20

• 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

The adoption within Gen IV of MSFR

designs have introduce new challenges

Preparative Chemistry and Salt Purification• Most suppliers of halide salts do not provide materials that can be used directly.

• The major impurities that must be removed to prevent severe corrosion of thecontainer metal are moisture/oxide contaminants.

• Once removed, these salts must be kept from atmospheric contamination byhandling and storage in sealed containers.

• During the US MSR program, a considerable effort was devoted to saltpurification by HF/H2 sparging of the molten salt. In addition to removingmoisture/oxide impurities, the purification also removes other halidecontaminants, such as chloride and sulfur.

• In our purifications the gaseous agent (HF) was in some cases replaced by solid ammonium hydrofluoride (NH4HF2, Tm ≈ 125 0C ), which is safer and more convenient in use for the removal of impurity oxide compounds from metal fluorides and for the conversion of U and Th oxides to fluorides.

MeO2 + 4NH4HF2 → MeF4 + 4NH4F ↑ + 2H2O ↑ (Tmax = 350-400 0C)

• To carry out these processes do not require expensive equipment and specialsafety measures. The purified anhydrous fluorides of metals was obtained, whichare used for the preparation of fluoride salt melts of different composition.

Production of anhydrous constituents->Melting -> Filtration -> Zone

recrystallization -> met.Th or Zr or Be treatment

22

Effect of Additions to Fuel Salt on Corrosion

The following tabulation classifies materials as those that

have been beneficial and those that have not:

Beneficial: Na, Be, Al, Cr, Li, Cu, NaH, Ca, Mo, Ti, Zr

Not beneficial: Si, Zn, Fe, AgNaI, KCl, MnO2, NiF2, FeF3, CrF2, NaCl, LiIO3

Elements of material can be oxidized by the following reactions:

Oxidation reaction: M � Mn+(F)n + ne

Reduction reaction: Ox (F)x + ne � Red(F)x-n

If the oxidation is UF4: 2UF4 + M <--> 2UF3 + MF2 (n=2)

Order that reflects the oxidation scale : Cr > Fe >> Ni > Mo > W

Because the products of oxidation of metals by fluoride

melts are quite soluble in corroding media, passivation

is precluded, and the corrosion rate depends on other

factors, including: oxidants, thermal gradients, salt flow

rate and galvanic coupling

UF4 Containing Corrosion Loop

77.5LiF-20ТhF4-2.5UF4+Te:

75LiF-20ТhF4 +5BeF2+хUF4+Te:

A simple voltammetric method was developed at ORNL for the determination of U(IV)/U(III) ratio.

The method involves the measurement of the potential difference between the equilibrium potential

of the melt, measured by an inert platinum electrode immersed in melt and voltammetric equivalent

of the standard potential of the U(IV)/U(III) couple E1/2.

For linear sweep voltammetry at a stationary electrode, the polarographic half-wave potential E1/2

corresponds to the potential on voltammogram at which the current is equal to 85.2% of the peak

current.

Additions of Alloy-N Specimens and Be to Fuel Salt

To create strongly reduction conditions in fuel

salt system, it is necessary to support low

values U(IV)/U(III)

25

Max 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

MSR Commercial Deployment Depends

Upon Resolving Multiple Materials Issues

• The experimental facility is developed to study

compatibility of Ni-based alloys under various

mechanical loads to the materials specimens with fuel

salts containing Cr3Te4 with redox potential

measurement

• LiF-BeF2-ThF4-UF4: 5 tests of 250 hrs each at fuel

salt temperature till to 750оС and [U(IV)]/[U(III)] ratio

from 0.7 to 500

• LiF-BeF2-UF4: 3 tests of 250 hrs each at fuel salt

temperature till to 800оС and [U(IV)]/[U(III)] ratio from

30 to 90

Test [UF3+UF4]]

mole %

[U(IV)]/

[U(III)]

ToC

Impurity content in the fuel salt after test, wt. %

Ni Cr Fe Cu Te

1 0.64 0.7 735 0.0034 0.0018 0.054 0.002 0.015

2 2.1 4 735 0.0041 0.0019 0.006 0.0012 0.0032

3 2.1 20 735 0.009 0.0055 0.003 0.001 0.015

4 2.0 500 735 0.26 0.024 0.051 0.019 0.013

5 2.0 100 750 0.22 0.031 0.065 0.055 0.034

U(IV)/U(III)

Alloy N exposed in salt containing LixTe and CryTex

undergoes grain IGC. The embrittlement is a function of the chemical activity of Te

associated with the telluride. Controlling the oxidation potential of the salt coupled

with the presence of Cr ions in the salt appears to be an effective means of limiting

Te embrittlement of Alloy N.

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

N- Alloys 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

R&D on Graphite Components

Country Germany Japan China

(Fangda Carbon)

China

（ICC, CAS）

Graphite NBG18 IG110 NG-CT-10 With fine particlas

Pore size >5mm 2.3mm 3.6mm <1mm

RemarksFor gas cooling

reactor

For gas cooling

reactorNo irradiation data In lab scale

31

• For areas of processing unit where there is direct

contact with the liquid metal (e.g. bismuth) traditional

structural materials such as alloys based on Fe, Co, Ni

are not suitable because of (1) increased solubility in

the liquid metal or (2) subjecting the mass transfer at

exposure system with a temperature gradient.

• Materials showed good compatibility with liquid metals

in the limited number of tests include graphite, and

refractory metals such as tungsten, rhenium,

molybdenum and tantalum. Except tantalum, these

materials are difficult to manufacture and compound.

All rapidly oxidize in air at temperatures of atmospheric

processes and require protection.

Materials for processing unit

32

• 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 its melting point and solubility for AnF3

in the solvent for this particular temperature

LiF NaF KF BeF2 ThF4 T, K А -В●10-3

Method

46.5 11.5 42 0 0 823-973 5.59 3.949 isothermal

saturation

73 0 0 27 0 825-1000 3.927 3.099 isothermal

saturation

66 0 0 34 0 800-900 3.231 3.096 γ-

spectrometry

15 58 0 27 0 825-925 3.639 2.750 γ-

spectrometry

17 58 0 25 0 800-900 3.253 2.578 γ-

spectrometry

78 0 0 0 22 873-973 2.58 1.73 γ-

spectrometry

75 0 0 5 20 873-1023 2.06 1.34 γ-

spectrometry

77 0 0 17 6 848-998 3.61 2.91 γ-

spectrometry

The data on PuF3solubility in molten salt fluorides appear to

follow a linear relationship within the experimental accuracy

of the measurements when plotted as logarithm of molar

concentration of actinide trifluoride vs. 1/T(K)

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

AnF3 and LnF3 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

36

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

37

MSR safeguards present new challenges

for the designer and the safeguards authorities

– Safeguards approaches and technology may require a major shift from current practices

• Unclear how much of the existing technology or that currently being developed can be applied

• Technology gaps must be determined

– Designers, safeguards experts, and the owner (operator) need to get involved early in order to address the challenges

•MSR technology presents the ideal opportunity to apply Safeguards-By-Design concepts

•“Difficult to safeguard” does not mean “less proliferation resistant”

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

39

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….