Thorium or Uranium fuel cycle for advanced nuclear reactors?

Wir schaffen Wissen – heute für morgen

Thorium or Uranium fuel cycle for advanced nuclear

reactors?Fuel recycling, multi-recycling, breeding and burning.

[email protected]

Laboratory for Reactor Physics and Systems Behaviour

Université de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE

15. Mai 2013

15. Mai 2013 Seite 2Seite 2Université de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE

15. Mai 2013

15. Mai 2013 Seite 3

Content

Seite 3

General introduction of actinides

Open and closed fuel cycle

Art of breeding and burning

Safety

Molten salt reactors


15. Mai 2013

15. Mai 2013 Seite 4Seite 4

Elements and their origin

Originated by: Big Bang,

Source: www.sciencegeek.net and E -K. Thielemann etal./Prog. Part. Nucl. Phys. 46 (2001) 5-22

Stellar, and Supernova nucleosynthesis.



Actinides

� Actinides, the heaviest elements in the periodic table are all unstable.

� But three of them have relatively long half-life: 235U, 238U, and 232Th

� Accordingly,they can be still found on earth.

� 235U is fissile and isnow main working horse.

� 238U and 232Th are fertileand can be transmuted, if irradiated,into fissile 239Pu and 233U.

138

139

140

141

142

143

144

145

146

147

148

149

150

151

152

153

154

CfBkCmAmPuNp U

PaTh

0.E+00

2.E+09

4.E+09

6.E+09

8.E+09

1.E+10

Hal

f l if

e (y

ears

)

neutron number

Cf

Bk

Cm

Am

Pu

Np

U

Pa

Th

Actinides half-lives in linear scale. Université de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE


Binding energy (per nucleon)

Fusion

Fission

The Liquid Drop Model:

The Coulumbic electrostatic repulsion between protons is reduced by fission.Université de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE

15. Mai 2013

� The fission of a heavy nucleus requires a total input energy to initiallyovercome the strong force (fission barrier) which holds the nucleus into a spherical or nearly spherical shape.

� This barrier is: 6.5 MeV for 232Th5.5 MeV for 238U5.3 MeV for 235U4.6 MeV for 233U4.0 MeV for 239Pu (synthetic ir reactor bred nuclides)

� Binding energy of the last neutron is released after the neutron absorption.

� The energy is: 5.1 MeV for 232Th4.9 MeV for 238U 6.4 MeV for 235U 6.6 MeV for 233U6.4 MeV for 239Pu

Seite 7Seite 7

Fission barrier & binding energy of last neutron


15. Mai 2013

Fission of 232Th, 238U, and 235U (natural Ac)

Seite 8

� 232Th + n (E=0eV) => 233Th* => 233Th + γthe binding energy is 5.1 MeV, but the barrier for fission is 6.5 MeV.232Th is not fissile but fissionable by high energy neutrons (1.4 MeV)

� 238U + n (E=0eV) => 239U* => 239U + γthe binding energy is 4.9 MeV, but the barrier for fission is 5.5 MeV.238U is not fissile but fissionable by high energy neutrons (0.6 Mev)

� 235U + n (E=0eV) => 236U* => fissionthe binding energy is 6.4 MeV, but the barrier for fission is 5.3 MeV.235U is fissile


15. Mai 2013

Fission of 235U as the only starting point

Seite 9

� 235U is the only natural fissile isotope.

� In nature it is always accompanied by 238U (238U 99.3%, 235U only 0.7%)

� It can be used in a thermal (moderated) reactorseither in natural isotopic composition or enriched.

� Strongly enriched it can be used also in fast spectrum reactors.(Enrichment is limited usually to 20% max., because of proliferation risk)

� 235U reactors as the first stage reactors can irradiate 232Th and 238U. (Th-U cycle can not be started without 235U or 239Pu from U-Pu cycle)

� 232Th and 238U are fissionable but under irradiation they are predominantly transmuted by neutron capture.


15. Mai 2013 Seite 10

Initial and secondary stages & open versus closed cycle

Uranium-Plutonium cycle

Initial stage

Open cycle

Reactor with Uranium fuel

cycle

U

Pu

U

MA

Pu

FP

ReactorUranium fuel

cycle

with Natural

orenrichedUranium

U

MAPu

FP

Secondary stage / Pu recycling

Closed cycle

Feed

Recycling


cycle

FPFeed


cycle

Recycling

MA

Partly closed cycle / Pu multi-recycling Fully closed cycle


15. Mai 2013 Seite 11

Initial and secondary stages & open versus closed cycle

Thorium-Uranium cycle

Secondary stage / Pu recyclingInitial stage

Open cycle

Reactor with Thorium fuel

cycle

Th

Pu

Th

MA

233U

FP

ReactorUranium fuel

cycle

with Natural

orenrichedUranium

U

MAPu

FP

Fully closed cyclePartly closed cycle / Pu multi-recycling

Closed cycle

Feed

Recycling


cycle

FPFeed


cycle

Recycling

MA


15. Mai 2013

Secondary fissile elements 233U and 239Pu

Seite 12

�:

233U + n (E=0eV) => 234U* => fission the binding energy is 6.6 MeV, but the barrier for fission is 4.6 MeV.233U is fissile

�:

239Pu + n (E=0eV) => 240Pu* => fissionthe binding energy is 6.4 MeV, but the barrier for fission is 4.0 MeV.239Pu is fissile

� Fissionable 232Th and 238U transmutations to fissile 233U and 239Puare called breeding or conversion.

UPaThnTh day 23392

2723391

min2223390

10

23290 → →→+

−− ββ

PuNpUnU day 23994

4.223993

min2423992

10

23892 → →→+

−− ββ



Fission probability for 233U, 235U, 239Pu, 232Th and 238U

1 .E - 0 3

1 .E - 0 2

1 .E - 0 1

1 .E + 0 0

1 .E - 0 3 1 .E - 0 1 1 .E + 0 1 1 .E + 0 3 1 .E + 0 5 1 .E + 0 7N e u t r o n e n e r g y ( e V )

No

rmal

ized

neu

tro

n f

lux

(n

/m2 /s

)

T h e r m a l s p e c t r a M S R

F a s t S p e c t r a M S R


0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

1.E-03 1.E-01 1.E+01 1.E+03 1.E+05 1.E+07Neutron energy (eV)

Fis

sion

pro

babi

lity

by in

tera

ctio

n

Pu239

U235

U233

U238

Th232


Capture probability for 233U, 235U, 239Pu, 232Th and 238U

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1


Cap

ture

pro

babi

lity

by in

tera

ctio

n Pu239

U235

U233

U238

Th232

� Fissile 233U, 235U and 239Pu transmutations by neutron captureare called parasitic neutron absorption.

� It is the main source of Minor Actinides (MA) build-up (mainly 239Pu).



Actinides chain for Th-U and U-Pu cycles

C:11.0%

C: 9.2%

RR: 100%

RR: 100%

M: 42.9%U234M: 35.5%

C:99.3%

C:81.2%

RR: 12.5%

RR: 11.0%

M: 14.2%U235M: 9.3%

RR: 12.3%

RR: 8.9%

C:18.8%

C:26.7%

M: 19.3%U236M: 10.7%

C:98.2%

C:90.3%

RR: 2.4%

RR: 2.4%

M: 4.2%Np237M: 2.7%

C:99.6%

C:92.1%

RR: 2.4%

RR: 2.1%

M: 4.7%Pu238M: 3.1%

C:94.0%

C:42.6%

RR: 2.3%

RR: 1.9%

M: 0.6%Pu239M: 0.9%

RR: 2.2%

RR: 0.8%

U237

7d

2d

Np238

Equilibrium U233 chainRR:

M:C:

total reaction rate with neutrons relative to U233 (without n,2n).

mass relative to U233. capture probability.

Light blue: thermal MSRDark blue: fast MSR

C:37.7%

C:34.2%

M: 0.3%Pu240M: 0.8%

C:99.9%

C:77.4%

RR: 0.9%

RR: 0.3%

Gatewayto Pu241Pu242Am andCm build-up

n,γβ- half-liven,2n

2.3d

U239

Np238

U238

23min

Pa231

26h

1d

Th231

U232

Pa232

α

M: 100%U233M: 100%

Pa233

Th233Th232

22min

27d



Recycling issues for Th-U and U-Pu cycles

� Th-U cycle issue:Hard gamma for instance from 212Bi, 208Tl, and 212Po.

� U-Pu cycle issue:Higher production of minor actinides,for instance Am and Cm (they emitsenergetic alfa and Cm also neutrons).It is difficult to fabricate them in solid fuel.

� =>In U-Pu cycle it is difficult to recyclethe MA, but Pu recycling is feasible.

� =>In Th-U cycle it is difficult to recycle U.Other fuel components are feasible.

� Molten Salt Reactors (MSR)with liquid fuel (no fabrication) and Pyrochemical reprocessing methods can accommodate both MA from U-Puand U from Th-U cycles much easier.


15. Mai 2013 Seite 17

Production of long-living radioactive

waste?

Limited resources

especially of uranium

235U?

Non-absolutesafety?

LWR

Open and closed cycles problematic issues

� Closed cycle is typically usedfor non-moderated reactorswith fast neutron spectrum.

� Open cycle is typically usedfor moderated reactors (LWR)with thermal neutron spectrum.


15. Mai 2013 Seite 18

Production of long-living radioactive

waste?

Closed fuelcycle withactinidesrecycling

Limited resources

especially of uranium

235U?

U-Pu fast (Th-U fast or th.)

breeder reactor

Non-absolutesafety?

Acceptable safety level ofbreeder with

closedcycle?

� Recycling of spent fuel (or reprocessing) includes two nontrivial steps: partitioning and fabrication !

Open and closed cycles problematic issues



Burning, breeding, transmutation

� Burning is equivalent to transmutation by fission and representssplitting of an actinide atom into two Fission Products (FP) atoms.

� Breeding, refers to a transmutation of fissionable to fissile nuclide.

� Transmutation generally means change of neutron or proton number of a given Ac or FP isotope by neutron irradiation and possible consequential β- decay.

� Major objective for transmutation are Minor Actinides (MA), which represent long-living radioactive waste.

� Transmutation creates less (shorter) radioactive isotopes (FP)or more fissile MA isotopes, which are later burned.

� Each reactor tends to a certain equilibrium fuel composition.If given fuel component should be transmuted, its share must be above the respective equilibrium share.

� Higher MA load can degrade the safety-related reactor parameters.



Art of breeding – trivial neutron economy

� In ideal reactor fissile 233U or 239Pu acts as fuel (or actually as catalysers), fertile 232Th or 238U are transmuted (or actually burned).

� The transmutation rate is equal or higher than the fission rate (self-sustaining).

� Trivial neutron economy: ideal reactor needs from each fission:1 neutron to maintain the fission chain reaction and another 1 neutron to breed new fissile isotope from the fertile one.

� Accordingly, the art of breeding is equivalent to good neutron economy.

Breeding �� Neutron economy (reactivity)

� In fast spectrum:233U produces in average 2.5 neutrons per fission.239Pu produces in average 2.9 neutrons per fission.

� Thus there is by 80% more neutrons to be “wasted”in the 238U-239Pu cycle (0.9 n) than in the 232Th-233U cycle (0.5 n).

� Therefore, the neutron economy of Thorium breeder needs to be much better.Université de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE


η - fission probability X neutrons originated per fission

� Product of fission probability and number of neutrons originated per fission illustrate how many neutrons are available in the reactor.

� Assuming the trivial neutron economy we need: 1 neutron to maintain the fission chain reaction and another 1 neutron to breed new fissile isotope from the fertile one.

0

0.3

0.6

0.9

1.2

1.5

1.8

2.1

2.4

2.7

3


ΞΞ ΞΞ (n

umbe

r of n

ew n

eutro

ns p

rodu

ced

by in

tera

ctio

n w

ith n

eutro

n)

Pu239

U235U233

U238Th232

2 necessary neutrons


η


Art of breeding – less trivial neutron economy

2.5152.5342.9212.795Neutrons originated per fission

232Th-233U232Th-233U238U-239Pu238U-239PuCycle

Molten Salt

Reactor(BOC)

Lead Cooled Reactor(EOC)


Sodium Cooled Reactor(BOC)

Reactor in equilibrium stateBOC = beginning of cycleEOC = end of cycle

1.9911.9401.8491.869Useful neutron consumption:

0.1070.0860.2190.199Fission of other AC

0.8780.8930.6920.669Fission of 239Pu or 233U

0.0150.0210.0890.132Fission of 238U or 232Th

1.0001.0001.0001.000Next fission

0.9910.9400.8490.869Fertile capture of 238U or 232Th

0.5240.5941.0720.927Neutron “ to be wasted ” :

0.3020.3930.7530.561Total parasitic capture

0.2220.2010.3190.366Leaking from the core

0.0330.1180.1400.071Parasitic capture of structural

0.0000.0950.1590.086Parasitic capture of FPs

0.1810.0970.2480.210Parasitic capture of other Ac

0.0890.0830.2050.194Parasitic capture of 239Pu or 233U

Following numbers refer to fully closed iso-breeding cycle



Thorium breeding capability in fast reactor is lower

2.5152.5342.9212.795Neutrons originated per fission

232Th-233U232Th-233U238U-239Pu238U-239PuCycle

Molten Salt

Reactor(BOC)



Sodium Cooled Reactor(BOC)

Reactor in equilibrium stateBOC = beginning of cycleEOC = end of cycle

1.9911.9401.8491.869Useful neutron consumption:

0.9910.9400.8490.869Fertile capture of 238U or 232Th

0.5240.5941.0720.927Neutron “ to be wasted ” :

0.3020.3930.7530.561Total parasitic capture

0.2220.2010.3190.366Leaking from the core

� In LFR (Lead cooled Fast Reactor) to achieve the operational iso-breedingconfiguration in the 232Th-233U cycle, it was necessary to reduce the leakage from 0.319 to 0.201.

� Accordingly: the 238U-239Pu cycle core is 120cm high and the 232Th-233U cycle core is 170cm high (40% higher to reduce leakage)

� Parasitic capture in 232Th-233U cycle is lower (TRU build-up and FP capture).

� So it works, but the neutron economy is more tight.

� However, Thorium is the only capable isotope to breed in thermal spectraUniversité de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE


Art of burning – transmutation

� In burner (transmuter) fertile 232Th or 238U should be avoided.

� They can be replaced by inert matrix (Doppler effect reduction).

� Should one of them be necessary, 232Th is better for transmutation.

� It originates 233U, which has higher fission probability and therefore induces lower amount of higher isotopes.

� If LWR MA should be burned, their mass should be above equilibrium MA content for given reactor (fertile material). This may deteriorate reactor’s safety characteristics.

� Criticality safety may be improved by external neutron source: for instance by the accelerator driven neutron spalation source. 0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1


Cap

ture

pro

babi

lity

by in

tera

ctio

n

Pu239

U233

U238

Th232



(Nuclear) Safety & MSR

'Safety' is the achievement of proper operating conditions, prevention of accidents and mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation hazards.http://www-ns.iaea.org/standards/concepts-terms.asp

Hazards/Risks:

1) Potential risks(in the absence of all protective measures)

2) Residual risks (remain despite provisions made to prevent accidents)

3) Acceptable risks (CH case: after Fukushima it may be difficult to convince public that the residual risk of wide-ranging land contamination is acceptablebecause of its very low frequency - what does it mean 10-5 or 10-6?)

http://www.iaea.org/ns/tutorials/regcontrol/assess/assess3212.htmUniversité de Genève, SÉMINAIRE DE PHYSIQUE CORPUSCULAIRE


Risk = probability X consequences (Farmers curve)

Options to improve the position of solid-fuel reactors :

Decreasing installed power or fuel burn-up(small modular reactors)

Adding barriers or decreasing the frequency of their fai lure(passive systems, TRISO particle fuel, underground reactors)

http://www.iaea.org/ns/tutorials/regcontrol/assess/assess3212.htm

Farmer curve illustration (log-log scale)

Solid-fuel reactors



Risk = probability X consequences (Farmers curve)

Online removal of highly mobile gaseous and volatile FP(Ac and remaining FP are embedded in the salt)

Lower consequences may allow reduction of classical b arriers(safety standards for reactor X reprocessing plant )

http://www.iaea.org/ns/tutorials/regcontrol/assess/assess3212.htm

Farmer curve illustration (log-log scale)

Liquid-fuel reactors:


15. Mai 2013

1.E-03

1.E-02

1.E-01

1.E+00

1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07Neutron energy (eV)

No

rma

lize

d n

eu

tro

n f

lux

(n

/m2 /s

)

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

Cro

ss-s

ectio

n X

S (b

arn)

Thermal spectra MSRFast Spectra MSRTh232 capture XSU233 fission XS

Seite 28

� Thermal reactor needs coolant (as moderator), its removal stops the chain reaction

� Fast reactor doesn’ t need coolant and its removal from the core is dangerous(positive void effect).

� Doppler effect (XS resonances broadening) depends on the Fissile/Fertile ratio.

Criticality safety (two basic parameters)

Coolant removal safe,

but MA recycling

impossible

Coolant removal danger,

but MA recycling

possible

Doppler effect

(resonances

broadening)



Source term and main accident types comparison

Barriers

Accidents driven by fission chain reaction in original core geometry

Criticality accidents

Decay heat removal

accidents

Recriticality accidents

Source term

Highly mobile gaseous and volatile FP

Accidents driven by failure ofdecay heat removal

Accidents driven by fission chain reaction in compacted (melted) core geometry

Level: high - low



Source term


Decay heat removal

accidents


Source term

SFR

Decay heat removal

accidents

LWR ADS


Source term

Decay heat removal

accidents

Recriticality accidents(fast ADS)




Source term

Storage of immobilized FP

On line removal of gaseous and

volatile FP

Decay heat removal

accidents

Source term

Decay heat removal

accidents

LWR spent fuel short-term storage MSR


(fast MSR?)



MSR can be also

designed as an

Accelerator dri

ven system



?

Historical Overview of MSR

1950s • Aircraft Reactor Experiment (ARE) *

1960s • Molten Salt Reactor Experiment (MSRE) *

1970s • Molten Salt Breeder Reactor (MSBR) *

1980s • Denatured Molten Salt Reactor (DMSR) *

1990s • Accelerator-driven transmutation of Nuclear Waste (ATW) **

2000s • Generation IV, Amster, Sphinx, Tier, Fuji…

2010s • MSFR, Mosart…

Future • Th-U breeding, TRU transmutation, H2 or process heat production… ?

* ORNL, ** LANL



Molten salt features and related advantages

Pure molten fluoride salts are transparent,

e.g. LiF-BeF 2. This feature may disappears with

addition of actinides.

Low vapor pressure of fluoride salts (reduced stresses on vessel and piping)

High-temperature operation(potential for hydrogen production)

Chemically inert substance (no fire or explosion hazard )

Freezing is inherent and passive (dispersion and freezing after leakage)

Radiation resistant substance (unlimited use only with purification)

Fuel is molten and in liquid state (it can be drained, no melting accidents)

Categories :Safety

Economics

Sustainability

Solubility of the Ac and FP (limited risk of radiotoxicity release )

Source: Ch. Forsberg ORNL

freezing < 600°C 1400°C < boi lin g



Neutronics characteristics and related advantages

Low absorption materials & no cladding

Categories :Safety

Economics

Sustainability

Online refueling and reprocessing

Flexible fuel composition(without blending and fabrication,

enables actinides recycling)

Low fuel load(low excess reactivity)

Fuel presence in salt (negative thermal feedback coefficient)

Low source term(low radiotoxic risk)

Available “excess” neutrons(thorium breeding and/or actinide burning, fixed fuel cost)

Online criticality maintenance (high availability)

Excellent neutron economy



MSR drawbacks

� Structural materials corrosion and irradiation embrittleme nt.(high Ni content alloys should be applied and redox potential must be controlled to prevent corrosion, furthermore in fast MSR the alloys suffer from radiation embrittlement.)

� Graphite, if applied, has limited lifespan.(its mechanical stability also suffers by fast neutron irradiation)

� Complicated molten salt reprocessing techniques. (fluoride volatilization techniques, Electro-separation processes, Molten salt / liquid metal reductive extraction)

� Fuel salt chemical treatment and possible proliferatio n risk. (redox potential control, on-line refueling, He bubbling to remove gaseous and volatile fission products, proliferation risk related to 233Pa or 233U relatively easy separation)


15. Mai 2013 Seite 36

MSR Dynamic features

III. Fission energy is predominantly released directly into the salt.

I. Delayed neutrons precursors areconvected (drifted) by the fuel flow.

720709699689679668658648638627617607597586576566

R0R

19

18

17

16

15

14

13

12

11

10

9

8

7

6

5

4

3

2

1

20

Temperaturedistribution inone graphite

channel

Radius

Node

II. Moderator is cooled by the liquid fuel.


15. Mai 2013 Seite 37

MSR rough classification

1) According to neutron spectra (thermal, epithermal, fast)

2) According to the purpose(burner, breeder, or both at once)

3) According to the leakage utilization (core zones and fluids number)

Physical number ofdifferent salts / loops

Single fluid Two fluids

Separation of fissile and fertile fluids / regions

Single fluid Virtual One and a half fluid

One and a half fluid

Two fluids

Fertile salts placement In core only or no fertile

In core only In core and blanket In blanket only

Core design One region Two regions One regions One regions

Blanket or similar region No Yes Yes Yes

Proliferation risk Lower Lower Higher Higher

Example MOSART (without blankets)

or MSRE

MSBR MSFR MOSART(with blankets)


15. Mai 2013 Seite 38

Novelty of the study:

� Equilibrium closed cycle is simulated by ERANOS based EQL3Dprocedure (modified at POLIMI by C. Fiorina for MSR) for full core geometry and each salt-graphite ratio.

� Salt and graphite feedback coefficients are evaluated for classical and inverted channel geometry.

� Criticality conditions are obtained, if possible, by adjusting the core radius.

� Salt composition was adopted from the European FP7 project EVOL (77.5 molar % of LiFenriched in 7Li (99.999 at%) and 22.5 molar % of Ac)

Single-fluid MSR fuel cycle parametric study


15. Mai 2013 Seite 39

H=1.892R

50 cm

50 cm

R+50cmR

Hastelloy

Salt & Graphite

Single-fluid MSR fuel cycle parametric study

Fuel-to-moderator ratio from 5% to 100% for classical and inverted channel geometries

X % = Salt volumetric ratio. Graphite Salt

80% 70% 60% 50% 40%

30% 20% 15% 10% 5%

80% 70% 60% 50% 40%

30% 20% 15% 10% 5% Inverted channel layout

Classical channel layout

22% 13% 37% 100% 90%

MSRE MSBR fiss. MSBR fert. MSFR & MOSART

Core geometry as a function of its radius R, which was used to obtain criticality, if possible.


15. Mai 2013 Seite 40

LCF

Pkeff ++

=

Neutron balance equation:

Assuming k eff =1 and P=νF:

Reaction rates ratios

F

L

F

C +=−1ν

Capture break-down

Infinite multiplication factor

Results: criticality and reaction rates

0.98

1.00

1.02

1.04

1.06

1.08

1.10

1.12

0 10 20 30 40 50 60 70 80 90 100Salt share in the core (%)

k inf

Eranos: kinf classical channels Eranos: kinf inverted channelsSerpent: kinf classical channelsSerpent: kinf inverted channels

1.0

1.1

1.2

1.3

1.4

1.5

1.6


Rea

ctio

n ra

te ra

tios

-0.1

0.0

0.1

0.2

0.3

0.4

0.5

Rea

ctio

n ra

te ra

tios

ν - 1C/FL/F (sec. axis) from EranosL/F (sec. axis) from Eq. 2

0.00

0.03

0.06

0.09

0.12

0.15

0.18


Cap

ture

to to

tal f

issi

on ra

tio

0.91

0.94

0.97

1.00

1.03

1.06

1.09

Cap

ture

to to

tal f

issi

on ra

tio

233U234U235-8UPuMASaltGraphite232Th (sec. axis)


15. Mai 2013 Seite 41

Breedingarea

Burningarea

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14


233 U

/232 T

h m

ass

ratio

0.00

0.02

0.04

0.06

0.08

0.10

0.12

0.14

σc,

Th2

32 /

(σc,

U23

3+σ

f,U23

3)

Mass ratio Classical channelMass ratio Inverted channelMicro. XS ratio Classical channelMicro. XS ratio Inverted channel

Results: masses and dimensions

233U/232Th iso-breeding mass ratio

Core radius & salt volume Absolute masses

Other mass ratios

0

1

2

3

4

5

6


Equ

ilibriu

m c

ore

radi

us (m

)

0

11

22

33

44

55

66

Sal

t vol

ume

in th

e co

re (m

3 )

Radius: classical channelsRadius: inverted channelsVolume classical channelsVolume inverted channels

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7


Mas

s re

lativ

e to

23

3 U

0.000

0.001

0.002

0.003

Mas

s re

lativ

e to

23

3 U

234U 235U-238U PuMA233Pa232U (sec. axis)

0

2

4

6

8


Mas

s (to

ns o

r kg)

0

45

90

135

180

Mas

s (to

ns o

r kg)

232U (prim. axis in kg)233U (prim. axis in tons)234U (prim. axis in tons)235-8U (prim. axis in tons)Pu (prim. axis in tons)MA (prim. axis in tons)232Th (sec. axis in tons)233Pa (sec. axis in kg)


15. Mai 2013 Seite 42

Results: partial effects and safety-related coeff.

Graph. density and temp. effectsSalt density and temp. effects

Total estimated effect

inverted classical Assumptions: 1) graphite expansion coefficient

is not zero2) the expansion is not restricted

by e.g. the reactor vessel

-10

-7.5

-5

-2.5

0

2.5

5


Sal

t rea

ctiv

ity e

ffect

s (p

cm/K

)

Salt ρ Eranos-inv.Salt ρ Eranos-class.Salt T. Eranos-inv.Salt T. Eranos-class.

Cell calculation-1.5

-1

-0.5

0

0.5

1

1.5

2

2.5


Gra

phite

reac

tivity

effe

cts

(pcm

/K)

Graph. ρ Eranos-inv.Graph. ρ Eranos-class.Graph. T. Eranos-inv.Graph. T. Eranos-class.

Cell calculation

-10

-7.5

-5

-2.5

0

2.5

5


Est

. tot

al fe

edba

ck c

oeff.

(pcm

/K)

Salt tot. Eranos-inv.Salt tot. Eranos-class.Graph. tot. Eranos-inv.Graph. tot. Eranos-class.

Estimate for infinite lattice


15. Mai 2013 Seite 43

MSR designing = separation techniques selection

1) Fluoride volatilization techniques (Fluorination)a) volatilization and fraction distillation

of solid spent fuelb) volatilization of the molten salt mixture

2) Electro-separation processes

3) Molten salt / liquid metal reductive extraction

4) Gaseous and volatile FP removal by He bubbling


15. Mai 2013 Seite 44

New idea: “Recycling without reprocessing”

� The aim of almost all classical reprocessing schemes isto remove FP and keep everything else.

� Unfortunately, FP can be usually removed as one of the last components (...and it may require all previously named techniques).

� Disadvantage, at least form the classical point of view, is that the Uranium is the first to be removed by the volatilization technique.

� Lets make from the disadvantage an advantage!

� Lets recycle just uranium and throw away the rest…!?

� For that we need “cheep” once through carrier salt with reasonable melting point, Ac solubility, and neutronic economy.(probably without enriched 7Li, e.g. NaF-BeF2 ).

� Generalized idea: lets try to use the reactor physic “freedom”of MSR designing to help to solve the chemistry and materials drawbacks.


15. Mai 2013 Seite 45

Cheeps salt? Will it be neutronically possible?

“Cheep” once through salt may possibly be NaF-BeF 2(with interesting eutectic melting point 486°C at NaF-BeF2-UF4 73-17-20)

Once through salt will surely absorb more neutrons.

Criticality maintenance? (through fuel addition: to be simulated)

Using LWR Pu as the start-up fuel?(and for fuel addition: to be simulated)

Other questions: How long will be the waste radiotoxic? It will consist mainly from FP. What is a reference radiotoxicity level?Carrier salt compatibility with possible final treatment, for instance vitrification?


15. Mai 2013 Seite 46

For instance for two-fluid fast MSR

TRU&FP& carrier salt

Newcarrier salt

each 5-10y

each 5-10y

F2

UF6

F2

UF6

each 5-10y

each 5-10y

New idea: “Recycling without reprocessing”


15. Mai 2013 Seite 47

0

0.2

0.4

0.6

0.8

1

0 2 4 6 8 10 12 14 16 18 20Time (EFPY)

Mas

s re

lativ

e to

U23

3 (%

)

NP237PU238PU239PU240PU241PU242

0

0.2

0.4

0.6

0.8

1

0 2 4 6 8 10 12 14 16 18 20Time (EFPY)

Mas

s re

lativ

e to

U23

3 (%

)

NP237PU238PU239PU240PU241PU242

Can we throw out the salt with TRU?

Probably yes. Thanks to the Th-U cycleproperties the TRU mass seems acceptable.

Fast MSR

U recyclingby volatilization

Thermal MSR

β-

Equilibrium U233 chainRR:

M:C:

total reaction rate with neutrons relative to U233 (without n,2n).

mass relative to U233. capture probability.

Light blue: thermal MSRDark blue: fast MSR

M: 100%U233M: 100%

C:11.0%

C: 9.2%

M: 42.9%U234M: 35.5%

C:99.3%

C:81.2%

M: 14.2%U235M: 9.3%

C:18.8%

C:26.7%

M: 19.3%U236M: 10.7%

C:98.2%

C:90.3%

M: 4.2%Np237M: 2.7%

C:99.6%

C:92.1%

M: 4.7%Pu238M: 3.1%

C:94.0%

C:42.6%

M: 0.6%Pu239M: 0.9%

C:37.7%

C:34.2%

M: 0.3%Pu240M: 0.8%

C:99.9%

C:77.4%


Seite 48Seite 48

� Thorium and Uranium are the only primordial Actinides left on Earth

� 238U and 232Th can be used to breed fissile 239Pu and 233U.

� 238U- 239Pu cycle is more efficient for breeding, but only in fast spectra.

�238U- 239Pu cycle runs, with low breeding, in current LWR, since it is in thermal

spectrum, amount of induced MA, or actually waste, is relatively high.

� 232Th- 233U cycle is less efficient for breeding, but can operate

in both fast and thermal spectra and produces less MA, or actually less waste.

� Breeders are usually fast reactors and criticality safety may be not solved yet.

� Molten salt reactors can profit from several appealing features based on:

fuel liquid state and 232Th- 233U cycle.

� New idea recycling without reprocessing:

the uranium can be easily recycled in case of the once-through salt.

Summary & Conclusions


15. Mai 2013

Seite 49Seite 49

Thank you


15. Mai 2013


Two very different schools of reactor design have emerged since the first reactors were built. One approach, exemplified by solid fuel reactors, holds that a reactor is basically a mechanical plant; the ultimate rationalization is to be sought in simplifying the heat transfer machinery.

The other approach, exemplified by liquid fuel reac tors, holds that a reactor is basically a chemical plant; the ultimate rationalization is to be sought in simplif ying the handling and reprocessing of fuel.

R.C. Briant & Alvin Weinberg, “Molten Fluorides as Power Reactor Fuels,” Nuc. Sci. Eng, 2, 797-803 (1957).


Thorium or Uranium fuel cycle for advanced nuclear reactors?

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