Jiří Křepel :: FAST reactor group :: Paul Scherrer Institut Fuel cycle …samofar.eu/wp-content/uploads/2017/07/Krepel_Fuel-cycle... · 2017-07-03 · WIR SCHAFFEN WISSEN – HEUTE

WIR SCHAFFEN WISSEN – HEUTE FÜR MORGEN

Fuel cycle aspects of MSR Jiří Křepel :: FAST reactor group :: Paul Scherrer Institut

MSR Summer school, July 2-4, 2017, Lecco (Como Lake), Italy [email protected]

Fuel cycle aspects of MSR, J. Krepel, MSR Summer school, July 2-4, 2017, Lecco (Como Lake), Italy

– process chain to obtain energy

Page 2

What is fuel cycle?

Example of closed “fuel cycle”. The actual waste is He and higher elements.

https://www.euronuclear.org/info/encyclopedia/n/nuclear-fuel-cycle.htm

Back end: o Interim storage o Transportation o Reprocessing o Partitioning o Transmutation o Waste disposal

This presentation covers the reactor physics aspects of irradiation and recycling.

Nuclear Fuel cycle Front end:

o Exploration o Mining o Milling o Conversion o Enrichment o Fabrication








=> Elements and their origin

Page 3

Nuclear fuel (resources)

Originated by: Big Bang, Stellar, and Supernova nucleosynthesis.

Periodic table of elements

http://www.sciencegeek.net/tables/tables.shtml

http://www.sciencegeek.net/tables/tables.shtml


The Liquid Drop Model:

The Coulombic electrostatic repulsion is a barrier for fusion.

Page 4

“Nuclear” Energy and Nuclear forces

Fusion

Fission

The reduced Coulombic electrostatic repulsion “drives” the fission. http://www.nuclear-power.net/nuclear-power/fission/liquid-drop-model/

Actinides

http://www.nuclear-power.net/nuclear-power/fission/liquid-drop-model/


=> are all unstable

1.E+00

1.E+02

1.E+04

1.E+06

1.E+08

1.E+10

138

139

140

141

142

143

144

145

146

147

148

149

150

151

152

153

154

Cf 98Bk 97Cm 96Am 95Pu 94Np 93U 92Pa 91Th 90 0.E+00

2.E+09

4.E+09

6.E+09

8.E+09

1.E+10

138

139

140

141

142

143

144

145

146

147

148

149

150

151

152

153

154

Cf 98Bk 97Cm 96Am 95Pu 94Np 93U 92Pa 91Th 90

Page 5

Actinides as nuclear fuel Actinides, the heaviest primordial elements in the periodic table, are all unstable.

But three of them have relatively long half-life: 235U: 0.7 x109 years 238U: 4.5 x109 years 232Th: 14 x109 years

Accordingly: they are still present in nature.

For 1 kg of 238U there are 3-4 kg of 232Th and 7.2 g of 235U on the Earth.

235U is the only fissile nuclide and its reserves are the smallest. Actinides half-life in logarithmic scale. Actinides half-life in linear scale.


– Oklo reactor Natural uranium evolution

0

5

10

15

20

25

30

35

-5000 -4000 -3000 -2000 -1000 0

Ura

nium

235

con

tent

(%)

Time (million years)

Oklo reactor appearance

https://en.wikipedia.org/wiki/Oklo

① Nuclear reactor zones, ② Sandstone, ③ Uranium ore layer, ④ Granite

Page 6

Earth origin

235U and 238U half-lifes differ. Accordingly the 235U content in natural uranium is evolving.

1.7 billions years ago it enabled water moderated natural nuclear fission reactor in Oklo (Africa).

Why not earlier? Several Dissolution–Precipitation cycles were necessary for the geological concentration of uranium.

What about fast reactor? What if the geological concentration in the earth outer core was faster? If so, it may be still running on U-Pu cycle.

Nowadays there is only 0.72% of 235U in natural uranium.


= maximal resources utilization

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+07Fi

ssio

n pr

obab

ility

by

inte

ract

ion

Neutron energy (eV)

U235

U238

Th232

Page 7

Sustainability

235U is the only primordial fissile nuclide and it is now the main working horse.

232Th and 238U are fissionable by fast neutrons (238U up to 5x more than 232Th).

Both of them are mainly capturing neutrons, what leads to their transmutation.

Reserves of actinides on the earth are not renewable. Aim: their max. utilization!

235U

238U

232Th

Data from JEFF 3.1 library

Relative size (surface) of actinides reserves.


is low

Page 8

Sustainability of initial 235U fueled reactors

CANDU PWR, BWR SFR, LFR, GFR Burn-up of the fuel in FIMA* % (GWd/thm): 0.65-0.75 (6.5-7.5) 3.3-5.0 (33-50) 10-15 (100-150) Burn-up in % of the original mass of natural uranium: 0.65-0.75 0.33-0.5 0.24-0.36 Sustainability? Any 235U fueled reactor has low sustainability (not even 1% of natural uranium is utilized)

* FIMA = FIssion MAterial = actinides = heavy metals


= 238U and 232Th catalytic burning

Page 9

Sustainability

Fertile fuel

Chain reaction

Catalyzer

One neutron is needed for next fission.

One of the new neutrons may be also captured by fertile 238U or 232Th.

Then they will be transmuted to fissile 239Pu or 233U.

This transmutation is also called conversion or breeding.

239Pu or 233U may actually act as an intermediary (catalyzer)

and 238U or 232Th indirectly as a fuel.

Very tight neutron economy.

neutron losses

Image modified from: www.oxfordreference.com/media/images


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+07

Fiss

ion

prob

abili

ty b

y in

tera

ctio

n

Neutron energy (eV)

Pu239

U235

U233

U238

Th232

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+07

Fiss

ion

prob

abili

ty b

y in

tera

ctio

n

Neutron energy (eV)

Pu239

U235

U233

U233 (ENDF/B-VII.0)

U238

Th232

Page 10

233U and 239Pu: synthetic (secondary) fissile elements Transmutation of fertile 232Th and 238U create fissile 233U and 239Pu:

In general, transmutation which increases fission probability is called Breeding.

(BTW: burning = fission)

High fission probability up to 90% is the biggest advantage of 233U. (for 239Pu it is 60-75%)

(JEFF 3.1 X ENDF/B VII.0)

UPaThnTh day 23392

2723391

min2223390

10

23290 → →→+

−− ββ

PuNpUnU day 23994

4.223993

min2423992

10

23892 → →→+

−− ββ



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+07

Fiss

ion

prob

abili

ty b

y in

tera

ctio

n

Neutron energy (eV)

Pu239Pu240U235U236U233U234

Transmutation of fissile 233U, 235U, and 239Pu create fertile 234U, 234U, and 240Pu:

When fissile nuclide captures neutron the products is typically fertile, thus it is called: Parasitic capture.

The secondary fertile element needs to absorb one additional neutron to became fissile!

Page 11

234U, 236U, and 234Pu: secondary fertile elements

UnU 23492

10

23392 →+

PunPu 24094

10

23994 →+

UnU 23692

10

23592 →+



(fission barrier X binding energy) There exist pairing effect described even by the Liquid Drop Model:

where: (or actually )

Hence the interacting neutron brings different binding energy to each nuclide.

Fission: binding energy > fission barrier. However, with growing nucleon number the barrier is decreasing => yes or no is not black and white.

Page 12

Fissile or fertile?

Un 23692

10 →+Un 234

9210 →+ Un 235

9210 →+

142 (even) no

143 (odd) yes

144 (even) no

Neutron nr.: 142 (even) Fissile: no

143 (odd) yes (poorly)

142 (even) no

143 (odd) yes

Pa23391

min22 →−βTh232

90 Thn 23390

10 →+ Uday 233

9227 →

−βNuclide:

U23392


Page 13

Uranium and Thorium fuel cycles

Cycle label: U-Pu Th-U

Main fertile: 238U 232Th

Main fissile: 239Pu 233U

Secondary fertile: 240Pu 234U

Secondary fissile: 241Pu (β-) 235U

Tertiary fertile: 242Pu or 241Am 236U

Tertiary fissile: 243Pu (β-) or 242Am 237U (β-)

4th fertile: 244Pu or 243Am 237Np or 238Pu

4th fissile: 245Cm or 244Am (β-) 239Pu

Th-U cycle produces less Minor Actinide - MA (Am, Cm, Np (from 235U) etc). It is based on 239Pu position. It has implication on the waste radiotoxicity.

Half-life

4.5e9

2.4e4

6500

14

3.7e5 or 432

Half-life

14e9

1.6e5

2.5e5

7.0e8

2.3e7


Page 14

233Pa and 239Np: intermediate products Transmutation of fertile 232Th and 238U goes through fertile 233Pa and 239Np:

It may happen that 233Pa and 239Np capture neutron:

The capture probability depends on cross-section and number of atoms N.

After some time equilibrium will establish where the 232Th and 238U capture rates (CR) and the 233Pa and 239Np decay rates (λN) are equal: CR= λN.

Based on the different decay constants λ , there will be 11x more 233Pa than 239Np in the core with the same transmutation rate.

UPaThnTh day 23392

2723391

min2223390

10

23290 → →→+

−− ββ

PuNpUnU day 23994

4.223993

min2423992

10

23892 → →→+

−− ββ

UPanPa h 23492

7.623490

10

23391 →→+

−β

PuNpnNp 24094

min6524093

10

23993 →→+

−β


2.1

2.2

2.3

2.4

2.5

2.6

2.7

2.8

2.9

3

3.1

1.E-03 1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

Neu

tron

pro

duct

ion

per f

issi

on

Neutron energy (eV)

Pu239U233U235U238Th232

Page 15

How many neutrons are available from fission?

Number of average neutrons per fission (ν-bar or ) is function of interacting neutron energy and differs between actinides.

From 5 basic isotopes (232Th, 238U, 233U, 235U, and 239Pu), it is highest for 239Pu: around 2.9 neutrons.

Second best is the 233U with only 2.5 neutrons.

235U with 2.43 neutrons is the worst from the major fissile isotopes.

232Th and 238U, if fissioned, also produce neutrons.


ν


– fission probability X neutrons originated per fission

Page 16

η

Eta (η) as the neutron generation factor describes generally the number of neutrons emitted by isotope/fuel per neutron absorption.

It was introduced by Enrico Fermi around spring 1941* as a part of the 4-factors formula for the fuel as a whole and solely for thermal neutrons.

It is often used for discussion of single isotope breeding capability.

In this case, it is a product of fission probability and ν-bar:

X

*Enrico Fermi, Physicist, book by Emilio Serge


Recalling 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.

Hence η should be higher than 2 in the respective spectrum.

It does not accounts for 238U and 232Th fission

and for different properties of secondary fertile and fissile isotopes. Page 17

η – fission probability X neutrons originated per fission

0

0.3

0.6

0.9

1.2

1.5

1.8

2.1

2.4

2.7

3

1.E-03 1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

η(n

umbe

r of n

ew n

eutr

ons

prod

uced

by

inte

ract

ion

with

neu

tron

)

Neutron energy (eV)

Pu239

U235

U233 (JEFF 3.1)

U233 (ENDBF VII)

U238

Th232

2 necessary neutrons

1.9 line in U-Pu cycle because of 238U 10% fission


Page 18

GenIV reactors = Sustainable reactors

Fuel recycling

Resources utilization (breeding)

Sustainability

Economics Proliferation

Waste reduction

Safety

Reactors with high Neutron economy

(fast spectrum)

Sustainability may conflict with other GenIV features.

MSR has potential to combine all four features.


0

2

4

6

8

10

12

14

16

18

20

0 20 40 60 80 100

Nat

. Ura

nium

util

izat

ion

(%)

Burnup (FIMA %)

235U enrichment of the feed

0.72% 1.5% 2.5% 5% 10% 20%

Page 19

Recycling ≤ Sustainability

Even if all actinides from spent fuel will be recycled, the utilization of natural resources can be still relatively low.

The “make-up fuel” (US English) or actually the feed (EU English) should not be enriched uranium.

Whenever enriched uranium is used as the feed, sustainability is strongly decreased.

Even its 100% utilization by recycling will not help.

Lets recall here than in 235U fueled reactor without recycling it is at maximum 0.75%.


80

82

84

86

88

90

92

94

96

98

100

0 2 4 6 8 10 12 14 16 18 20

Reso

urce

s util

izat

ion

(%)

Reprocessing frequency (FIMA %)

Reprocessing losses 0.01% 0.05% 0.1% 0.5% 1% 2%

Page 20

With natural uranium or thorium feed high sustainability can be achieved by recycling.

Nonetheless, due to the reprocessing losses, it will be always below 100%.

It depends on reprocessing method losses (L) and on the reprocessing frequency (F) (both expressed in fuel %).

Typical fuel burnup in solid fuel fast reactor is 10% FIMA. In MSR the discharge burnup may be lower.

Homework: please cross-check if it was derived correctly:

( )( )( )FL

FLlossesnUtilizatio−−−

−−=−=

111111

Sustainability = 238U and 232Th catalytic burning


Page 21

GenIV: Sustainability versus Safety

Sustainability often requires fast neutron spectrum.

In fast neutron spectrum coolant does not moderate the neutrons.

Coolant removal or fuel compaction leads to reactivity increase.

GFR has quite low positive void. It is, however, hard to cool in case of coolant depressurization.

SFR has strong positive void; nonetheless, it can be minimized by neutron leakage maximization in voided core. Still, there is an issue with sodium fire.

LFR has very strong void coefficient, but lead is not so easy to void.

In general SFR and LFR are low pressure system and the metallic coolant has retention potential for some problematic fission products.

MSR combines coolant and fuel in one liquid. It can be designed with negative void coefficient and fuel compaction / collection may be prevented. (moderated MSR breeder may have positive graphite temperature feedback coefficient)


Page 22

Is Gen IV the last one? No, let’s add some more Fuel / Cycle: 235U (U-Pu) U-Pu / closed Th-U / closed D-T / Li Sustainability*: 100-200 years 5 000 years 20 000 years 200 000+ years Radioactivity: Gen III+ Gen IV Gen IV+ Gen V Activated mat.: yes yes yes yes 1t burned fuel: 1000kg FP 1000kg FP 1000kg FP 800kg He + byproducts: 200kg Pu 50kg MA 15kg 237Np+238Pu MA=Np+Am+Cm 20kg MA

*Just kidding; it is hard to estimate energy consumption even in 100 years from now.


Page 23

Two basic fuel cycle issues related to sustainability

How to start it 1. Any reactor capable of burning for 238U and 232Th should be started

by 235U or by products from 235U fueled reactor.

How to maintain it 2. Any sustainable reactor for 238U and 232Th catalitic burning

should be capable to operate with equilibrium fuel composition. (secondary fertile and fissile, tertiary fissile and fertile, etc.…)

3. There should exist technology to regularly separate the fission products (FPs) from the fuel.

Separation of FPs is usually not possible without complete fuel decomposition. Hence, what other industries call recycling is called “reprocessing”.


Page 24

Initial and secondary stages & open versus closed cycle

Uranium-Plutonium cycle


Page 25

Initial and secondary stages & open versus closed cycle

Thorium-Uranium cycle


D 235U 238U 232Th Th+high 235U enriched U (HEU)

Fuel composition - equilibrium cycles

U-Pu Own Pu vector MA 238U U-Pu equilibrium cycle

234-236U

Th-U 233U 237Np 232Th Th-U equilibrium cycle238-242Pu

Page 26

Example of initial fuel composition equivalent

Fissile material Fertile material

Fuel composition - initial cycles (10% 235U equivalent)

A 235U 238U 10% 235U enriched U (LEU)

B LWR Pu vector MA 238U MOX fuel (+MA)

XC 235U 238U 232Th Th+20% 235U enriched U (MEU)

E LWR Pu vector MA 232Th Th+Pu (+MA)Limited PuF3 solubility in some fluorides salts


Page 27

Equilibrium cycle operation – neutron economy

238U 238U

235U235UPuA B

FP

Fresh Average Spent FuelInitial fuel = Fresh fuel

Converter > A B

Opencycle

A

Closedcycle

Convertor, e. g. PWR or IMSR, is operated usually in open fuel cycle.

Breeder profit from neutronics advantages only in the closed cycle. For Iso-breeding (EU) or Break-even (US) reactor => A=B.

Extreme breeder can be operated in Breed-and-Burn mode. It can have high fuel utilization even without reprocessing.

238U 238U

PuA B

Breeder ≤A B

Pu

FP

Opencycle

A

Fresh Average Spent FuelInitial fuel = Fresh fuel

Closedcycle

238U 238U

PuA~0

B

Breed & Burn << A B

FPOpencycle

*

Fresh Average Spent FuelInitial fuel = Average fuel!

Nextreactor


Page 28

Breeding capability – excess reactivity in equilibrium

Breeding capability can be estimated from excess reactivity in equilibrium fuel cycle.

If fuel cycle properties like: power (or flux), reprocessing scheme, and feed are fixed, reactor operation will converge to equilibrium.

In equilibrium mass flows, reaction rates, and reactivity are stabilized.

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-live n,2n

2.3d

U239

Np238

U238

23min

Pa231

26h

1d

Th231

U232

Pa232

α

M: 100%U233M: 100%

Pa233

Th233Th232

22min

27d


Page 29

Comparison of 16 reactors: 8 thermal & 8 fast

The simplified designs were adopted as is without optimization. If the core consists of assemblies with identical geometry but

different fuel composition only one assembly was simulated. If the geometry differs, all cases have been simulated, but only one selected is presented.

FHR 192.0 W/gHM

HTR 96.8 W/gHM

RBMK 13.7 W/gHM

PHWR 32.1 W/gHM

MSR-FLIBE 41.1 W/gHM

MSR 41.1 W/gHM

HPLWR 25.3 W/gHM

LWR 41.1 W/gHM

MSFR 41.1 W/gHM

MSFR-FLIBE 41.1 W/gHM

LFR 54.8 W/gHM

SFR 48.8 W/gHM

GFR 40.1 W/gHM

MFBR 178.6 W/gHM

NaCl-AcCl4 salt 54.8 W/gHM

AcCl4 salt 54.8 W/gHM


Page 30

Assumptions for equilibrium cycle simulation

Infinite lattice cell level simulation.

Reactor specific power given by burnup in FIMA % (FIssile MAterial %) and fuel residence time.

Neglecting fission products.

Zero reprocessing losses (L=0).

Continuous feed of fertile material (232Th or 238U).

ENDF/B-VII.0 nuclear data library.

With these assumptions we obtained equilibrium fuel composition and equilibrium reactivity.

The excess reactivity should be high enough to compensate for: neglected reprocessing losses, neutron leakage and fission products parasitic captures.


-6000-4000-2000

02000400060008000

100001200014000160001800020000220002400026000

Reac

tivity

(PCM

)

Hypoth. ρ233U234U235-238U237NpPu233PaOther AcStructuralExcess ρ

Th-Ucycle

Page 31

Excess reactivity in eql. cycle for Th-U cycle Excess reactivity for eql. fuel composition quantifies the closed cycle capability.

Comparison of 16 reactors is based on infinite lattice calculations with no FPs.

Th-U cycle: low 233U capture, power effect due to 233Pa capture (FHR, MFBR,…).

Krepel, J. at. al., Chapter for IAEA document, Near Term and Promising Long Term Options for Deployment of Thorium Energy, in preparation.


Page 32

Excess reactivity break-down method

Neutron balance eq.: Four assumptions: 1) 2) 3) main fertile represents at least 90% of all (n,2n) reactions Valid only in equilibrium: 4) total other-than-fertile actinides destruction (total fission rate without the fertile isotope fission rate) should be in equilibrium equal to the total other-than-fertile actinides production (capture or (n,2n) reactions on the main fertile element) Result:

totalnn

totalC

totalF

totalnn

totalP

RRRRR

k2,

2,inf

2++

+=

totalF

totalP RvR = other

CTh

CtotalC RRR +=

232

ThF

totalF

Thnn

ThC RRRR

232232232

2, −=+

Available neutrons Bonus from fertile Parasitic captures

( )Thnn

totalF

otherC

Thnn

totalF

Thnn

ThF

Thnn

totalF

otherC

Thnn

ThF

totalF

RRR

RRRR

RRRRRR

232232

232232

232

232232

2,2,

2,

2,

2,

2222

222

+−

+

++

−≅

+

−++−=

νννν

ν

νρ

Thnn

totalnn RR

232

2,2, =


-9000-6000-3000

0300060009000

12000150001800021000240002700030000330003600039000

Reac

tivity

(PCM

)

Hypoth. ρ239Pu240Pu241-242PuAmCm239NpOther AcStructuralExcess ρ

U-Pucycle

Page 33

Excess reactivity in eql. cycle for U-Pu cycle Low 239Pu fission probability: 239Pu: 65-75% X 233U: 90% => thermal reactors.

Excess reactivity is higher for fast reactors: 239Pu: ν=2.9 X 233U: ν=2.5

U-Pu cycle has better neutron economy, Th-U cycle better neutron efficiency.


-3000

0

3000

6000

9000

12000

15000

18000

21000

24000

27000

30000

33000

36000

39000

Reac

tivity

(PCM

)

Hyp. rho239Pu240Pu241-2PuCmAm239NpOther AcFPsSaltExcess ρ

U-Pucycle

Page 34

8 Fast MSR (salts) comparison - inclusive FPs

8 selected salts were compared (infinite medium of fast reactor with FPs).

U-Pu and Th-U equilibrium closed cycles were evaluated (by excess reactivity).

It confirmed that for U-Pu cycle chlorides are preferable.

The reactivity excess in chlorides may enable breed and burn mode.

Th-U cycle has two favorites 7LiF and Na37Cl carrier salts. 233U: ν=2.5, fission probability 90% X 239Pu: ν=2.9, fission probability 65-75%

-2000

0

2000

4000

6000

8000

10000

12000

14000

16000

18000

20000

22000

24000

26000

Reac

tivity

(PCM

)

Hyp. rho233U234U235-9U237NpPu233PaOther AcFPsSaltExcess ρ

Th-Ucycle


Illustration of burnup distribution. The fuel is in reality mixed.

Liquid fuel reactorFresh Fuel Spent Fuel

Page 35

Breed and burn fuel cycle mode

238U 238U

PuA~0

B

Breed & Burn << A B

FPOpencycle

*

Fresh Average Spent FuelInitial fuel = Average fuel!

Nextreactor

Solid fuel reactorFresh Fuel Spent Fuel

0

1/T

0 T

Fuel

dist

ribut

ion

p(t)

Irradiation time (t)

Solid fuelLiquid fuel

In case of a super breeder, fuel based only on fertile 238U or 232Th can be loaded to the reactor.

The fissile fuel will be produced (Breed) in the reactor.

Later during its burning it will supply enough neutrons to breed new fuel from new fresh fertile assemblies.

In solid fuel case it looks like this =>

The situation for liquid is different. Everything will be homogenized.

Different burnup distributions:

(T is average irradiation time)


0.90

0.95

1.00

1.05

1.10

1.15

1.20

0 10 20 30 40

Equi

libiru

m k

-infin

ity [-

]

Average discharge burn-up [% FIMA]

20UCl3-80UCl4UCl455NaCl-40ThCl-5PuCl60NaCl-40UCl368NaCl-32UCl3ThCl455NaCl-45ThCl4

Page 36

MSR in Breed-and-Burn (B&B) mode Method can be developed on the

burnup distributions (which differ between solid and liquid fuels).

The average k-infinity for given T can then be computed using a simple cell depletion calculation:

Main results on cell level:

• B&B mode is possible only with enriched 37Cl based MSR.

• U-Pu cycle is better that Th-U.

• B&B in Th-U cycle may require fissile support (e.g. LWR Pu).

Hombourger, B. et. al., Fuel cycle analysis of a molten salt reactor for breed-and-burn mode, ICAPP 2015, Nice

B&B mode represent open fuel cycle with up to 20% resources utilization.

( ) ( )∫∞

∞∞ =0

dttktpk TT

(it is proportional to average irradiation time T)


Page 37

Chlorides disadvantage: density and migration area

Chlorides salts have lower specific Ac density and higher migration area.

Chlorides area transparent for neutrons (absence of scattering).

High migration area => high leakage => blanket or reflector or bigger reactor.


Page 38

MSR Breed-and-Burn: core level

Concept SOFT-1980 MSFR B&B - PSI MCFR

B&B / salt No / nat. chlorides No / fluorides Yes / enr. chlorides Yes/ enr. chlorides

Core dimensions 5.23 m 2.25 m x 2.25 m 4 m x 4 m ?

Core volume 75 m3 9 m3 50 m3 ?

Blanket / cycle None / U-Pu 7.3 m3 / Th-U None/ U-Pu, Th-U+Pu None/ U-Pu

Reflector CaCl2-NaCl & steel Axial only - Hastelloy Yes – lead / Enr. lead Yes - ?

Processing Volatile & Soluble FP Volatile & Soluble FP Volatile FP only Volatile FP + ?

Processing flow 0.25 L/s 3-8 L/day 2 L/day ?

Cycle time ?/continuous electrolysis 6-16 years 52 years ?

B&B – PSI test design


- in ideal case = never-ending fire

Page 39

Breeding reactor

Fuel: 238U or 232Th

Reactor (“catalyzed” by 239Pu or 233U)

Energy

Works only if emissions (FPs) are (continuously) removed

http://www.clipartpanda.com/categories/ fire-clip-art-free-download


Page 40

Issue with closed cycle: reprocessing & fabrication

Since FPs absorb neutrons, they sooner or later poison the reactor.

Thus the fuel, which is highly radiotoxic, must be reprocessed, which is demanding and complicated.

In U-Pu cycle recycling of Pu and U is technologically mastered and practically available in several countries.

The by-products of the U-Pu, minor actinides Am and Cm emit α (heat source) and neutrons (mainly Cm). Their recycling may thus strongly complicate the fabrication of solid fuel.

Similar technological experience as for U-Pu is missing for Th-U cycle.

Furthermore, irradiated Th fuel emits more hard gammas from the 232U decay chain, mainly from 208Tl and 212Bi.

Recycling of solid Th fuel may be demanding.

Liquid fuel (no fabrication) can accommodate both MA from U-Pu and 232U from Th-U cycles (recycling by volatilization) more easily.


Page 41

Fission products sensitivity

Fast spectrum systems are less sensitive to fission products FPs…

Why?

The FPs cross-section are higher in thermal spectrum. (Isn't it valid for all cross-section, not only for FPs?)

There is also second reason: FPs to fissile fuel ratio.

In typical thermal burner (e.g. LWR or HTR see the figure) initial fissile load may be between 5-8%. After discharge there are usually 2% left.

In fast breeder reactor (SFR) the fissile isotopes can represent 10% of fuel and it is the same after discharge.

The FPs to fissile ratio is thus: 5/2 for LWR and 10/10 for SFR. (FPs replace fertile absorbers)

Hence, thermal reactors, especially burners, are more sensitive to fission products.

HTR fuel with initial 8% enrichment


Page 42

Fuel irradiation time and need of recycling

In solid fuel reactors the irradiation time is limited by: 1. Limited cladding lifetime caused by irradiation. 2. Fissile element load in burners (breeders can be self-sustaining). 3. Gaseous Fission Products (FPs) pressure. 4. Core poisoning by FPs neutron capture. In liquid fuel reactor: 1. There is no cladding. 2. Breeders are self-sustaining and fuel or Th can be continuously added. 3. Gaseous and volatile FPs are continuously removed form the core. 4. Remaining FPs are still poisoning the core by neutron capture. In MSR case there is not another reason for fuel reprocessing

than FPs removal.


Page 43

Sal components and FPs removal

MSR fluoride salts components: 1. Carrier salt (LiF, LiF-BeF2, NaF-BeF2, NaCl, etc.) 2. Fertile actinides (232Th and 238U). 3. Fissile fuel (mainly U or Pu vector). 4. By-products (MA). 5. FPs.

FPs removal There is not a simple method how to separate FPs from the fuel salt. Furthermore, even if several methods are combined,

FPs are usually the last separated component. Practically in every MSR design study or simulation,

the spent fuel salt is removed from the core for reprocessing being immediately replaced by the same cleaned salt.

Since the whole fuel salt mix must be removed from the core, the question is what should be recycled, why, and for what price?


Page 44

Motivation for salt recycling and recycling strategies

Why to recycle salt components: 1. From a reactor physics point of view, it is important to recycle 233U

or 239Pu as the main fissile elements; the other components are not substantive.

2. From a sustainability point of view, it may be important to recycle the main fertile elements Th or U and possibly some rare elements (Li, Be).

3. From economy point of view, it may be interesting to recycle all components. Nevertheless, it will depend on their price and on the reprocessing costs. In some cases their direct disposal, e.g. by vitrification, can be cheaper.


Page 45

Operations with liquid fuel

Two extremes: on-line recycling – everything in-situ and ASAP, and off-line recycling – everything ex-situ and with years of delay.

Salt removal from the

core

Removed salt share

Fissile fuel recycling (U-vector)

Fissile fuel return after

reprocessing

Carrier salt cleaning

Carrier salt return after

reprocessing

Reprocessing waste

immobilization

Continuous or

Batch-wise

From 0.1% to whole

salt volume

In-situ or

Ex-situ

ASAP or with months or

years of delay

In-situ or

Ex-situ

ASAP or with months or

years of delay

In-situ or

Ex-situ

Four possible basic operation with the liquid fuel: 1. Salt removal from the core.

(no direct impact to the core neutronics) 2. Salt cleaning inside of the core.

(direct impact to the core neutronics) 3. Salt cleaning or reprocessing outside of the core.

(no direct impact to the core neutronics) 4. Salt refilling into the core.

(direct impact to the core neutronics)

Recycling strategies:


Page 46

Comparison of 7 similar salt treatment schemes (Th-U) Assumptions: Reprocessing unit capacity 25 l/day. The volume for reprocessing is taken

from core (cases 6 and 7) or from temporary storage tank (cases 1-5).

Main conclusions: 1. Reactivity swing is positive and

proportional to the reprocessing time. (decreasing Th mass = +2.2 PCM/kg; increasing FPs mass = -2.0 PCM/kg)

2. Continuous Th refilling can be used as reactivity control, independently off the selected salt clean up treatment.

3. The strategy with longest reprocessing time has lowest average FPs content. (it has also highest breeding gain)

4. Its disadvantage is the biggest salt volume (initial load) necessary for reactor operation.

100

300

500

700

900

1100

1300

0 3 6 9 12 15 18 21 24

Rea

ctiv

ity (P

CM

)

Time (EFPM)

18000l each 24M9000l each 12M4500l each 6M2250l each 3M750l each 1M25l each 1daycont., Th refill each 1M

Strategy Nr.

Salt clean-up from FPS

Th refilling Min. salt volume for operation

1 18000l each 24M each 24M 36 m3 2 9000l each 12M each 12M 27 m3 3 4500l each 6M each 6M 22.5 m3 4 2250l each 3M each 3M 20.25 m3 5 750l each 1M each 1M 18.75 m3 6 25l each 1day each 1day 18 m3 7 continuous each 1M 18 m3

Reactivity swing for 7 recycling strategies Krepel, J. at. al., Comparison of Several Recycling Strategies and Relevant Fuel Cycles for Molten Salt Reactor. ICAPP 2015 Nice


Page 47

Combined recycling: fissile in-situ ASAP, the rest ex-situ Volatilization method can be

used for fluoride salt MSR in Th-U cycle to separate the Trans-Thorium elements (U, Np, Pu).

Partial selection is possible.

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

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-live n,2n

M: 100%U233M: 100%

Pa233

Th233Th232

22min

27d

Assumption for the simulation: repetitive application of reprocessing: U 0% other Ac 1% loss. FPs 99% removal efficiency are replaced every 12 months by Th.

Scenario I. (U+Np+Pu+…)

Scenario II. (U+Np)

Scenario III. (U only)


Combined recycling: Fissile in-situ ASAP, the rest ex-situ Fast spectrum MSR core

020406080

100120140160180200

0 10 20 30 40 50 60 70 80 90 100

Mas

s in

the

fuel

sal

t (kg

)

Burned mass (t) or 2.6GWth reactor operation (EFPY)

NP237PU238PU239Other Ac

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70 80 90 100

Cum

ulat

ed re

mov

ed m

ass

(kg)



0

5

10

15

20

25

0 10 20 30 40 50 60 70 80 90 100

Mas

s in

the

fuel

sal

t (kg

)



0

200

400

600

800

1000

1200

1400

1600

1800

0 10 20 30 40 50 60 70 80 90 100

Cum

ulat

ed re

mov

ed m

ass

(kg)



0

20

40

60

80

100

120

140

0 10 20 30 40 50 60 70 80 90 100M

ass

in th

e fu

el s

alt (

kg)



0

200

400

600

800

1000

1200

1400

1600

1800

0 10 20 30 40 50 60 70 80 90 100

Cum

ulat

ed re

mov

ed m

ass

(kg)



Scenario I. (U+Np+Pu+…) Scenario II. (U+Np) Scenario III. (U only)

Ac mass in the core

Cumulative Ac mass in the waste 100EFPY Krepel, J. at. al., Molten Salt Reactor with Simplified Fuel Recycling and Delayed Carrier Salt Cleaning. ICONE 2014 Prague


Combined recycling: radiotoxicity of cumulated waste

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

1.E+11

1.E+12

1 10 100 1000 10000 100000 1000000

Radi

otox

icity

of c

umua

ted

Ac (S

v)

Time (years)

1% losses for all Ac(except U)

All Ac recyclingOnly U recyclingU+Np recyclingDifference due to

232Th and 235U decay chains

Difference due to 238U decay chain

Difference due to 238U and 237Np

decay chains

Fast spectrum MSR core

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

0 16 251 3981 63096 1000000

Radi

otox

icity

of c

umua

ted

Ac (S

v)

Time (years)

232Th decay chain CM244

PU240

U236

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

0 16 251 3981 63096 1000000

Radi

otox

icity

of c

umua

ted

Ac (S

v)

Time (years)

237Np decay chainPU241

AM241

CM245

TH229

U233

NP237

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

0 16 251 3981 63096 1000000

Radi

otox

icity

of c

umua

ted

Ac (S

v)

Time (years)

238U decay chain PU238

RA226

TH230

PU242

U234

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

0 16 251 3981 63096 1000000

Inge

stio

n ra

diot

oxic

ity (S

v/kg

of f

uel)

Time (years)

235U decay chainAC227

PA231

PU239

U235

Mainly 238Pu

Mainly 239Pu and 240Pu Mainly

226Ra and 229Th

All Ac recycling case

Page 49


Page 50

Conclusion

Sustainability of 235U fueled reactors is low.

232Th and 238U can be burned in fast (232Th possibly also in thermal) breeders; with fuel recycling high sustainability may be achieved.

Fast spectrum breeder with solid fuel may have safety issues.

These issues may be eliminated by liquid fuel.

The liquid fuel state also provide fuel cycle flexibility.

Several cleaning, reprocessing, and refilling / removing techniques may be applied to liquid fuel.

Solubility and other thermochemical properties may be the limiting factor.

MSR may combine sustainability with acceptable safety, economy and proliferation resistance.

Page 51

Wir schaffen Wissen – heute für morgen

MSR is a very promising energy source. It can combine unparalleled safety features with high fuel utilization. It can also provide us enough time for mastering of the nuclear fusion!


Page 52

Sulfur production in Chloride MSRs

Sulfur embrittles steels & nickel alloys Main production paths:

Enrichment in Cl-37 substantially decreases Sulfur production.

Cl35 + n → P32 + α → S32

Cl37 + n → P34 + α → S34

0,18

1,8

18

180

1800

1

10

100

1000

10000

0 20 40 60 80 100

Equi

vale

nt m

ass f

or 5

0 m

3 [k

g]

Mas

s fra

ctio

n Su

lfur [

wpp

m]

Effective Full Power Years [EFPY]

Pure Chlorine-37Natural Chlorine1% Cl-35

[1] IANOVICI, E., TAUBE, M., Chemical behaviour of radiosulphur obtained by 35Cl(n, p)35S during in-pile irradiation, J. Inorg. Nucl. Chem. 37 12 (1975) 2561.

[2] IANOVICI, E., TAUBE, M., Chemical State of Sulphur Obtained by the 35Cl(n,p)35S Reaction during In Pile Irradiation, EIR-Bericht 267, Eidgenössisches Institut für Reaktorforschung, Würenlingen, Switzerland (1974).

Investigation on the speciation of Sulfur were carried out at EIR in the seventies.

beta + beta - stable

beta- decay

37Cl 35Cl 36Cl 34Cl

36S 34S 35S

34P 35P 33P 32P

33S 32S

(n,α)

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