Wir schaffen Wissen – heute für morgen

Fast Reactor Physics

Konstantin Mikityuk, FAST reactors group @ PSI

http://fast.web.psi.ch

Thorium Energy Conference 2013

CERN Globe of Science and Innovation

Geneva, Switzerland, October 27-31, 2013

2

Outline.

Fast reactors: breeding.

Fast reactors: past and future.

Fast reactors: few R&D projects in Europe.

Fast reactors: could Th become a fuel?

Sustainability

Safety

Proliferation resistance

Radiotoxicity and decay heat

Summary: advantages and disadvantages of Th for FR

3

Fast reactors: breeding.

4

Fast critical reactor

A fast neutron critical reactor is a category of nuclear reactor in which the fission

chain reaction is sustained by fast neutrons.

Such a reactor needs no neutron moderator, but must use fuel that is relatively

rich in fissile material when compared to that required for a thermal reactor.

1x10-2

1x10-1

1x100

1x101

1x102

1x103

1x104

1x105

1x106

1x107

Energy (eV)

0x100

1x1014

2x1014

3x1014

4x1014

5x1014

6x1014

7x1014

8x1014

Flu

x pe

r un

it le

thar

gy (

cm-2s-1

)

SFR

PWR

SFR PWR

5

Breeding

238 239

239

239

92U

93Np

94Pu

91Pa

90Th 232 233

233

233

β–

β–

β–

β–

Tho

rium

fuel

cyc

le

Ura

nium

fuel

cyc

le

(n,γ)

(n,γ)

fertile

fertile fissile

fissile

23

.5 m

2

.35

d

22

m

27

d

A production of new fissile isotopes in the nuclear

reactor is a kind of transmutation called a breeding and

non-fissile isotopes (U-238 and Th-232), which give

birth to the new fissile isotopes, are called fertile.

6

Neutron balance in a critical reactor

A_fissile

P = A_fissile + A_fertile + A_parasitic + LR

P = A + LR

keff = Production rate / (Absorption rate + Leakage Rate) = 1

A_fissile A_fissile A_fissile

h = 1 + BR + L

h – Number of n’s emitted per neutron absorbed in fissile fuel

BR – Breeding Ratio: Number of fissile nuclei created

per fissile nucleon destroyed

L – Number of neutrons lost per neutron absorbed in fissile fuel

7

Breeding: h for main fissiles

1x10-2

1x10-1

1x100

1x101

1x102

1x103

1x104

1x105

1x106

1x107

Neutron energy, eV

0

1

2

3

4

h

Pu-239

U-235

U-233

0x100

1x1014

2x1014

3x1014

4x1014

5x1014

6x1014

7x1014

8x1014

Flu

x pe

r un

it le

thar

gy (

cm-2s-1

)

SFR

PWR

Average number of fission neutrons emitted per neutron absorbed as a

function of absorbed neutron’s energy for three fissile isotopes

Best for breeding

8

Breeding

Burning of Pu-239 and U-233 in a fast neutron spectrum (>105 eV) provides

the highest number of fission neutrons per neutron absorbed in fuel.

The extra neutrons can be absorbed by fertile isotopes with a rate which is

equal or even higher than the fissile burning rate.

The fast neutron spectrum reactor with BR>1 is called a breeder and with

BR=1—an iso-breeder.

Fast neutron spectrum allows to efficiently “burn” fertile U-238 or Th-232—

via transmutation to fissile Pu-239 or U-233.

9

Fast reactors: past and future.

10

First "nuclear" electricity – fast reactor.

In 1949 EBR-I – Experimental Breeder Reactor I – was designed at Argonne

National Laboratory. In 1951 the world’s first electricity was generated from

nuclear fission in the fast-spectrum breeder reactor with plutonium fuel

cooled by a liquid sodium.

First “nuclear” electricity : four 200-watt light bulbs. Courtesy of ANL.

11

Fast reactors: 1946 – 2013 MWth

Hg Hg NaK Na LBE

Clementine

EBR-I

BR-10

DFR

LAMPRE

EBR-II

Fermi-1

Rapsodie BOR-60

SEFOR

KNK-II

BN-350

Phénix

PFR

OK-550/BM-40A

JOYO

FFTF

BN-600

Super-Phénix

FBTR

MONJU

CEFR

1946 1952

1951 1964

1958 2002

1959 1977

1961 1963

1961 1994

1963 1972

1967 1983

1968 2013

1969 1972

1972 1991

1972 1999

1973 2009

1974 1994

1974 1990

1977 2013

1980 1992

1980 2013

1985 1996

1985 2013

1994 2010

2010 2013

USA

USA

Russia

UK

USA

USA

USA

France

Russia

USA

Germany

Kazakhstan

France

UK

Russia

Japan

USA

Russia

France

India

Japan

China

0.025

1.2

8

60

1

62.5

200

40

55

20

58

750

563

650

150

140

400

1470

2990

40

714

65

12

The Generation IV International Forum (GIF) is a cooperative international

endeavor organized to carry out the R&D needed to establish the feasibility

and performance capabilities of the next generation nuclear energy systems.

Argentina, Brazil, Canada, France, Japan, Korea, South Africa, the UK and

the US signed the GIF Charter in July 2001, Switzerland in 2002, Euratom in

2003, China and Russia both in 2006.

Six nuclear energy systems were selected for further development:

4. Very-high-temperature reactor (VHTR)

5. Supercritical-water-cooled reactor (SWCR)

6. Molten salt reactor (MSR)

1. Gas-cooled fast reactor (GFR)

2. Sodium-cooled fast reactor (SFR)

3. Lead-cooled fast reactor (LFR)

13

Sustainability

Safety

Economics

Reliability

Proliferation-resistance

Generation-IV systems: keywords

14

Fast reactors: few R&D projects in Europe.

15

European sodium-cooled fast reactor.

Reactor vessel

Na

Ar 1 bar

core

Primary

pumpÍ6

SGÍ6

545ºC

395ºC

490ºC

240ºC

Na

~1 barH2O

185 bar

Secondary

pumpÍ6

IHXÍ6

Na

~1 bar

340ºC

525ºC

Air

HXÍ6

35ºC

Na

~1 bar

DHR

HXÍ6

DH

R lo

opÍ

6

Power: 3600 MWth

Coolant: sodium@1 bar

Fuel: (U-Pu)O2

Clad: stainless steel

ESFR EURATOM FP7 project

16

Lead-cooled fast reactor demonstrator.

Core

Primary pumpÍ8

SG

Í8

Reactor vessel

H2O

180 bar

480ºC

Pb

335ºC

450ºC

Feedwater

pump

Ar

1 bar

HP

turbineLP

turbine

Condenser400ºC

DH

R c

on

de

nse

r

H2O

1 b

ar

Power: 300 MWth

Coolant: lead@1 bar

Fuel: (U-Pu)O2

Clad: Stainless steel

ALFRED Consortium:

Italy,

Romania,

Poland, …

17

Gas-cooled fast reactor demonstrator.

Power: 75 MWth

Coolant: helium@70 bar

Fuel: (U-Pu)O2

Clad: Stainless steel

core

Primary

blowerÍ2

HX

Í2

Guard vessel

N2

14 bar

DHR

blowerÍ3

DHR

HX

H2O

10 bar 50ºC

He70bar H2O

65 bar26

0ºC

H2O pool

1 bar 50ºC

530ºC

N2

1 bar

Reservoir

DHR loopÍ3Main loopÍ2

127ºC

197ºC

Air cooler

Í2

1 bar

Water

pumpÍ2

35ºC

125ºC

ALLEGRO Consortium:

Czech Republic,

Hungary,

Slovakia, …

18

Fast reactors: could Th be a fuel?

19

Sustainability.

Depleted U stock

Spent fuel cooling

Fuel fabrication

Fast

reactors

Geologic

repository Separation

of elements

U-dep

Ac

AcO2 + FP AcO2 + FP

FP + losses

“Ac” = “actinides”,

i.e. U + Np + Pu + Am + Cm + ...

“FP” = fission products

AcO2

(According to calculations) fast reactors can operate in an equilibrium closed U-

Pu fuel cycle with BR=1 (amount of fissile produced = amount of fissile

consumed) fed by only depleted (or natural) uranium

20

238 237

238

238

237

239

239

239

235 234 92U

93Np

94Pu

95Am

96Cm

240

240

241

241 242

242 244

244 243

245

FP

242 243

+1000

854 –140

6

854

1

1 853

–1

6

5 –1

5

–14 6

–678 181 102

–84 12

–62

10

–6

4 10

–1 9

9

4

4

–2 3

–3 0

28

–4 21

–3 3

17 17

17

2

2

1

–5 –8

–844 –1

–142 –1000

(Cm) (Am) (Pu) (Np) (U)

242 m

feed fuel

6.7

5 d

2.1

d

23

.5 m

2

.35

d

7 m

in

14

.3 y

4.9

8 h

26

min

16

h

16

h

–1

(n,2n)

β–

(n,γ)

β+

fission

M

mass number

α

EQL-U: mass balance in SFR (simplified model)

21

Sustainability.

Could the same reactors operate in an equilibrium closed Th-U fuel cycle?

(According to calculations) the answer is yes, but since no U-233 (main fissile

isotope for this cycle) is available, we face a problem

Th disadvantage: How to start thorium fast reactor? What fissile material to

use? Plutonium? Uranium-235? Uranium-233 generated somewhere else?

22

EQL-Th: mass balance in SFR (simplified model)

237

239

92U

93Np

94Pu

91Pa

90Th 233

233

233

+1000

–35

feed fuel

6 959

959

22

m

231

6 26

h

231 6

232

1.3

d

6

4 234

6.7

h

4 27

d

955

–877

232 1 79

–4

1

234

–35 49

235

–39 10

236

–2 8

8

6.7

5 d

–2 6

238

6 2.1

d

238

–4 1

1

–1

1

237

1

228 232

1

FP

–5 –2

–957 –0

–35 –999

(Pu) (Np) (U) (Pa) (Th)

Th advantage: very low amount of minor actinides

Th disadvantage: production of U-232—precursor of

gamma emitters

23

U-234

U-235

U-236

U-238

Np-237

Np-239

Pu-238

Pu-239

Pu-240

Pu-241

Pu-242

Am-241

Am-242m

Am-243

Cm-242

Cm-244

Cm-245

Cm-246

0.01 0.1 1 10 100

0.07

0.01

0.04

81.59

0.10

0.31

10.17

5.78

0.66

0.55

0.36

0.02

0.15

0.01

0.11

0.03

0.02

EQL-U and EQL-Th fuel compositions in SFR (%wt)

Th-228

Th-230

Th-232

Pa-231

Pa-233

U-232

U-233

U-234

U-235

U-236

Np-237

Pu-238

Pu-239

Pu-240

0.01 0.1 1 10 100

0.04

0.04

85.64

0.06

0.12

0.05

9.56

2.98

0.60

0.63

0.13

0.10

0.02

0.01

24

EQL-U and EQL-Th neutron balance

U236

U238

Np237

Np239

Pu238

Pu239

Pu240

Pu241

Pu242

Am241

Am242m

Am243

Cm244

Cm245

Cm246

Structures

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4

Th230

Th232

Pa231

Pa233

U232

U233

U234

U235

U236

Np237

Pu238

Pu239

Pu240

Pu241

Pu242

Structures

0.0 0.2 0.4 0.6 0.8 1.0 1.2

k-inf = 1.30533 k-inf = 1.17023

Blue bars are isotope-wise contributions to absorption (sum up to 1)

Red bars are isotope-wise contributions to production (sum up to k-inf)

Th disadvantage:

lower k-infinity

25

Safety.

We look at just two reactivity effects: Doppler effect and (sodium) void effect

having in mind other reactivity effects (less fuel type dependent)

strongback

diagrid

core

control rods

vessel

Thermal expansion effects (not considered) Void reactivity effect

26

EQL-U and EQL-Th fuel reactivity effects in SFR

-3.0 -2.5 -2.0 -1.5 -1.0 -0.5 0.0

Doppler effect ($)

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

5.0

5.5

6.0

Voi

d ef

fect

($)

Th-232

U-233U-235

NaCladding

NaCladding

U-238

Pu-239

Pu-240

Pu-241

i

i

i

i

i

P

A

P

A

0

0

Th advantage:

stronger Doppler and

weaker void effects

Infinite medium (no leakage

component)

Doppler (Nominal → 3100 K)

Void (Nominal → 0 g/cm3)

Isotope-wise decomposition:

27

1x10-1

1x100

1x101

1x102

1x103

1x104

1x105

1x106

1x107

Neutron energy, eV

0

1

2

3

4

h

0x100

2x10-3

4x10-3

6x10-3

8x10-3

(u

) (c

m-2s-1

)

SFR

EQL-U and EQL-Th fuel reactivity effects in SFR

Why void effect is weaker in case of EQL-Th?

Sodium removal leads to

spectral hardening—shift to the right

Pu-239: grows quicker

U-233: grows slower

28

Proliferation resistance.

238 239

239

239

92U

93Np

94Pu

91Pa

90Th 232 233

233

233

β–

β–

β–

β–

Tho

rium

fuel

cyc

le

Ura

nium

fuel

cyc

le

(n,γ)

(n,γ)

fertile

fertile fissile

fissile

23

.5 m

2.35 d

Th disadvantage: fissile precursor has higher half

life, potential to be separated 22

m

27 d

Th advantage: misuse of U-233 is protected by

presence of U-232

231

232

232

β–

231

29

EQL-U and EQL-Th fuel RT and DH (no FP)

10 100 1000 10000 100000 1000000

Time, years

1E-006

1E-005

1E-004

1E-003

1E-002

Dec

ay h

eat,

W/g

SFR-U

SFR-Th

1

10

100

1000

10000

Rad

ioto

xici

ty, S

v/g

SFR-U

SFR-Th

Th advantage: Radiotoxicity and decay heat of EQL fuel are lower for ~10000y

30

Summary.

31

Summary... Th disadvantages

Past and current fast reactors were/are based on U-Pu cycle.

Operational experience with thorium-uranium fuel is low.

Experience in fuel manufacturing and reprocessing is lower for Th-U

fuel compared to U-Pu.

Fissile fuel for Th-U cycle (U-233) is not available.

U-232—precursor of hard gamma emitters—is produced in Th-U cycle

(n2n reaction is higher in fast spectrum).

k-infinity of equilibrium fuel is lower for Th-U cycle compared to U-Pu

one. This means that to reach iso-breeding the blankets of fertile

material can be required.

Fissile precursor of U-233 (Pa-233) has higher half life (compared to

Np-239)—potential to be separated and decayed to pure U-233.

32

Summary... Th advantages

Calculational analysis with state-of-the-art codes shows that fast

reactor can operate as an iso-breeder in Th-U cycle closed on all

actinides.

There is very low amount of minor actinides in EQL-Th fuel cycle.

Doppler effect is stronger and void effect is weaker in EQL-Th fuel

compared to EQL-U.

Misuse of U-233 is protected by presence of U-232 (predecessor of

hard gamma emitters).

Radiotoxicity and decay heat of EQL-Th fuel are lower during the first

10000 years of cooling compared to the EQL-U fuel.

Thank you. Questions?