1- Fast reactor basic features

M. Salvatores

The physics of fast vs thermal neutrons

Flexibility: breeding and/or burning for different missions in the fuel cycle

CONSULTANTS’ MEETING: EDUCATION & TRAINING SEMINAR ON FAST REACTOR SCIENCE AND TECHNOLOGY

ITESM CAMPUS SANTA FE, MEXICO CITY

29 JUNE – 03 JULY 2015

Classification of nuclear power systems based on system technology

…a wide offer!

Or: (n+nucleus with A nucleons)= (A+1)excited

(A+1) +γ: (n, γ) capture reaction

A+n+γ: (n,n‘) inelastic scattering

(A-1) +2n: (n,2n) scattering

etc

Reaction channels

Fission and other neutron-nuclei interaction reactions. A reminder

Fission

Energy distribution of neutrons emitted in fission

• The most fundamental technological difference between nuclear fission reactors

concerns the means by which the problem of sustaining a chain reaction is achieved.

• One solution is to slow down neutrons to so-called "thermal" energies (around 0.025 eV)

by using a « moderator »:

Moderator can be placed around a fuel lump to

slow-down fission neutrons from fast, MeV/keV,

to thermal, below eV, energies

This has the advantage of allowing a chain reaction to be sustained using natural or

slightly enriched uranium, and almost all of the world's operating power reactors employ

this solution -these are known as "thermal reactors". If water is the moderator, we speak

of « Light Water Reactors », LWR

• The disadvantage of this approach is that only 0.7% of uranium produces useful energy.

This can be overcome by increasing the proportion of fissile atoms by enrichment, or by

using plutonium, and by constructing the reactor without a moderator. In this case the

average energy of the neutrons in the core is much greater than in thermal reactors (they

are known as "fast" neutrons).

Thermal or fast neutrons?

• Significant elastic scattering of the neutrons in both spectra

• However, in FRs neutron moderation is much less since high A materials are used

• If sodium is chosen as coolant in FR, it is also the most moderating material

• In LWRs, neutrons are moderated primarily by hydrogen

• Slowing-down power in FR is ~1% that observed for typical LWR:

Minimum energy of a neutron after elastic collision is determined by the

parameter α: Emin= αE where:

• Thus, fast neutrons are either absorbed or leak from the reactor before they can reach thermal

energies

Neutron Moderation and Cross Sections Comparison

2

1A

1Aα

Hydrogen 0

Oxygen 0.779

Na 0.840

U 0.983

(Fission neutrons are or are not slow down)

Comparison of LWR and SFR neutron spectra

Fast reactor moderating materials Slowing down power

2

1A

1Aα

Comparison of fast reactor spectra

Energy dependence of neutron cross-sections: Pu-239

Energy dependence of neutron cross-sections: U-238

235U 239Pu 233U

Thermal Fast Thermal Fast Thermal Fast

f (barn) 582 1,81 743 1,76 531 2,79

2,42 2,43 2,87 2,94 2,49 2,53

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Fis

sio

n/A

bs

orp

tio

n

PWR

SFR

The fission/absorption ratios are consistently

higher for the fast spectrum SFR.

Thus, in a fast spectrum, actinides are

preferentially fissioned, not transmuted into

higher actinides

dEEEEdEEEABSORPTION/FISSION fcf

Integral cross sections comparison

Implications of fast spectrum physics

CR

Neutron balance comparison

SFR

Conversion ratio defined as TRU production/TRU destruction

For the equilibrium actinide concentration in the core, the neutron production

(per fission) in the fuel can be computed for a specified composition

consisting of i-components with proportions xi :

1/)( faffuel

eqD

(neutron/fission)

fi

i

i

ci

i

i

fi

i

i

fii

i

i

x

x

;x

x

fuel

eqD

where:

Consequences of the neutron balance features

In the case of a PWR, Nex is ~0

In the case of an FR, Nex is >1.0, i.e. there is an excess of neutrons that can be

used:

to convert the non-fissile U-238 isotope (99.3% of uranium) into

Pu-239, which is fissile

and/or to « transmute » specific elements (e.g. nuclear wastes)

The « neutron excess » Nex (in neutrons/fission) can be obtained as follows:

Nex = -Cpar- CFP- L

where:

Cpar is the total « parasitic » captures (per fission) in the core (structures,

coolant);

CFP is the total captures (per fission) of the fission products and

L is the total neutrons leaking out of the core (per fission)

fuel

eqD

As first discovered by Enrico Fermi back in 1944, the nuclear characteristics of TRU cross

sections in a fast neutron spectrum, as discussed previously, allow a great FR flexibility:

Breed (Conversion Ratio, i.e. ratio of TRU production/TRU destruction, CR>1)

Burn TRU, i.e. CR<1

Breed (e.g. Pu) and burn (wastes: e.g. MA)

CR~1: Self-sustaining cycles (i.e. fissile production=fissile destruction).

Wide coolant and fuel type choice according to the objective, e.g. short Doubling Time

CTD (i.e. the time required for a breeder reactor to produce enough material to fuel a

second reactor) : Na and dense (e.g. metal) fuels

Wide range of MA content and different Pu vectors or TRU compositions can be

handled.

Fast reactors flexibility

FRs have a unique potential to keep a large range of fuel

cycle options open leaving a limited legacy of highly

radiotoxic and radioactive material.

Fuel cycle issues should however be carefully analyzed

naf

n

af

n

af

n

iaf

i

89

8

core

n i

n

i

n

ii

F

ACIBGGAINBREEDINGINTERNAL 1

n

nFF is the total fission rate in the core

region n

is the total absorption rate of fissile

isotope i in core region n

is the total capture rate of fissile isotope (i-1) in core region n

n

iA

n

iC 1

The internal breeding gain is defined as the ratio of the net gain (i.e. production minus

destruction) of fissile material to the net destruction of fissile material in the core:

i: fissile isotope index

n: core region index

Breeder FR (CR>1)

The ω values characterize the reactivity of each isotope in a scale where the

„value“ of Pu-239 is 1 and that of U-238 is zero.

Adding an external blanket in order to „capture“ the neutrons in

excess, one can reach a total breeding gain (TBG) within a wide

range of values.

TBG = IBG +

A good breeding translates (together with high power density)

into „short“ doubling times CDT, i.e. the time required for a breeder

reactor to produce enough material to fuel a second reactor :

TBG365fP

T/fT1MTD

th

c

blanket

n i

n

i

n

1ii

F

AC

M mass of Pu

Pth thermal power

Tc out-of-pile time

T core residence time

f loading factor

In any case, a core with IBG~0 will help to achieve very long irradiation times

CDT

Relation between TRU consumption rate and TRU fraction in critical Advanced

Burner Fast Reactors:

-0.1

0.1

0.3

0.5

0.7

0.9

1.1

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

TRU fraction

No

rmalized

TR

U c

on

su

mp

tio

n r

ate

Metal, MA/Pu~1 feed

Oxide, MA/Pu~1 feed

Metal, LWR-TRU feed

Oxide, LWR-TRU feed

70-80% of max. theoretical

consumption can be

obtained with:

TRU/(U+TRU) ~0.4-0.6

both for metal or oxide

fuelled cores and for a wide

range of Pu/MA ratios

Varying the ratio TRU/(TRU+U) one can reach the maximum theoretical consumption of TRU:

Burner fast reactors CR<1 (or IBG<<0):

Alternatives: the ADS

Potential safety problems in the case of a critical core loaded with only TRU and

with a high content of Minor Actinides.

In these types of cores, the absence of Uranium produces both a very low fraction

of delayed neutrons and a very low Doppler reactivity coefficient (in general,

mostly due to U-238 capture).

Moreover, a high content of Minor Actinides like Am isotopes and Np induces a

deterioration of the void reactivity coefficient (in case of liquid metal coolants).

Sub-critical systems (or Accelerator Driven Systems ADS), were “rediscovered”

(~1985), since they could provide a possible way out from these potential

difficulties. Concept still to be demonstrated.

Alternatives (other than ADS):

Deep burners: HTRs; IMF-LWRs and Increased Moderation LWRs:

In all cases, the neutron spectrum can be considerably softer than a standard PWR,

and consequently low fission-to-absorption ratios for « gateway » isotopes (Pu-242, Cm-244

etc.) will induce very significant build-up of higher mass actinides (up to Cf isotopes)

As a reminder, there is a high Cf-252 production during the irradiation of TRU fuel in a

standard LWR with respect to a FR:

0

0.02

0.04

0.06

0.08

0.1

0.12

0 2 4 6 8 10 12

Cf-

25

2 k

g in

th

e c

ore

number of cycles

0.0E+00

1.0E-06

2.0E-06

3.0E-06

4.0E-06

5.0E-06

0 2 4 6 8 10 12 14 16

Cf-

25

2 k

g in

th

e c

ore

number of cycles

Cf-252 inventory in the core. Case of full TRU

multirecycle in a LWR

Cf-252 inventory in the core. Case of full TRU

multirecycle in a FR

Reactor

type PWR FR

Fuel type

Parameter

MOX

(Pu

only)

Homog

TRU

recycle

Pu only

Homog. TRU

recycle,

CR=1 and

MA/Pu~0.1

Homog.TRU

recycle,

CR=0.5 and

MA/Pu~1

Decay heat 1 x3 x0.5 x2.5 x38

Neutron

source

1 x8000 ~1 x150 x4000

Fuel cycle issues for FR with low conversion ratio

Feasibility issues can arise when considering not only the core feasibility but also fuel cycle performances.

E.g. in the case of decay heat and neutron production after post-irradiation cooling (at fuel fabrication)

At present, this option is a reference for FRs in some OECD countries with non-

proliferation concerns

Alternative: LWRs with CR~1 (harder spectrum).

Many studies in the past. Difficulty to overcome problem of positive void

coefficient, in particular for “degraded” Pu vectors.

New studies in Japan; the RBWR of Hitachi (CR~1 and negative void

coefficient)

Moreover, little hope to burn MA, since they can degrade further the void

coefficient value.

Isogenerator or break-even FRs: CR~1 (or TBG~0)

Fast reactors and multiple recycle allow sustainability in terms of resources optimal

utilization.

Uranium utilization without reprocessing has been envisaged since an early proposal by

Teller, and more recently by the Travelling Wave Reactor proposal of Terra Power

However, no miracle solution can be found with any once-through cycle

Moreover, used fuel if put in a repository will have comparable characteristics (i.e. activity,

residual heat, radiotoxicity etc.) as the used fuel of a standard PWR based once through

cycle.

A comprehensive analysis has been performed at ANL

Fast Reactors and close fuel cycle

Once-Through

Nuclear

Systems

PWR TWR

Uranium

utilisation % 0.6 ~2-5

Travelling Wave Reactor, TWR

Fast reactors allow a great flexibility in the choice of nuclear energy deployment

strategies

That flexibility can be used to design evolutionary fast reactor cores that can burn or

breed TRU according to the objective. This can be done in principle in the same vessel

(reversibility concept) and without degrading the core safety characteristics

Practically any type of Pu vector and Pu/MA composition ratio can be accepted in the

core

Different fuel forms (oxide or dense fuels) can be used, according to the objective

Fast reactors should be conceived within close cycle strategies, in order to maximize

benefits with respect to sustainability and waste minimization

Conclusions (1)

Fuel cycle issues are crucial in order to assess the feasibility and the economy of a specific

strategy:

Fuel reprocessing with very small losses in the TRU recovery is mandatory (e.g. 99.9%

recovery of any TRU isotope)

Build-up of higher mass actinides (Cm, Bk, Cf isotopes) can be a heavy burden at fuel

handling, fuel fabrication etc., with a potential impact on reactor availability and fuel

cycle optimization. This should be investigated in practical applications.

Multirecycle can hardly be avoided: any once-through approach will be limited by the

maximum achievable fuel burn-up

Molten salt systems with fast neutron spectrum and on-line fuel processing or other mobile

fuel concepts (not discussed here) could offer extra gains in terms of potential fuel cycle

simplifications and it could still be worthwhile to (re)-explore them

Conclusions (2)

Back-up

Thermal and fast reactors: EPR (thermal)

Cutaway of reactor pressure vessel of EPR (AREVA). EPR fuel SA (AREVA)

Thermal and fast reactors: SFR designed before 1990s

Source: A. Walter and A. Reynolds, Fast Breeder Reactors

Fertile blankets surround the core to increase BR

PWR with different moderator-to-fuel ratios

Vm/Vf = 3

(overmoderated)

= 2

(standard)

= 1.1

(high CR PWR)

FR

U235 0.236 0.317 0.403 0.312

U238 7.39 6.42 5.62 8.65

Np237 51.9 34.0 21.2 5.74

Pu238 10.5 4.77 2.30 0.552

Pu239 0.546 0.561 0.564 0.337

Pu240 77.0 40.0 16.8 1.89

Pu241 0.330 0.318 0.292 0.207

Pu242 33.2 23.2 12.9 3.23

Am241 74.3 48.7 27.9 8.55

Am242 0.232 0.220 0.194 0.211

Am243 89.3 69.2 44.1 9.82

Cm242 4.03 4.21 3.97 1.067

Cm243 0.183 0.175 0.176 0.074

Cm244 15.6 14.1 11.1 1.53

Cm245 0.147 0.147 0.152 0.127

Ratio of capture-to-fission

average cross sections for

different types of spectra

Vm

Vf