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