S. Mirnov

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FUNFI Fusion for Neutrons and Sub-critical Nuclear Fission
Villa Monastero, Varenna Italy September 12-15 2011
What we should do for transition from current
tokamaks to fusion-fission reactor
(From fusion romance to reality)
S. Mirnov
GNC RF TRINITY 142 190 Troitsk Moscow Reg. RUSSIA
2011
Introduction. Requirements shown by Russian
atomic engineers to the parameters FNS-1
1.Decision of steady state tokamak operations.
The superconductor and warm magnets systems
Current drive.
2.The problem of plasma-wall interaction
Phenomenological limit of power load on tokamak first wall.
The neutron flux limits in FNS-1.
Lithium as a tokamak PFC protector
3.Creation of hot plasma zone with high neutron emission
The control of plasma density
Impurity control in the center of tokamak plasma
4. Other candidates on the role of fusion neutron source type of FNS-1
The mirror traps (GDT)
Stellarators (LHD)
Superconductor tokamaks (Tor Supra, EAST, KSTAR)
Spherical torus. (NSTX, MAST)
Conclusions
Fusion-Fission roots in Russia
1956 Sakharov A. Memoirs Vintage Books, New York (1990)
142
I.N.Golovin, V.I.Pistunovich, G.E.Schatalov Preprint IAE 1973
Physical basis of tokamak-reactor with NBI (First hybrid FF)
I.N.Golovin, G.E.Schatalov, B.N.Kolbasov. Isvestiya AS USSR
1975 Energy and transport № 6 p28-34. .
USA/USSR Symposium on Fusion-Fission Reactors,
Lawrence Livermore Laboratory CONF-760733 July 13-16
1976.
USSR/USA Seminar 14March-1April 1977 M Atomisdat 1978
ITER Physics basis. Nuclear Fusion v39 N12 1999
Progress in ITER Physics basis. Nuclear Fusion v47 N6 2007
τE, 98 = 0,0365.Iр 0,97.BТ 0,08.PH -0,63.n0,41.M0,20.R1,93.(a/R)0,23.k0,67 sec
Potential using of fusion-fission
in the electricity production (for Russia)
Joint commission of 2009Y :
Plasma physicists (TRINITI,
KURCHATOV, IOFFE Inst.) and
Atomic engineers (DOLLEJAL,
KURCHATOV Inst. suggested
the three steps scheme of
Russian FNS development
1.Creation of the hydrogen TNS prototype – TNS-0
on the T-15 basis for the investigation of steady
state tokamak physics
2. Creation of the tokamak- breeder prototype –
TNS-1 - with neutron power 20-50MW and fuel
production equal 20-100 kg p/y
(2020-2025yy)
3.Creation of the commercial tokamak- breeder
prototype – TNS-2 with neutron power 100200MW and fuel production up to 1000 kg p/y
(2025-2035yy)
The Russian fission community has made special
requirements to the fusion neutron source FNS-1 for the first
step of investigation.
1. quasi-steady state (with availability Kt >80%) DT fusion
reactor operation
2. surface neutron load pn not lower 1017n/m2sec (0.2MW/m2)
3. total neutron output not lower Pn=20MW.
The main aims of FNS-1 :
demonstration of industrial probabilities of such neutron
source, with ability of nuclear fuel producing on the level
100kg for the test of different versions of experimental
blanket modules.
(So called “ Lopatkin’s requirements” at name of deputy
director of Dollejal institute A.V.Lopatkin )
Middle-scale device with the total cost not higher than $1bn.
No activity in direction of sub-critical Nuclear Fission from safety
problem.
What we should do for
transition from current fusion
devices to FNS-1?
Comparison of the total neutron energy production Qn per day: for FNS-1,
ITER (project), JET (should be multiplied up to 2000 times dotted array–
proposal steady state operational regime), and NIF (proposal regime,
should be multiplied up to 200000 times )
The superconductor or warm
magnets systems?
The creation of FNS-1 on the basis of a warm
magnets system (Cu, Al) is considered with
superconductor if one takes into account that a
limited time period (about 1 year ) will be sufficient
for the FNS-1main task – production of 100kg of
fuel by a steady- state fast neutron source with the
total neutron power not lower than 20MW and
density 0.2MW/m2.
Within this time the system will accumulate in its
units the relatively low neutron fluency (3 1024n/m2).
The 30% of such neutron fluency will be enough for
production of the planned 100 kg of fuel
Magnet (Cu) system of VNS-1 project (TRINITI-Kurchatov 1993Y)
1-poloidal, 2-toroidal Cu coils. 3-concret tank
a/R=0.7/2 m, B0 =3.5T, Jp =4MA P=300MW
The problem of plasma-wall
interaction
Phenomenological limit of
power load on the tokamak first
wall
A-dynamics of neutron production- Pfus
and PH,
B-PH in linear scale, crosses
- PH/S.
The tokamak praxis shows the existence
of rather hard phenomenological limit of
power load on the tokamak first wall
qc ≈0.2±0.1МW/m2
What happened after the violation of this
limit?
TFTR . Evolution of DT fusion power and plasma stored energy for
a series of plasmas with mixed D and T NBI. One or two lithium pellets
were injected into the plasma prior to NBI.
This limit will confine the maximum value of the neutron flux
density pn on the first wall of the tokamak-reactor.
In the most favorable case of burning tokamak the relation
between the total power of neutron production Qn to the plasma
heating power by α-particles PH (as is known) is equal to 4.
It means that for the burning reactor type DEMO limit
qc= 0.2МW/m2 is equivalent to limit pn=0.8МW/m2
This limit seems too low for a pure fusion reactor, but not for a
fusion-fission one.
For example, the fusion-fission burning tokamak scale ITER with
natural uranium blanket can produce up to 3GW electric power
and at the same time produce the fuel for fission reactors needed
for generation additional 10GW of electricity.
The neutron flux limit in FNS-1
In FNS-1 case the situation is complicated by the fact
that two sources of fast neutrons should be used in it
(as in burning tokamak).
The FNS-1 has two neutron producers – one part of
neutron output should be generated during the direct
interaction of high energy ion beam (100-300keV) with
the target - DT (1:1) tokamak plasma (QBF)
The second part should be the result of DT fusion in the
hottest target plasma (QTF). In real prototypes of FNS-1
– TFTR and JET both these fusion sources were
considerable.
In the steady state conditions the total heat flux to the first wall will
be the summary results of both sources of fusion α-particles and
NBI power.
If we write QBF as αРН , where α is function, connected NB power
PH with fusion output from direct interaction of ion beam with
target plasma, the summary neutron output can be written, as
Qn= 0.8(αРН + QTF)
and power of total heat flux as
РН + 0.2 (αРН + QTF).
That means:
pn/q = 0.8(α + QTF/РН)/ (1 + 0.2 (α+ QTF/ РН))
For burning tokamak (PH=0) we have again pn/q =4.
For the opposite case QTF =0 (case of “pure target”)
pn/q = 0.8α / 1 + 0.2α
The calculated α values as function of electron
temperature Te for different energies of injected Datoms to the clean (Zeff ≈1) DT (1:1) target plasma
(V.I. Pistunovich 1976)
The achievement of values α ≈1 in average-scale
tokamaks seems realistic.
In this case pn/q from can be 0.67 and if we take into
account the limit qc = 0.2МW/m2, the pn value should be
equal 0.13 МW/m2, which is lower of neutron density
requirements 0.2МW/m2 for FNS-1.
It is obvious that in order to increase the neutron
load tokamak with visible QTF output should be
chosen.
In particular, with QTF/РН = 1 the pn value can be
increased up to 0.26 МW/m2.
Another possible way is increasing qc by mitigation of
heat contact plasma-wall effect.
It is known that the most aggressive form of such contact is
the first wall bombardment by hot plasma during
development of plasma boundary instabilities (ELMs, Blob’s)
with high local energy loads and, as a result, with active
erosion of the first wall materials.
To smooth the local energy loads some experimentalists try
to reradiate plasma energy flux by injecting radiated
impurities into the plasma periphery.
The role of such kind impurity can play Li. It should be
pointed out that Li pellet injection was actively used in TFTR
Lithium as a tokamak PFC
protector
J.Bohdansky and J.Roth
Temperature dependence of sputtering behavior of Cu-Li alloys
Nucl. Instr.and Methods in Physics Research B23 (1987) 518
The main surprise of numerous tokamak lithium experiments
was the discovery of the poor lithium penetration to the hot
core from plasma periphery (lithium screening).
The effective ion charge in plasma center - Zeff(0) which had
been equal to 2 or higher (TFTR, T-11M, FTU, T-10, CDX-U,)
dropped down to 1 after first wall lithiaton.
The mechanism of lithium screening is not fully clear.
Probably that is consequence of deep gap between lithium
first (5.8eV) and second (75eV) ionization potentials.
The lithium screening effect can be served as a basis of
concept of permanent lithium circulation close the
tokamak first wall for their protection from the direct
plasma bombardment
FTU-experiment (M.L.Apicella et al.
2005)
Lithium splashing problem
Capillary Porous Structure
with (A) and without Li (В)
Scheme of Li circulation “emitter-collector”
FTU experiment. Li CPS limiter after plasma exposition
No Surface Damage
Li migration
by CPS
W
CPS
Creation of hot plasma zone with
high neutron emission
The control of plasma density
The record TFTR DT shot with NBI PH= 40MW
( PH/S=q ≈0.5MW/m2, pnmax ≈0.12MW/m2)
In steady state tokamak all fluxes of neutral atoms in
the plasma column should be balanced by their
outward diffusion, neutralization and pumping.
The atoms of NBI and He should be removed by the
divertor with expanded divertor plates (for example, by
“snowflake” type) probably with lithium coating.
For the He pumping and creation of Li-circulation in
divertor SOL a mushroom like pump limiter shown in
could be used.
The mushroom like pump limiter for the He pumping and
creation of Li-circulation in divertor or limiter SOL
Impurity control in the center of
tokamak plasma
As is shown by the experience of ECRH use in small
and average – scale tokamaks, the local ECRH
permits not only increasing Te, which increases the
fast ion relaxation time but also promoting impurity
removal from plasma center and increasing the
diffusion of DT ions. It is supposed to stabilize Zeff in
plasma center on the level of 1-1.2, which is needed
for achievement of α≈1.
The next way of the α increase (Fig.5) can be
bringing up the energy of the injected atoms from
110-120keV (TFTR, JET) up to 150keV.
Other candidates on the role of fusion
neutron source type of FNS-1
The mirror traps (GDT)
Stellarators (LHD)
Superconductor tokamaks (Tor Supra, EAST,
KSTAR)
Spherical torus. (NSTX, MAST)
The comparison of different types of magnetic
fusion devices as candidates on the role of FNS-1
Spherical torus.
Jp≈(5εaB0/qψ(95%))[1+k2(1+2δ2-1.2δ3)] (1.17- 0.65ε)/(1-ε2)2 /2,
The formal substitution in this equation ε = 0.7-0.8 and best
parameters of high performers “classical” (ε<0.4) tokamaks qψ(95%)=3, k=3, δ=0.8 3, k=3, -gives the above mentioned profit
in Jp up to 3-5 times. That is a mistake.
The right side of this equation consists of two kinds of
parameters. Their first part seems independent, hard
determined by experimentalist. That is ε, a, B0, which are the
“material” condition of experiment defined its scale and cost.
The second group of “plasma connected parameters” - qψ(95%),
k and δ seems as internal depended.
. Really each experimentalist should find an
optimal combination of these parameters with
the main goal to obtain the maximal plasma
current.
If this combination is universal, the maximal
Jpmах should be proportional
εaB0(1.17- 0.65ε)/(1-ε2)2,
The maximal values of Jp, received on 4 leaded spherical torus –
as function of εaB0 (1.17- 0.65ε)/(1-ε2)2 [14-19],
cross NSTXnew (B0=0.3T, Jp=1MA) – private communication
The next serious disillusion was brought by the
unexpected effect of spontaneous breaking of discharges
in spherical torus, which was observed simultaneously in
both leaded devices – NSTX and MAST.
The nature of this effect is not yet clear today. Probably that is
effect, connected with too high βТ in spherical torus.
As a result, the shot duration was limited in these devices
by 1-1.8 sec instead of 5sec, proposed by projects.
That means the spherical torus with ε = 0.7-0.8 lost the main
advantage of tokamaks – the potential ability of steady state
operations.
NSTX and MAST- effect of discharge interruption
J.E.Menard, M.G.Bell,e.a. Proc.21st IAEA Fusion Energy Conf., Chengdu, China, 2006 OV/2-4.
The f(ε)= (1.17- 0.65ε)/(1-ε2)2 function v.s. ε and R/a.
(For consistency with a “classical tokamak” region it is divided into1.17).
Few “ classical tokamaks”, spherical torus
and some FNS projects are shown by arrows.
Conclusions
The FNS program, developed parallel to ITER
activity, can essential approach the entry DTfusion in the commercial power
The initial technology requirements
(p≥0.2MW/m2, P≥20MW/m2, Kt≥0.8), shown by
Russian atomic engineers to the parameters
of the first stage of fusion neutron source
FNS-1, can be met under condition of
successful improvement of the existing
middle-scale tokamaks parameters in the
following main directions:
1. development of steady state (or quasi-steady state)
tokamak operations by learning to use non ohmic
current drive methods with a simultaneous
organization of a closed D,T- loop circulation and
He removal,
2. increase of energy NBI up to 150 and more keV,
3. lower level Zeff up to 1-1.2
(with possible use of Li technologies),
4. active use of ECRH for:
a) control of plasma density,
b) control of a impurity level,
c) heating of electrons and weakening of ion beams
relaxation,
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