FUNFI Fusion for Neutrons and Sub-critical Nuclear FissionVilla 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-11.Decision of steady state tokamak operations. The superconductor and warm magnets systemsCurrent 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 emissionThe control of plasma densityImpurity control in the center of tokamak plasma4. Other candidates on the role of fusion neutron source type of FNS-1The 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

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

ITER Physics basis. Nuclear Fusion v39 N12 1999Progress in ITER Physics basis. Nuclear Fusion v47 N6 2007

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 100-

200MW 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 tanka/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 B-PH in linear scale, crosses

and PH, - 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 D-

atoms 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

by CPS

CPS

W

Li migration

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

Jp≈(5εaB0/qψ(95%))[1+k2(1+2δ2-1.2δ3)] (1.17- 0.65ε)/(1-ε2)2 /2,

Spherical torus.

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 DT-

fusion 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,