# Reference Scenario for an Advanced Deuterium Power Plant System Page: 2 of 4

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TABLE 1.

The percentages of power in the neutrons from the plasma for various tritium percentages extracted (50 to 90%) and for 50%

and 100% of this tritium returned as helium-3.

Tritium 0% 50% 60% 70% 80% 90%

Removal

Percentage 0 50 100 50 100 50 100 50 100 50 100

Returned

2.45 MeV 5.7 6.3 5.6 6.4 5.6 6.6 5.6 6.7 5.6 6.9 5.6

Neutron %

14.1 Mev 32.6 18.1 16.2 14.8 12.9 11.3 9.7 7.7 6.4 4.0 3.2

Neutron %III. REFERENCE TOKAMAK POWER PLANT.

Large tokamak power plants are used to illustrate the approach,

because it is possible to use "ITER rules" [4] to develop a

consistent system. In reality, at large scale - 5000 to 6000 MW

thermal - other systems may turn out to be superior. The design

of the plant and the general approach is consistent with the

Wildcat study [5], calculations by Houlberg and Attenberger [6],

and recent calculations by Paul Rutherford [7]. A preliminary

calculation suggests that it should be possible to extract 70%

or more of the tritium using radio-frequency waves at _

15% of the electrical output. Several mechanisms have been

proposed by which ion cyclotron waves can be employed to

produce fluxes of cyclotron resonant particles. C. S. Chang

has studied a transport mechanism resulting from

asymmetric friction driven by waves with a non-symmetric

ki spectrum. Other studies have considered transport due to

the direct effect of RF on particle orbits through

perpendicular heating, or coupling of parallel momentum.

Particle fluxes from each of the mechanisms scale in a

similar way. To obtain a rough estimate of the power

required we approximate the RF drive tritium flux as [ref 1,

eq (14)]

,rf Pr P

20

where ppis the poloidal gyroradius, C0is the characteristic

tritium particle energy and P is the RF power deposition

density in the transported particles. We assume that most

of the tritium produced is pumped out before it reacts. If

that is the case then in steady state the particle flux at each

radius must balance the tritium production inside that

radius. A more detailed analysis is planned.

A. Example Tokamak parameters

R = 9 m, R/a = 2.75, a = 3.27 m, K = 2.4, 6= 0.65, 3N =5,

q95 = 3.1. Bc01 = 14 T, B = 7.31 T, the separation from the

toroidal coil to the plasma on the inside is 1.03 m.

I = 84.5 MA, 3= 17.7 %.

ne = neo [1 - a2/r2]05 ; at start up nD is 0.7 ne, at equilibrium

it is 0.65 ne, ,

The average electron density is 2.64 x 1020, the

Greenwald limit is 2.52 x 1020.

Te = Teo [1 - a2/r2]1 7 , T, = To [1 - a2/r2]20.

The average electron temperature is 34 keV, while the average

ion temperature is 37 keV.

At start up Zeff = 1.3 and at equilibrium it equals 1.4.Note that T, > T, for r/a < 0.7. In the central region it is

assumed that 30% of the fast ion power went to the ions and

70% to the electrons. A large chunk of the ion power is then

transferred by collisions to the electrons, because they need to

handle the bremsstrahlung and synchrotron and line radiation.

It is assumed that, in the center, the ion thermal conductivity

is less than that of the electrons (about a factor of five). The

pressure due to fast particles is assumed to be 7% of the total

thermal plasma pressure.

B. Power and power fluxes

The equilibrium charged particle power is 4,600 MW + 560

MW of neutrons. With the Wildcat blanket gains, there would

be a total of 1,940 MW of neutron generated power.

Bremsstrahlung power is 1,500 MW and synchrotron radiation

power is 930 MW at a wall reflectivity of 0.85. In addition there

will be auxiliary power of a few 100 MWs and line radiation

( hopefully mainly from the edge region).

In the equilibrium case, assuming that it would be possible

to transport 50% of the synchrotron radiated power out of

the vacuum vessel, the thermal load on the wall would be

about 1.95 MW/m2. This is a little high, but the neutron

flux would only be 0.3 MW/m2. Furthermore, the 14.1

MeV flux would be only about 0.1 MW/m2. One can hope

that in this situation the wall and shield might survive at

least 30 years. When allowance is made for the start up

phase, in the first plant, the 14.1 MeV part is on average

about 1.5 times higher, but this would still only amount to

a 14.1 MeV neutron fluence of 4.5 MW.y/m over thirty

years. The total neutron fluence is 10.5 MW.y m2 over 30

years. In later plants it will be even less.

In a deployment of such power plants, the first plant might

run for 3 years with 50% tritium removal, then 2 years with

75% removal before reaching the ultimate 90% tritium

removal. With helium-3 recycled the year after it is produced,

this system reaches an equilibrium with 50% of the extracted

tritium returned as helium-3 after 15 years. Delaying recycling

helium-3 for a year should account for any real-world

innefficiencies. In another variant, 90% of the tritium might

be extracted from the start.

At year 16, two new plants could start up using surplus

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Sheffield, J. Reference Scenario for an Advanced Deuterium Power Plant System, article, September 17, 2001; Tennessee. (digital.library.unt.edu/ark:/67531/metadc723736/m1/2/: accessed February 15, 2019), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.