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December 20, 2011

Inertial Electrostatic Confinement Fusion 2011 Workshop

US-Japan Inertial Electrostatic Confinement Fusion 2011 Workshop presentations

Joel Rogers presented simulations of a p-11B Polywell power reactor that would reach breakeven at diameter 13 meter. (15 pages)

Last year Joel Rogers had an inferior design simulated that had a breakeven diameter of 300 meters.

Some of the improvements are described in Joel Rogers 2010 patent (Modular Apparatus for Confining a Plasma)

Aneutronic fusion is the holy grail of fusion power research. A new method of operating Polywell was developed which maintains a nonMaxwellian plasma energy distribution. The method extracts downscattered electrons and replaces them with electrons of a unique higher energy. The confined electrons create a stable electrostatic potential well which accelerates and confines ions at the optimum fusion energy, shown in the graph below. Particleincell(PIC) simulations proceeded in two steps; 1) operational parameters were varied to maximize power balance(Q) in a small scale steadystate reactor; and 2) the small scale simulation results were scaled up to predict how big a reactor would need to be to generate net power. Q was simulated as the ratio of fusionpoweroutput to drivepowerinput. Fusionpower was computed from simulated ion density and ion velocity. Powerinput was simulated as the power required to balance nonfusing ion losses. The predicted breakeven reactor size was 13 meter diameter. Bremsstrahlung losses were also simulated and found manageable.




Reactor Break-Even Radius
● Bussard's Scaling Formula: Q1/Q2 = (R1/R2)5 [8]
● Break-Even Formula: Q(R=35cm)/Q(Rb) = (R/Rb)5
● Q(Rb) ≡ 1
● Solving for Break-Even Radius: Rb = R/Q1/5
● Rb = 0.35m/(4.1e-7)0.2 = 6.6 meter = smaller than ITER

* Bremsstrahlung losses ≈ 1/3 fusion output power

● To reduce Pb the reactor design can change:
● Reducing Te to 1% Ee would reduce Pb by 4.5X. [4]
● Boron fraction nb/np 20 -> 10% would reduce Pb by ~2X.
● Reducing Te might increase reactor size (Rb).
● Not yet tested in simulation.
● Radiation might be reduced to 5% of fusion power.

● New method efficiently recycles electron energy.
● Simulation predicts break-even Rb = 6.6m
● Additional design issues still need attention:
● Electron power drain must be reduced.
● Bremsstrahlung power drain must be reduced.
● A 3D simulation is needed for more realistic Pin.
● The future of aneutronic fusion power is bright.



IEC Fusion Thruster

Adam Israel (University of Syndey) presented a charge exchange thruster design

A new type of electric propulsion system is presented whereby an asymmetric hollow
cathode glow discharge is created within the thruster and ions are accelerated toward the exit nozzle. A novel characteristic of this thruster is the neutralization of the ions via charge-exchange reactions with the background gas, which is the dominant ion-neutral reaction at operating voltages of the order of ten- kV and currents of mA. The thruster is entirely self-contained with ions created, accelerated and neutralized within cathode. Langmuir probe measurements have shown the existence of a sharp potential 'ramp' within the cathode with a maximum potential drop of approximately 90% of the applied cathode voltage, resulting in highly energetic and collimated beams of neutral atomic hydrogen ejected from the cathode. Doppler-shift spectroscopy was used to measure the speed of the exiting neutral atoms which gave a specific impulse and thrust to be of the order of 10,000 seconds and 100 microNewtons respectively for hydrogen gas.

Overview of University of the Wisconsin IEC Research Program-2011 (22 pages)

•Considerable progress has been made in experimental facilities
–Adjustable Arm for FIDO and TOF measurements
–Argos chamber for pulsed neutron generation
–300 kV switch for rapid changeover of 4 IEC devices
–New design for 300 kV feed-through to avoid insulating stalk failures

•IEC Technology spinoff has been used for a materials irradiation facility


Future Goals of UW IEC Program
•Understand the role of negative ions in the spatial distribution of DD fusion events.
•Apply the TOF adjustable arm diagnostic to D3He fusion.
•Test plasma facing component materials at higher temperatures and a wider range of fluences.
• Investigate D3He fusion in 6-Gun SIGFE device.
•Analyze test results from STOROID and PING pulsed neutron facilities.
•Compare VICTER theoretical predictions with experimental data from HOMER.
•Increase the He+ source strength of HELIOS.
•Test pulsing effects on SIGFE.


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