Our paper gives an overview of the Princeton Field-Reversed Configuration (PFRC) fusion reactor concept and includes the status of development, the proposed path toward a reactor, and the commercialization potential of a PFRC reactor.
The Journal of Fusion Energy features papers examining the development of thermonuclear fusion as a useful power source. It serves as a journal of record for publication of research results in the field. This journal provides a forum for discussion of broader policy and planning issues that play a crucial role in energy fusion programs.
Our latest paper on DFD applications, “A Fusion-Propelled Transportation System to Produce Terrestrial Power Using Helium-3 From Uranus”, is now available from AIAA. This paper was part of the Future Flight Propulsion track and AIAA SciTech 2023. For those with AIAA membership, there is a video recording of the presentation as well! Download the paper here.
Our goal with this paper is to create a framework within which we can study the potential cost of electricity produced on Earth using helium-3 mined from Uranus. The scarcity of terrestrial helium-3, along with the radioactivity of methods to breed it, lead to extraterrestrial sources being considered as a means to enable clean helium-3 fusion for grid-scale electricity on Earth.
This paper builds on the work of Bryan Palaszewski who has published numerous papers on mining the atmospheres of the outer planets. Palaszewski’s work assumed fission-based power and propulsion systems, with a much lower (worse) specific power than we anticipate from a PFRC-based Direct Fusion Drive. We consider both transport and mining vehicles that are instead fusion-powered, including a fusion ramjet. This ramjet may be able to be both the mining vehicle and the orbital transfer vehicle to bring the refined helium-3 to the interplanetary transport,
The results allow us to estimate levelized cost of electricity, LCOE, for the electricity produced on Earth as a function of assumed cost of the fusion transports and mining system, cost of the PFRC reactors, amount of helium-3 stored on each transport and numbers of trips per year, etc. You can learn more about LCOE from the NREL website. Uranus is likely the most economical outer planet for mining due to its lower gravity and radiation environment and high concentration of helium in its atmosphere, about 15%. We find that with our set of assumptions, the resulting cost of electricity could potentially be competitive with wind and solar.
Future work will include analysis of the fusion ramjet trajectories between mining and transfer altitudes, and research into sizing a mining payload using membranes and adsorption to separate the helium-3 from the helium, rather than depend on heavy cryogenic techniques.
The internet was abuzz last week with the news that the National Ignition Facility had achieved that elusive goal: a fusion experiment that achieved net (scientific) energy gain. This facility, which uses 192 lasers to compress a peppercorn-sized pellet of deuterium and tritium, released 3 MJ of energy from 2 MJ of input heat.
We have to use the caveat that this is “scientific” gain because it does not account for the total amount of energy needed to make the laser pulse. As a matter of fact, the lasers require 400 MJ to make those 2 MJ that reach the plasma. If we account for this energy, we can call it the “wall plug” gain or “engineering” gain since it includes all the components needed. This gain for laser-induced fusion is still less than 1%, because the lasers are very inefficient.
Nonetheless, this is great news for all fusion researchers. Since we often get asked: Has anyone achieved net (scientific) gain yet? Now we can say: Yes! It is physically possible to release net energy from a fusing plasma, to get more energy output than direct energy input. This advance has been achieved through various new technology: machine learning to select the best fuel pellets, wringing more energy from the lasers, more exact control over the laser focusing. Modern technology, especially computing for predicting plasma behavior, explains why progress in fusion energy development is now accelerating.
Tokamaks have also come close to net gain, and in fact the JT-60 tokamak achieved conditions that could have produced net gain, if it had used tritium .
The reason JT-60 did not use tritium in those shots is very relevant to our fusion approach, the PFRC. Tritium is radioactive, rare, expensive to handle, and releases damaging neutrons during fusion. Tritium is also part of the easiest fusion reaction to achieve in terms of plasma temperature, the deuterium-tritium reaction. It makes sense for fusion experiments to use such a reaction, but this reaction presents many difficulties to a future working power reactor.
The PFRC is being designed to burn deuterium with helium-3, rather than with tritium, precisely to make the engineering of a reactor easier. The deuterium-helium-3 reaction releases no neutrons directly. Some deuterium will fuse with other deuterium to produce neutrons and tritium, but the PFRC is small enough easily expel tritium ash. This results in orders of magnitude less neutrons per square meter reaching the walls. Once we have scientific gain, like the NIF has now demonstrated for laser fusion, we have an easier path to engineering gain — that is, net electricity.
So while the laser fusion milestone doesn’t directly impact our work on the PFRC, it is important to the field. We will continue to follow the progress of all our peers as we work to achieve higher plasma temperatures in our own experiments!
 T. Fujita, et al. “High performance experiments in JT-60U reversed shear discharges,” Nuclear Fusion 39 1627 (1999). DOI: 10.1088/0029-5515/39/11Y/302
Last week, I attended the American Physical Society Division of Plasma Physics (APS DPP) 2022 Meeting. As the name entails, it was a meeting full of plasma physics with applications ranging from astrophysics to nuclear fusion energy. There were many great talks and posters on plasma physics research by companies, national labs, and universities, and one could sense an overall feeling of excitement around fusion shared by many attendees.
I had a pleasant time in Spokane, WA. Pictures from outside of the conference center (with many conference attendees standing nearby), including the nice view from the conference center, are shown below.
I presented a talk on the Princeton Field-Reversed Configuration (PFRC) fusion reactor concept, and how we can leverage public-private partnerships for its development. The talk discussed technical details of the PFRC, including the past modeling and experiments, current investigation, and future research & development plans. The talk also described the markets and commercialization opportunities for this reactor concept, including disaster relief and asteroid deflection. Here I am at the podium speaking.
I also presented a poster on our recent investigations of x-ray diagnostics on the PFRC-2 experiment for electron temperature and density measurements, which was mounted on a poster board in the conference center. Many people came by to ask about my poster as well as about general PFRC questions, which kept me talking for the majority of the 3-hour poster block session! It was great to discuss ideas and results with many scientists and students at the conference.
Dr. Sangeeta Vinoth also had a poster at this conference on collisional-radiative model developments to extract electron temperature measurements from spectroscopy, which she presented virtually. APS DPP 2022 was an exciting conference to attend, and I’m looking forward to seeing updates from presenters at this conference. That also includes us, as we have more research and investigation to do — stay tuned!
Last week, PSS Mike Paluszek visited ITER, the international fusion research experiment under construction in France. In light of Mike’s recent visit to ITER, we wanted to showcase an application of our tokamak Fusion Reactor Design function to the design of ITER. This function is part of the Fusion Energy Toolbox for MATLAB, a toolbox that includes a variety of physics and engineering tools for designing fusion reactors and studying plasma physics. We will also compute design parameters for ITER’s successor, the DEMOnstration power plant (DEMO), a fusion reactor currently in the design phase which is planned to achieve net electricity output.
We first apply the Fusion Reactor Design function to ITER. Note that ITER is expected to produce 500 Megawatts (500 MW) of fusion power, but this will not be converted into electric power, the power that goes into the electrical grid. DEMO, on the other hand, is planned to produce 500 MW of electric power from 2000 MW of fusion power. The Fusion Reactor Design function asks for the net electric power output of the reactor, P_E, as an input, so we generate a value for P_E for ITER by using the same ratio of electric-to-fusion power as in DEMO, giving us a P_E of 125 MW for ITER. The inputs used for the ITER design are shown below (see references [1,2]), where we use a data structure “d_ITER”:
d_ITER.a = 2; % plasma minor radius (m)
d_ITER.B_max = 13; % maximum magnetic field at the coils (T)
d_ITER.P_E = 125; % electric power output of the reactor (MW)
d_ITER.P_W = 0.57; % neutron wall loading (MW/m^2)
d_ITER.H = 1; % H-mode enhancement factor
d_ITER.consts.eta_T = 0.25; % thermal conversion efficiency
d_ITER.consts.T_bar = 8; % average ion temperature (keV)
d_ITER.consts.k = 1.7; % plasma elongation
d_ITER.consts.f_RP = 0.25; % recirculating power fraction
The first five inputs were described in our original post on the Fusion Reactor Design function. The function can be called to perform a parameter sweep over any of these inputs. We also specify values for some constants: the thermal conversion efficiency ‘eta_T’, the average ion temperature ‘T_bar’, the plasma elongation ‘k’, which is a measure of how elliptical the plasma cross-section is, and the recirculating power fraction ‘f_RP’. We can perform a parameter sweep over the minor radius (from a = 1.8 meters to a = 2.2 meters, with 100 points in between) and display a table of results simply with two lines of code:
d_ITER = FusionReactorDesign(d_ITER,'a',1.8,2.2,100); % run function
d_ITER.parameters % show table of resulting parameters
Looking at the results table from d_ITER.parameters, we see overall agreement with parameters for ITER [1,2]. The plasma major radius (essentially the tokamak radius) R_0 output is about 5 m, which is in the ballpark of the 6.2 m radius of ITER design, and the magnetic field at R_0 (on plasma axis) output is 4.8 Tesla, close to the ITER design value of 5.3 Tesla. The plasma current output is 17.5 MegaAmps, which is also close to ITER’s design of 15 MegaAmps.
The Fusion Reactor Design function also outputs plots that show whether or not the reactor satisfies key operational constraints for tokamaks, see the figure below. The first three curves check various constraints to ensure the plasma is stable, which we see are met as they are located in the unshaded region (though the green curve is marginally close to the constraint boundary). The blue curve’s position deep into the shaded region indicates that the reactor is far from producing enough electric current to sustain itself. The designers of ITER anticipated this, which is why ITER will additionally use a pulsed inductive current and test a combination of other techniques to drive the plasma current.
We now consider DEMO, which is in the design phase with the goal of net electrical power output. Similarly to running the ITER case, we set up a data structure (now called ‘d_DEMO’) with known DEMO input parameters  and perform a parameter sweep over the minor radius ranging from a = 2.7 meters to a = 3.1 meters:
d_DEMO.a = 2.9; % plasma minor radius (m)
d_DEMO.B_max = 13; % maximum magnetic field at the coils (T)
d_DEMO.P_E = 500; % electric power output of the reactor (MW)
d_DEMO.P_W = 1.04; % neutron wall loading (MW/m^2)
d_DEMO.H = 0.98; % H-mode enhancement factor
d_DEMO.consts.eta_T = 0.25; % thermal conversion efficiency
d_DEMO.consts.T_bar = 12.5; % average ion temperature (keV)
d_DEMO.consts.k = 1.65; % plasma elongation
d_DEMO.consts.f_RP = 0.25; % recirculating power fraction
d_DEMO = FusionReactorDesign(d_DEMO,'a',2.7,3.1,100); % run function
d_DEMO.parameters % show table of resulting parameters
The outputs for the DEMO case also show overall agreement with DEMO parameters . The plasma major radius R_0 output is 7.8 m, which is not far from the 9 m design radius for DEMO. The resulting on-axis magnetic field output is 6.2 T, close to the 5.9 T of the DEMO design. The plasma current output is now 21 MegaAmps, which is less than 20% away from the design value of 18 MegaAmps. It is important to note that in each of these parameters, we see an increase going from ITER to DEMO, which is consistent both in our model’s output and the actual design parameters in the papers [1-3].
The operational constraints plot for DEMO is shown in the figure below. DEMO is a larger reactor than ITER, and given the favorable scaling of tokamak operation with size, we expect improved results for operational constraints in DEMO. The three curves which check plasma stability are all satisfied. Unlike in the case of ITER which had the green curve close to the shaded region, the green curve in the case of DEMO stays safely in the unshaded region. The blue curve is still in the unshaded region, but much closer to the boundary of the unshaded region than ITER (now ~1.8, much closer to 1 than in the case of ITER which was ~4). This shows an improvement for DEMO compared to ITER as it is closer to producing enough self-sustaining plasma current, though it will still need some help from other current-generating techniques which will be tested on ITER.
Further upgrades of the Princeton Field Reversed Configuration-2 (PFRC-2) are underway with the goal of achieving the milestone of ion heating. The PFRC-2 is predicted to have substantial ion heating once the RF antenna frequency is lowered and the magnetic field is increased. To lower the RF frequency, we have installed additional capacitors in the tank circuit of PFRC-2. The picture below shows three capacitors, each with capacitance of 2 nanoFarads (2 nF), installed in a custom-built copper box.
The copper box is also shown in the bottom part of the image below, where it will be connected with a robust cable to the top box, which is called the tuning box. The tuning box is an aluminum box with one fixed capacitor and two tunable capacitors which can be adjusted to change the resonance frequency of the circuit.
Changes have also been made to the inside of the tuning box in order to prevent electrical arcing, which is a common issue when working with high-power and high-voltage circuits. To help prevent arcing, conical structures of brass have been fabricated and installed. The brass structure is shown alone in the first image below and is shown enveloping the cable connection in the second image below. The shape of these structures allows a better spread of the charge in the tuning box so as to lower the chances of electrical breakdown. Taking these preventative design decisions is key to ensuring reliable operation once the upgraded system is running.
This is a really excellent article on nuclear fusion, “Small-scale fusion tackles energy, space applications,” by M. Mitchell Waldrop, written January 28, 2020, Vol 117, No. 4 for the Proceedings of the National Academy of Sciences of the United States of America (PNAS). The article quotes team Dr. Cohen and Mr. Paluszek and provides an excellent and technically accurate discussion of FRCs, heating methods, and fusion fuel physics.
The Princeton Field Reversed Configuration-2 (PFRC-2) upgrade in field and frequency is underway. We are currently installing new coils around the experiment to increase the magnetic fields and new capacitors to help lower the RF operating frequency – all to reach our target milestones of measuring ion heating! This is an essential next step in our development of Direct Fusion Drive.
The power supplies are stacked in their rack, ready to supply power to the belt coils. The supplies must be programmed to energize for each pulse as they are not cooled and the coils would otherwise overheat. The belt coil holder component on the right was 3D printed at PPPL.
The new 2 nF capacitors, shown above (left image), must be enclosed in a custom copper box that will be part of the tank circuit of PFRC-2. Each component must be carefully designed, including the lengths of the connecting cables, for us to get the right frequency without exceeding voltage limits of the materials.
The above image is of the cable that will connect the tank circuit and the PFRC-2. These cables are very robust, and stiff so that the layout must be carefully planned. We will continue to post updates as we work towards that 2 MHz frequency milestone!
Electron density profiles on PFRC with USPR: Ultrashort Pulse Reflectometry (USPR) is a plasma diagnostic technique that would be used on the Princeton Field-Reversed Configuration (PFRC) to measure electron density profiles. Such profile measurements provide insight into the structure of PFRC plasma and can improve our estimates of confinement time. Our University partner is University of California, Davis, PI Dr. Neville Luhmann.
Evaluating RF antenna designs for PFRC plasma heating and sustainment: We intend to analyze RF antenna performance parameters critical to the validity of robust PFRC-type fusion reactor designs. Team member University of Rochester will support TriForce simulations and contractor Plasma Theory and Computation, Inc. will support RMF code simulations. Our national lab partner is Princeton Plasma Physics Laboratory, PI Dr. Sam Cohen.
Stabilizing PFRC plasmas against macroscopic low‐frequency instabilities: This award will use the TriForce code to simulate several plasma stabilization techniques for the PFRC-2 experiment. Our lab partner is PPPL and the team again includes the University of Rochester.
These awards will help us advance PFRC technology. Contact us for more information!
The Fusion Energy Toolbox for MATLAB is a toolbox for designing fusion reactors and for studying plasma physics. It includes a wide variety of physics and engineering tools. The latest addition to this toolbox is a new function for designing tokamaks, based on the paper in reference . Tokamaks have been the leading magnetic confinement devices investigated in the pursuit of fusion net energy gain. Well-known tokamaks that either have ongoing experiments or are under development include JET, ITER, DIII-D, KSTAR, EAST, and Commonwealth Fusion Systems’ SPARC. The new capability of our toolboxes to conduct trade studies on tokamaks allows our customers to take part in this exciting field of fusion reactor design and development.
The Fusion Reactor Design function checks that the reactor satisfies key operational constraints for tokamaks. These operational constraints result from the plasma physics of the fusion reactor, where there are requirements for the plasma to remain stable (e.g., not crash into the walls) and to maintain enough electric current to help sustain itself. The tunable parameters include: the plasma minor radius ‘a’ (see figure below), the H-mode enhancement factor ‘H’, the maximum magnetic field at the coils ‘B_max’, the electric power output of the reactor ‘P_E’, and the neutron wall loading ‘P_W’, which are all essential variables to tokamak design and operation. H-mode is the high confinement mode used in many machines.
This function captures all figure and table results in the original paper. We implemented a numerical solver which allows the user to choose a variable over which to perform a parameter sweep. A ‘mode’ option has been incorporated which allows one to select a desired parameter sweep variable (‘a’, ‘H’, ‘B_max’, ‘P_E’, or ‘P_W’) when calling the function. Some example outputs of the function are described below.
As an example, we will consider the case of tuning the maximum magnetic field at the coils ‘B_max’. The figure below plots the normalized operation constraint parameters for a tokamak as functions of B_max from 10 Tesla to 25 Tesla. The unshaded region, where the vertical axis is below the value of 1, is the region where operational constraints are met. We see that for magnetic fields below about 17.5 Tesla there is at least one operation constraint that is not met, while for higher magnetic fields all operation constraints are satisfied, thus meeting the conditions for successful operation. This high magnetic field approach is the design approach of Commonwealth Fusion Systems for the reactor they are developing .
Note, however, that there is a material cost associated with achieving higher magnetic fields, as described in reference . This is illustrated in the figure below, which plots the cost parameter (the ratio of engineering components volume V_I to electric power output P_E) against B_max. There is a considerable increase in cost at high magnetic fields due to the need to add material volume that can structurally handle the higher current loads required.
In this post we illustrated the case of a tunable maximum magnetic field at the coils, though as mentioned earlier, there are other parameters you can tune. This function is part of release 2022.1 of the Fusion Energy Toolbox. Contact us at email@example.com or call us at +01 609 275-9606 for more information.
Thank you to interns Emma Suh and Paige Cromley for their contributions to the development of this function.