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Pebblebed NTRs: Solid Fuel, but Different

Hello, and welcome back to Beyond NERVA!

Today, we’re going to take a break from the closed cycle gas core nuclear thermal rocket (which I’ve been working on constantly since mid-January) to look at one of the most popular designs in modern NTR history: the pebblebed reactor!

This I should have covered between solid and liquid fueled NTRs, honestly, and there’s even a couple types of reactor which MAY be able to be used for NTR between as well – the fluidized and shush fuel reactors – but with the lack of information on liquid fueled reactors online I got a bit zealous.

Beads to Explore the Solar System

Most of the solid fueled NTRs we’ve looked at have been either part of, or heavily influenced by, the Rover and NERVA programs in the US. These types of reactors, also called “prismatic fuel reactors,” use a solid block of fuel of some form, usually tileable, with holes drilled through each fuel element.

The other designs we’ve covered fall into one of two categories, either a bundled fuel element, such as the Russian RD-0410, or a folded flow disc design such as the Dumbo or Tricarbide Disc NTRs.

However, there’s another option which is far more popular for modern American high temperature gas cooled reactor designs: the pebblebed reactor. This is a clever design, which increases the surface area of the fuel by using many small, spherical fuel elements held in a (usually) unfueled structure. The coolant/propellant passes between these beads, picking up the heat as it passes between them.

This has a number of fundamental advantages over the prismatic style fuel elements:

  1. The surface area of the fuel is so much greater than with simple holes drilled in the prismatic fuel elements, increasing thermal transfer efficiency.
  2. Since all types of fuel swell when heated, the density of the packed fuel elements could be adjusted to allow for better thermal expansion behavior within the active region of the reactor.
  3. The fuel elements themselves were reasonably loosely contained within separate structures, allowing for higher temperature containment materials to be used.
  4. The individual elements could be made smaller, allowing for a lower temperature gradient from the inside to the outside of a fuel, reducing the overall thermal stress on each fuel pebble.
  5. In a folded flow design, it was possible to not even have a physical structure along the inside of the annulus if centrifugal force was applied to the fuel element structure (as we saw in the fluid fueled reactor designs), eliminating the need for as many super-high temperature materials in the highest temperature region of the reactor.
  6. Because each bead is individually clad, in the case of an accident during launch, even if the reactor core is breached and a fuel release into the environment occurs, the release of either any radiological components or any other fuel materials into the environment is minimized
  7. Because each bead is relatively small, it is less likely that they will sustain sufficient damage either during mechanical failure of the flight vehicle or impact with the ground that would breach the cladding.

However, there is a complication with this design type as well, since there are many (usually hundreds, sometimes thousands) of individual fuel elements:

  1. Large numbers of fuel beads mean large numbers of fuel beads to manufacture and perform quality control checks on.
  2. Each bead will need to be individually clad, sometimes with multiple barriers for fission product release, hydrogen corrosion, and the like.
  3. While each fuel bead will be individually clad, and so the loss of one or all the fuel will not significantly impact the environment from a radiological perspective in the case of an accident, there is potential for significant geographic dispersal of the fuel in the event of a failure-to-orbit or other accident.

There are a number of different possible flow paths through the fuel elements, but the two most common are either an axial flow, where the propellant passes through a tubular structure packed with the fuel elements, and a folded flow design, where the fuel is in a porous annular structure, with the coolant (usually) passing from the outside of the annulus, through the fuel, and the now-heated coolant exiting through the central void of the annulus. We’ll call these direct flow and folded flow pebblebed fuel elements.

In addition, there are many different possible fuel types, which regulars of this blog will be familiar with by now: oxides, carbides, nitrides, and CERMET are all possible in a pebblebed design, and if differential fissile fuel loading is needed, or gradients in fuel composition (such as using tungsten CERMET in higher temperature portions of the reactor, with beryllium or molybdenum CERMET in lower temperature sections), this can be achieved using individual, internally homogeneous fuel types in the beads, which can be loaded into the fuel support structure at the appropriate time to create the desired gradient.

Just like in “regular” fuel elements, these pebbles need to be clad in a protective coating. There have been many proposals over the years, obviously depending on what type of fissile fuel matrix the fuel uses to ensure thermal expansion and chemical compatibility with the fuel and coolant. Often, multiple layers of different materials are used to ensure structural and chemical integrity of the fuel pellets. Perhaps the best known example of this today is the TRISO fuel element, used in the US Advanced Gas Reactor fuel development program. The TRI-Structural ISOtropic fuel element uses either oxide or carbide fuel in the center, followed by a porous carbon layer, a pyrolitic carbon layer (sort of like graphite, but with some covalent bonds between the carbon sheets), followed by a silicon carbide outer shell for mechanical and fission product retention. Some variations include a burnable poison for reactivity control (the QUADRISO at Argonne), or use different outer layer materials for chemical protection. Several types have been suggested for NTR designs, and we’ll see more of them later.

The (sort of) final significant variable is the size of the pebble. As the pebbles go down in size, the available surface area of the fuel-to-coolant interface increases, but also the amount of available space between the pebbles decreases and the path that the coolant flows through becomes more resistant to higher coolant flow rates. Depending on the operating temperature and pressure, the thermal gradient acceptable in the fuel, the amount of decay heat that you want to have to deal with on shutdown (the bigger the fuel pebble, the more time it will take to cool down), fissile fuel density, clad thickness requirements, and other variables, a final size for the fuel pebbles can be calculated, and will vary to a certain degree between different reactor designs.

Not Just for NTRs: The Electricity Generation Potential of Pebblebed Reactors

Obviously, the majority of the designs for pebblebed reactors are not meant to ever fly in space, they’re mostly meant to operate as high temperature gas cooled reactors on Earth. This type of architecture has been proposed for astronuclear designs as well, although that isn’t the focus of this video.

Furthermore, the pebblebed design lends itself to other cooling methods, such as molten salt, liquid metal, and other heat-carrying fluids, which like the gas would flow through the fuel pellets, pick up the heat produced by the fissioning fuel, and carry it into a power conversion system of whatever design the reactor has integrated into its systems.

Finally, while it’s rare, pebblebed designs were popular for a while with radioisotope power systems. There are a number of reasons for this beyond being able to run a liquid coolant through the fuel (which was done on one occasion that I can think of, and we’ll cover in a future post): in an alpha-emitting radioisotope, such as 238Pu, over time the fuel will generate helium gas – the alpha particles will slow, stop, and become doubly ionized helium nuclei, which will then strip electrons off whatever materials are around and become normal 4He. This gas needs SOMEWHERE to go, which is why just like with a fissile fuel structure there are gas management mechanisms used in radioisotope power source fuel assemblies such as areas of vacuum, pressure relief valves, and the like. In some types of RTG, such as the SNAP-27 RTG used by Apollo, as well as the Multi-Hundred Watt RTG used by Voyager, the fuel was made into spheres, with the gaps in between the spheres (normally used to pass coolant through) are used for the gas expansion volume.

We’ll discuss these ideas more in the future, but I figured it was important to point out here. Let’s get back to the NTRs, and the first (and only major) NTR program to focus on the pebblebed concept: the Project Timberwind and the Space Nuclear Propulsion Program in the 1980s and early 1990s.

The Beginnings of Pebblebed NTRs

The first proposals for a gas cooled pebblebed reactor were from 1944/45, although they were never pursued beyond the concept stage, and a proposal for the “Space Vehicle Propulsion Reactor” was made by Levoy and Newgard at Thikol in 1960, with again no further development. If you can get that paper, I’d love to read it, here’s all I’ve got: “Aero/Space Engineering 19, no. 4, pgs 54-58, April 1960” and ‘AAE Journal, 68, no. 6, pgs. 46-50, June 1960,” and “Engineering 189, pg 755, June 3, 1960.” Sounds like they pushed hard, and for good reason, but at the time a pebblebed reactor was a radical concept for a terrestrial reactor, and getting a prismatic fueled reactor, something far more familiar to nuclear engineers, was a challenge that seemed far simpler and more familiar.

Sadly, while this design may end up have informed the design of its contemporary reactor, it seems like this proposal was never pursued.

Rotating Fluidized Bed Reactor (“Hatch” Reactor) and the Groundwork for Timberwind

Another proposal was made at the same time at Brookhaven National Laboratory, by L.P. Hatch, W.H. Regan, and a name that will continue to come up for the rest of this series, John R. Powell (sorry, can’t find the given names of the other two, even). This relied on very small (100-500 micrometer) fuel, held in a perforated drum to contain the fuel but also allow propellant to be injected into the fuel particles, which was spun at a high rate to provide centrifugal force to the particles and prevent them from escaping.

Now, fluidized beds need a bit of explanation, which I figured was best to put in here since this is not a generalized property of pebblebed reactors. In this reactor (and some others) the pebbles are quite small, and the coolant flow can be quite high. This means that it’s possible – and sometimes desirable – for the pebbles to move through the active zone of the reactor! This type of mobile fuel is called a “fluidized bed” reactor, and comes in several variants, including pebble (solid spheres), slurry (solid particulate suspended in a liquid), and colloid (solid particulate suspended in a gas). The best way to describe the phenomenon is with what is called the point of minimum fluidization, or when the drag forces on the mass of the solid objects from the fluid flow balances with the weight of the bed (keep in mind that life is a specialized form of drag). There’s a number of reasons to do this – in fact, many chemical reactions using a solid and a fluid component use fluidization to ensure maximum mixing of the components. In the case of an NTR, the concern is more to do with achieving as close to thermal equilibrium between the solid fuel and the gaseous propellant as possible, while minimizing the pressure drop between the cold propellant inlet and the hot propellant outlet. For an NTR, the way that the “weight” is applied is through centrifugal force on the fuel. This is a familiar concept to those that read my liquid fueled NTR series, but actually began with the fluidized bed concept.

This is calculated using two different relations between the same variables: the Reynolds number (Re), which determines how turbulent fluid flow is, and the friction coefficient (CD, or coefficient of drag, which deptermines how much force acts on the fuel particles based on fluid interactions with the particles) which can be found plotted below. The plotted lines represent either the Reynolds number or the void fraction ε, which represents the amount of gas present in the volume defined by the presence of fuel particles.

Hendrie 1970

If you don’t follow the technical details of the relationships depicted, that’s more than OK! Basically, the y axis is proportional to the gas turbulence, while the x axis is proportional to the particle diameter, so you can see that for relatively small increases in particle size you can get larger increases in propellant flow rates.

The next proposal for a pebble bed reactor grew directly out of the Hatch reactor, the Rotating Fluidized Bed Reactor for Space Nuclear Propulsion (RBR). From the documentation I’ve been able to find, from the original proposal work continued at a very low level at BNL from the time of the original proposal until 1973, but the only reports I’ve been able to find are from 1971-73 under the RBR name. A rotating fuel structure, with small, 100-500 micrometer spherical particles of uranium-zirconium carbide fuel (the ZrC forming the outer clad and a maximum U content of 10% to maximize thermal limits of the fuel particles), was surrounded by a reflector of either metallic beryllium or BeO (which was preferred as a moderator, but the increased density also increased both reactor mass and manufacturing requirements). Four drums in the reflector would control the reactivity of the engine, and an electric motor would be attached to a porous “squirrel cage” frit, which would rotate to contain the fuel.

Much discussion was had as to the form of uranium used, be it 235U or 233U. In the 235U reactor, the reactor had a cavity length of 25 in (63.5 cm), an inner diameter of 25 in (63.5 cm), and a fuel bed depth when fluidized of 4 in (10.2 cm), with a critical mass of U-ZrC being achieved at 343.5 lbs (155.8 kg) with 9.5% U content. The 233U reactor was smaller, at 23 in (56 cm) cavity length, 20 in (51 cm) bed inner diameter, 3 in (7.62 cm) deep fuel bed with a higher (70%) void fraction, and only 105.6 lbs (47.9 kg) of U-ZrC fuel at a lower (and therefore more temperature-tolerant) 7.5% U loading.

233U was the much preferred fuel in this reactor, with two options being available to the designers: either the decreased fuel loading could be used to form the smaller, higher thrust-to-weight ratio engine described above, or the reactor could remain at the dimensions of the 235U-fueled option, but the temperature could be increased to improve the specific impulse of the engine.

There was als a trade-off between the size of the fuel particles and the thermal efficiency of the reactor,:

  • Smaller particles advantages
    • Higher surface area, and therefore better thermal transfer capabilities,
    • Smaller radius reduces thermal stresses on fuel
  • Smaller particles disadvantages
    • Fluidized particle bed fuel loss would be a more immediate concern
    • More sensitive to fluid dynamic behavior in the bed
    • Bubbles could more easily form in fuel
    • Higher centrifugal force required for fuel containment
  • Larger particle advantages
    • Ease of manufacture
    • Lower centrifugal force requirements for a given propellant flow rate
  • Larger particle disadvantages
    • Higher thermal gradient and stresses in fuel pellets
    • Less surface area, so lower thermal transfer efficiency

It would require testing to determine the best fuel particle size, which could largely be done through cold flow testing.

These studies looked at cold flow testing in depth. While this is something that I’ve usually skipped over in my reporting on NTR development, it’s a crucial type of testing in any gas cooled reactor, and even more so in a fluidized bed NTR, so let’s take a look at what it’s like in a pebblebed reactor: the equipment, the data collection, and how the data modified the reactor design over time.

Cold flow testing is usually the predecessor to electrically heated flow testing in an NTR. These tests determine a number of things, including areas within the reactor that may end up with stagnant propellant (not a good thing), undesired turbulence, and other negative consequences to the flow of gas through the reactor. They are preliminary tests, since as the propellant heats up while going through the reactor, a couple major things will change: first, the density of the gas will decrease and second, as the density changes the Reynolds number (a measure of self-interaction, viscosity, and turbulent vs laminar flow behavior) will change.

In this case, the cold flow tests were especially useful, since one of the biggest considerations in this reactor type is how the gas and fuel interact.

The first consideration that needed to be examined is the pressure drop across the fuel bed – the highest pressure point in the system is always the turbopump, and the pressure will decrease from that point throughout the system due to friction with the pipes carrying propellant, heating effects, and a host of other inefficiencies. One of the biggest questions initially in this design was how much pressure would be lost from the frit (the outer containment structure and propellant injection system into the fuel) to the central void in the body of the fuel, where it exits the nozzle. Happily, this pressure drop is minimal: according to initial testing in the early 1960s (more on that below), the pressure drop was equal to the weight of the fuel bed.

The next consideration was the range between fluidizing the fuel and losing the fuel through literally blowing it out the nozzle – otherwise known as entrainment, a problem we looked at extensively on a per-molecule basis in the liquid fueled NTR posts (since that was the major problem with all those designs). Initial calculations and some basic experiments were able to map the propellant flow rate and centrifugal force required to both get the benefit of a fluidized bed and prevent fuel loss.

Rotating Fluidized Bed Reactor testbed test showing bubble formation,

Another concern is the formation of bubbles in the fuel body. As we covered in the bubbler LNTR post (which you can find here), bubbles are a problem in any fuel type, but in a fluid fueled reactor with coolant passing through it there’s special challenges. In this case, the main method of transferring heat from the fuel to the propellant is convection (i.e. contact between the fuel and the propellant causing vortices in the gas which distributes the heat), so an area that doesn’t have any (or minimal) fuel particles in it will not get heated as thoroughly. That’s a headache not only because the overall propellant temperature drops (proportional to the size of the bubbles), but it also changes the power distribution in the reactor (the bubbles are fission blank spots).

Finally, the initial experiment set looked at the particle-to-fluid thermal transfer coefficients. These tests were far from ideal, using a 1 g system rather than the much higher planned centrifugal forces, but they did give some initial numbers.

The first round of tests was done at Brookhaven National Laboratory (BNL) from 1962 to 1966, using a relatively simple test facility. A small, 10” (25.4 cm) length by 1” (2.54 cm) diameter centrifuge was installed, with gas pressure provided by a pressurized liquefied air system. 138 to 3450 grams of glass particles were loaded into the centrifuge, and various rotational velocities and gas pressures were used to test the basic behavior of the particles under both centrifugal force and gas pressure. While some bobbles were observed, the fuel beds remained stable and no fuel particles were lost during testing, a promising beginning.

These tests provided not just initial thermal transfer estimates, pressure drop calculations, and fuel bed behavioral information, but also informed the design of a new, larger test rig, this one 10 in by 10 in (25.4 by 25.4 cm), which was begun in 1966. This system would not only have a larger centrifuge, but would also use liquid nitrogen rather than liquefied air, be able to test different fuel particle simulants rather than just relatively lightweight glass, and provide much more detailed data. Sadly, the program ran out of funding later that year, and the partially completed test rig was mothballed.

Rotating Fluidized Bed Reactor (RBR): New Life for the Hatch Reactor

It would take until 1970, when the Space Nuclear Systems office of the Atomic Energy Commission and NASA provided additional funding to complete the test stand and conduct a series of experiments on particle behavior, reactor dynamics and optimization, and other analytical studies of a potential advanced pebblebed NTR.

The First Year: June 1970-June 1971

After completing the test stand, the team at BNL began a series of tests with this larger, more capable equipment in Building 835. The first, most obvious difference is the diameter of the centrifuge, which was upgraded from 1 inch to 10 inches (25.4 cm), allowing for a more prototypical fuel bed depth. This was made out of perforated aluminum, held in a stainless steel pressure housing for feeding the pressurized gas through the fuel bed. In addition, the gas system was changed from the pressurized air system to one designed to operate on nitrogen, which was stored in liquid form in trailers outside the building for ease of refilling (and safety), then pre-vaporized and held in two other, high-pressure trailers.

Photographs were used to record fluidization behavior, taken viewing the bottom of the bed from underneath the apparatus. While initially photos were only able to be taken 5 seconds apart, later upgrades would improve this over the course of the program.

The other major piece of instrumentation surrounded the pressure and flow rate of the nitrogen gas throughout the system. The gas was introduced at a known pressure through two inlets into the primary steel body of the test stand, with measurements of upstream pressure, cylindrical cavity pressure outside the frit, and finally a pitot tube to measure pressure inside the central void of the centrifuge.

Three main areas of pressure drop were of interest: due to the perforated frit itself, the passage of the gas through the fuel bed, and finally from the surface of the bed and into the central void of the centrifuge, all of which needed to be measured accurately, requiring calibration of not only the sensors but also known losses unique to the test stand itself.

The tests themselves were undertaken with a range of glass particle sizes from 100 to 500 micrometers in diameter, similar to the earlier tests, as well as 500 micrometer copper particles to more closely replicate the density of the U-ZrC fuel. Rotation rates of between1,000 and 2,000 rpm, and gas flow rates from 1,340-1,800 scf/m (38-51 m^3/min) were used with the glass beads, and from 700-1,500 rpm with the copper particles (the lower rotation rate was due to gas pressure feed limitations preventing the bed from becoming fully fluidized with the more massive particles).

Finally, there were a series of physics and mechanical engineering design calculations that were carried out to continue to develop the nuclear engineering, mechanical design, and system optimization of the final RBR.

The results from the initial testing were promising: much of the testing was focused on getting the new test stand commissioned and calibrated, with a focus on figuring out how to both use the rig as it was constructed as well as which parts (such as the photography setup) could be improved in the next fiscal year of testing. However, particle dynamics in the fuidized bed were comfortably within stable, expected behavior, and while there were interesting findings as to the variation in pressure drop along the axis of the central void, this was something that could be worked with.

Based on the calculations performed, as well as the experiments carried out in the first year of the program, a range of engines were determined for both 233U and 235U variants:

Work Continues: 1971-1972

This led directly into the 1971-72 series of experiments and calculations. Now that the test stand had been mostly completed (although modifications would continue), and the behavior of the test stand was now well-understood, more focused experimentation could continue, and the calculations of the physics and engineering considerations in the reactor and engine system could be advanced on a more firm footing.

One major change in this year’s design choices was the shift toward a low-thrust, high-isp system, in part due to greater interest at NASA and the AEC in a smaller NTR than the original design envelope. While analyzing the proposed engine size above, though, it was discovered that the smallest two reactors were simply not practical, meaning that the smallest design was over 1 GW power level.

Another thing that was emphasized during this period from the optimization side of the program was the mass of the reflector. Since the low thrust option was now the main thrust of the design, any increase in the mass of the reactor system has a larger impact on the thrust-to-weight ratio, but reducing the reflector thickness also increases the neutron leakage rate. In order to prevent this, a narrower nozzle throat is preferred, but also increases thermal loading across the throat itself, meaning that additional cooling, and probably more mass, is needed – especially in a high-specific-impulse (aka high temperature) system. This also has the effect of needing higher chamber pressures to maintain the desired thrust level (a narrower throat with the same mass flow throughput means that the pressure in the central void has to be higher).

These changes required a redesign of the reactor itself, with a new critical configuration:

Hendrie 1972

One major change is how fluidized the bed actually is during operation. In order to get full fluidization, there needs to be enough inward (“upward” in terms of force vectors) velocity at the inner surface of the fuel body to lift the fuel particles without losing them out the nozzle. During calculations in both the first and second years, two major subsystems contributed hugely to the weight and were very dependent on both the rotational speed and the pellet size/mass: the weight of the frit and motor system, which holds the fuel particles, and the weight of the nozzle, which not only forms the outlet-end containment structure for the fuel but also (through the challenges of rocket motor dynamics) is linked to the chamber pressure of the reactor – oh, and the narrower the nozzle, the less surface area is available to reject the heat from the propellant, so the harder it is to keep cool enough that it doesn’t melt.

Now, fluidization isn’t a binary system: a pebblebed reactor is able to be settled (no fluidization), partially fluidized (usually expressed as a percentage of the pebblebed being fluidized), and fully fluidized to varying degrees (usually expressed as a percentage of the volume occupied by the pebbles being composed of the fluid). So there’s a huge range, from fully settled to >95% fluid in a fully fluidized bed.

The designers of the RBR weren’t going for excess fluidization: at some point, the designer faces diminishing returns on the complications required for increased fluid flow to maintain that level of particulate (I’m sure it’s the same, with different criteria, in the chemical industry, where most fluidized beds actually are used), both due to the complications of having more powerful turbopumps for the hydrogen as well as the loss of thermalization of that hydrogen because there’s simply too much propellant to be heated fully – not to mention fuel loss from the particulate fuel being blown out of the nozzle – so the calculations for the bed dynamics assumed minimal full fluidization (i.e. when all the pebbles are moving in the reactor) as the maximum flow rate – somewhere around 70% gas in the fuel volume (that number was never specifically defined that I found in the source documentation, if it was, please let me know), but is dependent on both the pressure drop in the reactor (which is related to the mass of the particle bed) and the gas flow.

Ludewig 1974

However, the designers at this point decided that full fluidization wasn’t actually necessary – and in fact was detrimental – to this particular NTR design. Because of the dynamics of the design, the first particles to be fluidized were on the inner surface of the fuel bed, and as the fluidization percentage increased, the pebbles further toward the outer circumference became fluidized. Because the temperature difference between the fuel and the propellant is greater as the propellant is being injected through the frit and into the fuel body, more heat is carried away by the propellant per unit mass, and as the propellant warms up, thermal transfer becomes less efficient (the temperature difference between two different objects is one of the major variables in how much energy is transferred for a given surface area), and fluidization increases that efficiency between a solid and a fluid.

Because of this, the engineers re-thought what “minimal fluidization” actually meant. If the bed could be fluidized enough to maximize the benefit of that dynamic, while at a minimum level of fluidization to minimize the volume the pebblebed actually took up in the reactor, there would be a few key benefits:

  1. The fueled volume of the reactor could be smaller, meaning that the nozzle could be wider, so they could have lower chamber pressure and also more surface area for active cooling of the nozzle
  2. The amount of propellant flow could be lower, meaning that turbopump assemblies could be smaller and lighter weight
  3. The frit could be made less robustly, saving on weight and simplifying the challenges of the bearings for the frit assembly
  4. The nozzle, frit, and motor/drive assembly for the frit are all net neutron poisons in the RBR, meaning that minimizing any of these structures’ overall mass improves the neutron economy in the reactor, leading to either a lower mass reactor or a lower U mass fraction in the fuel (as we discussed in the 233U vs. 235U design trade-off)

After going through the various options, the designers decided to go with a partially fluidized bed. At this point in the design evolution, they decided on having about 50% of the bed by mass being fluidized, with the rest being settled (there’s a transition point in the fuel body where partial fluidization is occurring, and they discuss the challenges of modeling that portion in terms of the dynamics of the system briefly). This maximizes the benefit at the circumference, where the thermal difference (and therefore the thermal exchange between the fuel and the propellant) is most efficient, while also thermalizing the propellant as much as possible as the temperature difference decreases from the propellant becoming increasingly hotter. They still managed to reach an impressive 2400 K propellant cavity temperature with this reactor, which makes it one of the hottest (and therefore highest isp) solid core NTR designs proposed at that time.

This has various implications for the reactor, including the density of the fissile component of the fuel (as well as the other solid components that make up the pebbles), the void fraction of the reactor (what part of the reactor is made up of something other than fuel, in this particular instance hydrogen within the fuel), and other components, requiring a reworking of the nuclear modeling for the reactor.

An interesting thing to me in the Annual Progress Report (linked below) is the description of how this new critical configuration was modeled; while this is reasonably common knowledge in nuclear engineers from the days before computational modeling (and even to the present day), I’d never heard someone explain it in the literature before.

Basically, they made a bunch of extremely simplified (in both number of dimensions and fidelity) one-dimensional models of various points in the reactor. They then assumed that they could rotate these around that elevation to make something like an MRI slice of the nuclear behavior in the reactor. Then, they moved far enough away that it was different enough (say, where the frit turns in to the middle of the reactor to hold the fuel, or the nozzle starts, or even the center of the fuel compared to the edge) that the dynamics would change, and did the same sort of one-dimensional model; they would end up doing this 18 times. Then, sort of like an MRI in reverse, they took these models, called “few-group” models, and combined them into a larger group – called a “macro-group” – for calculations that were able to handle the interactions between these different few-group simulations to build up a two-dimensional model of the reactor’s nuclear structure and determine the critical configuration of the reactor. They added a few other ways to subdivide the reactor for modeling, for instance they split the neutron spectrum calculations into fast and thermal, but this is the general shape of how nuclear modeling is done.

Ok, let’s get back to the RBR…

Experimental testing using the rotating pebblebed simulator continued through this fiscal year, with some modifications. A new, seamless frit structure was procured to eliminate some experimental uncertainty, the pressure measuring equipment was used to test more areas of the pressure drop across the system, and a challenge for the experimental team – finding 100 micrometer copper spheres that were regularly enough shaped to provide a useful analogue to the UC-ZrC fuel (Cu specific gravity 8.9, UC-ZrC specific gravity ~6.5) were finally able to be procured.

Additionally, while thermal transfer experiments had been done with the 1-gee small test apparatus which preceded the larger centrifugal setup (with variable gee forces available), the changes were too great to allow for accurate predictions on thermal transfer behavior. Therefore, thermal transfer experiments began to be examined on the new test rig – another expansion of the capabilities of the new system, which was now being used rigorously since its completing and calibration testing of the previous year. While they weren’t conducted that year, setting up an experimental program requires careful analysis of what the test rig is capable of, and how good data accuracy can be achieved given the experimental limitations of the design.

The major achievement for the year’s ex[experimentation was a refining of the relationship between particle size, centrifugal force, and pressure drop of the propellant from the turbopump to the frit inlet to the central cavity, most especially from the frit to the inner cavity through the fuel body, on a wide range of particle sizes, flow rates, and bed fluidization levels, which would be key as the design for the RBR evolved.

The New NTR Design: Mid-Thrust, Small RBR

So, given the priorities at both the AEC and NASA, it was decided that it was best to focus primarily on a given thrust, and try and optimize thrust-to-weight ratios for the reactor around that thrust level, in part because the outlet temperature of the reactor – and therefore the specific impulse – was fixed by the engineering decisions made in regards to the rest of the reactor design. In this case, the target thrust was was 90 kN (20,230 lbf), or about 120% of a Pewee-class engine.

This, of course, constrained the reactor design, which at this point in any reactor’s development is a good thing. Every general concept has a huge variety of options to play with: fuel type (oxide, carbide, nitride, metal, CERMET, etc), fissile component (233U and 235U being the big ones, but 242mAm, 241Cf, and other more exotic options exist), thrust level, physical dimensions, fuel size in the case of a PBR, and more all can be played with to a huge degree, so having a fixed target to work towards in one metric allows a reference point that the rest of the reactor can work around.

Also, having an optimization point to work from is important, in this case thrust-to-weight ratio (T/W). Other options, such as specific impulse, for a target to maximize would lead to a very different reactor design, but at the time T/W was considered the most valuable consideration since one way or another the specific impulse would still be higher than the prismatic core NTRs currently under development as part of the NERVA program (being led by Los Alamos Scientific Laboratory and NASA, undergoing regular hot fire testing at the Jackass Flats, NV facility). Those engines, while promising, were limited by poor T/W ratios, so at the time a major goal for NTR improvement was to increase the T/W ratio of whatever came after – which might have been the RBR, if everything went smoothly.

One of the characteristics that has the biggest impact on the T/W ratio in the RBR is the nozzle throat diameter. The smaller the diameter, the higher the chamber pressure, which reduces the T/W ratio while increasing the amount of volume the fuel body can occupy given the same reactor dimensions – meaning that smaller fuel particles could be used, since there’s less chance that they would be lost out of the narrower nozzle throat. However, by increasing the nozzle throat diameter, the T/W ratio improved (up to a point), and the chamber pressure could be decreased, but at the cost of a larger particle size; this increases the thermal stresses in the fuel particles, and makes it more likely that some of them would fail – not as catastrophic as on a prismatic fueled reactor by any means, but still something to be avoided at all costs. Clearly a compromise would need to be reached.

Here are some tables looking at the design options leading up to the 90 kN engine configuration with both the 233U and 235U fueled versions of the RBR:

After analyzing the various options, a number of lessons were learned:

  1. It was preferable to work from a fixed design point (the 90 kN thrust level), because while the reactor design was flexible, operating near an optimized power level was more workable from a reactor physics and thermal engineering point of view
  2. The main stress points on the design were reflector weight (one of the biggest mass components in the system), throat diameter (from both a mass and active cooling point of view as well as fuel containment), and particle size (from a thermal stress and heat transfer point of view)
  3. On these lower-trust engines, 233U was looking far better than 235U for the fissile component, with a T/W ratio (without radiation shielding) of 65.7 N/kg compared to 33.3 N/kg respectively
    1. As reactor size increased, this difference reduced significantly, but with a constrained thrust level – and therefore reactor power – the difference was quite significant.

The End of the Line: RBR Winds Down

1973 was a bad year in the astronuclear engineering community. The flagship program, NERVA, which was approaching flight ready status with preparations for the XE-PRIME test, the successful testing of the flexible, (relatively) inexpensive Nuclear Furnace about to occur to speed not only prismatic fuel element development but also a variety of other reactor architectures (such as the nuclear lightbulb we began looking at last time), and the establishment of a robust hot fire testing structure at Jackass Flats, was fighting for its’ life – and its’ funding – in the halls of Congress. The national attention, after the success of Apollo 11, was turning away from space, and the missions that made NTR technologically relevant – and a good investment – were disappearing from the mission planners’ “to do” lists, and migrating to “if we only had the money” ideas. The Rotating Fluidized Bed Reactor would be one of those casualties, and wouldn’t even last through the 1971/72 fiscal year.

This doesn’t mean that more work wasn’t done at Brookhaven, far from it! Both analytical and experimental work would continue on the design, with the new focus on the 90 kN thrust level, T/W optimized design discussed above making the effort more focused on the end goal.

Multi-program computational architecture used in 1972/73 for RBR, Hoffman 1973

On the analytical side, many of the components had reasonably good analytical models independently, but they weren’t well integrated. Additionally, new and improved analytical models for things like the turbopump system, system mass, temp and pressure drop in the reactor, and more were developed over the last year, and these were integrated into a unified modeling structure, involving multiple stacked models. For more information, check out the 1971-72 progress report linked in the references section.

The system developed was on the verge of being able to do dynamics modeling of the proposed reactor designs, and plans were laid out for what this proposed dynamic model system would look like, but sadly by the time this idea was mature enough to implement, funding had run out.

On the experimental side, further refinement of the test apparatus was completed. Most importantly, because of the new design requirements, and the limitations of the experiments that had been conducted so far, the test-bed’s nitrogen supply system had to be modified to handle higher gas throughput to handle a much thicker fuel bed than had been experimentally tested. Because of the limited information about multi-gee centrifugal force behavior in a pebblebed, the current experimental data could only be used to inform the experimental course needed for a much thicker fuel bed, as was required by the new design.

Additionally, as was discussed from the previous year, thermal transfer testing in the multi-gee environment was necessary to properly evaluate thermal transfer in this novel reactor configuration, but the traditional methods of thermal transfer simply weren’t an option. Normally, the procedure would be to subject the bed to alternating temperatures of gas: cold gas would be used to chill the pebbles to gas-ambient temperatures, then hot gas would be used on the chilled pebbles until they achieved thermal equilibrium at the new temperature, and then cold gas would be used instead, etc. The temperature of the exit gas, pebbles, and amount of gas (and time) needed to reach equilibrium states would be analyzed, allowing for accurate heat transfer coefficients at a variety of pebble sizes, centrifugal forces, propellant flow rates, etc. would be able to be obtained, but at the same time this is a very energy-intensive process.

An alternative was proposed, which would basically split the reactor’s propellant inlet into two halves, one hot and one cold. Stationary thermocouples placed through the central void in the centrifuge would record variations in the propellant at various points, and the gradient as the pebbles moved from hot to cold gas and back could get good quality data at a much lower energy cost – at the cost of data fidelity reducing in proportion to bed thickness. However, for a cash-strapped program, this was enough to get the data necessary to proceed with the 90 kN design that the RBR program was focused on.

Looking forward, while the team knew that this was the end of the line as far as current funding was concerned, they looked to how their data could be applied most effectively. The dynamics models were ready to be developed on the analytical side, and thermal cycling capability in the centrifugal test-bed would prepare the design for fission-powered testing. The plan was to address the acknowledged limitations with the largely theoretical dynamic model with hot-fired experimental data, which could be used to refine the analytical capabilities: the more the system was constrained, and the more experimental data that was collected, the less variability the analytical methods had to account for.

NASA had proposed a cavity reactor test-bed, which would serve primarily to test the open and closed cycle gas core NTRs also under development at the time, which could theoretically be used to test the RBR as well in a hot-fore configuration due to its unique gas injection system. Sadly, this test-bed never came to be (it was canceled along with most other astronuclear programs), so the faint hope for fission-powered RBR testing in an existing facility died as well.

The Last Gasp for the RBR

The final paper that I was able to find on the Rotating Fluidized Bed Reactor was by Ludewig, Manning, and Raseman of Brookhaven in the Journal of Spacecraft, Vol 11, No 2, in 1974. The work leading up to the Brookhaven program, as well as the Brookhaven program itself, was summarized, and new ideas were thrown out as possibilities as well. It’s evident reading the paper that they still saw the promise in the RBR, and were looking to continue to develop the project under different funding structures.

Other than a brief mention of the possibility of continuous refueling, though, the system largely sits where it was in the middle of 1973, and from what I’ve seen no funding was forthcoming.

While this was undoubtedly a disappointing outcome, as virtually every astronuclear program in history has faced, and the RBR never revived, the concept of a pebblebed NTR would gain new and better-funded interest in the decades to come.

This program, which has its own complex history, will be the subject for our next blog post: Project Timberwind and the Space Nuclear Thermal Propulsion program.

Conclusion

While the RBR was no more, the idea of a pebblebed NTR would live on, as I mentioned above. With a new, physically demanding job, finishing up moving, and the impacts of everything going on in the world right now, I’m not sure exactly when the next blog post is going to come out, but I have already started it, and it should hopefully be coming in relatively short order! After covering Timberwind, we’ll look at MITEE (the whole reason I’m going down this pebblebed rabbit hole, not that the digging hasn’t been fascinating!), before returning to the closed cycle gas core NTR series (which is already over 50 pages long!).

As ever, I’d like to thank my Patrons on Patreon (www.patreon.com/beyondnerva), especially in these incredibly financially difficult times. I definitely would have far more motivation challenges now than I would have without their support! They get early access to blog posts, 3d modeling work that I’m still moving forward on for an eventual YouTube channel, exclusive content, and more. If you’re financially able, consider becoming a Patron!

You can also follow me at https://twitter.com/BeyondNerva for more regular updates!

References

Rotating Fluidized Bed Reactor

Hendrie et al, “ROTATING FLUIDIZED BED REACTOR FOR SPACE NUCLEAR PROPULSION Annual Report: Design Studies and Experimental Results, June, 1970- June, 1971,” Brookhaven NL, August 1971 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19720017961.pdf

Hendrie et al, “ROTATING FLUIDIZED BED REACTOR FOR SPACE NUCLEAR PROPULSION Annual Report: Design Studies and Experimental Results, June 1971 – June 1972,” Brookhaven NL, Sept. 1972 https://inis.iaea.org/collection/NCLCollectionStore/_Public/04/061/4061469.pdf

Hoffman et al, “ROTATING FLUIDIZED BED REACTOR FOR SPACE NUCLEAR PROPULSION Annual Report: Design Studies and Experimental Results, July 1972 – January 1973,” Brookhaven NL, Sept 1973 https://inis.iaea.org/collection/NCLCollectionStore/_Public/05/125/5125213.pdf

Cavity Test Reactor

Whitmarsh, Jr, C. “PRELIMINARY NEUTRONIC ANALYSIS OF A CAVITY TEST REACTOR,” NASA Lewis Research Center 1973 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19730009949.pdf

Whitmarsh, Jr, C. “NUCLEAR CHARACTERISTICS OF A FISSIONING URANIUM PLASMA TEST REACTOR WITH LIGHT -WATER COOLING,” NASA Lewis Research Center 1973 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19730019930.pdf

Categories
Development and Testing History Nuclear Thermal Systems

The Nuclear Lightbulb – A Brief Introduction

Hello, and welcome back to Beyond NERVA! Really quickly, I apologize that I haven’t published more recently. Between moving to a different state, job hunting, and the challenges we’re all facing with the current medical situation worldwide, this post is coming out later than I was hoping. I have been continuing to work in the background, but as you’ll see, this engine isn’t one that’s easy to take in discrete chunks!

Today, we jump into one of the most famous designs of advanced nuclear thermal rocket: the “nuclear lightbulb,” more properly known as the closed cycle gas core nuclear thermal rocket. This will be a multi-part post on not only the basics of the design, but a history of the way the design has changed over time, as well as examining both the tests that were completed as well as the tests that were proposed to move this design forward.

Cutaway of simplified LRC Closed Cycle Gas Core NTR, image credit Winchell Chung of Atomic Rockets

One of the challenges that we saw on the liquid core NTR was that the fission products could be released into the environment. This isn’t really a problem from the pollution side for a space nuclear reactor (we’ll look at the extreme version of this in a couple months with the open cycle gas core), but as a general rule it is advantageous to avoid it most of the time to keep the exhaust mass low (why we use hydrogen in the first place). In ideal circumstances, and with a high enough thrust-to-weight ratio, eliminating this release could even enable an NTR to be used in surface launches.

That’s the potential of the reactor type we’re going to be discussing today, and in the next few posts. Due to the complexities of this reactor design, and how interconnected all the systems are, there may be an additional pause in publication after this post. I’ve been working on the details of this system for over a month and a half now, and am almost done covering the basics of the fuel itself… so if there’s a bit of delay, please be understanding!

The closed cycle gas core uses uranium hexafluoride (UF6) as fuel, which is contained within a fused silica “bulb” to form the fuel element – hence the popular name “nuclear lightbulb”. Several of these are distributed through the reactor’s active zone, with liquid hydrogen coolant flowing through the silica bulb, and then the now-gaseous hydrogen passing around the bulbs and out the nozzle of the reactor. This is the most conservative of the gas core designs, and only a modest step above the vapor core designs we examined last time, but still offers significantly higher temperatures, and potentially higher thrust-to-weight ratios, than the VCNTR.

A combined research effort by NASA’s Lewis (now Glenn) Research Center and United Aircraft Corporation in the 1960s and 70s made significant progress in the design of these reactors, but sadly with the demise of the AEC and NASA efforts in nuclear thermal propulsion, the project languished on the shelves of astronuclear research for decades. While it has seen a resurgence of interest in the last few decades in popular media, most designs for spacecraft that use the lightbulb reactor reference the efforts from the 60s and 70s in their reactor designs- despite this being, in many ways, one of the most easily tested advanced NTR designs available.

Today’s blog post focuses on the general shape of the reactor: its basic geometry, a brief examination of its analysis and testing, and the possible uses of the reactor. The next post will cover the analytical studies of the reactor in more detail, including the limits of what this reactor could provide, and what the tradeoffs in the design would require to make a practical NTR, as well as the practicalities of the fuel element design itself. Finally, in the third we’ll look at the testing that was done, could have been done with in-core fission powered testing, the lessons learned from this testing, and maybe even some possibilities for modern improvements to this well-known, classic design.

With that, let’s take a look at this reactor’s basic shape, how it works, and what the advantages of and problems with the basic idea are.

Nuclear Lightbulb: Nuclear Powered Children’s Toy (ish)

Easy Bake Oven, image Wikimedia

For those of us of a certain age, there was a toy that was quite popular: the Easy-Bake Oven. This was a very simple toy: an oven designed for children with minimal adult supervision to be able to cook a variety of real baked goods, often with premixed dry mixes or simple recipes. Rather than having a more normal resistive heating element as you find in a normal oven, though, a special light bulb was mounted in the oven, and the waste heat from the bulb would heat the oven enough to cook the food.

Closed cycle gas core bulb, image DOE colorized by Winchell Chung

The closed cycle gas core NTR takes this idea, and ramps it up to the edges of what materials limits allow. Rather than a tungsten wire, the heat in the bulb is generated by a critical mass of uranium hexafluoride, a gas at room temperature that’s used in, among other things, fissile fuel enrichment for reactors and other applications. This is contained in a fused silica bulb made up of dozens of very thin tubes – not much different in material, but very different in design, compared to the Easy-Bake Oven – which contains the fissile fuel, and prevents the fission products from escaping. The fuel turns from gas to plasma, and forms a vortex in the center of the fuel element.

Axial cross-section of the fuel/buffer/wall region of the lightbulb, Rodgers 1972

To further protect the bulb from direct contact with the uranium and free fluorine, a gaseous barrier of noble gas (either argon or neon) is injected between the fuel and the wall of the bulb itself. Because of the extreme temperatures, the majority of the electromagnetic radiation coming off the fuel isn’t in the form of infrared (heat), but rather as ultraviolet radiation, which the silica is transparent to, minimizing the amount of energy that’s deposited into the bulb itself. In order to further protect the silica bulb, microparticles of the same silica are added to the neon flow to absorb some of the radiation the bulb isn’t transparent to, in order to remove that part of the radiation before it hits the bulb. This neon passes around the walls of the chamber, creating a vortex in the uranium which further constrains it, and then passes out of one or both ends of the bulb. It then goes through a purification and cooling process using a cryogenic hydrogen heat exchanger and gas centrifuge, before being reused.

Now, of course there is still an intense amount of energy generated in the fuel which will be deposited in the silica, and will attempt to melt the bulb almost instantly, so the bulb must be cooled regeneratively. This is done by liquid hydrogen, which is also mostly transparent to the majority of the radiation coming off the fuel plasma, minimizing the amount of energy the coolant absorbs from anything but the silica of the bulb itself.

Finally, the now-gaseous hydrogen from both the neon and bulb cooling processes, mixed with any hydrogen needed to cool the pressure vessel, reflectors of the reactor, and other components, is mixed with microparticles of tungsten to increase the amount of UV radiation emitted by the fuel. This then passes around the bulbs in the reactor, getting heated to their final temperature, before exiting the nozzle of the NTR.

Overall configuration, Rodgers 1972

The most commonly examined version of the lightbulb uses a total of seven bulbs, with those bulbs being made up of a spiral of hydrogen coolant channels in fused silica. This was pioneered by NASA’s Lewis Research Center (LRC), and studied by United Aircraft Corp of Mass (UA). These studies were carried out between 1963 and 1972, with a very small number of follow-up studies at UA completing by 1980. This design was a 4600 MWt reactor fueled by 233U, an isp of 1870 seconds, and a thrust-to-weight ratio of 1.3.

A smaller version of this system, using a single bulb rather than seven, was proposed by the same team for probe missions and the like, but unfortunately the only papers are behind paywalls.

During the re-examination of nuclear thermal technology in the early 1990s by NASA and the DOE, the design was re-examined briefly to assess the advantages that the design could offer, but no advances in the design were made at the time.

Since then, while interest in this concept has grown, new studies have not been done, and the design remains dormant despite the extensive amount of study which has been carried out.

What’s Been Done Before: Previous Studies on the Lightbulb

Bussard 1958

The first version of the closed cycle gas core proposed by Robert Bussard in 1946. This design looked remarkably like an internal combustion firing chamber, with the UF6 gas being mechanically compressed into a critical density with a piston. Coolant would be run across the outside of the fuel element and then exit the reactor through a nozzle. While this design hasn’t been explored in any depth that I’ve been able to determine, a new version using pressure waves rather than mechanical pistons to compress gas into a critical mass has been explored in recent years (we’ll cover that in the open cycle gas core posts).

Starting in 1963, United Aircraft (UA, a subsidiary of United Technologies) worked with NASA’s Lewis Research Center (LRC) and Los Alamos Scientific Laboratory (LASL) on both the open and closed cycle gas core concepts, but the difficulties of containing the fuel in the open cycle concept caused the company to focus exclusively on the closed cycle concepts. Interestingly, according to Tom Latham of UA (who worked on the program), the design was limited in both mass and volume by the then-current volume of the proposed Space Shuttle cargo bay. Another limitation of the original concept was that no external radiators could be used for thermal management, due to the increased mass of the closed radiator system and its associated hardware.

System flow diagram, Rodgers 1972

The design that evolved was quite detailed, and also quite efficient in many ways. However, the sheer number of interdependent subsystems makes is fairly heavy, limiting its potential usefulness and increasing its complexity.

In order to get there, a large number of studies were done on a number of different subsystems and physical behaviors, and due to the extreme nature of the system design itself many experimental apparatus had to be not only built, but redesigned multiple times to get the results needed to design this reactor.

We’ll look at the testing history more in depth in a future blog post, but it’s worth looking at the types of tests that were conducted to get an idea of just how far along this design was:

RF Heating Test Apparatus, Roman 1969

Both direct current and radio frequency testing of simulated fuel plasmas were conducted, starting with the RF (induction heating) testing at the UA facility in East Hartford, CT. These studies typically used tungsten in place of uranium (a common practice, even still used today) since it’s both massive and also has somewhat similar physical properties to uranium. At the time, argon was considered for the buffer gas rather than neon, this change in composition will be something we’ll look at later in the detailed testing post.

Induction heating works by using a vibrating magnetic field to heat materials that will flip their molecular direction or vibrate, generating heat. It is a good option for nuclear testing since it is able to more evenly heat the simulated fuel, and can achieve high temperatures – it’s still used for nuclear fuel element testing not only in the Compact Fuel Element Environment Test (CFEET) test stand, which I’ve covered here https://beyondnerva.com/nuclear-test-stands-and-equipment/non-nuclear-thermal-testing/cfeet-compact-fuel-element-environmental-test/ , but also in the Nuclear Thermal Rocket Environmental Effects Simulator, which I covered here: https://beyondnerva.com/nuclear-test-stands-and-equipment/non-nuclear-thermal-testing/ntrees/ . One of the challenges of this sort of heating, though, is the induction coil, the device that creates the heating in the material. In early testing they managed to melt the copper coil they were using due to resistive heating (the same method used to make heat in a space heater or oven), and constructing a higher-powered apparatus wasn’t possible for the team.

This led to direct current heating testing to achieve higher temperatures, which uses an electrical arc through the tungsten plasma. This isn’t as good at simulating the way that heat is distributed in the plasma body, but could achieve higher temperatures. This was important for testing the stability of the vortex generated by not only the internal heating of the fuel, but also the interactions between the fuel and the neon containment system.

Spectral flux from the edge of the fuel body, Rodgers 1972 (will be covered more in depth in another post)

Another concern was determining what frequencies of radiation silicon, aluminum and neon were transparent to. By varying the temperature of the fissioning fuel mass, the frequency of radiation could, to a certain degree, be tuned to a frequency that maximized how much energy would pass through both the noble gas (then argon) and the bulb structure itself. Again, at the time (and to a certain extent later), the bulb configuration was slightly different: a layer of aluminum was added to the inner surface of the bulb to reflect more thermal radiation back into the fissioning fuel in order to increase heating, and therefore increase the temperature of the fuel. We’ll look at how this design option changed over time in future posts.

More studies and tests were done looking at the effects of neutron and gamma radiation on reactor materials. These are significant challenges in any reactor, but the materials being used in the lightbulb reactor are unusual, even by the standards of astronuclear engineering, so detailed studies of the effects of these radiation types were needed to ensure that the reactor would be able to operate throughout its required lifetime.

Fused silica test article, Vogt 1970

Perhaps one of the biggest concerns was verifying that the bulb itself would maintain both its integrity and its functionality throughout the life of the reactor. Silica is a material that is highly unusual in a nuclear reactor, and the fact that it needed to remain not only transparent but able to contain both a noble gas seeded with silica particles and hydrogen while remaining transparent to a useful range of radiation while being bombarded with neutrons (which would change the crystalline structure) and gamma rays (which would change the energy states of the individual nuclei to varying degrees) was a major focus of the program. On top of that, the walls of the individual tubes that made up the bulbs needed to be incredibly thin, and the shape of each of the individual tubes was quite unusual, so there were significant experimental manufacturing considerations to deal with. Neutron, gamma and beta (high energy electron) radiation could all have their effect on the bulb itself during the course of the reactor’s lifetime, and these effects needed to be understood and accounted for. While these tests were mostly successful, with some interesting materials properties of silica discovered along the way, when Dr. Latham discussed this project 20 years later, one of the things he mentioned was that modern materials science could possibly offer better alternatives to the silica tubing – a concept that we will touch on again in a future post.

Another challenge of the design was that it required seeding two different materials into two different gasses: the neon/argon had to be seeded with silica in order to protect the bulb, and the hydrogen propellant needed to be seeded with tungsten to make it absorb the radiation passing through the bulb as efficiently as possible while minimizing the increase in the mass of the propellant. While the hydrogen seeding process was being studied for other reactor designs – we saw this in the radiator liquid fueled NTR, and will see it again in the future in open cycle gas core and some solid core designs we haven’t covered yet – the silica seeding was a new challenge, especially because the material being seeded and the material the seeded gas would travel through was the same as the material that was seeded into the gas.

Image DOE via Chris Casilli on Twitter

Finally, there’s the challenge of nuclear testing. Los Alamos Scientific Laboratory conducted some tests that were fission-powered, which proved the concept in theory, but these were low powered bench-top tests (which we’ll cover in depth in the future). To really test the design, it would be ideal to do a hot-fire test of an NTR. Fortunately, at the time the Nuclear Furnace test-bed was being completed (more on NERVA hot fire testing here: https://beyondnerva.com/2018/06/18/ntr-hot-fire-testing-part-i-rover-and-nerva-testing/ and the exhaust scrubbers for the Nuclear furnace here: https://beyondnerva.com/nuclear-test-stands-and-equipment/nuclear-furnace-exhaust-scrubbers/ ). This meant that it was possible to use this versatile test-bed to test a single, sub-scale lightbulb in a controlled, well-understood system. While this test was never actually conducted, much of the preparatory design work for the test was completed, another thing we’ll cover in a future post.

A Promising, Developed, Unrealized Option

The closed cycle gas core nuclear thermal rocket is one of the most perrenially fascinating concepts in astronuclear history. Not only does it offer an option for a high-temperature nuclear reactor which is able to avoid many of the challenges of solid fuel, but it offers better fission product containment than any other design besides the vapor core NTR.

It is also one of the most complex systems that has ever been proposed, with two different types of closed cycle gas systems involving heat exchangers and separation systems supporting seven different fuel chambers, a host of novel materials in unique environments, the need to tune both the temperature and emissivity of a complex fuel form to ensure the reactor’s components won’t melt down, and the constant concerns of mass and complexity hanging over the heads of the designers.

Most of these challenges were addressed in the 1960s and 1970s, with most of the still-unanswered questions needing testing that simply wasn’t possible at the time of the project’s cancellation due to shifting priorities in the space program. Modern materials science may offer better solutions to those that were available at the time as well, both in the testing and operation of this reactor.

Sadly, updating this design has not happened, but the original design remains one of the most iconic designs in astronuclear engineering.

In the next two posts, we’ll look at the testing done for the reactor in detail, followed by a detailed look at the reactor itself. Make sure to keep an eye out for them!

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References

McLafferty, G.H. “Investigation of Gaseous Nuclear Rocket Technology – Summary Technical Report” 1969 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19700008165.pdf

Rodgers, R.J. and Latham, T.S. “Analytical Design and Performance Studies of the Nuclear Light Bulb Engine” 1972 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19730003969.pdf

Latham, T.S. “Nuclear Light Bulb,” 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920001892.pdf

Categories
Development and Testing Forgotten Reactors History Nuclear Thermal Systems

Liquid Fueled NTRs: An Introduction

Hello, and welcome back to Beyond NERVA! Today we continue our look into advanced NTR fuel types, by diving into an extended look at one of the least covered design types in this field: the liquid fueled NTR (LNTR).

This is a complex field, with many challenges unique to the phase state of the fuel, so while I was planning on making this a single-part series, now there’s three posts! This first one is going to discuss LNTRs in general, as well as some common problems and challenges that they face. I’ll include a very brief history of the designs, almost all of them dating from the 1950s and 1960s, which we’ll look at more in depth in the next couple posts.

Unfortunately, a lot of the fundamental problems of an LNTR get deep – fast, for a lot of people, but the fundamental concepts are often not too hard to get in the broad strokes. I’m gonna try my best to explain them the way that I learned them, and if there’s more questions I’ll attempt to point you to the references I’ve used as a layperson, but I honestly believe that this architecture has suffered from a combination of being “not terrible, not great” in terms of engine performance (1300 s isp, 19/1 T/W).

With that, let’s get into liquid fueled NTRs (LNTR), their history, and their design!

Basic Design Options for LNTR

LNTRs are not a very diverse group of reactor concepts, partially due to the nature of the fuel and partially because they haven’t been well-researched overall. All designs I’ve found use centrifugal force to contain molten fuel inside a tube, with the central void in the spinning tube being the outlet point for the propellant. The first design used a single, large fuel mass in a single fuel element, but quickly this was divided into multiple individual fuel elements, which became the norm for LNTR through the latest designs. One consequence of this first design was the calculation of the neutronic moderation capacity of the H2 propellant in this toroidal fuel structure, and the authors of the study determined that it was so close to zero that it was worth it to consider the center of the fuel element to be a vacuum as far as MCNP (the standard neutronic modeling code both at the time, and in updated form now) is concerned. This is something worth noting: any significant neutron moderation for the core must come from the reflectors and moderator either integrated into the fuel structure (complex to do in a liquid in many cases) or the body of the reactor, the propellant flow won’t matter enough to cause a significant decrease in neutron velocities.

They do seem to fall into two broad categories, which I’ll call bubblers and radiators. A bubbler LNTR is one where the fuel is fed from the outside of the fuel element, through the molten fuel, and into the central void of the fuel element; a radiator LNTR passes propellant only through the central void along the long axis of the fuel element.

A bubbler has the advantage that it is able to use an incredible amount of surface area for heat transfer from the fuel to the propellant, with the surface area being inversely proportional to the size of the individual bubbles: smaller bubbles, more surface area, more heat transfer, greater theoretical power density in the active region of the reactor. They also have the advantage of being able to regeneratively cool the entire length of the fuel element’s outside surface as a natural consequence of the way the propellant is fed into the fuel, rather than using specialized regenerative cooling systems in the fuel element canister and reactor body. However, bubblers also have a couple problems: first, the reactor will not be operating continuously, so on shutdown the fuel will solidify, and the bubbling mechnaism will become clogged with frozen nuclear fuel; second, the breaching of the bubbles to the surface can fling molten fuel into the fast-moving propellant stream, causing fuel to be lost; finally, the bubbles increase mixing of the fuel, which is mostly good but can also lead to certain chemical components of the fuel being carried at a greater rate by either vaporizing and being absorbed into the bubbles or becoming entrained in the fuel and outgassing when the bubble breaches the surface. In a way, it’s sort of like boiling pasta sauce: the water boils, and the bubbles mix the sauce while they move up, but some chemical compounds diffuse into the water vapor along the way (which ones depend on what’s in the sauce), and unless there’s a lid on the pot the sauce splatters across the stove, again depending on the other components of the sauce that you’re cooking. (the obvious problem with this metaphor is that, rather than the gaseous component being a part of the initial solution they’re externally introduced)/

Radiators avoid many of the problems of a bubbler, but not all, by treating the fuel almost like a solid mass when its under centrifugal force: the propellant enters from the ship end, through the central void in the fuel element, and then out the aft end to enter the nozzle through an outlet plenum. This makes fuel retention a far simpler problem overall, but fuel will still be lost through vaporization into the propellant stream (more on this later). Another issue with radiators is that without the propellant passing all the way through the fuel from the outer to inner diameter, the thermal emissions will not only go into the propellant, but also into the fuel canister and the reactor itself – more efficiently, actually, since H2 isn’t especially good at capturing heat,k and conduction is more efficient than radiation. This requires regenerative cooling both for the fuel canister and the reactor as well most of the time – which while doable also requires a more complex plumbing setup within the reactor body to maintain material thermal limits on even relatively high temperature materials, much less hydrides (which are good low-volume, low-mass moderators for compact reactors, but incredibly thermally sensitive).

As with any other astronuclear design, there’s a huge design envelope to play with in terms of fuel matrix, even in liquid form (although this is more limited in liquid designs, as we’ll see), as well as moderation level, number and size of fuel elements, moderator type, and other decisions. However, the vast majority of the designs have been iterative concepts on the same basic two ideas, with modifications mostly focusing on fuel element dimensions and number, fuel temperature, propellant flow rates, and individual fuel matrix materials rather than entirely different reactor architectures.

It’s worth noting that there’s another concept, the droplet core NTR, which diffuses the liquid fuel into the propellant, then recaptures it using (usually) centrifugal force before the droplets can leave the nozzle, but this is a concept that will be covered alongside the vapor core reactor, since it’s a hybrid of the two concepts.

A (Very) Brief History of LNTR

Because we’re going to be discussing the design evolution of each type of LNTR in depth in the next two posts, I’m going to be incredibly brief here, giving a general overview of the history of LNTRs. While they’re often mentioned as an intermediate-stage NTR option, there’s been a surprisingly small amount of research done on them, with only two programs of any significant size being conducted in the 1960s.

Single cavity LNTR, Barrett 1963

The first proposal for an LNTR was by J. McCarthy in 1954, in his “Nuclear Reactors for Rockets.” This design used a single, large cylinder, spun around the long axis, as both the reactor and fuel element. The fuel was fed into the void in the cylinder radially, bubbling through the fuel mass, which was made of uranium carbide (UC2). This design, as any first design, had a number of problems, but showed sufficient promise for the design to be re-examined, tweaked, and further researched to make it more practical. While I don’t have access to this paper, a subsequent study of the design placed the maximum specific impulse of this type of NTR in the range of 1200-1400 seconds.

Multiple Fuel Element LNTR, Nelson et al 1963

This led to the first significant research program into the LNTR, carried out by Nelson et al at the Princeton Aeronautical Engineering Laboratory in 1963. This design changed the single large rotating cylinder into several smaller ones, each rotating independently, while keeping the same bubbler architecture of the McCarthy design. This ended up improving the thrust to weight ratio, specific impulse, power density, and other key characteristics. The study also enumerated many of the challenges of both the LNTR in general, and the bubbler in specific, for the first time in a detailed and systematic fashion, but between the lack of information on the materials involved, as well as lack of both computational theory and modeling capability, this study was hampered by many assumptions of convenience. Despite these challenges (which would continue to be addressed over time in smaller studies and other designs), the Princeton LNTR became the benchmark for most LNTR designs of both types that followed. The final design chosen by the team has a vacuum specific impulse of 1250 s, a chamber pressure of 10 atm, and a thrust-to-weight ratio of about 2:1, with a reactor mass of approximately 100 metric tons.

Experimental setup for bublle behavior studies, Barrett Jr 1963

Studies on the technical details of the most challenging aspect of this design, that of bubble motion, would continue at Princeton for a number of years, including experiments to observe the behavior of the particular bubble form needed while under centrifugal acceleration, but challenges in modeling the two-phase (liquid/gas) interactions for thermodynamics and hydrodynamics continued to dog the bubbler design. It is unclear when work stopper on the bubbler design, but the last reference to it that I can find in the literature was from 1972, in a published Engineering Note by W.L. Barrett, who observed that many of the hoped-for goals were overly optimistic, but not by a huge margin. This is during the time that American astronuclear funding was being demolished, and so it would not be surprising that the concept would go into dormancy at that point. Since the restarting of modest astronuclear funding, though, I have been unable to find any reference to a modern bubbler design for either terrestrial or astronuclear use.

Perhaps the main reason for this, which we’ll discuss in the next section, is the inconveniently high vapor pressure of many compounds when operating in the temperature range of an LNTR (about 8800 K). This means that the constituent parts of the fuel body, most notably the uranium, would vaporize into the propellant, not only removing fissile material from the reactor but significantly increasing the mass of the propellant stream, decreasing specific impulse. This, in fact, was the reason the Lewis Research Center focused on a different form of LNTR: the radiator.

Work on the radiator concept began in 1964, and was conducted by a team headed by R Ragsdale, one of the leading NTR designers ar Lewis Research Center. To mitigate the vapor losses of the bubbler type, the question was asked if the propellant actually had to pass through the fuel, or if radiant heating would suffice to thermalize the hydrogen propellant while minimizing the fuel loss from the liquid/gas interaction zone. The answer was a definite yes, although the fuel temperature would have to be higher, and the propellant would likely need to be seeded with some particulate or vapor to increase its thermal absorption. While the overall efficiency would be slightly lower, only a minimal loss of specific impulse would occur, and the thrust to weight ratio could be increased due to higher propellant flow (only so much propellant can pass through a given volume of bubbler-type fuel before unacceptable splattering and other difficulties would arise). This seems to have reached its conclusion in 1967, the last date that any of the papers or reports that I’ve been able to find, with a final compromise design achieving 1400 s of isp, a thrust-to-core-weight-ratio of 4:1, at a core temperature of 5060 K and a reactor pressure of 200 atm (2020 N/m^2).

However, unlike with the bubbler-type LNTR, the radiator would have one last, minor hurrah. In the 1990s, at the beginning of the Space Exploration Initiative, funding became available again for NTR development. A large conference was held in 1991, in Albuquerque, NM, and served as a combination state-of-research and idea presentation for what direction NTR development should go in, as well as determining which concepts should be explored more in depth. As part of this, presentations were made on many different fundamental reactor architectures, and proposals for each type of NTR were made. While the bubbler LNTR was not represented, the radiator was.

LARS cross-section, Powell 1991

This concept, presented by J Powell of Brookhaven National Lab, was the Liquid Annular Reactor System. Compared to the Lewis and Princeton designs, it was a simple reactor, with only seven fuel elements, These would be spaced in a cylinder of Be/H moderator, and would use a twice-through coolant/propellant system: each cylinder was regeneratively cooled from nozzle-end to ship-end, and then the propellant, seeded with W microparticles, would then pass through the central void and out the nozzle. Interestingly enough, this design did not seem to reference the work done by either Princeton or Lewis RC, so there’s a possibility that this was a new design from first principles (other designs presented at the conference made extensive use of legacy data and modeling). This reactor was only conceptually sketched out in the documentation I’ve found, operated at higher temperatures (~6000 K) and lower pressures (~10 atm) than the previous designs to dissociate virtually all of the hydrogen propellant, and no estimated thrust-to-core-weight ratios.

It is unclear how much work was done on this reactor design, and it also remains the last design of any LNTR type that I’ve been able to come across.

Lessons from History: Considerations for LNTR Design

Having looked through the history of LNTR design, it’s worth looking at the lessons that have been learned from these design studies and experiments, as well as the reasons (as far as we can tell) that the designs have evolved the way they did. I just want to say up front that I’m going to be especially careful about when I use my own interpretation, compared to a more qualified someone else’s interpretation, on the constraints and design philosophies here, because this is an area that runs into SO MANY different materials, neutronics, etc constraints that I don’t even know where to begin independently assessing the advantages and disadvantages.

Also, we’re going to be focusing on the lessons that (mostly) apply to both the bubbler and radiator concepts. The following posts, covering the types individually, will address the specific challenges of the two types of LNTR.

Reactor Architecture

The number of fuel elements in an LNTR is a trade-off.

  • Advantages to increasing the number of fuel elements
    • The total surface area available in the fuel/propellant boundary increases, increasing thrust for a given specific impulse
    • The core becomes more homogeneous, making a more idealized neutronic environment (there’s a limit to this, including using interstitial moderating blocks between the fuel elements to further thermalize the reactor, but is a good rule of thumb in most cases)
  • Advantages to minimizing the number of fuel elements
    • The more fuel elements, the more manufacturing headache in making the fuel element canisters and elements themselves, as well as the support equipment for maintaining the rotation of the fuel elements;
      • depending on the complexity of the manufacturing process, this could be a significant hurdle,
      • Electronic motors don’t do well in a high neutron flux, generally requiring driveshaft penetration of at least part of the shadow shield, and turbines to drive the system can be so complex that this is often not considered an option in NTRs (to be fair, it’s rare that they would be needed)
    • The less angular velocity is needed for each fuel element to have the same centrifugal force, due to the larger radius of the fuel element
    • For a variety of reasons the fuel thickness increases to maintain the same critical mass in the reactor – NOTE: this is a benefit for bubbler-type LNTRs, but either neutral or detrimental to streamer-type NTRs.

Another major area of trade-off is propellant mass flow rates. These are fundamentally limited in bubbler LNTRs (something we’ll discuss in the next post), since the bubbles can’t be allowed to combine (or splattering and free droplets will occur), the more bubbles the more the fuel expands (causing headaches for fuel containment), and other issues will present themselves. On the other hand, for radiator – and to a lesser extent the bubbler – type LNTRs, the major limitation is thermal uptake in the propellant (too much mass flow means that the exhaust velocity will drop), which can be somewhat addressed by propellant seeding (something that we’ll discuss in a future webpage).

Fuel Material Constraints

One fundamental question for any LNTR fuel is the maximum theoretical isp of a design, which is a direct function of the critical temperature (when the fuel boils) and at what rate the fuel would vaporize from where the fuel and propellant interact. Pretty much every material has a range of temperature and pressure values where either sublimation (in a solid) or vaporization (in a liquid) will occur, and these characteristics were not well understood at the time.

This is actually one of the major tradeoffs in bubbler vs radiator designs. In a bubbler, you get the propellant and the maximum fuel temperature to be the same, but you also effectively saturate the fuel with any available vapor. The actual vapor concentrations are… well, as far as I can tell, it’s only ever been modeled with 1960s methods, and those interactions are far beyond what I’m either qualified or comfortable to assess, but I suspect that while the problem may be able to be slightly mitigated it won’t be able to be completely avoided.

However, there are general constraints on the fuels available for use, and the choice of every LNTR has been UC2, usually with a majority of the fuel mass being either ZrC or NbC as the dilutent. Other options are available, potentially, such as 184W-U or U-Si metals, but they have not been explored in depth.

Let’s look at the vapor pressure implications more in depth, since it really is the central limitation of LNTR fuels at temperatures that are reasonable for these rockets.

Vapor Pressure Implications

A study on the vapor pressure of uranium was conducted in 1953 by Rauh et al at Argonne NL, which determined an approximate function of the vapor pressure of “pure” uranium metal (some discussion about the inhibiting effects of oxygen, which would not be present in an NTR to any great degree, and also tantalum contamination of the uranium, were needed based on the experimental setup), but this was based on solid U, so was only useful as a starting point.

Barrett Jr 1963

W Louis Barret Jr. conducted another study in 1963 on the implications of fuel composition for a bubbler-type LNTR, and the constraints on the potential specific impulse of this type of reactor. The author examined many different fissile fuel matrices in their paper, including Pu and Th compounds:

From this, and assuming a propellant pressure of 10^3 psi, a maximum theoretical isp was calculated for each type of fuel:

Barrett Jr 1963

Additional studies were carried out on uranium metal and carbon compounds – mostly Zr-C-U, Nb-C-U and 184W-C-U, in various concentrations – in 1965 and 66 by Kaufman and Peters of MANLABS for NASA Lewis Research Center (the center of LNTR development at the time), conducted at 100 atmospheres and ~4500 to ~5500 K. These were low atomic mass fraction systems (0.001-0.02), which may be too low for some designs, but will minimize fissile fuel loss to the propellant flow. Other candidate materials considered were Mo-C-U, B-C-U, and Me-C-U, but not studied at the time.

A summary of the results can be found below:

Perhaps the most significant question is mass loss rates due to hydrogen transport, which can be found in this table:

Kaufman, 1966

These values offer a good starting point for those that want to explore the maximum operating temperature of this type of reactor, but additional options may exist. For instance, a high vapor pressure, high boiling point, low neutron absorption metal which will mix minimally with the uranium-bearing fuel could be used as a liquid fuel clad layer, either in a persistent form (meant to survive the lifetime of the fuel element) or as a sacrificial vaporization layer similar to how ablative coatings are used in some rocket nozzles (one note here: this will increase the atomic mass of the propellant stream, decreasing the specific impulse of such a design). However, other than the use of ZrC in the Princeton design study in the inner region of that fuel element design (which was also considered a sacrificial component of the fuel), I haven’t seen anyone discuss this concept in depth in the literature.

A good place to start investigating this concept, however, would be with a study done by Charles Masser in 1967 entitled “Vapor-Pressure Data Extrapolated to 1000 Atmospheres of 13 Refractory Materials with Low Thermal Absorption Cross Sections.” While this was focused on the seeding of propellant with microparticles to increase thermal absorption in colder H2, the vapor-pressure information can provide a good jumping off point for anyone interested in investigating this subject further. The paper can be found here: https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19670030361.pdf.

Author speculation concept:

Another, far more speculative option is available if the LNTR can be designed as a thermal breeder, and dealing with certain challenges in fuel worth fluctuations (and other headaches), especially at startup: thorium. This is because Th has a much lower vapor pressure than U does (although the vapor pressure behavior of carbides in a high temperature, high pressure situation doesn’t seem to have been studied ThO2 and ThO3 outperform UC2 – but oxides are a far worse idea than carbides in this sort of reactor), so it may be possible to make a Th-breeder LNTR to reduce fissile fuel vapor losses – which does nothing for C, or Zr/Nb, but may be worth it.

This requires a couple things to happen: first, the reactor’s available reactivity needs to be able to remain within the control authority of the control systems in a far more complex system, and the breeding ratio of the reactor needs to be carefully managed. There’s a few reasons for this, but let’s look at the general shape of the challenge.

Many LNTR designs are either fast or epithermal designs, with few extending into the thermal neutron spectrum. Thorium breeds into 233U best in the thermal neutron spectrum, so the neutron flux needs to be balanced against the Th present in the reactor in order to make sure that the proper breeding ratio is maintained. This can be adjusted by adding moderator blocks between the fuel elements, using other filler materials, and other options common to NTR neutronics design, but isn’t something that I’ve seen addressed anywhere.

Let’s briefly look at the breeding process: when 232Th is bred into 233U, it goes through a two-week period where the nucleus undergoing the breeding process ends up existing as 233Pa, a strong neutron poison. Unlike the thorium breeding molten salt reactor, these designs don’t have on-board fuel reprocessing, and that’s a very heavy, complex system that is going to kill your engine’s dry mass, so just adding one isn’t a good option from a systems engineering point of view. So, initially, the reactor loses a neutron to the 232Th, which then changes to 233Th before quickly decaying into 233Pa, a strong neutron poison which will stay in the reactor until long after the reactor is shut down (and so waste energy will need to be dealt with, but radiation may/probably is enough to deal with that), and then it’s likely that the next time the engine is started up, that neutron poison has transmuted into an even more fissile material unless you load the fuel with 233U first (233U has a stronger fission capture cross-section than 235U, which in practical effect reduces the fissile requirements by ~33%)!

This means that the reactor has to go through startup, have a reasonably large amount of control authority to continue to add reactivity to the reactor to counterbalance the fission poison buildup of not only 233Pa, but other fission product neutron poisons and fissile fuel worth degradation (if the fuel element has been used before), and then be able to deal with a potentially more reactive reactor (if the breeding ratio has more of a fudge factor due to the fast ramp-up/ramp-down behavior of this reactor, varying power levels, etc, making it higher in effect than ~1.01/4).

The other potential issue is that if you need less fissile material in the core, every atom of fissile is more valuable in the core than a less fissile fuel. If the vapor entrainment ends up being higher than the effective breeding ratio (i.e. the effect of breeding when the reactor’s operating), then the reactor’s going to lose reactivity too fast to maintain. Along these lines, the 233Pa behavior is also going to need to be studied, because that’s not only your future fuel, but also a strong neutron poison, in a not-great neutronic configuration for your fuel element, so there’s a few complications on that intermediate step.

This is an addressable option, potentially, but it’s also a lot of work on a reactor that already has a lot of work needed to make feasible.

Conclusions

Liquid fueled NTRs (LNTRs) show great promise as a stepping stone to advanced NTR development in both their variations, the bubbler and radiator variants. The high specific impulse, as well as potentially high thrust-to-weight ratio, offer benefits for many interplanetary missions, both crewed and uncrewed.

However, there are numerous challenges in the way of developing these systems. Of all the NTR types, they are some of the least researched, with only a handful of studies conducted in the 1960s, and a single project in the 1990s. These projects have focused on a single family of fuels, and those have not been able to be tested under fission power for various neutronic and reactor physics behaviors necessary for the proper modeling of these systems.

Additionally, the interactions between the fuel and propellant in these systems is far more complex than it is in most other fuel types. Only two other types of NTR (the droplet/colloid core and open cycle gas core NTRs) face the same level of challenge in fissile fuel retention and fuel element mass entrainment that the LNTR faces, especially in the bubbler variation.

Finally, they are some of the least well-known variations of NTR in both popular and technical literature, with only a few papers ever being published and only short blurbs on popular websites due to the difficulty in finding the technical source material.

We will continue to look at these systems in the next two blog posts, covering the bubbler-type LNTR in the next one, and the radiator type in the one following that. These blog posts are already in progress, and should be ready for publication in the near term.

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References

General

Specific Impulse of a Liquid Core Nuclear Rocket, Barrett Jr 1963 https://arc.aiaa.org/doi/abs/10.2514/3.2141?journalCode=aiaaj

ANALYSES OF VAPORIZATION IN LIQUID URANIUM BEARING SYSTEMS AT VERY HIGH TEMPERATURES Kaufman and Peters 1965 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19660002967.pdf

VAPOR-PRESSURE DATA EXTRAPOLATED TO 1000 ATMOSPHERES (1.01~108 N/m2) FOR 13 REFRACTORY MATERIALS WITH LOW THERMAL ABSORPTION CROSS SECTIONS Masser 1967 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19670030361.pdf

VAPOR-PRESSURE DATA EXTRAPOLATED TO 1000 ATMOSPHERES FOR 10 REFRACTORY ELEMENTS WITH THERMAL ABSORPTION CROSS SECTIONS LESS THAN 5 BARNS Masser 1967 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19680016226.pdf

Bubbler

A Technical Report on the CONCEPTUAL DESIGN – STUDY OF A LIQUID-CORE NUCLEAR ROCKET, Nelson et al 1963 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19650026954.pdf

Radiator

“PERFORMANCE POTENTIAL OF A RADIANT-HEAT-TRANSFER LIQUID-CORE NUCLEAR ROCKET ENGINE,” Ragsdale 1967 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19670030774.pdf

HEAT- AND MASS-TRANSFER CHARACTERISTICS OF AN AXIAL-FLOW LIQUID-CORE NUCLEAR ROCKET EMPLOYING RADIATION HEAT TRANSFER, Ragsdale et al 1967 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19670024548.pdf

“FEASIBILITY OF SUPPORTING LIQUID FUEL ON A SOLID WALL NUCLEAR ROCKET CONCEPT,” Putre and Kasack 1968 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19680007624.pdf

The Liquid Annular Reactor System (LARS) Propulsion, Powell et al 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19910012832.pdf

Categories
Nuclear Thermal Systems Test Stands

Fluid Fueled NTRs: A Brief Introduction

 Hello, and welcome back to Beyond NERVA! This is actually about the 6th blog post I’ve started, and then split up when they ran more than 20 pages long, in the last month, and more explanatory material was needed before I discussed the concepts I was trying to discuss (this blog post has also been split up multiple times).

I apologize about the long hiatus. A combination of personal, IRL complications (I’ve updated the “About Me” section to reflect this, but those will not affect the type of content I share on here), and the professional (and still under wraps) opportunity of a lifetime have kept me away from the blog for a while. I want to return to Nuclear Thermal Rockets (NTRs) for a while, rather than continuing Nuclear Electric Propulsion (NEP) power plants, as a fun, still-not-covered area for me to work my way back into writing regularly for y’all again.

This is the first in an extensive blog series on fluid fueled NTRs, of three main types: liquid, vapor, and gas core NTRs. These reactors avoid the thermal limitations of the fuel elements themselves, increasing the potential core temperature to above 2550 K (the generally accepted maximum thermal limit on workable carbide fuel elements), increasing the specific impulse of these rockets. At the same time, structural material thermal limits, challenges in adequately heating the propellant to gain these advantages in a practical way, fissile fuel containment, and power density issues are major concerns in these types of reactors, so we’re going to dig into the weeds of the general challenges of fluid fueled reactors in general in this blog post (with some details on each reactor type’s design envelope).

Let’s start by looking at the basics behind how a nuclear reactor can operate without any solid fuel elements, and what the advantages and disadvantages of going this route are.

Non-Solid Fuels

A nuclear reactor is, at its basic level, a method of maintaining a fission reaction in a particular region for a given time. This depends on maintaining a combination of two characteristics: the number of fissile atoms in a given volume, and the number and energy of neutrons in that same volume (the neutron flux). As long as the number of neutrons and the number of fissile atoms in the area are held in balance, a controlled fission reaction will occur in that area.

Solid Core Fuel Element, image DOE

The easiest way to maintain that reaction is to hold the fissile atoms in a given place using a solid matrix of material – a fuel element. However, a number of things have to be balanced for a fuel element to be a useful and functional piece of reactor equipment. For an astronuclear reactor, there are two main concerns: the amount of power produced by the fission reaction has to be balanced by how much thermal energy the fuel element is able to contain, and the fuel element needs to survive the chemical and thermal environment that it is exposed to in the reactor. (Another for terrestrial reactors is that the fuel element has to contain the resulting fission products from the reaction itself, as well as any secondary chemical pollutants, but this isn’t necessarily a problem for astronuclear reactors, where the only environment that’s of concern is the more heavily shielded payload of the rocket.) 

This doesn’t mean that a reactor has to use a solid fuel element. As the increasingly well known molten salt reactor, as well as various other fluid fueled reactor concepts, demonstrate, the only requirement is the combination of the number of fissile atoms and the required energy level and density of neutrons to exist in the same region of the reactor. This, especially in Russian literature, is called the “active zone” of the reactor core. This can be an especially useful as a term, since the reactor core can contain areas that aren’t as active in terms of fission activity. (A great example of this is the travelling wave reactor, most recently investigated – and then abandoned – by Terrestrial Energy.) But more generally it’s useful to differentiate the fueled areas undergoing fission from other structures in the reactor, such as neutron moderation and control regions in the reactor. The key takeaway is that, as long as there is enough fuel, and the right density of neutrons at the right energy, then a sustained – and controlled – fission reactor has been achieved.

The obvious consequence is that the solid fuel element isn’t required – and in the case of a nuclear thermal rocket, where the efficiency of the rocket is directly tied to the temperature it can achieve, the solid fuel is in fact a major limitation to a designer. The downside to this is that, unlike solids, fluids tend to move, especially under thrust. Because the materials used in a solid fueled rocket are already at the extremes of what molecular bonds can handle, this means that either very clever cooling or very robust containment methods need to be used to keep the rest of the reactor from destroying itself.

Finally, one of the interesting consequences of not having a solid fuel element is that the reactor’s power density (W/m^2) and specific power (W/kg) can be increased in proportion to how much coolant can be used in theory, but in practice it can be challenging to maintain a high power density in certain types of fluid fueled reactors due to the high rate of thermal expansion that these reactors can undergo. There are ways around this, and fluid fueled reactors can have higher power densities than even closely related solid fueled variants, but the fact that fluids are able to expand much more than solids under high temperatures is an effect that should be taken into account. On the other hand, if the fluid expands too much, it can drop the power density, but not necessarily the specific mass of the system.

Types of and Reasons for Fluid Fuels

Fluid fuels fall into three broad categories: liquids, vapors, and gasses. There are intermediate steps, and hybrids between various phase states of fuel, but these three broad categories are useful. While liquid fuels are fairly self-explanatory (a liquid state fissile material is used to fuel the core, often uranium carbide mixed with other carbides, or U-Mo, but other options exist), the vapor and gas concepts are far less straightforward overall. The vapor core has two major variants: discrete liquid droplets, or a low pressure, relatively low temperature gaseous suspension similar to a cloud. The gas core could be more appropriately called a “plasma core,” since these are very high temperature reactors, which either mechanically hold the plasma in place, or use hydrodynamic or electrodynamic forces to hold the plasma in place.

However, they all have some common advantages, so we’ll look at them as a group first. The obvious reason for using non-solid fuel, in most cases, is that they are generally less thermally limited than solid fuels are (with some exceptions). This means that higher core temperatures, and therefore higher exhaust velocity (and specific impulse) can be achieved.

Convection pattern in radiator-type
liquid fuel element, image DOE

An additional benefit to most fluid fueled designs is that the fluid nature of the fuel helps mitigate or eliminate hot spots in the fuel. With solid fuels, one of the major challenges is to distribute the fissile material throughout the fuel as evenly as possible (or along a specifically desired gradient of fissile content depending on the position of the fuel element within the reactor). If this isn’t done properly, either through a manufacturing flaw or migration of the fissile component as a fuel element becomes weakened or damaged during use, then a hot spot can develop and damage the fuel element in both its nuclear and mechanical properties, leaning to a potentially failed fuel element. If the process is widespread enough, this can damage or destroy the entire reactor.

Fluid fuels, on the other hand, have the advantage that the fuel isn’t statically held in a solid structure. Let’s look at what happens when the fuel isn’t fully homogeneous (completely mixed) to understand this:

  1. A higher density of fissile atoms in the fuel results in more fission occurring in a particular volume.
  2. The fuel heats up through both radiation absorption and fission fragment heating.
  3. The fuel in this volume becomes less dense as the temperature increases.
  4. The increased volume, combined with convective mixing of cooler fuel fluids and radiation/conduction from the surface of the hotter region cools the region further.
  5. At the same time, the lower density decreases the fission occurring in that volume, while it remains at previous levels in the “normally heated” regions.
  6. The hot spot dissipates, and the fuel returns to a (mostly) homogeneous thermal and fissile profile.

In practice, this doesn’t necessarily mean that the fuel is the same temperature throughout the element – this very rarely occurs, in fact. Power levels and temperatures will vary throughout the fuel, causing natural vortices and other structures to appear. Depending on the fuel element configuration, this can be either minimized or enhanced depending on the need of the reactor. However, the mixing of the fuel is considered a major advantage in this sort of fuel.

Another advantage to using fluid fuels (although one that isn’t necessarily high on the priority list of most designs) is that the reactor can be refueled more easily. In most solid fueled reactors, the fissile content, fission poison content, and other key characteristics are carefully distributed through the reactor before startup, to ensure that the reactor will behave as predictably as possible for as long as possible at the desired operating conditions. In terrestrial solid reactors, refueling is a complex, difficult process, which involves moving specific fuel bundles in a complex pattern to ensure the reactor will continue to operate properly, with only a little bit of new fuel added with each refueling cycle.

PEWEE Test stand, image courtesy DOE

There were only two refuelable NTR testbeds in the US Rover program: Pewee and the Nuclear Furnace. Both of these were designed to be fuel element development apparatus, rather than functional NTRs (although Pewee managed to hit the highest Isp of any NTR tested in Rover without even trying!), so this is a significant difference. While it’s possible to refuel a solid core NTR, especially one such as the RD-0410 with its discrete fuel bundles, the likely method would be to just replace the entire fueled portion of the reactor – not the best option for ease of refueling, and one that would likely require a drydock of sorts to complete the work. To give an example, even the US Navy doesn’t always refuel their reactors, opting for long-lived highly enriched uranium fuel which will last for the life of the reactor. If the ship needs refueled, the reactor is removed and replaced whole in most cases. This reticence to refuel solid core reactors is likely to still be a thing in astronuclear reactors for the indefinite future, since placing the fuel elements is a complex process that requires a lot of real-time analysis of the particulars of the individual fuel elements and reactors (in Rover this was done at the Pajarito Site in Los Alamos).

Fluid fuels, though, can be added or removed from the reactor using pumps, compressed gasses, centrifugal force, or other methods. While not all designs have the capability to be refueled, many do, and some even require online fuel removal, processing and reinsertion into the active region of the core to maintain proper operation. If this is being done in a microgravity environment, there will be other challenges to address as well, but these have already been at least partially addressed by on-orbit experiments over the decades in the various space programs. (Specific behaviors of certain fluids will likely need to be experimentally tested for this particular application, but the basic physics and engineering solutions have been researched before).

Finally, fluid fuels also allow for easier transport of the fuel from one location to another, including into orbit or another planet. Rather than having a potentially damageable solid pellet, rod, prism, or ribbon, which must be carefully packaged to not only prevent damage but accidental criticality, fluids can be transported with far less risk of damage: just ensure that accidental criticality can’t occur, chemical compatibility between the fluid and the vessel it’s carrying, and package it strongly enough to survive an accident, and the problem is solved. If chemical processing and synthesis is available wherever the fuel is being sent (likely, if extensive and complex ISRU is being conducted), then the fuel doesn’t even need to be in its final form: more chemically inert options (UF4 and UF6 can be quite corrosive, but are easily managed with current materials and techniques), or less fissile-dense options (to reduce the chance of accidental criticality further) can be used as fuel precursors, and the final fuel form can be synthesized at the fueling depot. This may not be necessary, or even desirable, in most cases, but the option is available.

So, while solid fuels offer certain advantages over fluid fuels, the combination of being more delicate (thermally, chemically, and mechanically) combine to make fluid fuels a very attractive option. Once NTRs are in use, it is likely that research into fluid fueled NTRs will accelerate, making these “advanced” systems a reality.

Fuel Elements: An Overview

Now that we’ve looked at the advantages of fluid fuels in general, let’s look at the different types of fluid fuels and the proposals for the form the fuel elements in these reactors would take. This will be a brief overview of the various types of fuels, with more in-depth examinations coming up in future blog posts.

Liquid Fuel

A liquid fueled reactor is the best known popularly, although the most common type (the molten salt reactor) uses either fluoride or chloride salts, both of which are very corrosive at the temperatures an NTR operates at. While I’ve heard arguments that the extensive use of regenerative cooling can address this thermal limitation, this would still remain a major problem for an NTR. Another liquid fuel type, the molten metal reactor, has also been tested, using highly corrosive plutonium fuel in the best known case (the Liquid Annular Molten Plutonium Reactor Experiment, or LAMPRE, run by Los Alamos Scientific Lab from 1957 to 1963, covered very well here).

Early bubbler-type liquid NTR, Barrett 1963

The first proposal for a liquid fueled NTR was in 1954, by J McCarthy in “Nuclear Reactors for Rockets.” This design spun molten uranium carbide to produce centrifugal force (a common characteristic in liquid NTRs of all designs), and passed the propellant through a porous outer wall, through the fuel mass, and into the central void in the reactor before it was ejected out of the nozzle.The main problem with this reactor was that the tube was simply too large to allow for as much heat transfer as was ideal to take place, so the next evolution of the design broke up the single large spinning fuel element up into several thinner ones of the same length, increasing the total surface area available for heating the propellant. This work was conducted at Princeton, and would continue on and off until 1973. These designs I generally call “bubblers,” due to the propellant flow path.

Princeton multi-fuel-element bubbler, Nelson et al 1963

One problem with these designs is that the fuel would vaporize in the low pressure hydrogen environment of the bubbles, and significant amounts of uranium would be lost as the propellant went through the fuel. Not only is uranium valuable, but it’s heavy, reducing the exhaust velocity and therefore the specific impulse. Another issue is that there are hard limits to how much propellant can be passed through the fuel at any given time before it starts to splatter, directly tying thrust to fuel volume. 

In order to combat this, a team at NASA’s Lewis Research Center decided to study the idea of passing the propellant only through the central void in the fuel, allowing radiation to be the sole means of heating the propellant. Additional regenerative cooling structures were needed for this design, and ensuring the propellant got heated sufficiently was a challenge, but this sort of LNTR, the radiator type, became the predominant design. Vapor losses of the uranium were still a problem, but were minimized in this configuration.

It too would be cancelled in the late 1960s, but briefly revived by a team at Brookhaven National Laboratory in the early 1990s for possible use in the Space Exploration Initiative, however this program was not selected for further development.

Despite these challenges, liquid core NTRs have the potential to reach above 1300 s isp, and a T/W ratio of up to 0.5, so there is definite promise in the concept.

Droplet/Vapor Fuel

Picture a spray bottle, the sort used for household plants, ironing, or cleaning products like window cleaner. When the trigger is pulled, there’s a fine spray of liquid exiting the nozzle, which contains a mix of liquid and gas. Using a similar system to mix liquids and gasses is possible in a nuclear reactor, and is called a droplet core NTR. This reactor type is useful in that there’s incredible surface area available for radiation to occur into the propellant, but unfortunately it also means that separating the fuel droplets from the propellant upon leaving the nozzle (as well as preventing the fuel from coating the reactor core walls) is a major hydrodynamics challenge in this type of reactor.

Vapor core NTR, Diaz et al 1992

The other option is to use a vapor as fuel. A vapor is a substance that is in a gaseous state, but not at the critical point of the material – i.e. at standard temperature and pressure it would still be a liquid. One interesting property of a vapor is that a vapor is able to be condensed or evaporated in order to change the phase state of the substance without changing its temperature, which could be a useful tool to use for reactor startup. The downside of this type of fuel is that it has to be in an enclosed vessel in order to maintain the vapor state.

So why is this useful in an NTR? Despite the headaches we’ve just (briefly, believe it or not) discussed in the liquid fuels section, liquid fuel has a major advantage over gaseous fuel (our next section): the liquid phase is far better at containing its constituent parts than the gas phase is, due to the higher interatomic bond strength. At the same time, maintaining a large, liquid body can be a challenge, especially in the context of complex molecular structures in some of the most chemically difficult elements known to humanity (the actinides and transuranics). If the liquid component is small, though, it’s far easier to manage the thermal distribution, as well as offering greater thermal diffusion options (remember, the heat IN the fissile fuel needs to be moved OUT of it, and into the propellant, which is a direct function of available surface area).

The droplet core NTR offers many advantages over a liquid fuel in that the large-scale behavior of the liquid fuel isn’t a concern for reactor dynamics, and the aforementioned high surface area offers awesome thermal transfer properties throughout the propellant feed, rather than being focused on one volume of the propellant.

Vapors offer a middle ground between liquids and gasses: the fissile fuel itself is in suspension, meaning that the individual molecules of fissile fuel are able to circulate and maintain a more or less homogeneous temperature. 

This is another design concept that has seen very little development as an NTR (although NEP applications have been investigated more thoroughly, something that we’ll discuss the application and complications of, for an NTR in the future). In fact, I’ve only ever been able to find one design of each type designed for NTR use (and a series of evolving designs for NEP), the appropriately named Droplet Core Nuclear Rocket (DCNR) and the Nuclear Vapor Thermal Reactor (NTVR).

Droplet Core NTR, Anghaie et al 1992

The DCNR was developed in the late 1980s based on an earlier design from the 1970s, the colloid core reactor. The original design used ultrafine microparticles of U-C-ZR carbide fuel, which would be suspended in the propellant flow. This sort of fuel is something that we’ll look at more when covering gas core NTRs (metal microparticles are one of the fuel types available for a GCNTR), but the use of carbides increases the fuel failure temperature to the point that structural components would fail before the fuel itself would, leading to what could be called an early pseudo-dusty plasma NTR. The droplet core NTR took this concept, and applied it to a liquid rather than solid fuel form. We’ll look at how the fuel was meant to be contained before exiting the nozzle in the next section, but this was the main challenge of the DCNR from an engineering point of view.

The NVTR was a compromise design based on NERVA fuel element development with a different fissile fuel carrier. Here, the fuel (in the form of UF4) is contained within a carbon-carbon composite fuel element in sealed channels, with interspersed coolant channels to manage the thermal load on the fuel element. While significant thrust-to-weight ratio improvements were possible, and (in advanced NTR terms) modest specific impulse gains were possible, the design didn’t undergo any significant development. We’ll cover containment in the next section, and other options for architectures as well.

Gas Fuel

Finally, there are gas core NTRs. In these, the fuel is in gaseous form, allowing for the highest core temperatures of any core configuration. Due to the very high temperatures of these reactors, the uranium (and in general the rest of the components in the fuel) become ionized, meaning that a “plasma core” is as accurate a description as a “gas core” is, but gas remains the convention. The fuel form for a gas core NTR has a few variants, with the most common being UF6, or metal fuel which vaporizes as it is injected into the core. Due to the high temperatures of these reactors, the UF6 will often break down as all of the constituent molecules become ionized, meaning that whatever structures will come in contact with the fuel itself (either containment structures or nozzle components) must be designed in such a way to prevent being attacked by high temperature fluorine ions and hydrofluoric acid vapors formed when the fluorine ions come in contact with the propellant.

Containing the gas is generally done in one of three ways: either by compressing the gas mechanically in a container, by holding the gas in the middle of the reactor using the gas pressure from the propellant being injected into the core, or by using electromagnets to contain the plasma similarly to how a spherical tokamak operates. The first concept is a closed cycle   gas core (CCGCNTR, or GC-C), while the second two are called open cycle gas core NTRs (OCGCNTR or GC-O), because while the first one physically contains the fuel and prevents fission products, unburned fuel, and the previously mentioned free fluorine from exiting in the exhaust plume of the reactor, the open cycle’s largest problem in designing a workable NTR is that the vast majority (often upwards of 90%) if the uranium ends up being stripped away from the plasma body before it undergoes fission – a truly hot radioactive mess which you don’t want to use anywhere near anything sensitive to radiation and an insanely inefficient use of fissile material. There are many other designs and hybrids of these concepts, which we’ll cover in the gas core NTR series, and will look briefly at the containment challenges below.

Fluid Fuel Elements: Containment Strategies

Fluid fuels are, well, fluid. Unlike with a solid fuel element, as we’ve looked at in the past, a fluid has to be contained somehow. This can be in a sealed container or by using some outside force to keep it in place.

Another issue with fluid fuels can be (but isn’t always) maintaining the necessary density to achieve the power requirements for an NTR (or any astronuclear system, for that matter). All materials expand when heated, but with fluids this change can be quite dramatic, especially in the case of gas core NTRs. Because of this, careful design is required in order to maintain the high density of fissile fuel necessary to make a mass-efficient rocket engine possible.

This leads to a rather obvious conclusion: rather than the fuel element being a physical object, in a fluid fueled NTR the fuel element is a containment structure. Depending on the fuel type and the reactor architecture, this can take many forms, even in the same type of fuel. This will be a long-ish review of the proposed fuel containment strategies, and how they impact the performance of the reactors themselves.

One thing to note about all of these reactor types is that 235U is not required to be the fissile component in the fuel, in fact many gas core designs use 233U instead, due to the lower requirements for critical mass. (According to most Russian literature on gas core NTRs, this  reduces the critical mass requirements by 1/3). Other options include using 242mAm, a stable isomer of 242Am, which has the lowest critical mass of any fissile fuel. By using these types of fuels rather than the typical 235U, either less of the fuel mass needs to be fissile (in the case of a liquid fueled NTR), or less fuel in general is needed (in the case of vapor/gas core NTRs). This can be a double-edged sword in some systems with high fuel loss rates (like an open cycle gas core), which would require more robust and careful fuel management strategies to prevent power transients due to fuel level variations in the active zone of the reactor, but the overall reduction in fuel requirements means that there’s less fuel that can be lost. Many other fissile fuel types also exist, but generally speaking either short half-lives, high spontaneous fission rates, or expense in manufacture have prevented them from being extensively researched.

Let’s look at each of the design types in general, with a particular focus on gas core NTRs at the end.

Liquid FE

For liquid fuels, there’s one universal option for containing the fuel: by spinning the fuel element. However, after this, there’s two main camps on how a liquid fueled NTR interacts with the propellant. The original design, first proposed in the 1950s and researched at least through the 1960s, proposed the use of either one or several spinning cylinders with porous outer walls (frits), which would be used to inject the propellant into the reactor’s active region. For those that remember the Dumbo reactor, this may be familiar as a folded flow NTR, and does two things: first, it allowed the area surrounding the fuel elements to be kept at very low temperatures, allowing the use of ZrH and other thermally sensitive materials throughout the reactor, and second it increases the heat transfer area available from the fuel to the propellant. Experiments (using water as a uranium analog) were conducted to study the basics of bubble behavior in a spinning fluid to estimate fuel mass loss rates, and the impact of evaporation or vaporization of various forms of uranium (including U metal, UC2, and others) were conducted. 

This concept is the radiator type LNTR. Here, rather than the folded flow used previously, axial flow is used: the H2 is used as a coolant for reactor structures (including the nozzle) passing from the nozzle end to the ship end, and then injected through the central void of each of the fuel elements before exiting the nozzle. This design reduces the loss of fuel mass due to bubbling in the fuel, but adds an additional challenge of severely reducing the amount of surface area available for heat transfer from the fuel to the propellant. In order to mitigate this, some designs propose to seed the propellant with microparticles of tungsten, which would absorb the significant about of UV and X rays coming off the fuel, and turn it into IR radiation which is more easily absorbed by the H. At the designed operating temperatures, this reactor would dissociate the majority of the H2 into monatomic hydrogen, increasing the specific impulse significantly.

In all these designs, there is no solid clad between the fuel itself and the propellant, because this means that the hottest portion of the fuel element would be limited by how high the temperature can reach before melting the clad. Some early LNTR designs used a mix of molten UC2 and ZrC/NbC as a fuel element, with the ZrC meant to migrate to the upper areas of the fuel element and not only provide neutron moderation but reduce the amount of erosion from the propellant. It may be possible to use a liquid metal clad as a barrier to prevent mass erosion of the fissile fuel in a metal fueled reactor as well, and possibly even add some neutron moderation for the fuel element itself. However, the material would need to have not only a very high boiling point, high thermal conductivity, low reactivity to both hydrogen and the fuel, and low neutron capture cross section, it would also need to have a high vapor pressure in order to prevent erosion from the propellant flow (although I suppose adding additional clad during the course of operation would also be an option, at the cost of higher propellant mass and therefore lost specific impulse).

Droplet/Vapor FE

Now let’s look at the vapor core NTR.

NVTR fuel element, Diaz et al 1992

Containing the UF4 vapor in the NVTR vapor core NTR is done by using a sealed tube embedded in a fuel element, which is then surrounded by propellant channels to carry away the heat. Two configurations were proposed in the NTVR concept: the first used a large central cavity, sealed at both ends, to contain the vapor, and the second design dispersed the fuel cylinders in an alternating hexagonal pattern throughout the fuel element. The second option provides a more even thermal distribution not only within the fuel element itself, but across the entire active zone of the reactor core.

Droplet core NTRs are very different in their core structure. Rather than having multiple areas that the fissile fuel is isolated in, the droplet core sprays droplets of fissile fuel into a large cylinder, which is spun to induce centrifugal force. The fuel is kept away from the walls of the reactor core using a collection of high-pressure H2 jets, injecting the propellant into the fuel suspension and maintaining hydrostatic containment on the fuel. The last section of the reactor core, instead of using hydrogen, injects a liquid lithium spray to bind with the uranium, which is then carried to the walls of the reactor due to the lack of tangential force. The fuel is then recirculated to the top of the reactor vessel, where it is once again injected into the core.

This hydrostatic equilibrium concept is very similar to how many gas core NTRs operate (which we’ll look at below), and has proven to be the biggest Achilles’ Heel of these sorts of designs. While it may be theoretically possible to do this (the lower temperatures of the droplet core allow for collection and recirculation, which may provide a means of fissile fuel loss reduction), many of the challenges of the droplet core are very similar to that of the open cycle gas core, a far more capable engine type.

Gas Core

Gas core containment is possibly the most complex topic in this post, due to the sheer variety of possible designs and extreme engineering requirements. We’ll be discussing the different designs in depth in upcoming blog posts, but it’s worth doing an overview of the different designs, their strengths and weaknesses, here.

Closed Cycle

One half of the lightbulb configuration, McLafferty et al 1968

The simplest design to describe is the closed cycle gas core, which in many ways resembles a vapor core NTR. In most iterations, a sealed cylinder with a piston at one end (similar in many ways to the piston in an automobile engine), is filled with UF6 gas. This is compressed in order to reach critical geometry, and fission occurs in the cylinder. The walls of the cylinder are generally made out of quartz, which is transparent to the majority of the radiation coming off the fissioning uranium, and is able to resist the fluorination from the gas (other options include silicon dioxide, magnesium oxide, and aluminum oxide). Additionally, while the quartz will darken under the heat, the radiation actually “anneals” the quartz to keep it transparent, and coolant is run through the cylinder to maintain the material within thermal limits; a vortex is induced during fission which, when properly managed, also keeps the majority of the uranium (now in a charged state) from coming in contact with the walls of the chamber as well, reducing thermal load on the material. Some designs have used pressure waves in place of the piston to induce fission, but the fluid-mechanical result is very similar. This results in a lightbulb-like structure, hence the common nickname “nuclear lightbulb.” One variation mentioned in Russian literature also uses a closed uranium loop, circulating the fissile fuel to minimize the fission product buildup and maintain the fissile density of the reactor.

The main advantage to these types of designs is that all fission products and particle radiation are contained within the bulb structure, meaning that fission product and radiation release into the environment is eliminated, with only gamma and x-ray radiation during operation being a concern. However, due to the fact that there’s a solid structure between the fuel element and the propellant, this engine is thermally limited more than any other gas core design, and its performance in both thrust and specific impulse suffers as a result.

Open Cycle

The next very broad category is an open cycle gas core. Here, there is usually no solid structure between the fissioning uranium and the propellant, meaning that core temperatures can reach astoundingly high temperatures (sometimes limited only by the melting temperature of the materials surrounding the active reactor zone, such as reflectors and pressure vessel). Sadly, this also means that actually containing the fuel is the single largest challenge in this type of reactor, and the exhaust tends to be incredibly radioactive as a result, On the plus side, this sort of rocket can achieve isp in the tens of thousands of seconds (similar to or better than electric propulsion), and also achieve high thrust.

Perhaps the easiest way to make a pure open cycle gas core NTR is to allow the fuel and the propellant to fully mix, similarly to how the droplet core NTR was done, and either ensure all (or most) of the fissile fuel is burned before leaving the rocket nozzle. Insanely radioactive, sure, but with a complete mixing of the fissioning atoms and the propellant the theoretically most efficient transfer of energy is possible. However, the challenge of fully fissioning the fuel in such a short period of time is significant, and I can’t find any evidence of significant research into this type of gas core reactor.

Due to the challenges of burning the fissile fuel completely enough during a single pass through the reactor, though, it is generally considered required to maintain a more stable fissile structure within the reactor’s active region. Maintaining this sort of structure is a challenge, but is generally done through gasdynamic effects: the propellant injected into the reactor is used to push the fuel back into the center of the reactor. This involves the use of a porous outer wall of the reactor, where the hydrogen propellant is inserted at a high enough pressure and evenly enough spaced intervals to counterbalance both the tendency of the plasma to expand until it’s not able to undergo fission and the tendency of the fuel to leave the nozzle before being burned.

Soviet-type Vortex Stabilized open cycle, image Koroteev et al 2007

The next way is to create a low pressure stagnant area in the center of the core, which will contain the fissile fuel. In order to maintain this type of pressure differential, a solid structure is usually needed, generally made out of a high temperature refractory metal. In a way this is a hybrid closed/open cycle gas core (even though the plasma isn’t in direct contact with the structure of the reactor itself), because the structure itself is key to generating this low pressure zone necessary for maintaining this plasma body fuel element. This type of NTR has been the focus of Russian gas core research since the 1970s, and will be covered more in the future.

Spherical gas core diagram, image NASA

As I’m sure most of you have guessed, fuel containment is a very complex and difficult problem, and one that’s had many solutions over the years (which we’ll cover in a future post). Most recent gas core NTR designs in the US are based on the spherical gas core. Here, the plasma is held in the center of the active zone using jets of propellant from all sides. This is generally called a porous wall gas core NTR, and while it takes advantage of any vortex stabilization that may occur in the fuel, it does not rely on it; in many ways, it’s a lot like an indoor skydiving arena with air jets blowing from all sides. This design, first proposed in the 1970s, uses high pressure propellant to contain the fuel in the reactor, and in many designs the flow can be adjusted to deal with the engine being under thrust, pushing the fuel toward the nozzle in traditional design configurations. Most designs suffer from massive erosion of the fuel by shear forces from the propellant eroding the fuel from the outside edge, but in some conceptual sketches this can be gotten around using non-traditional nozzle configurations which have a solid structure along the main thrust axis of the rocket. (More on that in a future post. I’m still trying to track down the sources to fully explain that pseudo-aerospike concept).

Hybrid gas core diagram, Beveridge 2017

The most promising designs as far as fuel loss rates minimize the amount of plasma required to maintain the reaction. This is what’s known as a hybrid solid-gas NTR, first proposed by Hyland in the 1970s, and also one of the designs which has been most recently investigated by Lucas Beveridge. Here, the fissile fuel is split between two components: the high-temperature plasma fuel is used for final heating of the propellant, but isn’t able to sustain fission independently. Instead, a sphere of solid fuel encases the outside of the active zone of the reactor. This minimizes the amount of fuel that can be easily eroded while ensuring that a critical mass of fissile material is contained in the active region of the reactor. This really is less complicated than it sounds, but is difficult to summarize briefly without delving into the details of critical geometry, so I’ll try to explain it this way: the interior of the reactor is viewed by the neutrons in the reactor as a high-density low temperature fuel area, surrounding a low density high temperature fuel area, with the coolant/moderator passing through the high density area and flowing around the low density area, making a complete reactor between these parts while minimizing how much of the low density fuel is needed and therefore minimizing the fuel loss. I wish I was able to make this more clear in less than a couple pages, but sadly I’m not that good at summarizing in non-technical terms. I’ll try and do better on the hybrid core post coming in the future.

All of these designs suffer from massive fuel loss, leading to highly radioactive exhaust and incredibly inefficient engines which are absurdly expensive to operate due to the amount of highly enriched fissile fuel needed. (Because everything going into the reactor needs to fission as quickly as possible, every component of the fuel itself needs to undergo fission as easily as possible.) This is the major Achilles heel of this NTR type: despite the massive potential promise, the fuel loss, and radioactive plume coming off these reactors, make them unusable with current engineering.

There’s going to be a lot more that I’m going to write about this type of NTR, and I skipped a lot of ideas, and variations on these ideas, so expect a lot more in the coming year on this subject.

Cooling the Reactor/Heating the Propellant

Finally there’s cooling, which usually comes in one of two varieties:

  1. cooling using the propellant, as in most NTR designs that we’ve seen, to reject all the heat from the reactor
  2. cooling in a closed loop, as is done in an NEP system
Hybrid gas core with secondary cooling diagram, Beveridge 2017

While the ideal situation is to reject all the heat into the propellant, which maximizes the thrust and minimizes the dry mass of the system, this is the exception in many of these systems, rather than the norm. There’s a couple reasons for this: containing a fluid with fast-moving (or high pressure) hydrogen is challenging because the gas wants to strip away the mass that it comes in contact with (far easier in a fluid than a solid), H2 is insanely difficult to contain at almost any temperature, and these reactors are designed to achieve incredibly high temperatures which can outstrip the available heat rejection area that the reactor designs allow.

Complicating the issue further, hydrogen is mostly transparent to the radiation that a nuclear reactor puts off (mostly in the hard UV/X/gamma spectrum), meaning that it takes a lot of hydrogen to reject the heat produced in the reactor (a common complaint in any gas-cooled reactor, to be fair), and that hydrogen doesn’t get heated that much on an atom-by-atom basis, all things considered.

There’s a way around this, though, which many designs, from LARS on the liquid side to basically every gas core design I’ve ever seen use: microparticle or vapor seeding. This is a form of hybrid propellant, which I mention in my NTR propellants page. Basically, a metal is ground incredibly fine (or is vaporized), and then included in the propellant feed. This captures the high-wavelength photons (due to its higher atomic mass, and greater opacity to those wavelengths as a result), which are re-emitted at a lower frequency which is more easily absorbed by the propellant. While the US prefers to use tungsten microparticles in their designs, the USSR and Russia have also examined two other types of metals: lithium and NaK vapor. These have the advantage of being lower mass, impacting the overall propellant mass less, and also far easier to control fluid insertion rates (although microparticles can act as fluidized materials due to their small size, and maintain suspension in the H2 propellant well). This is a subject that I’ll cover in more depth in the future in the gas core NTR post.

(Side note: I’ve NEVER seen data on non-hydrogen propellant in a liquid-fueled NTR, but this problem would be somewhat ameliorated by using a higher atomic mass fuel, but which one is used will determine both how much more radiation would be directly absorbed, and what kind of loss in specific impulse would accompany this substitution. Also, using other elements/molecules would significantly change the neutronic structure and hydrodynamic behavior of the reactor, a subject I’ve never seen covered in any paper.)

Sadly, in many designs there simply isn’t the heat capacity to remove all of the reactor’s thermal energy through the propellant stream. Early gas core NTRs were especially notorious for this, with some only able to reject about 3% of the reactor’s thermal energy into the propellant. In order to prevent the reactor and pressure vessel from melting, external radiators were used – hence the large, arrowhead-shaped radiators on many gas core NTR designs.

This is unfortunate, since it directly affects the dry mass of the system, making it not only heavier but less power efficient overall. Fortunately, due to the high temperatures which need to be rejected, advanced high temperature radiators can be used (such as liquid droplet radiators, membrane radiators, or high temperature liquid metal radiators) which can reject more energy in less mass and surface area.

Another example, one which I’ve never seen discussed before (with one exception) is the use of a bimodal system. If significant amounts of heat are coming off the reactor, then it may be worth it to use a power conversion system to convert some of the heat into electricity for an electric propulsion system to back up the pure thermal system. This is something that would have to be carefully considered, for a number of reasons:

  1. It increases the complexity of the system: power conversion system, power conditioning system, thrusters, and support subsystems for each must be added, and each needs extensive reliability testing.
  2. It will significantly increase the mass of the system, so either the thrust needs to be significantly increased or the overall thrust efficiency needs to offset the additional dry mass (depending on the desire for thrust or efficiency in the system).
    1. Knock on mass increases will be extensive, with likely additions being: an additional primary heat loop, larger radiators for heat rejection, main truss restructuring and brackets, additional radiation shielding for certain radiation sensitive components, possible backup power conditioning and storage systems, and many other subsystem support structures.
  3. This concept has not been extensively studied; the only example that I’ve seen is the RD-600, which used a low power mode with an MHD that the plasma passed directly through in a closed loop system (more on this system in the future); this is obviously not the same type of system being discussed here. The only other similar parallel is with the Werka-type dusty plasma fission fragment rocket, which uses a helium-xenon Brayton turbine to provide about 100 kWe for housekeeping and system electrical power. However, this system only rejected less than 1% of the total FFRE waste heat.
    1. The proper power conversion system needs to be selected, thruster selection is in a similar position, and other systems would go through similar selection and optimization processes would need to be done. This is made more complex due to the necessity to match the PCS and thermal management of the system to the reactor, which has not been finalized and is currently very inefficient in terms of fissile material. If a heat engine is used, the quality of the heat reduces, meaning larger (and heavier) radiators are needed, as well.

Fluid Fuels: Promises of Advanced Rockets, but Many Challenges to Overcome

As we’ve seen in this brief overview of fluid fueled NTRs, the diversity in advanced NTR designs is broad, with an incredible amount of research having been done over the decades on many aspects of this incredibly promising, but challenging, propulsion technology. From the chemically challenging liquid fuel NTR, with several materials and propellant feed challenges and options, to the reliable vapor core, to the challenging but incredibly promising gas core NTR, the future of nuclear thermal propulsion is far more promising than the already-impressive solid core designs we’ve examined in the past.

Coming up on Beyond NERVA, we will examine each of these types in detail in a series of blog posts, and the information both in this post and future posts will be adapted into more-easily referenced web pages. Interspersed with this, I will be working on filling in details on the Rover series of engines and tests on the webpage, and we may also cover some additional solid core concepts that haven’t been covered yet, especially the pebble-bed designs, such as Timberwind and MITEE (the pebble-bed concept is also sometimes called a fluidized bed, since the fuel is able to move in relation to the other pellets in the fueled section of the reactor in many designs, so can be considered a hybrid system in some ways).

With the holiday season, life events, and concluding the project which has kept me from working as much as I would have liked on here in the coming months, I can’t predict when the next post (the first of three on liquid fueled NTRs) will be published, but I’ve already got 7 pages written on that post, six on the next (bubblers), and 6 on the final in that trilogy (radiator LNTR) with another 4 on vapor cores, and about 10 pages on the basic physics principles of gas core reactor physics (which is insanely complex), so hopefully these will be coming in the near future!

As ever, I look forward to your feedback, and follow me on Twitter, or join the Beyond NERVA Facebook page, for more content!

References

This is just going to be a short list of references, rather than the more extensive typical one, since I’m covering all this more in depth later… but here’s a short list of references:

Liquid fuels

“Analysis of Vaporization of Liquid Uranium, Metal, and Carbon Systems at 9000 and 10000 R,” Kaufman et al 1966 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19660025363.pdf

“A Technical Report on Conceptual Design Study of a Liquid Core Nuclear Rocket,” Nelson et al 1963 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19650026954.pdf

“Performance Potential of a Radiant Heat Transfer Liquid Core Nuclear Rocket Engine,” Ragsdale 1967 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19670030774.pdf

Vapor and Droplet Core

“Droplet Core Nuclear Reactor (DCNR),” Anghaie 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920001887.pdf

“Vapor Core Propulsion Reactors,” Diaz 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920001891.pdf

Gas Core

“Analytical Design and Performance Studies of the Nuclear Light Bulb Engine,” Rogers et al 1973 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19730003969.pdf

“Open Cycle Gas Core Nuclear Rockets,” Ragsdale 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920001890.pdf

“A Study of the Potential Feasibility of a Hybrid-Fuel Open Cycle Gas Core Nuclear Thermal Rocket,” Beveridge 2017 https://etd.iri.isu.edu/ViewSpecimen.aspx?ID=439

Categories
Fission Power Systems Nuclear Thermal Systems

Carbides: Nuclear Thermal Fuels of the Past and Future

Hello, and welcome to Beyond NERVA!

Today, we’re looking at a different kind of fuel element than the ones we’ve been examining so far on this blog, one that promises higher operation temperatures and therefore more efficient NTRs: carbide fuel elements. We’ll also look at a few different options for NTR designs using carbide fuels: the first one being from Russia (and the only NTR to be tested outside the US), the RD-0410/0411 architecture (two different sizes of a very similar reactor type); the second is the grooved ring tricarbide NTR (a modern US design involving a unique fuel element geometry); and, finally, the SULEU reactor (Superior Use of Low Enriched Uranium, another modern US design with many unique reactor architecture and safety features).

700px-NaCl_polyhedra
NaCl cubic structure, which is very similar to the structure of UC. Image via Wikipedia

So, to begin, what are carbides? Carbides are a solid solution of carbon and at least one other, less electronegative element. These materials are known for very high temperature melting points, and are often used in high speed tooling. Tungsten carbide, for instance, is used for both high-speed wood and metal bits, blades, and other tools.

In the NERVA reactors, niobium carbide and zirconium carbide were used as fuel element cladding, to prevent the fuel elements from being aggressively eroded by the hot hydrogen propellant. By the time of the XE-Prime test, the fuel particles suspended in the graphite matrix of the fuel element were uranium carbide, individually coated with zirconium carbide.

These are monocarbide compositions, though. There are other options: tricarbides (with three electronegative components, leading to a different lattice structure, as well as different mechanical and thermal properties) and carbide nitrides (a composite material containing both carbide and nitride structures; nitrides being a similar concept to carbides, but with N instead of C) – a possibility that is apparently of great interest to Russian NTR designers, but more on that later.

Even during Rover, however, the advantages of making the fuel elements themselves out of carbides were known, and research on the fuel elements began as far back as the 1960s in the US. This research included two of the test chambers in the nuclear furnace tests (examined in the Hot Fire Part 2 blog post to a small extent); but these were considered a more advanced follow-on technology, while the graphite fuel elements with encapsulated fuel particles were the ones that were intended to be used for the planned Mars missions.

Carbides have many advantages over many other materials. One example is that carbides are able to be built up with many different processes, most notably chemical vapor deposition (CVD), where a series of chemical precursors are used to deposit the different components in the carbide structure at much lower temperature than the melting – or decomposition – point of the carbide. Another advantage is that they tend to be relatively dimensionally stable when under high heating, meaning they don’t swell that much.

The USSR, on the other hand, decided very early on to commit to using carbide fuel elements for their NTR, and came up with a novel reactor architecture to both take advantage of the high temperatures of the carbide fuel elements, and to deal with the problems that they posed.

One major disadvantage to carbides is that they are prone to cracking… to a rather severe degree. This means that any cladding material needs to be able to handle this cracking. This was seen in the fuel elements in the NF-1 test, where every (U, ZrC)Carbide fuel element had a great deal of splitting; and was one of the reasons that this fuel was not considered the best option for early NTRs, until these issues were worked out.

Another disadvantage to carbides is the difficulty in manufacturing a consistent carbide, especially if multiple different types of electronegative components are used. Often there will be clusters of different monocarbides in what is supposed to be a tricarbide solution, meaning that the physical properties (notably, the fissile properties of the fuel itself) vary at different points in the fuel element. This can be made even worse if the fuel element is exposed to the hot hydrogen propellant stream as the H2 strips away the carbon (forming CH4, C2H2, and a number of other hydrocarbons); it also changes the chemical properties of the solution, sometimes allowing droplets of metal to form at well above their melting point, resulting in various other problems.

Oxides: The Familiar Fissile Chemical Composition

Carbides have been used for nuclear fuel elements for a very long time. The fuel pellets in later Rover and NERVA engines were encapsulated carbide beads spread in a graphite matrix. This allowed the fissile fuel itself to become hotter before decomposition occurred. To understand the advantages, though, we have to compare them to the other uranium-bearing compound that is more frequently used: uranium oxide.

Nuclear_Fuel_Pellets_(14492225000)
UO2 fuel pellets. Courtesy of Areva.

In the oxide fuel pellets, the O2 would separate from the U, causing metallic crystals to form in the fuel pellet, changing its neutronic and chemical properties. To make matters worse, the O2 could then migrate outside the pyrocarbon or ZrC coating, causing chemical reactions in the surrounding graphite. All of this can occur below its melting point of 2,865 C (3,138 K). This changes the neutronic behavior within the fuel elements in different amounts at different locations within the reactor, causing control issues for the operators, and requiring more design work from the engineers to ensure the reactor can deal with these problems.

Another problem with UO2 is that it has very poor thermal conductivity. Temperature gradients of more than a thousand degrees C are seen in terrestrial fuel pellets of UO2 roughly the thickness of a pencil. There are many ways around this,the latest being the use of CERMET fuels, which use very small pellets of UO2, surrounded by refractory metals that are much better thermal conductors; but these metals themselves also limit the temperature the fuel element can operate at (with the new reactor designs that use beryllium for its’ moderation properties, the relatively low, 1,287 C melting point of Be determines the maximum specific impulse of the rocket).

The Advantages to Carbide Fuel Elements

(Chemistry warning! I’ll keep it as light as possible, but…)

Ta Hf Nb and ZrC absorption data, NASA
Neutron absorption spectra of HfC, TaC, NbC and ZrC, ENDF data, image courtesy NASA

Carbides, on the other hand, have some of the highest melting points known to humanity. Tantalum hafnium carbide (Ta4HfC5) has a melting point of 3942 C, the highest known melting point. How high the melting point is depends on a number of factors, including what materials are used and the ratio between those elements in the structure of the carbide itself.

Unfortunately, both tantalum and hafnium have fairly high neutron absorption cross sections, so they are not ideal materials for carbide nuclear fuel elements. These are typically made out of some combination of uranium carbide and either niobium carbide, zirconium carbide, or both.

Another advantage to using carbide fuel elements is that this allows the actual fissile fuel to be more evenly spread throughout the fuel element, creating a more homogeneous (i.e. consistent) fission power profile across the fuel element. This is an advantage to reactor designers, since the more heterogeneous the reactor, the more headache it is for the designer to ensure stable fission behavior in the fuel element. The more consistently the fissile material is spread, the more controllable it is, and the more evenly the power is produced, making the behavior of the reactor more predictable. This has been known since the beginning of nuclear power, and is why later Rover fuel elements were moving away from the coated pellets mixed into graphite style of fuel and toward a composite fuel element, where the uranium carbide fuel was spread in a webwork throughout the graphite matrix of the fuel element.

The Complications of Carbide Fuel Elements

What the actual melting temperature is for a given material is… complicated, though, for a number of reasons.

The first depends on what proportion everything is in, and this is difficult to get consistent. As noted in a recent paper on a unique NTR geometry (which we’ll look at in the next post), getting the perfect stoichiometric ratio (i.e. the ratio between carbon, uranium, and any other elements present) is virtually impossible, so compromises need to be made. Too much carbon, and the temperature drops slightly. Too little carbon, and the material doesn’t mix as well, causing areas that have lower melting points, or higher thermal conductivity, or a number of other undesirable properties.

The second problem is in mixing: a fuel element designer wants to have a material that’s consistent all the way through the fuel element, not discrete little clumps of different materials as one moves through the fuel element. Because of the way that carbide fuel elements are made (DC sintering, a similar process to spark plasma sintering that’s used for CERMET fuel elements), the end result is grains of NbC, ZrC, and UC2 side by side, rather than a mixture (a solid solution, to be precise) of (Nb, Zr, U)C; and each grain has different thermal, neutronic, and chemical properties. It is possible to heat the fuel element, and have the constituents become this ideal solid solution, as was discovered using CFEET for carbide fuel element testing (more on that in the next post as well). This offers hope for more consistent mixing of the elements in the fuel itself, but establishing the correct ratios remains a problem.

Erosion effects Pelaccio et al
Corrosion effects in carbide fuels, Pelaccio et al

There’s one more big problem with carbide fuel elements, though: hydrogen corrosion. Unlike in graphite composite or CERMET fuel elements, the carbon that is stripped away by hot hydrogen is actually chemically bound to the uranium, zirconium, and niobium in the fuel element, not as a material matrix surrounding the chemical components that support fission in the fuel element. This means that if there’s a clad failure, the local ratio of carbon will change, causing free metal to form, either as a pure metal or an alloy, unevenly across the fuel element. This means that hot spots can develop, or parts of the fuel element will melt far below the melting temperature of the carbide the fuel element was originally made of. Flecks or droplets of metal can be eroded into the hot hydrogen stream, potentially causing damage downstream of the fuel element failure. In a worst case scenario, uranium could collect in areas of the reactor that it’s not meant to, creating a power peak in a spot that could be… inconvenient, to say the least.

These are challenges that carbide fuel element designers have always faced, and continue to face today. Careful chemical synthesis will definitely help, but there are limits to this. Preheating the fuel elements after sintering to ensure a more consistent solid solution is already showing considerable advantages in composition, and in material properties as well. Cladding the fuel element with carefully selected clad materials (often ZrC, which is already a component of the carbide fuel element, with a similar coefficient of thermal expansion and good modulus of elasticity), and ensuring consistent high quality application (usually through chemical vapor deposition these days, which has increased in quality and consistency a lot since the days of Project Rover) of the coating will eliminate (or at the least minimize) the erosion effects of graphite.

Another option that I’ve seen mentioned, but have been unable to find much information on, is an idea mentioned in Russian papers about their RD-041X engines: carbides and nitrides (which have a similar chemical composition, but with electronegative components ionically bonded with nitrogen, rather than carbon) in a solid solution. This leads to a more complex chemical structure, and may allow for less erosion of the carbon from the fuel element. Unfortunately, this literature is hard to find; and, when it is available, it hasn’t been translated from Russian. However, according to the most commonly available paper (linked in the references), adding a nitride component to the fuel element may boost the maximum fuel element temperature.

The Other Fuel: Plutonium Carbides

We don’t talk about plutonium much on this blog (yet), but plutonium carbides have been investigated to a certain degree as well. They may not be as attractive as uranium carbide for a number of reasons, but as a potential fuel element, they may show some promise.

Why are they less attractive? First is the neutron fission cross section of Pu is skewed much more to the fast spectrum in Pu than in U. This means that the more moderated the neutron flux, the more likely it is that when a neutron interacts with a nucleus of 239Pu, it won’t fission but continue up the transuranic chain. Many of these elements are also fissile, but again much more so in the fast spectrum. This means that more and more neutron poisons can build up in your core, requiring more and more reactivity to overcome. This also means that when it’s time to decommission the core, it will be much more radioactive than a similar U-fueled reactor (on average, there are of course a lot of factors that go into this). Finally, this means that the core has to have more fuel in it; and, unlike with uranium, there’s no “Low Enriched Plutonium,” the fraction of 238Pu (used in RTGs) or 240Pu (which is gamma-active, and a headache) is very low. This is convenient if you’re making fuel elements, but a very different regulatory game than LEU, with huge restrictions on who can work with the fuel element materials for development of an NTR.

Second, 239Pu is illegal to use in space, in accordance with international treaty. Now, LEU235 is also illegal, but that is more likely to change, since it involves having less concentrated fissile material in space, unlike the use of Pu, which is considered a major nuclear proliferation risk, even if it’s out in space. The treaty was written to prevent nuclear weapons in space sneaking in the back door, and Pu has been (in the public’s mind) intimately tied to nuclear weapons development from day 1.

Mixed carbide fuels (containing both uranium and plutonium) have been investigated as an alternative to MOX (mixed oxide) fuels for fast breeder reactors, either in the (U, Pu)C or the (U, Pu)2C3 phases. The usual benefits of carbides over oxides apply to this fuel form: higher metal density and better thermal conductivity being the main two. Due to a number of challenges, including very low oxygen requirements for fabrication, minimal experience with fabrication of mixed carbide fuels, and the general lack of information on the chemistry of PuC, this is a largely unknown field, but research is being conducted to extend our knowledge of these areas.

At present, these materials are a curiosity, although they could lead to advanced fuels for terrestrial use. Until their chemistry and materials properties are better known, however, it is unlikely we’ll see an NTR powered with mixed carbide fuel.

How are Carbides Used in NTRs?

Traditional” Carbide-fueled NTRs

Reactor cell w carbide, Finseth
Sketch of NF-1 Carbide Fuel Test Cell with Carbide Fuel Cross Section, Finseth 1991

In Rover, carbide fuel elements were researched that had a very similar form factor to the fuel elements. These were hexagonal in cross section, about 33 cm long, and clad in NbC. The main difference was that there was a single large hole, rather than nineteen small holes. An NTR was in the early concept design, but was never put through the reactor geometry refinement process.

Designs have been proposed over the years using hexagonal prism fuels similar to Rover carbide fuel elements, but none are currently under development, as far as I can see. This doesn’t exclude their use, even with LEU, but NASA and the DOE are currently pursuing other fuel element geometries.

The Other Tradition: Russian NRE

twistedRibbon05Russia has been in the nuclear thermal rocket business for as long as the United States, but their design philosophy is hugely different from the American one. Just like NASA and the DOE don’t use the term “nuclear thermal rocket” (NTR), instead preferring “nuclear thermal propulsion” (NTP), Roscosmos and Rosatom (who work together to develop the Russian program) use the term “nuclear rocket engine”, or NRE.

The design changes start with the fuel element design, extend through the basic geometry of the reactor and beyond, and have major implications for testing and materials options with this system.

First, let’s look at the fuel elements. One of the considerations for fuel element design is the amount of surface area that can be contacted by the propellant. Thermal transfer is determined by the thermal emissivity of the fuel element material, and the thermal conductivity and transparency of the propellant. The more surface area, the more heat is transferred, given those previously mentioned factors are equal. Rather than using a fuel prism as American NTP has done, with increasing number of holes through a hexagonal prism, the Russian NRE uses what is commonly known as a “twisted ribbon” design, where a rectangular prism (or any number of other designs, such as a cluster of rods, square prisms, or other shapes- see the image above for the variations that have been tested) is rotated along its long axis. A cluster of these fuel elements are placed in a tube (known as a calandria, similar to the design used in CANDU reactors, but with different geometry and materials), ending in a nozzle at the end of the bundle.

twistedRibbon01

Unlike with the American NTP designs, there isn’t a single fuel element cluster running down the center of the NRE. In fact, there’s NO fuel at the center of the reactor. The Russians don’t use a homogeneous reactor design, either for neutronic power or thermal energy. The center of the reactor, rather than containing fuel, contains moderator. Since the fuel elements (and therefore all the sources of heat for the reactor) are spread around the periphery of the reactor core, rather than being evenly distributed in the core, this means that a moderator with much lower melting temperatures can be used in the design (both zirconium and lithium hydrides are mentioned as options, neither of which would be able to withstand the temperatures of a homogeneous core NTR). This also means that a bimodal design (known in the Russian program as a “nuclear power and propulsion system,” or NPPS, rather than BNTP as NASA calls it) can integrate the working fluid channels more easily into the design without a complete redesign of either the fuel element or the header and footer support plates. We’ll cover BNTRs in a later post, including the NPPS, but it’s worth mentioning that this design offers more design flexibility than the traditional, hexagonal prism NTP fuel elements used in American designs.

Fuel Bundle
Loading of fuel bundles into core, Russian film

Finally, due to the fact that a number of fuel element bundles are radially spread across the reactor, an individual fuel bundle can be tested on its’ own in a prototypic neutronic and thermal environment, rather than needing to test the entire NTP core in a hot fire test, as is required for the American designs. This testing has been conducted both at the EWG-1 research reactor [with ten consecutive restarts, a total testing time of 4000 s (although how much was at full power, and what sort of transient testing was done, is unknown), at a maximum hydrogen exhaust temperature of 3100 K, achieving a theoretical specific impulse of 925 s and a power density for the system of 10 Mwt/L] and at the rocket test stand in Semipalatinsk (although those test results are still classified). The Russians have also done full-scale electric heating tests of NRE designs, settling on two: the RD-0410 (35 kN thrust, for unmanned probes – and possibly for proof-of-concept mission use) and RD-0411 (~392 kN of thrust, for crewed missions). Statistics for the RD-0410, based on these electrically heated tests, can be seen below:

NRE Performance, Zukhov et al
NRE Performance Specifications, Zhukov et al

Sadly, there isn’t much more information available about the current NRE designs and plans. We’ll come back to its’ variant, the NPPS, when we look at bimodal designs in the future.

Grooved Ring NTR: Not All American Designs are Hexagonal

Diagram with separate ring, Taylor et al
Fuel element and element cluster, Taylor et al 2017

This is a new NTR design, designed around the use of a (Zr, Nb, U)C fuel element of a very different shape than the traditional hexagonal prism, currently under development at NASA and the University of Tennessee. Just as with the twisted ribbon fuel elements, the fuel element geometry for this NTR has been changed to maximize surface area, and allow for more heat to be transferred to the propellant. This both maximizes the specific impulse and minimizes the amount of propellant needed for cooling purposes (however, H2 remains the best moderator available, and a minimum amount for neutronic reasons will always be needed, even if not for cooling the fuel elements).

The fuel elements are radially grooved discs of uranium tricarbide (Nb, Zr, U)C, although hafnium and tantalum were also investigated (and eliminated due to the much higher neutron absorption rates). The hydrogen flows from the outside of a stack of these fuel elements, separated with beryllium spacers, and then flows down a central channel.

Due to the unique geometry of this fuel element design, much optimization was needed for the groove depth, hydrogen flow rates, uranium density in the fuel element (in the initial design, 95% enriched HEU was used for ease of calculation, however with additional optimization and research into stoichiometric ratios of U with the other electronegative components, the authors believe less than 20% enrichment is possible), and other factors.

CFEET Testing
CFEET testing of fuel element, Taylor et al

Thermal testing, including hot hydrogen testing using CFEET, has been carried out at Marshall SFC, using vanadium as a surrogate for depleted uranium. The team hopes to continue to refine such factors as manufacturing consistency, improved mixing of the solid solution of the carbide, and other manufacturing issues in carbide fuels, before hopefully moving on to electrically heated carbide tests using depleted uranium (DU) to optimize the carbide chemistry of uranium itself.

This NTR offers the potential for 3000 C exhaust temperatures at 4 psi. Unfortunately, due to the preliminary nature of the work that has been carried out to date (this reactor design is less than a year old, unlike the designs that have gone through decades of development of not just the fuel elements themselves but also the engine system), thrust and theoretical specific impulse using this reactor design has not been determined yet.

This novel fuel element form offers promise, though, of a new NTR fuel element geometry that allows for better thermal transfer to the propellant, and the team are performing extensive material fabrication and optimization experiments to further our understanding of tricarbide fuel element performance and manufacture, in addition to developing this new fuel element form factor.

Tricarbide Foam Fuel Elements: You REALLY Want Surface Area? We Got It!

This is a very different carbide fuel form, with novel manufacturing practices yielding a truly unique fuel element.

Most solid core fuel elements are chunks of material, no matter what form they take (and we’ve seen quite a few forms in this post already), with the propellant flowing around or through them; either through holes that are milled or drilled, the surface of the twisted ribbon, or through grooves cut in a disc. That’s not the case here, however!

65 ppi RVC foam tricarbide, Youchistan et al
Youchistan et al

The team at Sandia National Laboratory, Ultramet, Inc., and the University of Florida have come up with a new take on carbide manufacture, utilizing chemical vapor deposition (CVD, a common method of carbide manufacture) on a matrix that starts life as open-pore polyurethane foam. This foam is then pyrolized (baked… ish) to form a carbonized skeleton of the foam structure. This is then heated, and CVI (chemical vapor infiltration, a variation of CVD) processes are used to impregnate the carbonized skeleton with uranium, zirconium, and niobium; turning the structure’s outer surfaces to (U, Zr, Nb)C carbide (a number of factors affect the depth of the penetration). Then, CVD is used to coat the new carbide structure with ZrC or NbC to clad the more chemically fragile tricarbide, and protect it from the H2 propellant that will flow through the open pores remaining after this carbidization and CVD coating process.

Foam cross section Youchistan et alThis concept has been tested using tantalum as a surrogate for uranium (a common choice for pre-depleted uranium electrically heated testing of carbide fuel elements), with two foam densities, 78% and 85%; leading to the discovery that there’s a trade-off: the 78% had better thermal transfer properties, but the 85% offers more volume for the fissile material, meaning that lower enrichment was possible.

The team members at Sandia made a preliminary MCNP model of an NTR for use with these fuel elements, with a number of unique features. This was a heterogeneous core (meaning uneven fuel distribution), with 60% porosity foam fuel, using yttrium hydride for the moderator (which has to be maintained below 1400 K by circulating hydrogen between it and the fuel), and with a Be reflector. For these initial modeling calculations, 93.5% enriched HEU was used. It was discovered that a 500 MWt NTR was possible using this fuel form, but due to the unoptimized, preliminary nature of this design, values for thrust and specific impulse are still up in the air.

INSPI at the University of Florida will be conducting electrically heated hot hydrogen tests on DU-containing tricarbide fuel foams in the temperature range of 2500-3000 K, as these fuel foams become available, although the timeline for this is unclear. However, research is continuing in this truly novel fuel form, and the possibilities are very promising.

Carbides: Great Promise, with Complications

As we’ve seen in this post, carbide fuel elements offer many advantages for designers of nuclear thermal rockets. Their high melting point allow for higher propellant exhaust temperatures, improving the specific impulse of an NTR. Their ability to have their properties manipulated by changing the composition and ratio of the components allows a material designer to optimize the fuel elements for a number of different purposes. Their strength allows for truly novel fuel forms that give an NTR designer a lot more flexibility in design. Finally, their similar coefficient of thermal expansion, and often good modulus of elasticity, make them important materials for use in all NTRs, not just those fueled with fissile-containing carbides.

However, the chemical and materials properties of these substances, manufacturing processes required to consistently produce them, and modes of failure (including the implications for these types of failure in an operating NTR) show that there’s still much work to be done in order to bring carbide fuel elements to the same level of technological maturity currently enjoyed by graphite composite fuel elements.

The promise of carbides, though, makes developing the chemistry of fissile-bearing carbides of all forms, perhaps most especially uranium tricarbides, a worthy goal for the advancement of nuclear power in space. This research has been ongoing for decades, continues worldwide, and is bearing fruit.

References

Uranium Dioxide

Uranium Dioxide Wikipedia page: https://en.wikipedia.org/wiki/Uranium_dioxide

Thermodynamic and Transport Properties of Uranium Dioxide and Related Phases, IAEA 1965 http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/24/071/24071477.pdf

Thermal Conductivity of Uranium Dioxide, IAEA 1966: http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/34/065/34065217.pdf

Uranium Carbide

Nuclear Thermal Propulsion Carbide Fuel Corrosion and Key Issues; Pelaccio et al 1994

https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19950011802.pdf

Evaluation of Novel Refractory Carbide Matrix Fuels for Nuclear Thermal Propulsion; Benensky et al 2018

https://www.researchgate.net/publication/324164284_Evaluation_of_Novel_Refractory_Carbide_Matrix_Fuels_for_Nuclear_Thermal_Propulsion?ev=publicSearchHeader&_sg=c-5LZwXyF_AvDFznQi5AHQdF_KlJYE7p8Qiii6M3H6nNFhlKWQ1oQ8Kh8B40UI13RMZ_7DTLgNp1KgE

Ultra High Specific Impulse Nuclear Thermal Rocket, Part II; Charmeau et al 2009

https://www.osti.gov/servlets/purl/950459

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion; Taylor et al 2017

https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20180002033.pdf

Mixed Carbides

Plutonium Tricarbide Isomers: A Theoretical Approach; Molpeceres de Diego, 2015 https://uvadoc.uva.es/bitstream/10324/13556/1/TFM-G413.pdf

Mastery of (U, Pu)C Carbide Fuel: From Raw Materials to Final Characteristics, Christelle Duguay, 2012

https://www.epj-conferences.org/articles/epjconf/pdf/2013/12/epjconf_MINOS2012_01005.pdf

Rover Carbide Fuel Elements

Nuclear Furnace-1 Test Report, LA-5189MS; Kirk et al 1973

https://ntrl.ntis.gov/NTRL/dashboard/searchResults/titleDetail/LA5189MS.xhtml

Performance of (U, Zr)C-Graphite (Composite) and of (U, ZR)C (Carbide) Fuel Elements in the Nuclear Furnace 1 Test Reactor, LA-5398-MS; Lyon 1973

https://www.osti.gov/servlets/purl/4419566

Nuclear Rocket Engine

Russian Nuclear Rocket Engine Design for Mars Exploration Zakirov et al 2007 https://www.researchgate.net/publication/222548572_Russian_Nuclear_Rocket_Engine_Design_for_Mars_Exploration

Ticarbide Grooved Ring NTR

Grooved Fuel Rings For Nuclear Thermal Rocket Engines tech brief; MSFC 2009 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20090008640.pdf

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element; Barkett et al 2013 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20130011208.pdf

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion; Taylor et al 2017 Conference paper: https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20180002033.pdf Presentation Slides: https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20180002060.pdf

Tricarbide Foam Fuel Element

A Tricarbide Foam Fuel Matrix for Nuclear Thermal Propulsion, SAND-2006-3797C; Youchison et al 2006

https://www.osti.gov/servlets/purl/1266203

Categories
Fission Power Systems Low Enriched Uranium Nuclear Thermal Systems Spacecraft Concepts

LEU NTP Part Three: Spacecraft Overview

Hello, and welcome to the Beyond NERVA blog! Today, we continue our in-depth look of NASA’s new nuclear thermal rocket. We briefly looked at the history of NTP (as NASA calls it, “nuclear thermal propulsion”) in part one, and in part two we took a deep dive into the materials that NASA is investigating for its’ new design, ceramic metal (CERMET) fuel elements. Today, we look at the stage and spacecraft itself, with a brief look at some information about the proposed engine design. The next post will focus on the testing and launch safety considerations for an NTP system (as well as some unique guidance, navigation, and control considerations), and we’ll close with a post about other options for using low enriched uranium to fuel a nuclear thermal rocket, this time using advanced carbide fuels.

As we saw in the first post, nuclear thermal rockets are nothing new. The US has built and tested them before, and even successfully tested one in flight configuration. In the second post, we looked more closely at the new materials technologies that are being used to make an even more capable NTR, CERMET fuels, but we also saw that there’s a problem: in order to use low enriched uranium (LEU), the fuel needs large amounts of isotopically separated tungsten, and this has been a major challenge for the supplier. To date, I have been able to find no information about deliveries of even 50% enriched 184W (the needed isotope), much less the more than 90% enriched tungsten needed for the fuel elements that NASA has designed.

So how does NASA plan to address this problem? Well, 184W would be useful for more than NTRs (tungsten is also used as a neutron reflector in the core of certain thermonuclear weapons designs, hence the lack of details on the development process and difficulties associated with it), so just because NASA has not been able to have the process developed doesn’t mean that it won’t be in the future (weapons programs have an easier time getting money than NASA’s nuclear program).

Advances in LEU Nuclear Propulsion

Even if this doesn’t pan out, there’s still other options. The one that caught the public’s attention last year was the signing of a new contract with BWXT. While far from a household name, in the DOE, US Navy, and NASA they are well-known. They helped build the USS Nautilus, and were an early (and currently are the major) supplier of fuel for the US Nuclear Navy. They offer commercial and research fuel resupply and disposal contracts on a number of reactor designs. Since the early 1980’s they have fabricated all of the Department of Energy’s experimental fuel elements (with the exception of KRUSTY, which was fabricated by Y12). They also are a prime contractor for many of NASA’s nuclear-related activities, participating in environmental impact assessments, technical consultancy, and other areas.

They have proposed a new, and thus far poorly described, design for an NTR, using CERMET fuels of a varying composition at different points in the core, to better manage and moderate the neutrons produced during fission. Based on what little information I’ve been able to gather since NETS 2018, according to Michael Eades using molybdenum/tungsten (MoW) as a matrix material for CERMET fuels is apparently as good as using tungsten-184 for both moderation and thermal limits, and this appears to be the path that BWXT will be using moving forward. However, I haven’t had the chance to go over the research yet, and apparently there are some significant changes, so today we’ll focus on the rest of the spacecraft.

It’s likely that the design will be similar to the one proposed by BWXT last year, however, which uses a technique known as “zoned moderation,” where different parts of the reactor are exposed to different neutron flux energies due to the distribution of moderator and reflectors throughout the core. There is no reason that this technique will not work using natural uranium as a fuel element matrix material rather than the beryllium and tungsten that was proposed for the earlier design.

BWXT Core Large
BWXT LEU NTP Core Configuration, image via NASA

Even this isn’t the end of the options, though… another fuel form, advanced tricarbides, also offer the potential for LEU use, in particular in the Superior Use of Low Enriched Uranium (SULEU) reactor design, which we’ll cover in a future blog post (not only is it a carbide-fueled reactor, but there are enough other nifty design features that this reactor definitely needs its’ own post).

The Beginnings of the Modern Astronuclear Thermal Era

Beginnings are important in nuclear engineering. A retired Lawrence Livermore engineer once told me “nuclear design is evolutionary,” and that’s especially the case with this system.

In many ways, the new dawn for nuclear propulsion was in 1990, at “Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop,” held in Albuquerque, NM from June 10-12. This conference happened during the death throws of the Strategic Defense Initiative (SDI, Reagan’s Star Wars program), when funding was being cut for every single program associated with SDI. There was a nuclear thermal rocket design that was part of SDI, Project Timberwind, which used a pebblebed reactor to increase fuel surface area, but this was a relatively early casualty of Congressional budget cuts. In addition, as noted by the DOD Office of the Inspector General, the program not only was over budget and consistently failed to meet benchmarks, but there were questions about the predicted performance of the engine as well.

pebbleBed01
Project Timber Wind Core Cross-Section, image via Atomic Rockets

In order to continue moving forward with nuclear thermal propulsion, the three main stakeholders in the US came together to present their concepts. The conference started by establishing what had come before, and also a “baseline” was established to compare new ideas to the legacy NERVA designs that were available at the time. New subsystems, techniques for handling cryogenic hydrogen, and materials all would combine to make even an NTR using the same fuel elements and reactor geometry would be greatly improved over what was available in 1973. After this, presentations were made about many different aspects of nuclear thermal propulsion, from launch safety concerns to materials advances to advanced concepts for liquid, vapor, and plasma fueled reactor designs. The focus was on getting the most bang for the very few bucks that would be coming down the pipeline, and on the difficulties of testing any design under the regulatory regime that was in place at the time (which was not hugely different from what we face now).

The only way at the time to be able to test an engine was to capture ALL of the exhaust that passed through the reactor, which meant that you had to be able to store it somewhere – a very big somewhere. It also meant that more exhaust translated rather directly into greater expense for testing, so the thrust of the designed engine was specified to be in the 25,000 klbf range, similar to what the Pewee engine provided during Project Rover. We aren’t going to be getting into testing options in this post (that’s the next one), but keep in mind that the ability to fully test an NTP system on Earth is going to be a critical requirement, and the size of the engine (and the amount of propellant that needs to be captured) directly affect how difficult (and expensive) it will be to test as a full system.

NERVAPewee2, AEC 1971
Pewee on test stand, 1971, courtesy DOE

SNRE Diagram, Borowski 2010
Small Nuclear Rocket Engine, Borowski NASA GRC

This conference also can be seen as the birthplace of the immediate predecessor of the LEU NTP system: Stan Borowski’s Small Nuclear Rocket Engine (updated version of the design available here). This engine is a Pewee-class, graphite composite modern design, and still remains an option for a small NTR, although one that would require HEU rather than the LEU that NASA is currently focusing on. This engine also allows for bimodal operation, where oxygen is injected into the hot hydrogen stream and then ignited, giving a big boost to the amount of thrust available (at the cost of specific impulse), which became the LANTR, or Lunar Oxygen Augmented Nuclear Thermal Rocket, for faster trips to and from the Moon (which is so close the increased thrust is a big boost to mission capabilities).This design was investigated for more than a decade as NASA’s primary NTP concept, and remains an area of active research at both NASA’s Glenn Research Center and at Oak Ridge National Laboratory, where many of the fuel fabrication techniques are being investigated in depth.

 

Many of the parts of the SNRE stage remain in the LEU NTP stage. These include non-nuclear components, the basic shape and volume of the stage, nuclear and thermal shielding, and while slightly changed the mission requirements are largely the same. Perhaps the only major-ish change is the difference in the proposed launch vehicle: in the early days of NASA’s Design Reference Mission for Mars 5.0, the Ares V rocket was still on the drawing boards – and in the mission plans. This rocket ended up being canceled with the end of the Constellation program, and a slightly smaller replacement, the Space Launch System, was proposed. The difference between the rockets necessitates a re-juggling of what is launched on each orbital launch, due to the decrease in payload capacity from 140 mt to low Earth orbit to 110 mt, but this is something that can be addressed relatively easily by using slightly smaller modules (although often it ends up requiring an extra launch). So, looking back over the design proposals leads to a lot of insights into NASA’s thinking and requirements for their new nuclear rocket design.

Since there’s still a lot of questions about the exact form that the engine itself will take, let’s go ahead and look at the rest of the NTP stage: shielding, non-nuclear components, propellant tankage, size, and mission requirements.

Radiation Shielding

Radiation shielding is essential on any nuclear system, but nuclear propulsion presents a number of challenges that are unique. Of course, the biggest part of this effort is to reduce crew dose during operation, but the engine components that aren’t in the core of the reactor (such as turbopumps, actuators, etc) will also be materially attacked by the radiation flux coming off the reactor, and the fuel itself can be heated as well (which causes local boiling and cavitaton in the turbopumps – and both are bad news). For a deep dive into this subject, I cannot recommend Winchell Chung’s Atomic Rockets page on the subject highly enough.

Because space is pretty much the definition of the middle of nowhere, the only thing that really needs to be shielded is the spacecraft itself. To save mass, the easiest thing to do is to stick your nuclear reactor on one end of the ship, your crew quarters on the other, with the fuel tanks in the middle. Then, place a radiation shield between the nuclear reactor and the rest of the ship. This is called a shadow shield, because the ship stays in the shadow of this radiation shield.

Shadow Shield, Caffrey MSFC 2017
Shadow Shield with Radiation Flux, Caffrey MSFC 2017

What is Radiation, and How is it Shielded?

There are four main types of radiation that come off a nuclear reactor: alpha, beta, and neutron radiation form the group known as particle radiation, and high energy photons like hard UV, x rays, and gamma rays, form the ray portion of the radiation flux. (These, obviously, are ionizing radiation types. Non-ionizing radiation, on the other hand, is not a danger to the crew, and is something to just be dealt with or exploited by the ship – infrared, for instance, also copiously comes off a nuclear reactor, but that heat energy is the entire point of running the thing!) This second type of radiation is made up entirely of photons, but of much higher frequency than visible light. The first type, however, is a salad of different particles: alpha particles are bare helium-4 nuclei (and as such have a charge of +2), beta radiation is a high-energy electron (charge -1), and neutron radiation is made up of – surprise! – neutrons (and as such have no charge)

The easiest way to consider shielding is to split the two types of radiation up and deal with them separately, since they have almost opposite requirements for stopping them. So, let’s look at particles first, and then rays.

Particle radiation is stopped through a process called “elastic scattering,” which is most easily pictured by a pair of balls, one moving and one stationary, hitting each other. Depending on the mass and velocity of each ball, they will reflect off each other, and momentum from the ball that WAS moving gets at least partially transferred into the ball that was stationary. How much is transferred depends on the relative masses of the balls: the closer the masses, the more energy can be transferred. So, to stop any of the particle radiation types, low-atomic-mass (low-Z) materials are ideal, usually something chock full of hydrogen. This lends itself to water, hydrates, and organic materials. However, the atoms in the material will obviously be bounced around, and over time the material will become degraded. As an additional challenge, ray-type radiation will break the hydrocarbon chains that make up organic shielding, and as such will degrade these types of materials even further (for a look more into these effects, check out the organically moderated reactor concept, only think of the challenges of slowing these particles to a stop). The other option doesn’t work for neutron radiation, but works on the other two: electromagnetic confinement. This is the approach used by the concept of a mini-magnetosphere for a ship, being explored by NASA and Rutherford-Appleton Laboratory, and diverts the particles before they come in contact with any materials using powerful electromagnets. This is a very advanced concept, and often is far more massy than using a passive material. In addition, alpha and beta particles aren’t able to leave the reactor’s pressure vessel anyway, so they generally aren’t a concern.

There ARE particles that are a concern, though: neutrons, which are uncharged but slowed and stopped the same way, and galactic cosmic rays, or GCRs. These are higher-Z nuclei that have been ejected from some high energy event, like a supernova, and come tearing through space at a significant percentage of the speed of light. They cause a large amount of damage on the atomic level, and are a major source of the radiation flux that astronauts receive. Unfortunately, because they’re moving so fast they’re virtually impossible to stop or divert, unless you have a strong electromagnetic field blanketing whatever you’re protecting (and even then, because they have so much mass, it’s hard to get them completely diverted, just slowed a bit).

Neutron Shielding

Neutrons are basically unheard of in space, however, so dealing with those is easier: you just have to worry about blocking the reactor from the rest of the ship. This can be done using a number of materials, often very high in hydrogen. During NERVA, a lot of study was put into neutron shielding, and many of the concepts were discovered to be impractical, either due to manufacturing difficulties or material mass, but two stand out: lithium hydride (LiH) and boron carbide (B4C). LiH is the most effective neutron shield per unit mass, and if the lithium is enriched such that only 6Li is used (lithium with an atomic mass of 6), it becomes a very effective neutron shield as well, since it wants to capture another neutron to become 7Li. The downsides are that it doesn’t work nearly as well in high-neutron flux environments, does not conduct heat well, is thermally limited to prevent dissociation of Li and H, and is highly reactive so requires some sort of cladding material to prevent chemical reactions. Boron carbide, on the other hand, is the most effective shield per unit volume, especially when 10B is used (one of the best neutron poisons available). The lack of hydrogen makes it less effective as a moderator of the neutrons that aren’t captured by the boron, though, and it has 20% more mass than a LiH shield of similar shielding characteristics. This is already an off-the-shelf product, though, and the use of 10B will not change these well-established manufacturing procedures, so it remains a very attractive option, especially if smaller, individual shields (spot shields) are needed for individual components, such as the stepping motors used to control any control drums used in the design.

Gamma and X-Ray Shielding

Rays, on the other hand, tend to be simpler to stop – assuming you can handle large amounts of mass! For lower-energy photons, when they come into contact with an atom, they are absorbed by an electron in the electron cloud, which jumps to a higher energy state, then drops down, emitting a slightly lower energy photon in the process. This effect is how neon lights are produced, or how chemicals can be identified by spectral emission. This is where lead shielding comes in for terrestrial reactors (and magnetite-heavy concrete, along with other design features), and lead is commonly used in shadow shield designs for the same reason. However, any high atomic mass element (HZE) can be a reasonably effective shadow shield, and depleted uranium (238U) is sometimes used as a shield for compact reactors due to its greater density and atomic number. The down side to this method of shielding, however, is that it’s heavy, and heavy is the LAST thing that you want on your spaceship. Unfortunately, for complicated reasons I’m not going to get into here, there’s no way to effectively reflect these high-energy photons, so this really is the only way that we are able to deal with them.

Keep in mind, for the main payload (in this case the crew quarters) there is plenty of other mass in the way. This includes the tanks of propellant, the material the tanks are made out of, structural components to transfer the thrust from the engine to the payload without destroying the ship, support equipment… all of these will absorb or reflect radiation to a greater or lesser extent. We’ll look at the propellant and tanks separately, but keep in mind that while the majority of the shielding is provided by the radiation shield, this doesn’t mean that this is the only shielding available.

The exact size and composition of the shield is going to change, depending on the final design of the engines that will be used, but there shouldn’t be a huge variation in the type or quantity of radiation coming off one form of solid core NTR versus another, so only minor tweaks to the shield composition should be necessary. For a good example of the types of changes that may be necessary, an analysis of the Kilopower reactor by McClure and Poston shows how shielding requirements change as fuel type changes [Insert link].

NTP-Specific Shielding

Flux from NTR in silicon and neutron, Caffery et al
Radiation flux from NTR, Caffery et al

Starting in 2014, a team of researchers at Oregon State University and Marshall Spaceflight Center led by Jarvis Caffery has been examining NASA’s shielding requirements for NTP. In a nutshell, the goal is to REDUCE the overall radiation exposure to the crew for the length of the mission by reducing the overall mission flight time, reducing the crew’s exposure to much more damaging galactic cosmic rays and HZE particle radiation from events like supernovae and neutron star collisions. This doesn’t mean that there aren’t short-term radiation limits that NASA has to work within, or career doses of radiation that are a severe limitation to current mission planners. Given current NASA radiation dose limits, it’s actually impossible to use chemically propelled rockets, because the crew would reach their lifetime dose limit either on the surface of Mars, or on the trip back. NASA is re-examining these limits, and recent legislation that has been proposed to study low-dose radiation exposure may end up significantly changing these requirements in the future.

Human Dose Limits, Caffery et al
Human Dose Limits, Caffery et al via NASA

Caffery et al suggest that in order to maximize the benefit of radiation shielding for the available mass budget, it may be best to concentrate on combined shielding around the crew habitat, to deal with the radiation flux coming off both the reactor and from the environment, rather than concentrating more mass in the shadow shield. However, they also note that using HZE shielding (like tungsten, lead or uranium) near the crew habitat is something to be avoided, since this is how you get brehmstrahlung, and either gamma or x-rays flood your crew cabin.

For the shielding between the reactor and the rest of the ship, this is certainly going to be more than one shield, and the main one is likely to be a composite shield, for a variety of reasons. Various parts of the rocket engine itself will need to be shielded to ensure the more sensitive components are exposed to as-low-as-practicable neutron fluxes. Perhaps the two most important are the stepping motors that will likely be used for the control systems and the turbopumps. Depending on the design that ends up being used, the turbopumps may be on the “hot” side of the main shadow shield, or if a significant part of the shielding occurs in the main structure of the ship and so the shadow shield is reduced in mass, these pumps may be exposed to too high a neutron (or gamma) radiation flux. These can be shielded by secondary shields – in this case possibly B4C, because it not only is a more effective shield per unit volume, but also moderates the neutrons that interact with it less than something rich in hydrogen, leading to lower neutron absorption rates into the mechanical assembly.

The main shield has many options, but there are definite limits to what can be done. Any one concept isn’t going to be good enough, there’s going to need to be a solution that addresses many tradeoffs and problems. This leads to the composite main shadow shield, a concept we’ve seen before in the Kilopower system.

LANL Kilopower screencap Rad Shield
Kilopower Radiation Shield, larger torii are DU, with inserts of LiH. Image courtesy Los Alamos NL

Looking at the Kilopower shield, there are layers of an HZE material (in this case tungsten, but DU is another good option), with thin layers of LiH sandwiched between. This means that the neutron moderation benefits of the LiH – and therefore the likelihood that the neutron will be slowed enough to be absorbed – are spread through the bulk of the shield.

LiH is one of the best neutron shields out there (the best by mass), especially when enriched with Li6, but it has a number of problems, including chemical and thermal stability. Especially in the case of the Li6-enriched variety, a lot of energy will be deposited here as the neutrons are first slowed, and then absorbed, which means that heating can be significant, and unfortunately LiH isn’t the best conductor out there. Dissociation of LiH into lithium metal and H2, which then will either form pockets of gas that weaken the surrounding material, or is lost through outgassing, can occur if the thermal load gets too high.

In order to mitigate this, and also increase the chances of neutron capture and energy deposition in the more thermally conducive HZE shielding plates, the LiH is spread through the shield. This allows for the LiH to be in a small enough sheet to allow for needed thermal dispersion into the more thermally conductive (and less sensitive) metal components. It also means that the secondary gamma emissions from the neutron moderation and capture have plenty of shielding to stop them before they reach the end of the shield.

A design using B4C would have less volume, but more mass. This material is already something that’s commonly used in machine tools all over the world, and even enriching the boron to increase its’ likelihood of absorbing neutrons won’t change those manufacturing techniques significantly. One option studied by McCafferty et al was a pebblebed design, where spheres of B4C would be packed into a casing made out of some structural material. This allows the already better thermal properties of B4C to be maximized, while maintaining the shielding properties of the material by minimizing available ray paths for radiation through the material. Due to its’ higher mass, this material hasn’t been studied as extensively as LiH, but offers some distinct advantages, and so this is was explored more thoroughly in the 2015 paper I linked above. With its’ machinability (and long industrial application), thermal conductivity and resistance, and lower-volume shielding properties, this is a material that will likely show up in many designs, if not necessarily for a main shield then definitely for secondary shields.

While the flux is going to be highest when the reactor is operating, this does not mean that the only radiation flux coming off the reactor is during operation. Once fission occurs in the reactor core, the entire reactor becomes irradiated – the longer it operates, and the higher power it operates at, the more radioactive the partially used fuel and exposed reactor components will become. This means that the highest radiation flux coming off the reactor will likely be in the final burn of the nuclear fuel’s life (and the reactor itself if it’s not designed for refueling), when it will also likely be pretty much exhausted of fuel. This is the worst case that must be designed for, and unfortunately the one most sensitive to decisions that haven’t been made yet.

Unfortunately, at the moment the design for the shield will be up in the air. Until a number of decisions have been finalized in the engine design process, including fuel type, enrichment, neutron spectrum, and others, only options and broad outlines are able to be proposed. Another challenge brought up by the authors is that the primary tools used to model time-dependent dosing calculations, the MCNP code released by Los Alamos National Labs, isn’t exactly the best at these sorts of calculations. Because of this, testing of any shielding system will be needed.

The Propellant Tanks

Any propulsion stage needs propellant to work, and in the case of NTRs the ideal propellant is one of the most difficult to work with: hydrogen. Hydrogen is the lightest of the elements, and as such has a bad habit of being able to seep through just about everything, weakening it in the process. Cryogenic cooling can significantly reduce its’ bulk, but it remains incredibly bulky even under cryogenics, and its’ very low liquefaction point means that maintaining it in cryogenic storage is a major challenge.

Hydrogen Boil Off Rates with TPS
Multi-month hydrogen boil off rates with MLI Thermal Protection, Rapp 2016

This is the hydrogen boil-off problem, and it’s something that has vexed every rocket designer to use LH2 since the beginning of the space age (and many chemical engineers in the decades since it has been discovered). In low Earth orbit (LEO), H2 tends to boil off at predictable rates due to a number of factors which define the quirks of various systems. In addition, as the hydrogen seeps through the structures of the tank and spacecraft, these get weakened and brittle, known as hydrogen embrittlement. Add in the large volume that H2 requires, and this can be one of the most challenging propellants to use in a rocket, and many of the challenges dealt with in the Rover program were actually related to using H2 propellant, which hadn’t been done in the US before.

Completed LH2 Tank on VAC at Michoud Assembly Facility
SLS LH2 main tank under construction 2016, image via NASA

There are two ways to deal with this problem: first is with a purely passive system, as is done in launch vehicles, and the second is to use an actively cooled system to minimize or eliminate hydrogen boiloff. The first option, unfortunately, isn’t an option for the NTP stage (due to very long mission times), but the passive cooling technology is still used, based on the LH2 tank design for the Space Launch System. This tank is made out a special aluminum/copper/lithium alloy (Al 2195), which is a high-strength, weldable alloy. Currently, new welding techniques are being used with this alloy on the construction of the SLS main tank at NASA’s Michoud facility by Boeing, which will also improve the quality of an NTP propellant tank as well.

 

STS Main Tank Cutaway
Space Shuttle External Tank, image via Wikipedia

Surrounding the main tank is a thermal protection system (TPS), commonly a foam insulation (in the Apollo SIV-B third stage, this was polyethylene foam internal to the tank, which resulted in less than 10% boil-off during the LEO insertion and TMI burn phases of the mission), and a sun shield as well (likely, in the case of a longer-term mission, seen as a gold foil coating on the stage around the propellant tanks). Additional TPS techniques have continued to be investigated, including use of H2 gasses during the boiling being used as a vapor to cool the rest of the TPS through careful venting of the overpressure coming off the H2 tank, and the use of cryocoolers to further cool the thermal shields and mitigate heat transfer. However, it seems unclear exactly which materials would be used for a long-term cryogenic storage TPS for the NTP, and this could be a major problem for such a long-duration mission as a manned Mars mission, which would require the H2 to be maintained for over a year. An example of what has been flown would be the Power Reactant Storage Hydrogen Tank, which lost 2.03% of its’ reactant per day over the 21 day lifetime of the system. This leads to a huge increase in the needed H2 for an extended mission, and a corresponding loss of payload capacity. Even more modern systems lead to a boil-off rate of about 6%/month, which is incredibly prohibitive for as extended mission as a Mars crewed mission.

 

Nuclear + Rockets: Always Complicated

When I started this project, it seemed (relatively) easy: nuclear power, while complex, isn’t unknowable. Rocket propulsion is far more complicated in so many ways than nuclear thermal rockets: only heat is necessary, not the finicky balance between fuel and oxidizer. Sure, there are a thousand details, but that’s what engineering is for.

There is a truth to this, but one of the simplest systems is a wonderful example of why this subject is so difficult to address. Propellant tanks are, in theory, fairly simple: it’s a fancy thermos, with given rates of boil-off that can be adjusted by improving the insulation of the system. Heat mitigation is primarily needed from the solar environment, which is a task that all spacecraft have to address

With a nuclear reactor, there are two additional vectors for thermal heating, both from the reactor. First, there’s gamma ray heating, caused by the remaining gamma radiation after the primary shadow shield and support equipment. A small amount of this is coming from the fission reactions themselves, but the bulk of the shield will absorb these particles. The larger component comes from the neutron flux coming off the reactor, either through elastic collisions of the neutrons and hydrogen in the tank – which slow the neutron and accelerate the hydrogen, heating it – or through the secondary gamma radiation caused by these collisions. As each neutron is slowed (thermalized), it is more likely to interact with the next atomic nucleus, increasing the number of reactions while reducing the energy of each of those interactions, until the neutron is finally captured. These resulting gamma rays are far more easily absorbed by heavier (higher-Z) atoms than lighter ones, so while it’s more unlikely that they will be absorbed by the propellant, the support structures and tank itself will be heated by these interactions, transferring the heat to the propellant through conduction.

Gamma and Neutron Heating, Taylor et al 2015
Energy deposition from gamma and neutron heating, Taylor et al 2015

According to a 2015 paper by B.D. Taylor et al of NASA’s Marshall Spaceflight Center, neutron interactions with liquid H2 drop to effectively zero after less than 50 cm of penetration into the tank itself (due to hydrogen’s excellent moderation properties), and gamma heating becomes the major source of nuclear-caused thermal heating after about 15 cm.

Thermal and convective behavior of NTR tank, Taylor 2015
Convective Behavior in NTR Propellant Tank, Taylor et al 2015

Due to the unique nature of the internal heating caused by the radiation flux (rather than the still-present external heating caused by background radiation that is a more well-understood problem), thermal stratification and complex convective cycles are far more likely to develop in an NTR’s propellant tanks. This can be mitigated by careful construction of baffling, and possibly with mixing equipment internal to the tank itself.

Enter Active Cooling: The Zero Boil-Off Tank

If LH2 boiloff was eliminated, not only would your propellant not be leaking away constantly, but the hydrogen embrittlement would be reduced or eliminated as well. In addition, according to some studies the launch mass of a LH2 system could be reduced by 20% or more if boiloff was eliminated for cislunar space missions, between the mass of the H2 and the required larger tankage requirements. While thermal shielding is a huge help (such as the gold foil seen on many spacecraft), the ambient temperature of space is still higher than the boiloff temperature, so active cooling is needed. In addition, the propellant will be warmed by both gamma rays and neutrons that weren’t absorbed by the shadow shield, and so needs to be actively cooled to prevent even faster boiloff. This problem is so severe, in fact, that NASA no longer plans to try and use just passive cooling techniques to get to Mars.

Enter the Zero-Boiloff Tank, a design that NASA began researching in 2006 with the Florida Solar Energy Center. This system uses a multi-stage cryocooler and hydrogen densification system to ensure continuous cooling of the cryo H2. This design started as a small (150 L, due to facility safety regulations) dewar, built at the FSEC, surrounded by a storage vessel. Later tests used a much larger tank, closer to what would be used for a rocket propellant tank, either for a stage or a propellant depot.

MHTB Schematic
MHTB Ground Test Article Schematic (Non-flight configuration), image courtesy NASA

This is a system that we’ll go more into depth on in its’ own post, so we’re going to look at it more briefly than we typically do here in the interests of blog post length.

CRYOTE 3D Cutaway, NASA
CRYOTE System Cutaway, image via NASA

In short, a ZBO tank uses integral cryocoolers to maintain the propellant below the boiling temperature of the H2. This definitely adds dry mass and complexity to the system, but by significantly reducing or eliminating boil-off, the overall mass needed for the system to complete the mission requirements is reduced by a large amount. This can be paired with vapor-cooled shielding and passive TPS to optimize the mass of the system.

 

This is still a very active area of research, since it directly impacts chemical as well as NTR systems. From the development of a small, desktop breadboard system, to a larger, outdoor system, an on-orbit technology demonstration mission (CRYOTE), and continued research into system components and optimization, much research is still being done to optimize system mass and usability. As such, the final design of the propellant tanks is still very much up in the air.

GTA at MSFC
Test Article at MSFC, Image Courtesy NASA

There is one advantage that the ZBO designs have over traditional tank designs for NTR use: the internal support structure will act as an additional shield down the center line of the spacecraft, protecting the payload more than just the LH2 remaining in the tanks.

LH2 Shielding Goes Away Through the Mission

As propellant is expended during a burn, there will be less mass between the payload and the reactor, meaning that secondary radiation protection will decrease the longer the engines burn.

Empty vs full prop tank radiation flux AR
Radiation Flux of Empty vs. Full Propellant Tanks, via Atomic Rockets

This is a problem for the payload, because the flux coming off the reactor increases the longer its’ burned (due to fission product decay, ambient delayed neutron flux, and increased reactivity requirements to overcome neutron poisoning in the fuel elements). As mentioned above, the internal structures of a zero boil-off tank mitigate this problem somewhat, but they aren’t so large that they would completely fill the entire center line of the spacecraft between the reactor and the payload. However, there have been some designs that retain a column of H2 in the tanks, even when “empty,” which mitigate this. There is a mass loss if this is done, but depending on acceptable radiation dose to the payload, and the radiation flux coming off the reactor, this may be a good decision for some spacecraft designs, especially smaller ones where the distance between the reactor and the payload bus is smaller than on larger spacecraft (for instance, a lunar shuttle vs. a Mars spacecraft).

LEU NTP Mars Vehicle

The LEU NTP stage’s primary mission is going to be a crewed mission to Mars. This doesn’t mean that the stage can’t be used for other missions, but every project needs a mission, and in this case that mission is NASA’s Mars Design Reference Mission 5.0 with an expected mission date for the 2037 launch window to Mars.

In order to complete this mission, a number of components are going to need to be assembled on-orbit: the core propulsion stage (CPS, containing not only the main engines, but reaction control systems, avionics, a solar electrical power system – no bimodal plans for the basic design – and cryogenic fluid management hardware), an in-line propellant tank (essentially the same as the CPS, but with the engines and shielding replaced with more LH2 tank, and a smaller RCS), a saddle truss with up to 5 LH2 drop tanks (one in-line, the rest attached to the outside of the truss), a smaller saddle truss for payload, a deep space habitat (based on the TransHab design), and an on-orbit manned spacecraft (the Orion module).

The basic design for the propulsion bus hasn’t changed since the design of the Nuclear Cryogenic Propulsion Stage, the immediate predecessor to the LEU NTP stage. One retired DOE engineer of my acquaintance loves to point out “all nuclear design is evolutionary in nature,” and this is one of those times that it clearly shows.

The numbers that are used in this section are based on the HEU version of this stage, optimized for Mars DRM 5.0. They will likely be slightly different, not only due to the new engines but also due to advances in mission design and vehicle optimization, by the time this system is taking its’ first crew to Mars, but they will be very close.

Core Propulsion Stage

Core Stack BB Card 2014
NCPS Core Propulsion Stage as of 2014 (Keep in mind, engine mass will likely be slightly different), image courtesy NASA

Being NASA, NTP design is modular in nature, with the idea being that the same propulsion and power module can be used for multiple mission types by adding additional propellant tankage and support equipment depending on the mission profile and destination. So, a Lunar shuttle may only need the core stage, while extensive additional tankage and payload would be necessary for a Mars mission.

The core propulsion stage for the NCPS (and likely the LEU NTP stage) is approximately 25 meters long and 8.4 meters wide, carries three CERMET-fueled 25 klbf NTP engines (Pewee class) for main ship propulsion, 47.2 metric tons of LH2 propellant, and 15.6 metric tons of reaction control system fuel and oxidizer (NTO/MMH). When launched, it will be fully fueled, with a wet mass of 109.5 mt (dry mass 46.2 mt). This pretty much maxes out the payload capabilities of the Space Launch System, which is the preferred method for lofting the stage into orbit. A composite truss structure provides the structural strength for the stage. The reaction control system is nitrous oxide/monomethel hydrazine hypergolic fueled, based on the Fregat RCS, with 328 s of specific impulse. The main engines are planned to be rated at 900 s isp, but as we’ve seen there are many questions remaining about the actual design of the engine that will be used.

In-Line Propellant Tank

In-line Prop Tank BB Card 2014
In-Line Propellant Tank (as of 2014), image courtesy NASA

Moving up the spacecraft structure, the next module to be launched will be an in-line LH2 tank (ILT), which at 25.7 m is just slightly longer than the CPS, but with the same diameter. This module is very similar to the CPS, but replacing the engines and shadow shield with additional tank volume. The RCS is also smaller on this stage, since it’s not on the end of the stack and therefore needs to apply less force than the rear-most section of the spacecraft. With a dry mass of 29.7 mt, 79.2 mt of usable LH2, and just over 2 mt of RCS fuel/oxidizer, the total wet mass of this module while on the ground is 108.2 mt – once again, constrained by the capabilities of the Space Launch System. A similar composite truss structure is used on this portion as the CPS, and docking adapters on each end are used to secure this module to the CPS aft, and the saddle truss forward.

Saddle Truss with Drop Tanks

Drop Tank BB Card, 2014
Drop Tank with Saddle Truss (as of 2014), image courtesy NASA

The third portion of the spacecraft is a long (27.8 m) saddle truss, which means that the structural components form a cylinder around a central hollow. In this case, that hollow holds an additional in-line drop tank, another part of the RCS, and has the capability to mount additional external drop tanks (this part of the spacecraft is far enough forward that these will be shielded by the shadow shield). With a total dry mass of 29.75 mt, and a total wet mass of 118.4 mt, this portion of the stack carries a minimum of 84 mt of LH2 propellant. Since this is a drop tank, and will be used for trans-Mars injection burns, the ZBO tank will not be used here, leading to LH2 boiloff of approximately 1.54 mt. Once again, this will take up the full launch capabilities of the SLS, and will be the second-to-last module launched.

Mission Payload

The final portion of the spacecraft is the mission payload. In this case, it consists of a smaller saddle truss (containing mission specific payload, an RCS, and a canister for holding cargo, approx 12.14 mt), a fully stocked deep space habitat (TransHab, 51.85 mt fully stocked), and the crewed spacecraft (in this case the Orion spacecraft, but the original design called for the MPCV, Orion’s predecessor, which massed 14.49 mt without fuel). This is the lightest weight of all the launched modules, with 78.8 mt of mass on the pad. It’s possible that this launch may carry additional fuel, but instead it may just take advantage of using a less capable (and therefore less costly) launch vehicle.

The Integrated Stack

MTV Copernicus (NCPS Config), NASA

While in low Earth orbit, and once fully assembled, the Mars crewed spacecraft will mass approximately 414.15 metric tons, delivered by four launches of the Space Launch System. Once assembled, the crew will be delivered to the spacecraft for the beginning of the trip to Mars. This will be the largest spacecraft ever meant to travel ANYWHERE except in low Earth Orbit, and will only be smaller than the International Space Station.

Mission Profile 2014
Tentative SLS-based construction and mission duration (as of 2014), image courtesy NASA

Another nice thing about this spacecraft is that, because it’s so long, and the mass is well-distributed, it will also be the first to use centrifugal artificial gravity. By rotating it end over end, it is possible to induce 1 gee of centrifugal acceleration after the trans-Mars injection (TMI) burns, and slow the rotation down to 0.38 gee by the time of Mars orbital insertion (MOI). Then, the rotation will be stopped, and the MOI burn will take place.

Variants of this design have been proposed since the middle of the 1900’s, both for pure nuclear thermal and bimodal thermal and electric propulsion. The bimodal variant (named by its’ creator, Stan Borowski at NASA’s Glenn Research Center) is the Copernicus – B, and has a single large Hall thruster mounted on the center of mass of the spacecraft. After TMI, and spacecraft spinup, the electric thruster is activated for the Earth-Mars cruise period, turning around at the midway point (RCS and navigational correction on a spinning spacecraft has been demonstrated before, it’s more difficult but completely doable). This reduces the travel time to Mars significantly over pure NTP, but at the cost of a much more complex reactor system (of a type that the US isn’t currently investigating strongly, although most astronuclear companies have considered the idea, including NASA’s prime contractor for NTP, BWXT), a power conversion system, added heat rejection equipment, and the electrical thrusters and propulsion. This design is more complex, however, and the current contracts for NTP focus heavily on the pure reactor core.

Stack Construction

Most mission designs for crewed Mars missions assume more than one vehicle: at least one, often two, NTP powered cargo ships are sent to Mars before the crewed vehicle described here. These cargo missions are usually planned for arriving at Mars, and having systems verification completed, before the manned mission. This does impose an approximately 22 month delay on the manned mission (the time it takes for another launch window to open from Earth to Mars), but on the other hand it ensures that the supplies and resources needed by the astronauts have been delivered safely. These designs use the CPS as described above, as well as the in-line fuel tank, but the additional saddle truss with drop tanks may or may not be necessary, depending on the mass requirements for delivery to Mars and the number of cargo missions. These follow a slower, minimum-energy (Hohmann transfer) TMI profile, whereas the crewed mission will follow a faster transit (both to reduce crew exposure to the interplanetary radiation environment and to maximize surface stay time).

An early (2009) construction plan for a two cargo ship mission (available here, based on the Ares V, the predecessor to the SLS) involved launching two core propulsion stages, to be mounted to two uncrewed cargo ships for a minimum energy transfer to Mars. This involved a total of four launches for the two craft, each of which would have a mass in LEO of about 236 mt. However, based on the launch estimates more recently provided compared to the launch requirements for this version of the mission’s manned vehicle (which requires three launches as opposed to the more recent estimate of four), it is likely that each of these vehicles may require three launches instead of two (the Ares V was designed for 140 mt to LEO, significantly more than the SLS). One other change, however, is that the overall mass of the crewed interplanetary transfer vehicle is only 326 mt, indicating that a significant amount of mass that current plans assume is on the crewed vehicle would be transferred by the cargo missions instead (my guess is that this is because they were planning on 140 mt to LEO for this design study, not 110 mt). These modules would be assembled in LEO before TMI, and until the first burn for leaving Earth orbit, the reactors would not achieve criticality. This makes the reactor effectively radiologically inert, and not a concern to operate around during launch and construction.

SLS Sensitivity Chart

These modules could be assembled at the ISS (assuming it’s still around by the time crewed Mars missions are being launched), or independently in LEO. Details on specific construction methods are sketchy, however with extensive experience in multi-module construction on orbit by most international players involved in the ISS, it shouldn’t pose too great of a challenge – just one with many technical details to work out.

LEU NTP: The Latest Plan to Get to Mars

Nuclear thermal propulsion offers the chance to open far more distant places than humanity has ever set foot to human exploration. While it’s theoretically possible to use chemical or electric propulsion, nuclear thermal propulsion offers far higher efficiency than chemical engines, with high thrust making orbital and interplanetary maneuvering far more rapid than the slow but steady burn of electric thrusters.

Currently, NASA’s plans to go to Mars heavily rely on this promising technology, which was demonstrated over 50 years ago (as we saw in part 1). New requirements in the types of fuel that are able to be used have led to major advances in materials engineering, and open up the possibility of using low enriched uranium (as we saw in part 2). By this point, the basic design for the interplanetary spacecraft is (hopefully) clear.

There remain issues to be dealt with, though: First, the engines need to go through a testing regime that will minimize radiological release to the environment, and be demonstrated to be able to be launched safely, and survive a launch failure without causing an environmental disaster or accidental criticality event; second, the core propulsion stages need to not only be launched, but also be used to their maximum effectiveness to get us to Mars. These will comprise the next two blog posts, which research is already well underway on. After that, I hope to address a different popular fuel form, carbide fuels, which offer even higher operating temperatures, and also address the Russian version of NTR, the RD-0410 “twisted ribbon” architecture, which China has also been experimenting with in recent years.

Additional Reading

Nuclear Thermal Propulsion

Performance Design and Qualification For Engine, NERVA, 75K, Full Flow; Aerojet Nuclear Systems Company, 1970

The Timber Wind Special Access Program Audit Report; US DOD Office of the Inspector General, 1992

The Proceedings of Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop, 1990

Affordable Development And Demonstration of a Small Nuclear Thermal Rocket (NTR) Engine and Stage: How Small is Big Enough?; Borowski et al, NASA Glenn RC, 2016

Robust Exploration and Commercial Missions to the Moon Using LANTR Propulsion and In-Situ Propellants Derived from Lunar Polar Ice (LPI) Deposits; Borowski et al, NASA GRC 2016

Survey of Fuel System Options for Low Enriched Uranium (LEU) Nuclear Thermal Propulsion; Benensky et al, University of Tennessee, 2017

Radiation and Shielding

Types of Radiation video, FermiLab

Gamma Ray Attenuation of Common Materials; McAllister PG Research Foundation 2012

In-Space Radiation Environment and Crew Quarters Shielding

Neutron Astronomy; Casadei, University of Birmingham, 2017

Human Radiation Exposure Tolerance and Expected Exposure During Colonization of The Moon and Mars; L. Joseph Parker, the Mars Society, 2016

Performance Study for Galactic Cosmic Ray Shield Materials; Kim et al, College of William and Mary, NASA Langley Research Center, 1995

Homepage, Rutherford Appleton Laboratory Mini Magnetosphere Project

In-Space Reactor Shielding

Application of Transport Techniques to the Analysis of NERVA Shadow Shields, Capo and Anderson, Westinghouse ANL, 1972

Shield Materials Recommended for Space Power Nuclear Reactors; Kaszubinski, NASA Lewis RC, 1973

The Evaluation of Lithium Hydride for Use in a Space Nuclear Reactor Shield, Including a Historical Perspective; Knolls Atomic Power Lab, Lockheed Martin, 2005

Investigation of Lithium Metal Hydride Materials for Mitigation of Deep Space Radiation, Rojdev and Atwell, NASA Johnson SFC 2016

Aluminum—Titanium Hydride—Boron Carbide Composite Provides Lightweight
Neutron Shield Material, NASA/AEC Fact Sheet, 1967

Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study – Phase I NIAC Final Report; Thibeault et al, Advanced Materials and Processing Branch NASA Langley Research Center, 2012

Shielding Development for Nuclear Thermal Propulsion; Caffrey et al, NASA Marshall SFC and Oregon SU, Conference paper (2015)  and (Presentation (2017)

Integrated NTP Vehicle Radiation Design, Caffrey et al, NASA Marshall SFC, 2018

Auxiliary Support Systems for NTR

 

Propellant Tanks and Zero Boil-Off

Passive Thermal Protection

Future Orbital Transfer Vehicle Technology, Vol II; Davis, Boeing for NASA, 1982

Transporting Hydrogen to the Moon or Mars and Storing It There, Rapp JPL (retired) 2016

Issues of Long-Term Cryogenic Propellant Storage in Microgravity; Muritov et al, NJ Institute of Technology, 2012

Zero Boil-Off

Long Term Zero Boil-Off Liquid Hydrogen Storage Tanks; Baik, Florida Solar Energy Center, 2005

Zero Boil Off Methods for Large-Scale Liquid Hydrogen Tanks Using Integrated Refrigeration and Storage; Notardonato et al, NASA Kennedy RC, 2014

Cryogenic Orbital Testbed (CRYOTE) Ground Test Article Final Report; Jounson et al NASA Glenn RC, 2015

Innovative Stirling-Cycle Cryocooler For Long Term In Space Storage of Cryogenic Liquid Propellants; SBIR Contract Page

Cryogenic Fluid Management Technology Development for Nuclear Thermal Propulsion; Taylor et al, NASA Marshall SFC, 2015

Nuclear Cryogenic Propulsion Stage

The Nuclear Cryogenic Propulsion Stage; Houts et al NASA Marshall SFC, 2014 (Conference Paper) and (Presentation)

Nuclear Cryogenic Propulsion Stage Conceptual Design and Mission Analysis; Kos et al NASA Marshall SFC, 2014 (Conference Paper) and (Presentation Slides)

Nuclear Cryogenic Propulsion Stage Affordable Development Strategy; Doughty et al, NASA Marshall SFC, 2014 (Conference Paper) and (Presentation Slides)

Launch Vehicles

Ares V Launch Vehicle (Early NCPS and LEU NTP Launch Vehicle)

Ares V Wikipedia Page

Ares V Fact Sheet, NASA 2006

Review of U.S. Human Spaceflight Plans Committee Final Report, 2009

Ares V: Application to Solar System Scientific Exploration; Jet Propulsion Laboratory, 2008

Space Launch System (Current NASA Super-Heavy Lift Vehicle)

NASA Space Launch System Main Page

NASA SLS Overview Page

NASA’s Mars Design Reference Mission 5.0 and Associated Considerations

Human Exploration of Mars Design Reference Architecture 5.0; Drake et al NASA Johnson SC, 2010

Nuclear Thermal Rocket/Vehicle Characteristics and Sensitivity Trades for NASA’s Design Reference Architecture (DRA) 5.0 Study; Borowski et al NASA Glenn RC, 2009

Nuclear Thermal Propulsion Truss, Analysis and Optimization; Scharber et al NASA Marshall SFC, 2016 (Conference Paper) (Presentation Slides)

Blog Updates

I hope to have these blog posts released in a more timely manner. Unfortunately, these posts often have me searching for weeks for obscure information that is difficult to find even when paper titles and authors are known, and this last year has been more… fulsome with events in my personal life, let’s say. Hopefully, the greatest challenges are now behind me, and I hope to be able to post more frequently.

Unfortunately, with the difficulty in putting out just the blog (and associated pages), the YouTube channel is now on indefinite hold. There are draft scripts for many different videos, which will likely be edited into pages for the site in the coming weeks and months, but I can’t reasonably see myself being able to edit those scripts, record them, and do the video editing, much less the animations required for the scripts, at any point in the near future.

On the bright side, as some of you may have seen, the Facebook group has hit over 100 members! Feel free to come join the conversation if you’re on FB! (At some point I may branch out onto other platforms as well, but for now it’s difficult enough just keeping up with the blog and FB groups!)

Categories
Development and Testing Low Enriched Uranium Nuclear Thermal Systems

LEU NTP Part Two: CERMET Fuel – NASA’s Path to Nuclear Thermal Propulsion

Hello, and welcome back to Beyond NERVA, for our second installment of our blog series on NASA’s new nuclear thermal propulsion (NTP) system.

In the last post, we looked briefly at nuclear thermal rockets (NTRs) in general, and NERVA’s XE-Prime engine, the only time a flight configuration NTR has ever been tested in the US. We also looked at the implications for modern manufacturing and methods that would be used in any new NTR, since we are hardly going to be falling back on 60’s era technology for things like turbopumps and cryogenic storage of fuels. Finally, we looked briefly at a new material for the fuel elements, a composite of ceramic fissile fuel and metal matrix called CERMET.

This post is a deep dive into CERMET itself, including its’ design and manufacture, a little bit of its history during the Rover program, its’ rebirth in the 1990s, the test stands currently used for non-nuclear testing and some current ideas to continue to improve its’ capabilities. This is going to be more of a materials and fuel elements deep dive post, the next post will look at the engines themselves, the hot-fire test options and plans will be covered in the following one, and our last post in the series will look at other low-enriched uranium designs that don’t use CERMET fuels, but instead use carbides.

Fuel elements are where the fission itself occurs, and as such tend to be perhaps the most important part of any nuclear reactor. In the case of nuclear thermal propulsion systems (NTR, NTP to NASA), these come in three broad categories: graphite composite ((GC) such as in NERVA, which we looked at in the last post), CERMET, and carbides (something we’ll look at down the road in this series). Each have their advantages and disadvantages, but all have the same goal: to heat the propellant gas passing through the reactor as much as possible, in order to produce the maximum thrust and efficiency that the engine can provide.

Fuel Element Temperature Map, Borowski
Graph of operating temperature vs. lifetime of various NTR fuel element material options, image courtesy NASA

CERMET is a higher-temperature option than the GC elements used during the majority of Rover (although CERMET FEs were tested as part of Rover), and allows for much more control in fabrication thanks to the unique structure of the material itself. In fact, it’s able to provide the possibility of using low enriched uranium for NTR propulsion, which makes it incredibly attractive to NASA.

CERMET composites are used in many different areas of manufacturing and industry, for tooling, bearings, and other materials where hardness, heat resistance, and thermal conductivity are all needed, and the combinations used vary wildly. Different CERMET combinations have different properties, and as such are an incredibly flexible material choice.

Even in the broader nuclear field, there are other CERMET fuel elements being developed, to make more accident-tolerant fuels for terrestrial reactors. These are obviously very different in design (U3O8-Al CERMET fuels are one of the IAEA’s accident tolerant fuels of interest, and are also outside the scope of this blog post), but keep in mind that every time you hear about CERMET nuclear fuel, it’s not necessarily flying humans to Mars, it may be coming soon to a nuclear power plant near you!

However, the focus of Beyond NERVA is space, so let’s turn back to the skies. How is it that CERMET will make NASA’s new nuclear thermal rocket work? To understand that, we first need to understand what CERMET is, and why NASA decided to pick it as a fuel type of interest 20 years ago.

CERMET Fuel Elements

CERMET micrograph, NASA
W-UO2 CERMET micrograph, image courtesy NASA

CERMET is an acronym for CERamic METal composite, and was one of the first fuel forms tested as part of Project Rover, primarily by Idaho National Laboratory (INL) and General Electric, in the 1960s, and were picked up again in the 1990s as an alternative to carbides for advanced nuclear thermal fuel elements. This fuel form offers increased temperature resistance, better thermal conductivity, and greater strength compared to the graphite fuel elements that ended up being selected for NERVA, but unfortunately they also required much more development. Other options for fuel elements included advanced graphite composite and carbide fuel elements of various types, which are introduced in the NTR-S page and will be examined in their own posts.

CERMET fuel elements are a way to gain the thermal resistance and chemical advantages of oxide fuels and the thermal conduction properties of metal fuels in a single fuel form. In order to have both, uranium oxide (UO2) fuel pellets measured in millimeters or micrometers are suspended in a metal matrix, usually tungsten. To protect the oxide from any potential chemical change, these microparticles of UO2 are usually coated before the fuel element itself is made. Then the metal matrix is made, usually using a hot isostatic press (HIP), where the powdered material is placed in a mold, then pressed and cooked, although other techniques are possible as well.

There is another characteristic that makes CERMET fuel attractive in the west: it offers the possibility to use low-enriched uranium instead of highly-enriched uranium by carefully selecting the metals that the matrix is made out of to maximize the amount of moderation available from the fuel elements themselves. Low enriched uranium (LEU) offers one major advantage: a lowering of the security burden required to handle nuclear material needed to test reactor components. The vast majority of NTR systems that have been proposed over the years have been fueled with highly enriched uranium (HEU), which is over 95% 235U. This isn’t quite to bomb-grade 235U, but it’s close, and relatively easy to complete the final few steps of isotopic enrichment needed to be able to construct a weapon. (There are many other safeguards in place that make the loss of HEU unlikely, not the least of which is that the reactor won’t even be on the planet anymore, but nuclear non-proliferation is a serious concern that must be addressed in depth – just not here! For a good, in-depth look into non-proliferation I recommend (among many others), the Nuclear Diner blog, most especially the posts on the Iran nuclear treaty, from a technical-policy point of view.) Due to this increased cost (security, permitting, site re-licensing, etc.), the vast majority of institutions are unable to assist NASA and the DOE with their testing of NTR components. This is a problem, because much of the experimental engineering testing work is often done by Master’s and Doctoral students working on their dissertations. Without access to the materials used in construction, this isn’t an option, leaving the testing to NASA and DOE personnel (who are far more expensive and busy), and slowing up the whole development process. By using LEU, these institutions (that are mostly already certified to work with LEU, and many even have research reactors) are able to more fully participate in the development of the next generation of NTRs.

Often, the assumption is that HEU is superior to LEU, because the majority of LEU is fertile, not fissile: it can absorb a neutron (becoming 239U), then go through two beta decays (239Np, 239Pu), and then become fissile plutonium 239, and then can undergo fission. Why not bring along only the stuff that can split already? Breeding is a far messier process in real life than on paper, after all, and the neutronic environment is far more predictable with (mostly) only one isotope of uranium present. However, breeding occurs in all fuel elements, to the point that by the time fuel is removed from a reactor in the current fleet, the majority of the energy isn’t coming from fissioning 235U, but 239Pu. The amount of breeding that occurs is called the breeding ratio, a ratio of 1:1 means that exactly as much fissile material is being produced as is being burned. Generally speaking, this ratio is higher than 1, in order to account for the buildup of fission byproducts (or poisons) produced over the course of the fuel element’s life. The breeding ratio for this type of reactor is likely not much above 1 (most aren’t, unless it’s meant to either fuel other plants or to produce weapons, neither of which is a goal with a rocket engine); one nuclear engineer of my acquaintance suggested a back-of-envelope guess of about 1.01 for the breeding ratio, but this will largely depend on the details of the fuel element that is finally selected, the reactor core geometry, and the amount of propellant being used (among other factors). With this being the case, assuming careful management of the reactor’s neutron budget (how many neutrons are bouncing off/being absorbed/causing fission/being generated, compared to what’s needed to ensure stable operation), the majority of the “useless” 238U can in fact be burned. A paper by Vishal Patel et al (sorry about the paywall, I try and avoid them but they’re very common in nuclear engineering) suggests that the overall system could actually mass less for the same power output, which would mean that it would be better from an engineering perspective to use LEU rather than HEU. These results were for one particular reactor geometry, but the PI did mention in private correspondence that this isn’t necessarily a difficult thing to achieve, as long as the designers don’t remain tied to one particular fuel element geometry, and so could apply to many different reactor architectures.

CERMET Composition and Manufacture

CERMET fuels have many different components to them, and as such many different physical and chemical properties that have to be accounted for. However, the primary concern from a materials point of view tends to be the thermal limitations of the materials used in the FE.

CERMET Material Melting and Vaporization Points, Stewart 2015
Image from “A Historical Review of CERMET Fuel Development and the Engine Performance Implications,” Stewart, 2015

As with any composite material, there are quite a few steps to making CERMET fuels. This will be a shallow but reasonably thorough look at the manufacturing challenges on each step of the way.

In order to construct a CERMET fuel element, first the fissile fuel granules need to be made. This is not too different from the process used to make terrestrial fuel elements, which are uranium oxide (UO2) based, the main difference is the size of the resulting fuel: instead of having fuel in a pellet the size of the last joint of your finger, it’s a roughly spherical granule ~100 um in diameter.

Angular UO2 Microparticles
Angular UO2 microparticles, image courtesy NASA

There are relatively few suppliers for this form of UO2, and the most common one (BWXT) does not offer it at the price that NASA can work with. Y12 has plenty available in the right size, but they’re angular and irregular in shape; this is a problem because the release of neutrons and fission products is difficult enough to calculate when the beads are spherical, due to their distribution in the overall matrix, if they aren’t spherical enough that will affect the direction and spectrum of the resulting neutron flux, and therefore the behavior of the reactor as a whole. NASA, fortunately, has the capability to spherize these too-angular granules, though (due to their experience and equipment for plasma spray coatings in the Plasma Spheroidization System in the Thermal Spray Laboratory), and both Oak Ridge NL and the Center for Space Nuclear Research are working on gellation processes that allow for these small particles to become spherical.

ZrO2 MSFC
W-ZrO2 CVD Coated Particles, image courtesy NASA

After the sphere is made, it (usually) has to be coated with a cladding material for three reasons: first, the hot hydrogen propellant will attack the oxide very aggressively; second, the metal matrix surrounding the fissile fuel is unable to completely trap the fission products in the fuel element, leading to irradiated exhaust; and finally the UO2 in the fuel particles tends to break down, so the clad keeps the now-crystallized U in basically the same place as it was before FE thermal damage. The first coatings experimented with were pyrolitic graphite, the same as is used in TRISO fuel. However, this still has a reasonably low melting temperature (for something in an NTR), so tungsten was experimented with next. Attempts to solidify W powder around the UO2 particles led to inconsistent or relatively poor quality results, and so other options have been explored. These include chemical vapor deposition (CVD, for a long time the preferred method), plasma deposition, and other options. In the last couple years, a new technique has been shown to offer better results, which uses fine grains of tungsten rather than the CVD spray. While not as consistent in its coating, it offers advantages to fission fragment capture and overall coating consistency that make it superior to the CVD coatings.

HIP process
Image courtesy NASA

 

After the fuel particles themselves are manufactured, it’s time to make the fuel element itself. This is done by pouring (at carefully selected ratios, and in this case in particular locations) the powdered tungsten and fuel particles into a mold (usually niobium), placed on a vibrating table to settle the particles, then compressed at high temperatures for extended periods of time. This process is known as Hot Isostatic Press (HIP) sintering, and continues to be used in many fuel element designs. However, the size of the granules, the amount of pressure and temperature applied for how long, and many other factors play into HIP sintering, and especially in a field where crystalline phase can be a major determining factor of if your reactor will work or not (in fuel, moderator, and even some structural components), having a consistent and high-quality matrix around the fuel particles is essential. Again, there are processes that have been proposed in recent years that offer benefits such as lower temperature and shorter time, but we’ll go into those below.

61 channel near-full size HIP can sealed
Modern HIP can, NASA

Initially, the result of these processes was a squat cylinder with coolant channels, which would then be milled and assembled into a fuel element. As time went on, and both techniques and materials understanding improved, the fuel elements began to be cast in longer and longer single units.

Finally, the external clad is applied to the fuel elements. Both chemical vapor deposition and milled inserts have been used over the years for the propellant channel clad, with bubbling in the early tests and differences with the thermal expansion coefficient of the different materials (the clad and the fuel element it’s bonded to would swell at different rates, leading to a number of materials problems) led to the use of milled inserts being used from an early stage. These inserts (usually tungsten or niobium) are then welded to end plates and external clad sheets, also usually niobium.

The Beginnings of CERMET Fuels

Originally developed by Argonne National Labs (ANL) and General Electric(GE) in the 1960s, what were then called composite fuel elements (CFEs) are a type of fuel that gained attention for NTRs in the early to mid 1990s due to the increased thermal conductivity that the metal matrix offers to the FE as a whole. GE developed what would ultimately become the GE710 fuel element from 1962 to 1968, using HEU. After over 300,000 hours of in-environment testing, this program collected a significant amount of data.

ANL 200 MW Reactor
Image courtesy DOE

According to Gordon Kruger (of General Electric at the time of his presentation to the joint NASA/DOD/DOE Nuclear Thermal Propulsion workshop in 1990, the “seed” source as it were for this section), there were two different ANL designs: one was a 100 klbf, 2,000 MWt NTR, with a thrust-to-weight ratio of 5:1 and offering 850 s of specific impulse, the second was a smaller, 200 MWt design. This was (as with most CERMET designs) a tungsten-uranium oxide (W-UO2) fuel element. The fuel particles themselves were chemically stabilized by doping them with gadolinium, and the clad for the fuel particles was W doped with Rhenium. The fuel element developed in this process is now called the ANL-2000 CERMET FE, and remains a popular one for NTR designers. It has a very high number of propellant channels (331 per FE) to allow for greater cooling capability of the fuel.

The GE design, on the other hand, was meant to be more versatile The base design was for a high temperature gas cooled reactor (HTGR), with helium as a working fluid, designed for a 10,000 hour life. Those same fuel elements, in a different core geometry, could instead burn much faster, and much hotter, for use as an NTR (with cryogenic H2 propellant), but the harder use (and harsher chemical environment) correspondingly shortened the life of the fuel elements. This is the GE 710 fuel element, which in a slightly modified form – known as the GE 711 – is still a strong contender for NTR designs, and was the front-runner for the LEU NTP that NASA is working on. With 64 propellant channels of larger diameter, this FE offers a trade-off of easier manufacture (due to the larger, less numerous boreholes) with the potential for greater thermal differences in the FE due to the greater distance between the channels.

Both these designs have many things in common, such as the hexagonal prism shape, and information sharing between the groups was a regular thing. As such, techniques used for the different stages of manufacture was common as well.

Non-Spherical Microparticles
UO2 particles

Both designs used spheres of UO2. These can still be manufactured by two places in the US (Oak Ridge National Labs and BWXT), but there are challenges to getting the pieces to be spherical when they’re that small, so the price is correspondingly high. This indicates at least something of a learning curve when it comes to this stage of manufacture, both for ensuring homogeneity of fissile fuel load (if it’s poorly mixed, hot spots and dead zones can form, leading to very bad things – or nothing at all), and for size and shape consistency. Because of the extreme temperatures, both during manufacture and operation, the gadolinium (Ga) doping experimented with at ANL became essential to stabilize the UO2, and to prevent the dissociation of the oxygen and uranium. Nursing the dissociation temperature up was a consistent effort throughout this process.

ZrO2 MSFCThe clad on the fuel pellets is a challenge another way, as well: applying an even coat of tungsten across the tiny spherical oxide pellets is a major technical challenge, and one that was addressed at the time with chemical vapor deposition (CVD), where the tungsten is liquefied and then sprayed (under a certain set of conditions) over the oxide spheres. Because the droplets are small, they have a high relative surface area, so they are able to coat a material that wouldn’t normally be able to resist the temperature of the molten substance (in this case tungsten, doped with rhenium to lower the melting point). This can lead to a very even coating, if the two substances are chemically compatible, and if the conditions are just right enough for the droplets to be able to spread out enough, and spread evenly enough across the surface. This is a very large challenge, and one that took a lot of time and energy from the teams designing the fuel elements. A competing process, pressure bonded cladding, was also examined for both the fuel particles and the clad for the fuel element itself.

Can component fit check pic

Once the fuel particles were fabricated, the metal matrix of the fuel element could then be fabricated. Hot isostatic press sintering (HIP) was the preferred method of manufacture for the fuel elements. This led to complications stabilizing the UO2 in the fuel (which isn’t able to stand the temperatures of molten tungsten, hence the sintering) used by both groups, hence the Gadolinium doping of the fuel pellets. The trade-off was always how to increase the density of the tungsten (and therefore the energy density and strength of the FE as a whole) while decreasing the amount of decay in the UO2, either by lessening the temperature or the time that the material is cooked, or by chemically stabilizing the oxide itself. Once sintering was complete, the mold is set aside to cool, then the CERMET plug is removed.

SPS SampleThe result of this exercise was known as a compact. This was then machined to drill propellant holes and do final shaping, and its fissile fuel load was assessed. It was labeled, and set aside until a sufficient collection of machined compacts had been completed. These were then stacked according to fissile fuel load, and then the tungsten fuel element end plates, external clad and propellant clad tubes were welded into place to form the overall hexagonal prism shape. These are then assembled in a number of different ways for either an HTGR or an NTR.

The most mature designs to come out of this development series was the GE 710 fuel element, with 19 working fluid channels, and the ANL 2000 designs with 312 coolant channels. In many ways, these form a baseline for CERMET fuels as the NERVA XE-Prime serves as a baseline for NTRs as a whole. Many CERMET NTR designs use this as their baseline fuel form, and for good reason. This fuel element was tested for HTGC reactor use in the 1970s, and showed promising results. However, gas cooled reactors were never popular in the US, and production ended.

The Rebirth of the Idea, and the Building of Test Stands

After the cancellation of the GE710 project, CERMET FE design went quiet for a number of decades, until the 1990s, when the idea was revived again after Project Timberwind (and the rest of the Strategic Defense Initiative) got shot down during defense cuts under President H.W. Bush.

In the early 1990s, focus shifted back from the pebblebed and toward other options. While it was acknowledged that graphite composite was better developed, and carbides offered higher-temperature operation, CERMET fuels were seen as a good compromise. At some point after the 1991 Nuclear Thermal Propulsion conference, focus shifted to CERMET fuels as being compatible enough with the legacy NERVA systems and data collected, while also being easier to work with than carbide fuels. A good overview of the decision process to proceed with CERMET fuels can be seen in Mark Stewart’s presentation for NETS 2015, “A Historical Review of CERMET Fuel Development and Engine Performance Implications” (paper and slides).

Many of the best-known designs for NTRs in the last 25-30 years have been the work of either Michael Houts at NASA’s Marshall Spaceflight Center or Stan Borowski, of NASA’s Glenn Research Center. Looking at the systemic implications of not only the rocket engineering side of things, but the mission analysis, development cost, and testing options available to develop NTRs, they firmly established a new baseline nuclear rocket, seen in popular artwork for over 30 years. Many of these designs were based around a smaller Rover-legacy advanced graphite composite fueled reactor known as the Small Nuclear Rocket Engine. Ths idea was to design an engine just big enough to be useful, and if it wasn’t powerful enough, just add another engine! We’ll look at this design more in depth at a later point, but it is important in that it was a mid-1990’s design that could use CERMET fuel, possibly the first modern one, and is in many ways the baseline for what a modern NTR can do.

In order to gather the information needed to develop the nuclear fuel elements, a number of test stands have been built by NASA in recent years to thermally and environmentally test experimental fuel elements, using depleted uranium (DU) and induction heating. The two most commonly used are the Nuclear Thermal Reactor Element Environmental Simulator (NTREES) and the CERMET Fuel Element Environment (CFEET) test stand. Since hot-fire tests were not an option anymore, and the experimental fuel elements still needed to be exposed to the thermal and environmental conditions of an operating NTR, these were seen as the best way to spend what little money had been allocated to nuclear spaceflight over a number of years.

NTREES

The Nuclear Thermal Reactor Element Environmental Simulator was first proposed by William Emrich of NASA’s Marshall Spaceflight Center in 2008, and was designed to simulate everything but the radiation environment that an NTR fuel element would experience. This was the next best thing possible, short of starting nuclear hot-fire tests again (which neither the regulations nor the budget would allow): many of the other questions that needed to be answered in order to build a new NTR was being addressed in other programs; for example, cryogenic hydrogen was a major challenge in Rover, but research had continued through chemical propulsion systems. The questions that remained mostly had to do with either core geometry or the fuel element itself, and most of those questions were chemical. By substituting other materials (such as ZrO2) with similar properties (thermal behavior, etc) to UO2 in initial tests, and then move on to the more difficult to use depleted uranium (DU) for more promising test runs (as we saw in the KRUSTY post, DU carries a far stricter burden as far as safety procedures and regulation), testing could continue- and be more focused on the last details that needed to be worked out chemically and thermally.

Houts NTREES Facility 2013

When the test stand was being designed, flexibility was one of the main foci of the design decisions that were made; after all, new equipment for nuclear thermal testing is incredibly rare, and funding for it is virtually impossible to come by, so one piece of test equipment can’t be specialized to just one design, to sit collecting dust on the shelf after that project is canceled and a new one comes along with requirements that make the old equipment obsolete.

NTREES consists of a pressure vessel, an induction heating arrangement for the test article, a data acquisition unit, and an exhaust treatment system. Hydrogen is introduced at the needed pressure and rate into the pressure vessel, where it encounters the test article. Measurements are taken through view ports in the side of the pressure vessel, and then the hot hydrogen is cooled by adding a large amount of nitrogen. This gas mixture is then passed through a mass spectrometer, and then further cooled and collected. The mass spectrometer is designed to be able to detect a wide range of atomic masses, so that uranium-bearing compounds can be detected to measure fissile fuel erosion; with pressure, temperature, and flow sensors they make up the inputs for the data acquisition system.

Chamber installation
Pressure Chamber during upgrade, image courtesy NASA MSFC

The bulk of the test stand is the pressure vessel, which is water cooled, ASME code stamped, and has a maximum operating pressure of 6.9 megapascals (MPa).  Because of the need for flexibility, NTREES can handle test articles up to 2.5 meters long, and 0.3 m in diameter. A number of sapphire view ports along each side of the pressure vessel are used for instrumentation and observation. Along the bottom are ports for the induction heater used to bring the test article up to temperature (one of these can also be modified for vacuum system use). The induction heater is a 1.2 MW unit, upgraded in 2014, although the upgrade wasn’t immediately able to be fully implemented until later due to having to wait for funding to upgrade the N2 cooling system to handle the power increase.

After the now-hot H2 leaves the test article, it enters a gas mixer, which adds cold nitrogen to cool the H2 rapidly, and to dilute it with a more inert gas to reduce explosive hazards. This sleeve is also water-cooled, which draws out even more heat from the gas. The lessons learned about handling gaseous and liquid hydrogen were well-learned, and multiple safety systems and design choices have gone into handling this potentially dangerous and reactive gas safely. Another example of this is at the hot end interface with the test article: there is more pressure on the nitrogen outside the H2 feed, so that N2 inbleeding prevents any H2 leakage at a seal which would be very prone to failure due to the high temperatures involved.

The mixer is also the first stage of the effluent cleanup system, designed to ensure that no potentially harmful chemical releases occur when the exhaust is released into the atmosphere. The second stage of the cleanup system is a water cooled sleeve that further chills the gas mixture (this system was upgraded in 2014 as well, to allow the system to carry away all the heat generated – and therefore be able to run longer-duration tests at higher temperatures). Finally, a filter and back-pressure system is used to clean the now-cool gas before it is exhausted through a smokestack on the outside of the facility.

After dilution, the gas stream passes in front of a far more flexible spectrometer than usual. Most spectrometers only examine a relatively small band of the periodic table, because they’re only needing to measure particular elements. In this case, the elements that could be in the exhaust stream are spread fairly well across the periodic table, and as such a more versatile spectrometer was needed to be able to accurately assess the effluent stream.

The data acquisition system consists of the mass spectrometer, pressure sensors, gas temperature sensors, flow sensors, thermocouples for general temperature measurements, H2 detectors in the chamber and the room, and pyrometers to measure the temperature of the test article itself, and the associated electronics to collect the information from these sensors.

The design of the facility was safety-oriented from the beginning, with every precaution being taken to handle the GH2 safely. If you’re interested, the systems are looked at more on the NTREES page.

When put together, this facility allows for chemical and thermal testing of NTR fuel elements for extended periods of time in an environment that is missing only one component to mimic the environment of an NTR core: radiation. This means that fuel elements can be easily tested for manufacturing technique verification, clad material choice, erosion rates of fuel element materials, and other questions that are primarily chemical or mechanical rather than nuclear in origin.

 

There is one other difference between this test stand and the environment that a nuclear fuel element will, and that’s the source and distribution of the heat. In NTREES, the induction heating coil is the source of the heat. Power distribution starts on the outside of the fuel element, and  While the coil can be customized to a certain extent to manage the thermal load for different test articles, the spiral pattern will still be there, and the heat will be generated in the fuel element following the rules of inductive heating, not nuclear heating.

 

In a nuclear fuel element, considerable effort is taken to ensure that there is an even distribution of heat across the fuel element (taking into account all factors), because having a “hot spot” in your fuel element (higher-than desired density of fissile material) can do bad things to your reactor. Because of this, the power density is carefully assessed during manufacture and assembly. In the fuel element, temperature tends to peak around the edge of the fuel element, but otherwise be consistently distributed throughout. This difference can be significant, especially for clad/matrix interfaces where local hot spots can exacerbate thermal expansion differences and clad failure.

The radiation environment in a nuclear reactor will cause additional swelling, and neutron damage, fission product buildup, and other effects will need to be accounted for as well. This difference is something that can be modeled, either through extrapolation from old data sets or from materials analysis in various radiation environments and beamlines in facilities around the world. While verification and validation tests in a reactor environment similar to an NTR core will be needed for whatever fuel elements are selected, this testing allows many of the hurdles to be addressed before this very expensive step is taken.

CFEET

Front photo with lables, Bradley
CFEET front view, NASA MSFC

The CERMET Fuel Element Environmental Test (CFEET) stand was originally proposed in 2012 by David Bradley at NASA’s Marshall Spaceflight Center as a lower-cost alternative to NTREES. One of the consistent problems in engineering is that to make something more flexible the complexity must increase. This increases the cost to both build and maintain the test stand, which results in a higher cost per test. Also, the larger the volume the test stand uses, the more supplies are needed (in the case of NTREES, GH2 and GN2, plus water for the cooling system), which also increases cost.

CFEET is a low-cost, small scale test stand for NTR fuel elements. It also exposes a test article to temperatures and hydrogen environment that they would experience in the core of an NTR, but again the radiation effects aren’t accounted for since this is purely an inductively heated test stand. Rather than have the extensive piping, effluent cleanup, and exhaust systems that NTREES uses, CFEET uses a simple vacuum chamber with a single RF coil for induction heating to test thermal properties and general reactions with the hydrogen (The hydrogen is pumped through the FE during testing, but I can’t find any information about flow rate of the gas).

CFEET Dimensions, BradleyThis means that the majority of CFEET fits on a (large) desktop. The vacuum chamber is only 16.9” tall and 10” in diameter, and it’s the largest component of the system. Rated to 10^-6 Torr, the chamber has a vacuum-rated RF feed-through port one one side, and opposite that port another, sapphire one for pyrometer readings. Additional ports connect the turbopumps and other equipment to the chamber.

The induction heating equipment is rated to 15 kW, with an output frequency of 20-60 kHz. While significantly lower output than NTREES, CFEET is still able to get test articles to reach temperatures over 2400 K. An insulating sleeve (with a hole formed in it to allow pyrometer readings) of various materials is used to minimize heat loss through radiation.

While CFEET is not able to simulate gas flow, as NTREES is, it is able to assess thermal, chemical, and mechanical properties of materials at temperature and in a pure-hydrogen atmosphere. Because the system is far simpler, and takes far fewer consumables to operate, it is far cheaper to use as a test bed.

More info on CFEET is available on the CFEET page!

What Have They Taught Us?

FE Post-Test W HfN
CERMET FE post-CFEET test, image via NASA

Both NTREES and CFEET have been used to help assess various manufacturing techniques for fuel elements, and also evaluate clad materials and thermal expansion issues. NTREES is able to assess erosion rates (both in mass and in chemical composition). While these aren’t the sexy tests, they have informed decisions about clad materials, manufacturing methods, and the inherent tradeoffs in different designs without having to go through the major expense of designing, building, testing, and then hot-fire testing a nuclear reactor.

Work has continued on investigating different microstructures within the FE, using depleted UO2 (dUO2) for chemical and thermal analysis. These tests have explored many different options as far as fine structure of the fuel forms available, and continue to inform CERMET fuel element design today.

Development Challenges for LEU NTP, and a New Direction

A major change occurred in 2012, however: it was decided by the White House that highly enriched uranium (HEU) would not be used for civilian purposes in the US, in order to reduce the risk of nuclear weapons proliferation, and that low enriched uranium (LEU) would be used for all civilian purposes, including medical and industrial isotope production.. This decision has resulted in thousands, if not tens of thousands, of pages of response, from dry, indifferent technical papers to proponents and opponents of the move screaming and raging in every direction. Because of this decision, NASA’s nuclear programs were forced to look at LEU systems, not the HEU ones that they’d always used. While there are a number of ways to make an NTR out of LEU instead of HEU, the two main options are CERMET and carbide fuel elements. Because CERMET was already under development, and there were ways to use LEU in CERMET fuel, this was the path that was decided by NASA’s management. However, LEU carbide designs (most notably SULEU, the Superior Utilization of Low Enriched Uranium carbide-based NTR) are also an option, and one that offers higher temperature operation as well, but since CERMET fuels are more developed within NASA’s design paradigm they remain the primary focus of NASA’s development.

One of the greatest fears in any development program is the problems that simply can’t be assessed within the budget, the timeframe, or both, of a program. Every program has them, and many engineering fish tales have been made out of solving them. When they haven’t been solved, though, they are the things that often define a program’s schedule… and its cancellation date.

For the LEU NTP program, the main challenge is in the fuel element matrix, and the isotopic purity of the tungsten (W) needed for the metal matrix of the fuel in particular. For an HEU reactor, the isotope of tungsten was less of a concern, because there was a more flexible neutron budget for the reactor due to the higher fuel load. With LEU, the neutron budget becomes tighter, and the more management of the neutron spectrum you can do within the FE, the fewer neutrons are lost to the structural components of the reactor. Isotopic enrichment of reactor components other than fuel elements is relatively common, and so this wasn’t seen as a major challenge.

Most of the analysis up to this point on LEU NTP has focused on this line of development. Tungsten-184 has a small enough neutron capture cross section that it can reflect a neutron many times within the fuel element itself, increasing the likelihood of a capture by the higher-cross sectioned fuel nuclei. In fact, a recent paper by Vishal Patel of the Center for Space Nuclear Research in Idaho Falls, ID (who has kindly answered many questions, often sent at odd hours of the night, while I was researching this post) demonstrated some surprising characteristics that are possible with LEU CERMET fuel… including an overall reduction in system mass! This is an especially surprising result, but he actually went on Facebook to discuss the finding in the first day or two that the paper came out, and the overall conclusion was interesting:

 So the reason all this ends up working is that you are constrained by thermal design concerns (need enough surface are for heat transfer) rather than neutronic reasons (needing enough volume to go critical). This is typical for reactors of this size and above. At much lower thrusts the neutronics eventually dominates and HEU looks better but no rocket person cares for those lower levels of thrust for this type of system. The idea of this study was to show the systems are comparable, choose whichever one you want (but the obvious first thought is proliferation and economics, so choose the one that fits your constraints). 

Unfortunately, tungsten enrichment is a major challenge, and one that we aren’t going to be able to discuss in detail, because 184W is useful in another nuclear technology: explosives. This is because W is a great neutron reflector, and so is used in fission explosives to increase the number of neutrons entering the core during the initial neutron pulse from the initiation of the nuclear detonation. According to NASA, the LEU FEs, as designed, required 90% enriched 184W. It was expected that a 1 mg sample at 50% purity would be available in October of 2016, but a mix of accidents (an inadvertent chemical release is mentioned in the Mid-Year Game Changing Development Status Report for 2017) and technical challenges (which are classified) has forced this requirement into the forefront of everyone involved in the NTP program’s mind.

Alternatives exist, however. BWXT, already a major supplier of experimental fuel elements, has suggested a different core design, where graded molybdenum (Mo) and tungsten can be used instead of (90%) pure 184W. This design is one that is still very new, and because of that (and since it’s being developed by a private company and not a public institution) there’s not much information available. New contracts were signed between NASA and BWXT in 2017 to fund the development of their FE design, and hopefully as time goes on more information will become available. According to one person knowledgeable about the program, hopefully the Nuclear and Emerging Technologies for Space 2018 (to be held in Las Vegas in February) will bring more information. I have been trying to find out more information on this design, but unfortunately there’s not much out there that I can see. I also don’t have the background to determine if the manufacturing techniques described above will be compatible with this particular FE design, or the reasons why they would or wouldn’t be. Being the end of the year, it would be surprising if we heard anything before NETS this year.

Another change that has been floating around since about 2011 is a new process for manufacturing the metal matrix of the fuel element: spark plasma sintering (SPS). This seems to have been most thoroughly explored at Idaho National Laboratory and the Center for Space Nuclear Studies in Idaho Falls, ID. Instead of using HIP sintering, where heat and pressure are used to coax the temperature for a consistent metal matrix down, the individual grains are welded together using electric arcing. This allows a lower sintering temperature to be achieved, allowing for less decomposition of the UO2 in the fuel particles.

This also allows for a new type of clad to be used. Rather than the difficulties that have been experienced with the CVD clad, a binder is used to apply tungsten microparticles. This is one of the newest techniques to be explored for fuel particle coating, and in order to take advantage of it SPS has to be used, because the HIP temperatures are too high. For more info on these developments I recommend this paper by Zhong et al from INL and this presentation by Barnes.

How This Changes the Core

BWXT Core
BWXT Core, image via BWXT

Any time a fuel element is changed, either in composition or enrichment, it can lead to significant changes to the core of the reactor. The biggest change in NASA’s NTP system is that tie tubes have been eliminated from the core. As discussed in the last post, the tie tubes perform many different functions, not just structural support for the fuel elements (which suffered persistent failures due to vibrations in the core), but also provided neutron moderation and supplied power to the turbopumps as well. Because of this, there have been designs for tie tubes for LEU NTR cores, although often these are placed around the periphery of the core rather than spread throughout like was originally planned for in the NERVA core. This changes the power distribution in the core, and makes it so that some reactor geometry design changes are necessary, but those are incredibly specific to the fuel elements used, and the results of extensive modeling of neutronic behavior and reactor physics.

Because the fuel elements are able to withstand higher temperatures, the entire reactor will run at elevated temperatures compared to the XE-Prime engine. This gives an increase in specific impulse over the graphite composite core type, although how much of one will largely depend on the particulars of the fuel elements and reactor power, and therefore core geometry, of the design that is finally tested.

More to Come!

Keep checking back for our next installment, which will look at the various reactor cores and engines themselves, for both the LEU NTP system and the Nuclear Cryogenic Propulsion Stage. We’ll also look at test stands and limitations for hot-fire ground testing, and how those will influence the decisions made for the new engines. Finally we’ll wrap up at a look at the advanced carbide designs that are being looked at (although not too closely on NASA’s part… yet!)

Sources and Additional Reading

A Summary of Historical Solid Core Nuclear Thermal Propulsion Fuels, Benensky 2013

  • If you only read one reference on this list, make it this one!

CERMET Fueled Reactors, Cowan et al 1987

A CERMET Fueled Reactor for Nuclear Propulsion, Kruger 1991

Hot Hydrogen Testing of W-UO2 Dioxide CERMET Fuel Materials for NTP, Hihcman et al 2014

Affordable Development and Optimization of CERMET Fuels for NTP Ground Testing, Hickman et al 2014

Design Evolution of HIP Cans for NTP CERMET Fuel Fabrication, Mireles 2014

Spark Plasma Sintering of Fuel CERMETs for Nuclear Reactor Applications, Zhong et al 2011

Low Enriched Nuclear Thermal Propulsion Systems, Houts et al 2017

NTP CERMET Fuel Development Status, Barnes 2017

2017 Game Changing Development program Mid-year Review Slides

Channel update:

My apologies for the delay on posting, the holidays have a way of creating slowdowns in material getting written. Hopefully I will be able to post more regularly soon. Research for the next post (on NASA’s plans for hot-fire test capability at Stennis Spaceflight Center, and the limitations that may place on testing) is underway, as well as research to prepare for results to hopefully be announced at NETS 2018. Sadly, I will not be able to attend, but look forward to all the papers that will be presented on these fascinating engines. I hope to publish on the latest in these new designs shortly after the conference ends. After that, a final post in the series on carbide fuel element LEU NTRs will wrap up this blog series.

At that point, the focus will shift back to trying to get the YT channel going. I haven’t touched Blender in a while, but I don’t think that it will be difficult to do what I need to do, I just need to sit down and learn. The scripts are largely written in draft form, I just need to go back over them for a final edit, then start doing the audio. The search still goes on for video clips to use, especially for Project Rover. Any links to clips that I would be able to use would be greatly appreciated!

 

Cpoyright 2018 Beyond NERVA. Contact for reprint permission.