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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
Development and Testing Low Enriched Uranium Nuclear Thermal Systems Test Stands

NTR Hot Fire Testing 2: Modern Designs, New Plans for the LEU NTP

Hello, and welcome back to Beyond NERVA in the second part of our two-part series on ground testing NTRs. In part one, we examined the testing done at the National Defense Research Site in Nevada as part of Project Rover, and also a little bit of the zero power testing that was done at the Los Alamos Scientific Laboratory to support the construction, assembly, and zero-power reactivity characterization of these reactors. We saw that the environmental impact to the population (even those living closest to the test) rarely exceeded the equivalent dose of a full-body high contrast MRI. However, even this low amount of radioisotope release is unacceptable in today’s regulatory environment, so new avenues of testing must be explored.

NERVAEngineTest, AEC
NRX (?) Hot-fire test, image courtesy DOE

We will look at the proposals over the last 25 years for new ways of testing nuclear thermal rockets in full flow, fission-powered testing, as well as looking at cost estimates (which, as always, should be taken with a grain of salt) and the challenges associated with each concept.

Finally, we’re going to look at NASA’s current plans for test facilities, facility costs, construction schedules, and testing schedules for the LEU NTP program. This information is based on the preliminary estimates released by NASA, and as such there’s still a lot that’s up in the air about these concepts and cost estimates, but we’ll look at what’s available.

Diagram side by side with A3
Full exhaust capture at NASA’s A3 test stand, Stennis Space Center. Image courtesy NASA

Pre-Hot Fire Testing: Thermal Testing, Neutronic Analysis, and Preparation for Prototypic Fuel Testing

Alumina sleeve during test, Bradley
CFEET test, NASA MSFC

We’ve already taken a look at the test stands that are currently in use for fuel element development, CFEET and NTREES. These test stands allow for electrically heated testing in a hydrogen environment, allowing for testing of he thermal and chemical properties of NTR fuel. They also allow for things like erosion tests to be done, to ensure clad materials are able to withstand not just the thermal stresses of the test but also the erosive effects of the hot hydrogen moving through them at a high rate.

However, there are a number of other effects that the fuel elements will be exposed to during reactor operation, and the behavior of these materials in an irradiated environment is something that still needs to be characterized. Fuel element irradiation is done using existing reactors, either in a beamline for out-of-core initial testing, or using specially designed capsules to ensure the fuel elements won’t adversely affect the operation of the reactor, and to ensure the fuel element is in the proper environment for its’ operation, for in-core testing.

 

TrigaReactorCore
TRIGA reactor core, image courtesy Wikimedia

A number of reactors could be used for these tests, including TRIGA-type reactors that are common in many universities around the US. This is one of the advantages of LEU, rather than the traditional HEU: there are fewer restrictions on LEU fuels, so many of these early tests could be carried out by universities and contractors who have these types of reactors. This will be less expensive than using DOE facilities, and has the additional advantage of supporting additional research and education in the field of astronuclear engineering.

 

 

Irradiation vessel design for ATF, Thody
Design of an irradiation capsule for use with the ATF, Thody OSU 2018

The initial fuel element prototypes for in-pile testing will be unfueled versions of the fuel element, to ensure the behavior of the rest of the materials involved won’t have adverse reactions to the neutronic and radiation environment that they’ll be subjected to. This is less of a concern then it used to be, because material properties under radiation flux have been continually refined over the decades, but caution is the watchword with nuclear reactors, so this sort of test will still need to be carried out. These experiments will be finally characterized in the Safety Analysis Report and Technical Safety Review documents, a major milestone for any fuel element development program. These documents will provide the reactor operators all the necessary information for the behavior of these fuel elements in the research reactor in preparation for fueled in-pile testing. Concurrently with these plans, extensive neutronic and thermal analysis will be carried out based on any changes necessitated by the in-pile unfueled testing. Finally, a Quality Assurance Plan must be formulated, verified, and approved. Each material has different challenges to producing fuel elements of the required quality, and each facility has slightly different regulations and guidelines to meet their particular needs and research guidelines. After these studies are completed, the in-pile, unfueled fuel elements are irradiated, and then subjected to post irradiation examination, for chemical, mechanical, and radiological behavior changes. Fracture toughness, tensile strength, thermal diffusivity, and microstructure examination through both scanning electron and transmission electron microscopy are particular areas of focus at this point in the testing process.

 

One last thing to consider for in-pile testing is that the containment vessel (often called a can) that the fuel elements will be held in inside the reactor has to be characterized, especially its’ impact on the neutron flux and thermal transfer properties, before in-pile testing can be done. This is a relatively straightforward, but still complex due to the number of variables involved, process, involving making an MCNP model of the fuel element in the can at various points in each potential test reactor, in order to verify the behavior of the test article in the test reactor. This is something that can be done early in the process, but may need to be slightly modified after the refinements and experimental regime that we’ve been looking at above.

Another consideration for the can will be its’ thermal insulation properties. NTR fuel elements are run at the edge of the thermal capabilities of the materials they’re made out of, since this maximizes thermal transfer and therefore specific impulse. This also means that, for the test to be as accurate as possible, the fuel element itself must be far hotter than the surrounding reactor, generally in the ballpark of 2500 K. The ORNL Irradiation Plan suggests the use of SIGRATHERM, a soft graphite felt, for this insulating material. Graphite’s behavior is well understood in reactors (and for those in the industry, the fact that it has about 4% of the density of solid graphite makes Wigner energy release minimal).

Pre-Hot Fire Testing: In-Pile Prototypic Fuel Testing

 

406px-High_Flux_Isotope_Reactor_Vertical_Cross_Section
High Flux Isotope Reactor (HFIR), Oak Ridge National Lab, image courtesy Wikimedia

Once this extensive testing regime for fuel elements has been completed, a fueled set of fuel elements would be manufactured and transported to the appropriate test reactor. Not only are TRIGA-type reactors common to many universities an option, but three research reactors are also available with unique capabilities. The first is the High Flux Isotope Reactor at Oak Ridge, which is one of the longest-operating research reactors with quite a few ports for irradiation studies at different neutron flux densities. As an incredibly well-characterized reactor, there are many advantages to using this well-understood system, especially for analysis at different levels of fuel burnup and radiation flux.

 

 

 

 

 

TREAT INL
Transient Reactor Test (TREAT) at Idaho NL. Image courtesy DOE

The second is a newly-reactivated reactor at Idaho National Laboratory, the Transient Reactor Test (TREAT). An air cooled, graphite moderated thermal reactor, the most immediately useful instrument for this sort of experiment is the hodoscope. This device uses fast neutrons to detect fission activity in the prototypic fuel element in real time, allowing unique analysis of fuel element behavior, burnup behavior, and other characteristics that can only be estimated after in-pile testing in other reactors.

 

800px-Advanced_Test_Reactor_001
Advanced Test Reactor, Idaho NL. Image courtesy DOE

The third is also at Idaho National Lab, this is the Advanced Test Reactor. A pressurized light water reactor, the core of this reactor has four lobes, and almost looks like a clover from above. This allows for very fine control of the neutron flux the fuel elements would experience. In addition, six of the locations in the core allow independent cooling systems that are separated from the primary cooling system. This would allow (with modification, and possible site permission requirements due to the explosive nature of H2) the use of hydrogen coolant to examine the chemical and thermal transfer behaviors of the NTR fuel element while undergoing fission.

Each of these reactors uses a slightly different form of canister to contain the test article. This is required to prevent any damage to the fuel element contaminating the rest of the reactor core, an incredibly expensive, difficult, and lengthy process that can be avoided by isolated the fuel elements from their surrounding environment chemically. Most often, these cans are made out of aluminum-6061, 300 series stainless steel, or grade 5 titanium (links in the reference section). According to a recent Oak Ridge document (linked in references), the most preferred material would be the titanium, with the stainless being the least attractive due to 59Fe and 60Co activation leading to the can to become highly gamma-active. This makes the transportation and disposal of the cans post-irradiation much more costly.

Here’s an example of the properties that would be tested by the time that the tests we’ve looked at so far have been completed:

Fuel Properties and Parameters to Test
Image courtesy Oak Ridge NL

NTR Hot Fire Testing For Today’s Regulatory Environment

It goes without saying that with the current regulatory strictures placed on nuclear testing, the same type of testing as done during Rover will not be able to be done today. Radioisotope release into the environment is something that is incredibly stringently regulated, so the open-air testing as was conducted at Jackass Flats would not be possible. However, there are multiple options that have been proposed for testing of an NTR in the ensuing years within the more rigorous regulatory regime, as well as cost estimates (some more reliable than others) and characterization of the challenges that need to be overcome in order to ensure that the necessary environmental regulations are met.

The options for current hot-fire testing of an NTR are: the use of upgraded versions of the effluent scrubbers used in the Nuclear Furnace test reactor; the use of boreholes as effluent capture and scrubbing systems (either already-existing boreholes drilled for nuclear weapons tests that have not been used for that purpose at Frenchman’s Flat, or new boreholes at the Idaho National Laboratory); the use of a horizontal, hydrogen-cooled scrubbing system (either using existing U-la or P-tunnel facilities modified for the purpose, or constructing a new facility at the National Nuclear Security Site); and the use of a new, full-exhaust-capture system at NASA’s current rocket test facilities at the John C. Stennis Space Center in Mississippi.

The Way We Did It Before: Nuclear Furnace Exhaust Scrubbers

Transverse view, Finseth
NF1 configuration, image from Finseth, 1991 courtesy NASA

The NF-1 test, the last test of Project Rover, actually included an exhaust scrubber to minimize the amount of effluent released in the test. Because this test was looking at different types of fuel elements than had been looked at in most previous tests, there was some concern that erosion would be an issue with these fuel elements more than others.

Effluent Cleanup System Flow Chart
Image from Nuclear Furnace 1 test report, Kirk, courtesy DOE

Axial view, FinsethThe hydrogen exhaust, after passing the instrumentation that would provide similar data to the Elephant Gun used in earlier tests, would be cooled with a spray of water, which then flashed to steam. This water was initially used to moderate the reactor itself, and then part of it was siphoned off into a wastewater holding tank while the rest was used for this exhaust cooling injection system. After this, the steam/H2 mixture had a temperature of about 1100 R.

After leaving the water injector system, the coolant went through radial outflow filter that was about 3 ft long, containing two wire mesh screens, the first with 0.078 inch square openings, the second one with 0.095 inch square openings.

Once it had passed through the screens, a steam generator was used to further cool the effluent, and to pull some of the H2O out of the exhaust stream. Once past this steam generator, the first separator drew the now-condensed water out of the effluent stream. Part of the radioactive component of the exhaust is at this point dissolved in the water. The water was drawn off to maintain an appropriate liquid level, and was moved into the wastewater disposal tank for filtering. A further round of exhaust cooling followed, using a water heat exchanger to cool the remaining effluent enough to condense out the rest of the water. The water used in this heat exchanger would be used by the steam generator that was used earlier in the effluent stream as its’ cool water intake, and would be discharged into the wastewater holding tank, but would not come in direct contact with the effluent stream. Once past the heat exchanger, the now much cooler H2/H2O mixture would go through a second separator identical in design to the first. At this point, most of the radioactive contaminant that could be dissolved in water had been, and the discharge from this unit was at this point pretty much completely dry.

A counterflow, U-tube type heat exchanger was then used to cool the effluent even more, and then a third separator – identical to the first two – was used to capture any last amounts of water still present in the effluent stream. During normal operation, though, basically no water would collect in this separator. The gas would then be passed through a silica gel sorption bed to further dry it. A back flow of gaseous nitrogen would be used to dry this bed for reuse. The gas, at this point completely dried, was then passed through another heat exchanger almost identical to the one that preceded the silica gel bed.

Charcoal Trap System
From NFI test report, Kirk, via DOE

After passing through a throttle valve (used to maintain back-pressure in the reactor), the gas was then passed through an activated charcoal filter trap, 60 inches long and 60 inches in diameter, to capture the rest of the radioactive effluent left in the hydrogen stream after being mixed with LH2 to further cool the gas to 250-350 R. Finally, the now-cleaned H2 is burned to prevent a buildup of H2 gas in the area- a major explosion hazard. This filter system was constantly adjusted after each power test, because pressure problems kept on cropping up for a number of reasons, from too much resistance to thermal disequilibrium.

So how well did this system do at scrubbing the effluent? Two of the biggest concerns were the capture of radiokrypton and radioxenon, both mildly radioactive noble gasses. The activated charcoal bed was primarily tasked with scrubbing these gasses out of the exhaust stream. Since xenon is far more easily captured than krypton in activated charcoal, the focus was on ensuring the krypton would be scrubbed out of the gas stream, since this meant that all the xenon would be captured as well. Because the Kr could be pushed through the charcoal bed by the flow of the H2, a number of traps were placed through the charcoal bed to measure gamma activity at various points. Furthermore, the effluent was sampled before being flared off, to get a final measurement of how much krypton was released by the trap itself.

Looking at the sampling of the exhaust plume, as well as the ground test stations, the highest dose rat was 1 mCi/hr, far lower than the other NTR tests. Radioisotope concentrations were also far lower than the other tests. However, some radiation was still released from the reactor, and the complications of ensuring that this doesn’t occur (effectively no release is allowed under current testing regimes) due to material, chemical, and gas-dynamic challenges makes this a very challenging, and costly, proposition to adapt to a full-flow NTR test.

Above Ground Test Option #1: Exhaust Scrubbing

The most detailed analysis of this concept was in support of the Space Nuclear Thermal Propulsion program, run by the Department of Energy – better known as Project Timber Wind. This was a far larger engine (111kN as opposed to 25 kN) engine, so the exhaust volume would be far larger. This also means that the costs associated with the program would be larger due to the higher exhaust flow rate, but unfortunately it’s impossible to make a reasonable estimate of the cost reduction, since these costs are far from linear in nature (it would cost significantly more than 20% of the cost estimated for the SNTP engine). However, it’s a good example of the types of facilities needed, and the challenges associated with this approach.

SNTP Test Facility
Image courtesy DOE

The primary advantage to the ETS concept is that it doesn’t use H2O to cool the exhaust, but LH2. This means that the potential for release of large amounts of (very mildly) irradiated water into the groundwater supply are severely limited (although the water solubility of the individual fission products would not change). The disadvantage, of course, is that it requires large amounts of LH2 to be on hand. At Stennis SC, this is less of an issue, since LH2 facilities are already in place, but LH2 is – as we saw in the last blog post – a major headache. It was estimated that either a combined propellant-effluent coolant supply could be used (~181,440 kg), or a separate supply for the coolant system (~136,000 kg) could be used (numbers based on a maximum of 2 hours burn time per test). To get a sense of what this amount of LH2 would require, two ~1400 kl dewars of LH2 would be needed for the combined system, about ¾ of the LH2 supply available at Kennedy Space Center (~3200 kl).

Once the exhaust is sufficiently cooled, it is a fairly routine matter to filter out the fission products (a combination of physical filters and chemical reactions can ensure that no radionucleides are released, and radiation monitoring can verify that the H2 has been cleaned of all radioactive effluent). In the NF-1 test, water was used to capture the particulate matter, and the H2O was passed through a silica gel bed to remove the fission products. An activated carbon filter was used to remove the noble gasses and other gaseous and aerosol fission products. After this, depending on the facility setup, it is possible to recycle a good portion of the H2 from the test; however this has massive power requirements for the cryocoolers and hydrogen densification equipment to handle this massive amount of H2.

Saddle Mountain facility diagram
Alternative test facility layout

Due to both the irradiation of the facilities and the very different requirements for this type of test facility, it was determined that the facilities built for the NRDS during Rover would be insufficient for this sort of testing, and so new facilities would need to be constructed, with much larger LH2 storage capabilities. One more recent update to the concept is brought up in the SAFE proposal (next section), using already existing facilities at the Nevada Test Site (now National Nuclear Security Site), in the U-la or P-tunnel complexes. These underground facilities were horizontal, interconnected tunnel complexes used for sub-critical nuclear testing. There are a number of benefits to using these (now-unused) facilities for this type of testing: first, the rhyolite that the P-tunnel facility is cut into is far less permeable to fission products, but remains an excellent heat sink for the thermal effects of the exhaust plume. Second, it’s unlikely to fracture due to overpressure, although back-pressure into the engine itself will constrain the minimum size of the tunnel. Third, a hot cell can be cut into the mountain adjacent to the test location, making a very well-shielded facility for cool-down and disassembly beside the test location, eliminating the need to transport the now-hot engine to another facility for disassembly.

After the gas has passed through a length of tunnel, and cooled sufficiently, a heat exchanger is used to further cool the gas, and then it’s passed through an activated charcoal filter similar to the one used in the NF-1 test. This filtered H2 will then be flared off after going through a number of fission product detectors to ensure the filter maintained its’ integrity. The U-la tunnels are dug into alluvium, so we’ll look at those in the next section.

One concern with using charcoal filters is that their effectiveness varies greatly depending on the temperature of the effluent, and the pressure that it’s fed into the filter. Indeed, the H2 can push fission products through the filter, so there’s a definite limit to how small the filter can be. The longer the test, the larger the filter will be. Activated charcoal is relatively cheap, but by the end of the test it will be irradiated, meaning that it has to be disposed of in nuclear waste repositories.

Cost estimates were avoided in the DOD assessment, due to a number of factors, including uncertain site location and the possibility of using this facility for multiple programs, allowing for cost sharing, but the overall cost for the test systems and facilities was estimated to be $500M in 1993 dollars. Most papers seem to think that this is the most expensive, and least practical, option for above ground NTR testing.

The Borehole Option: Subsurface Active Filtration of Exhaust

Many different options have been suggested over the years as to testing options. The simplest is to fire the rocket with its’ nozzle pointed into a deep borehole at the Nevada Test Site, which has had extensive geological work done to determine soil porosity and other characteristics that would be important to the concept. Known as Subsurface Active Filtration of Exhaust, or SAFE, it was proposed in 1999 by the Center for Space Studies, and continued to be refined for a number of years.

SAFE schematic
SAFE concept, Howe 2012, image courtesy NASA

In this concept, the engine is placed over an already existing (from below-ground nuclear weapons testing) 8 foot wide, 1200 foot deep borehole, with a water spray system being mounted adjacent to the nozzle of the NTR. The first section of the hole will be clad in steel, and the rest will simply be lined with the rock that is being bored into. The main limiting consideration will be the migration of radionucleides into the surrounding rock, which is something that’s been modeled computationally using Frenchman’s Flat geologic data, but has not been verified.

SAFE injector model
SAFE injection system model, Howe 2012

The primary challenges associated with this type of testing will be twofold: first, it needs to be ensured that the fission products will not migrate into groundwater or the atmosphere; and second, in order to ensure that the surrounding bedrock isn’t fractured – and therefore allows greater-than-anticipated migration of fission products to migrate from the borehole – it is necessary to prevent the pressure in the borehole from reaching above a certain level. A sub-scale test with an RL-10 chemical rocket engine and radioisotope tracers was proposed (this test would have a much smaller borehole, and use known radioisotope tracers – either Xe or Kr isotopes – in the fuel to test dispersion of fission products through the bedrock). This test would provide the necessary migration, permeability, and (given appropriate borehole scaling to ensure prototypic temperature and pressure regimes) soil fracture pressures to ensure the full filtration of the exhaust of an NTR.

The advantage to doing this test at Frenchman’s Flat is that the ground has already been extensively tested for the porosity (35%), permeability (8 darcys), water content (initial pore saturation 30%), and homogeneity (alluvium, so pretty much 100%) that is needed. In fact, a model already exists to calculate the behavior of the soil to these effects, known as WAFE, and the model was applied to the test parameters in 1999. Both full thrust (73.4 kg/s of H2O from both exhaust and cooling spray, and 0.64 kg/s of H2) and 30% thrust (20.5 kg/s H2O and 0.33 kg/s of H2) were modeled, both assuming 600 C exhaust injection after the steel liner. They found that the maximum equilibrium pressure in the borehole would reach 36 psia for the full thrust test, and 21 psia in the 30% thrust case, after about 2 hours, well within the acceptable pressure range for the borehole, assuming the exhaust gases were limited to below Mach 1 to prevent excess back-pressure buildup.

P-Tunnel setup

Other options were explored as well, including using the use of the U-la facility at the NNSS for horizontal testing. This is an underground set of tunnels in Nevada, which would provide safety for the testing team and the availability of a hot cell for reactor disassembly beside the test point (the P-tunnel facility is also cut into similar alluvial deposits, so primary filtration will come from the soil itself, and water cooling will still be necessary).

INL geology 2
INL geological composition, image courtesy DOE

Further options were explored in the “Final Report – Assessment of Testing Options for the NTR at the INL.” This is a more geologically complex region, including pahoehoe and rubble basalt, and various types of sediment. Another complication is that INL is on the Snake River plain, and above an aquifer, so the site will be limited to those places that the aquifer is more than 450 feet below the surface. However, the pahoehoe basalt is gas-impermeable, so if a site can be found that has a layer of this basalt below the borehole but above the aquifer, it can provide a gas-impermeable barrier below the borehole.

A 1998 cost estimate by Bechtel Nevada on the test concept estimated a cost of $5M for the non-nuclear validation test, and $16M for the full-scale NTR test, but it’s unclear if this included cost for the hot cell and associated equipment that would need to be built to support the test campaign, and I haven’t been able to find the specific report.

However, this testing option does not seem to feature heavily in NASA’s internal discussions for NTR testing at this point. One of the disadvantages is that it would require the rocket testing equipment, and support facilities, to be built from scratch, and to occur on DOE property. NASA has an extensive rocket testing facility at the John C. Stennis Space Center in Hancock County, MS, which has geology that isn’t conducive to subterranean testing of any sort, much less testing that requires significant isolation from the water table, and most NASA presentations seem to focus on using this facility.

The main reasons given in a late 2017 presentation for not pursuing this option are: Unresolved issues on water saturation effects on soil permeability, hole pressure during engine operation, and soil effectiveness in exhaust filtering. I have been unable to find the Bechtel Nevada and Desert Research Institute studies on this subject, but they have been studied. I would be curious to know why these studies would be considered incomplete.

One advantage to these options, though, which cannot be overstated, is that these facilities would be located on DOE land. As was seen in the recent KRUSTY fission-powered test, nuclear reactors in DOE facilities use an internal certification and licensing program independent of the NRC. This means that the 9-10 year (or longer), incredibly expensive certification process, which has never been approved for a First of a Kind reactor, would be bypassed. This alone is a potentially huge cost savings for the project, and may offset the additional study required to verify the suitability of these sites for NTR testing compared to certifying a new location – no matter how well established it is for rocket testing already.

Above Ground Test Option #2: Complete Capture

Flow Diagram Coote 2017
Image via Coote 2017, courtesy NASA

In this NTR test setup, the exhaust is slowed from supersonic to subsonic speeds, allowing O2 to be injected and mixed well past the molar equilibrium point for H2O. The resultant mixture is then combusted, resulting in hot steam and free O2. A water sprayer is used to cool the steam, and then passes through a debris trap filled with water at the bottom. It is then captured in a storage pool, and the remaining gaseous O2 is run through a desiccant filter, which is exhausted into the same storage pool. The water is filtered of all fission products and any unburned fuel, and then released. The gaseous O2 is recaptured and cooled using liquid nitrogen, and whatever is unable to be efficiently recaptured is vented into the atmosphere. The primary advantage to this system is that the resulting H2O can be filtered at leisure, allowing for more efficient and thorough filtration without the worry of over-pressurization of the system if there’s a blockage in the filters.

Subscale Concept Render
Subscale test stand render, image courtesy BWXT via NASA

There are many questions that need to be answered to ensure that this system works properly, as there are with all of the systems that have yet to be tested. In other to verify that the system will work as advertised, a sub-scale demonstrator will need to be built. This facility will use a hydrogen wave heater in place of the nuclear reactor, and test the rest of the components at a smaller scale wherever possible. Due to the specific needs of the exhaust capture system, especially the need to test complete combustion at different heat loads, the height of the facility may not be able to be scaled down (in order to ensure complete combustion, the gas flow will need to be subsonic before mixing and combustion). Thermal loading on structures is another major concern for the sub-scale test, since many components must be tested at the appropriate temperature, and the smaller structures won’t be able to passively reject heat as well. Finally, some things won’t be able to be tested in a sub-scale system, so what data will need to be collected in the full-scale system needs to be assessed.

One last thing to note is that this system will also be used to verify that high-velocity impacts of hot debris will not be a concern. This was, of course, seen in many of the early Rover tests, as fuel elements would break and be ejected from the nozzle at similar velocities to the exhaust. While CERMET fuels are (likely) more durable, this is an accident condition that has to be prepared for. In addition, smaller pieces of debris need to be able to be fully captured as well (such as flakes of clad, or non-nuclear components). These tests will need to be carried out on the sub-scale test bed to ensure for the regulators that any accident is able to be addressed. This adds to the complexity of the test setup, and encourages the ability to change the test stand as quickly and efficiently as possible – in other words, to make it as modular as possible. This also increases the flexibility of the facility for any other uses that it may be put to.

NTP Testing at Stennis Space Center

SSC overview
Stennis SC test facilities, image courtesy NASA

This last testing concept seems to be the front-runner for current NASA designs, to be integrated into the A3 test stand at NASA’s Stennis Space Center (SSC). SSC is the premier rocket test facility for NASA, testing both solid and liquid rocket engines. The test facilities are located in the “fee area,” a 20 square mile area (avg. radius 2.5 miles) surrounded by an acoustic “buffer zone” that averages 7.9 miles in radius (195 sq mi). With available office space, manufacturing spaces, and indoor and outdoor warehouse space, as well as a number of rocket engine test stands, the facility has much going for it. Most of the rocket engines being used by American launch companies have been tested here, going all the way back to the moon program. This is a VERY advanced, well-developed facility for the development of any type of chemical engine ever developed… but unfortunately, nuclear is different. Because SSC has not supported nuclear operations, a number of facilities will need to be constructed to support NTR testing at the facility. This raises the overall cost of the program considerably, to less than but around $850M (in 2017 dollars). A number of facilities will need to be constructed at SSC to support NTR testing, for both E3 and A3 test stands.

Diagram side by side with A3
Image from Houts presentation 2017, via NASA

As one of the newer facilities at SSC, the A3 test stand groundbreaking was held in August of 2007, and was completed in 2014. It is the only facility that is able to handle the thrust level (300+ Klbf at altitude, 1,000 Klbf nominal design) and simulated altitude (100 Kft) that testing a powerful upper stage requires. There are two additional facilities designed to operate at lower-than-ambient atmospheric pressures at SSC, the A2 test stand (650 Klbf at 60 Kft) and the E3 test facility (60 Klbf at 100 Kft). The E3 facility will be used for sub-scale testing, turbopump validation, and other tests for the NTP program, but the A2 test stand seems to not be under consideration at this time. The rest of the test stands at SSC are designed to operate at ambient pressure (i.e. sea level), and so they are not suitable for NTP testing.

The E3 facility would be used for sub-scale testing, first of the turbopumps (similar to the tests done there for the SSME), and sub-scale reactor tests. These would likely be the first improvements made at SSC to support the NTP testing, within the next couple years, and would cost $35-38M ($15-16M for sub-scale turbopump tests, $20-22M for the sub-scale reactor test, according to preliminary BWXT cost estimates). Another thing that would be tested at E3 would be a sub-scale engine exhaust capture system, which has been approved for both Phases 1&2, work to support this should be starting at any time ($8.74M was allocated to this goal in the FY’14 budget). From what I can see, work had already started (to an unknown extent) at E3 on this sub-scale system, however I have been unable to find information regarding the extent of the work or the scale that the test stand will be compared to the full system.

A3 under construction
A3 test stand under construction, image courtesy NASA

The A3 facility has the most that needs to be added, including facilities for power pack testing ($21M); a full-flow, electrically heated depleted uranium test (cost TBD); a facility for zero power testing and reactor verification before testing ($15M); an adjacent hot cell for reactor cool-down and disassembly (the new version of the EMAD facility, $220M); and testing for both sub-scale and full scale fission powered NTP testing (cost to be determined, it’s likely to be heavily influenced by regulatory burden). This does not include radiation shielding, and an alternate ducting system to ensure that the HVAC system doesn’t become irradiated (a major headache in the decomissioning of the original E-MAD facility). It is unlikely that design work for this facility will start in earnest until FY21, and construction of the facility is not likely to start until FY24. Assuming a 10 year site licensing lead time (which is typical), it is unlikely that any nuclear testing will be able to be done until FY29, with full power nuclear testing not likely until FY30.

Notional schedule
Notional Development Timeline

Documents relating to the test stands at SSC show that there has been some funding for this project since FY ‘16, but it’s difficult to tell how much of that has gone to analysis, environmental impact studies, and other bureaucratic and regulatory necessities, and how much has gone to actual construction. I HAVE had one person who works at SSC mention that physical work has started, but they were unwilling to provide any more information than that due to their not being authorized to speak to the public about the work, and their unfamiliarity with what is and isn’t public knowledge (most of it simply isn’t public). According to a presentation at SSC in July of 2017, the sub-scale turbopump testing may start in the next year or two, but initial design work for the A3 test stand is unlikely to start before FY’21.

NTP draft tech demonstration draft timeline
Draft Tech Development Roadmap, image via NASA

According to the presentation (linked below), there are two major hurdles the program needs to overcome on the policy and regulatory side. First, a national/agency level decision needs to be made between NASA, the DOE, and the NRC as to the specific responsibilities and roles for NTP development, especially in regards to: 1. reactor production, engine and launch vehicle integration strategy, and 2. ground, launch, and in-space operations of the NTR. Second, NTP testing at SSC requires a nuclear site license, which is a 9-10 year process even for a traditional light water power reactor, much less as unusual a reactor architecture as an NTR. This is another area that BWXT’s experience is being leaned on heavily, with two (not publicly available) studies having been carried out by them in FY16 on both a site licensing strategy and implementation roadmap, and on initial identification of policy issues related to licensing an NTP ground test at SSC.

Regulatory Burdens, Bureaucratic Concerns, and Other Matters

Originally, this post was going to delve into the regulatory and environmental challenges of doing NTR testing. An NTR is very different from any other sort of nuclear reactor, not only because it’s a once-through gas cooled reactor operating at a very high temperature, but also due to the performance characteristics that the reactor is expected to be able to provide.

Additionally, these are short-lived reactors – 100 hours of operation is more than enough to complete an entire crewed mission to Mars, and is a long lifetime for a rocket engine. However, as we saw during the Rover hot-fire testing, there are always issues that come up that aren’t able to be adequately tested beforehand (even with our far more advanced computational and modeling capabilities), so iteration is key. This means that the site has to be licensed for multiple different reactors.

Unfortunately, these subjects are VERY complex, and are very difficult to learn. Communicating with the NRC in and of itself is a subspecialty of both the nuclear and legal industries for reactor designers. The fact that the DOE, NASA, and the NRC are having to interact on this project just adds to the complexity.

So, I’m going to put that part of this off for now, and it will become its’ own separate blog post. I have contacted NASA, the DOE and the NRC looking for additional information and clarification in their various areas, and hopefully will hear back in the coming weeks or months. I also am reading the appropriate regulations and internal rules for these organizations, and there’s more than enough there for a quite lengthy blog post on its’ own. If you work with any of these organizations, and are either able to help me gather this information or get me in touch with someone that can, I would greatly appreciate it if you contact me.

Upcoming Posts!

For now, we’re going to leave testing behind as the main focus of the blog, but we will still look at the subject as it becomes relevant in other posts. For now, we’re going to do one final post on solid core pure NTRs, looking at carbide fueled NTRs, both the RD-0410 in Russia and some legacy and new designs from the US. After that, we’ll move on to bimodal NTR/chemical and bimodal NTR/thermal electric designs in the next post.

After that, with one small exception, we’ll leave NTRs behind for a while, and look at nuclear electric propulsion. I plan on doing pages for individual reactor designs during this time, both NTR and NEP, and add the as their own pages on the website. As I write posts, I’ll link to the new (or updated) pages as they’re completed.

Be sure to check out the rest of the website, and join us on Facebook! This blog is far from the only thing going on!

 

References:

In Pile Testing

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel; ORNL/TM-2017/376, Howard et al September 2017

https://info.ornl.gov/sites/publications/Files/Pub100562.pdf

DOE Order 414.1D, Quality Assurance; approved 4/2011

https://info.ornl.gov/sites/publications/Files/Pub100562.pdf

10 CFR Part 830, Nuclear Safety Management; https://info.ornl.gov/sites/publications/Files/Pub100562.pdf

High Flux Isotope Reactor homepage: https://neutrons.ornl.gov/hfir

Advanced Test Reactor Irradiation Facilities and Capabilities; Furstenau and Glover 2009

https://web.archive.org/web/20090508234733/http://anes.fiu.edu/Pro/s8Fur.pdf

Transient Reactor Test Facility homepage: http://www4vip.inl.gov/research/transient-reactor-test-facility/

Al 6061 Matweb page: http://asm.matweb.com/search/SpecificMaterial.asp?bassnum=ma6061t6

300 Stainless Steel; Pennsylvania Stainless, http://www.pennstainless.com/stainless-grades/300-series-stainless-steel/

Grade 5 Titanium Matweb page: http://asm.matweb.com/search/SpecificMaterial.asp?bassnum=mtp641

SIGRATHERM, SGL (manufacturer) website: https://www.sglgroup.com/cms/international/products/product-groups/cfrc_felt/speciality-graphites-for-high-temperature-furnaces/soft-felt.html?__locale=en

Nuclear Furnace ECS

Nuclear Furnace 1 Test Report; LA-5189-MS, by W.L. Kirk, 1973

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

DOE Fact Sheet, Appendix 2

https://digital.library.unt.edu/ark:/67531/metadc619748/m2/1/high_res_d/101088.pdf

Above Ground Effluent Treatment System

Space Nuclear Thermal Propulsion Final Report, R.A. Haslett, Grumman Aerospace Corp, 1995 http://www.dtic.mil/get-tr-doc/pdf?AD=ADA305996

Space Nuclear Thrmal Propulsion Test Facilities Subpanel Final Report, Allen et al, 1993 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19930015916.pdf

Subsurface Active Filtration of Exhaust (SAFE)

Ground Testing a Nuclear Thermal Rocket: Design of a sub-scale demonstration experiment, Howe et al, Center for Space Nuclear Research, 2012

http://large.stanford.edu/courses/2014/ph241/wendorff1/docs/aiaa-2012-3743.pdf

Subscale Validation of the Subsurface Active Filtration of Exhaust Approach to NTP Ground Testing, Marshall et al, NASA Glenn RC, 2015 (Conference Paper) https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20150022127.pdf and (Presentation Slides) https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20150021275.pdf

Final Report – Assessment of Testing Options for the NTR at INL, Howe et al, Idaho NL, 2013

https://inldigitallibrary.inl.gov/sites/sti/sti/5806466.pdf

Complete Exhaust Capture and NASA Planning

Stennis Space Center Activities and Plans Overview presentation, NASA

https://www.nasa.gov/sites/default/files/files/OverviewofSSC_CSActivitiesandPlans_508.pdf

Development and Utilization of Nuclear Thermal Propulsion; Houts and Mitchell, 2016 (slideshow)

https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20160002256.pdf

Low Enriched Uranium (LEU) Nuclear Thermal Propulsion: System Overview and Ground Test Strategy, Coote 2017 (slideshow)

https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/20170011172.pdf

NASA FY18 Budget Estimates:

https://www.nasa.gov/sites/default/files/atoms/files/fy_2018_budget_estimates.pdf

NTP Technical Interchange Meeting at SSC, June 2017 (slideshow)

https://www.nasa.gov/sites/default/files/atoms/files/ntp_tim_at_ssc_-_2728-jun-17.pptx

Categories
Development and Testing Fission Power Systems Nuclear Thermal Systems Test Stands

NTR Hot Fire Testing Part I: Rover and NERVA Testing

Hello, and welcome back to Beyond NERVA, where today we are looking at ground testing of nuclear rockets. This is the first of two posts on ground testing NTRs, focusing on the testing methods used during Project ROVER, including a look at the zero power testing and assembly tests carried out at Los Alamos Scientific Laboratory, and the hot-fire testing done at the National Defense Research Station at Jackass Flats, Nevada. The next post will focus on the options that both have and are being considered for hot fire testing the next generation of LEU NTP, as well as a brief look at cost estimates for the different options, and the plans that NASA has proposed for the facilities that are needed to support this program (what little of it is available).

We have examined how to test NTR fuel elements in nun-nuclear situations before, and looked at two of the test stands that were developed for testing thermal, chemical, and erosion effects on them as individual components, the Compact Fuel Element Environment Simulator (CFEET) and the Nuclear Thermal Rocket Environment Effects Simulator (NTREES). These test stands provide economical means of testing fuel elements before loading them into a nuclear reactor for neutronic and reactor physics behavioral testing, and can catch many problems in terms of chemical and structural problems without dealing with the headaches of testing a nuclear reactor.

However, as any engineer can tell you, computer modeling is far from enough to test a full system. Without extensive real-life testing, no system can be used in real-life situations. This is especially true of something as complex as a nuclear reactor – much less a rocket engine. NTRs have the challenge of being both.

NTS_-_EMAD_Facility_002
Engine Maintenance and Disassembly Facility, image via Wikimedia Commons

Back in the days of Project Rover, there were many nuclear propulsion tests performed. The most famous of these were the tests carried out at Jackass Flats, NV, in the National Nuclear Test Site (Now the National Criticality Experiment Research Center), in open-air testing on specialized rail cars. This was far from the vast majority of human habitation (there was one small – less than 100 people – ranch upwind of the facility, but downwind was the test site for nuclear weapons tests, so any fallout from a reactor meltdown was not considered a major concern).

The test program at the Nevada site started with the arrival of the fully-constructed and preliminary tested rocket engines arriving by rail from Los Alamos, NM, along with a contingent of scientists, engineers, and additional technicians. After doing another check-out of the reactor, they were hooked up (still attached to the custom rail car it was shipped on) to instrumentation and hydrogen propellant, and run through a series of tests, ramping up to either full power or engine failure. Rocket engine development in those days (and even today, sometimes) could be an explosive business, and hydrogen was a new propellant to use, so accidents were unfortunately common in the early days of Rover.

After the test, the rockets were wheeled off onto a remote stretch of track to cool down (from a radiation point of view) for a period of time, before being disassembled in a hot cell (a heavily shielded facility using remote manipulators to protect the engineers) and closely examined. This examination verified how much power was produced based on the fission product ratios of the fuel, examined and detailed all of the material and mechanical failures that had occurred, and started the reactor decommissioning and disposal procedures.

As time went on, great strides were made not only in NTR design, but in metallurgy, reactor dynamics, fluid dynamics, materials engineering, manufacturing techniques, cryogenics, and a host of other areas. These rocket engines were well beyond the bleeding edge of technology, even for NASA and the AEC – two of the most scientifically advanced organizations in the world at that point. This, unfortunately, also meant that early on there were many failures, for reasons that either weren’t immediately apparent or that didn’t have a solution based on the design capabilities of the day. However, they persisted, and by the end of the Rover program in 1972, a nuclear thermal rocket was tested successfully in flight configuration repeatedly, the fuel elements for the rocket were advancing by leaps and bounds past the needed specifications, and with the ability to cheaply iterate and test new versions of these elements in new, versatile, and reusable test reactors, the improvements were far from stalling out – they were accelerating.

However, as we know, the Rover program was canceled after NASA was no longer going to Mars, and the development program was largely scrapped. Scientists and engineers at Westinghouse Astronuclear Laboratory (the commercial contractor for the NERVA flight engine), Oak Ridge National Laboratory (where much of the fuel element fabrication was carried out) and Los Alamos Scientific Laboratory (the AEC facility primarily responsible for reactor design and initial testing) spent about another year finishing paperwork and final reports, and the program was largely shut down. The final report on the hot-fire test programs for NASA, though, wouldn’t be released until 1991.

Behind the Scenes: Pre-Hot Fire Testing of ROVER reactors

Pajarito map
Pajarito Test Area, image courtesy LANL

These hot fire tests were actually the end result of many more tests carried out in New Mexico, at Los Alamos Scientific Laboratory – specifically the Pajarito Test Area. Here, there were many test stands and experimental reactors used to measure such things as neutronics, reactor behavior, material behavior, critical assembly limitations and more.

Honeycomb grainy closeup
Honeycomb, with a KIWI mockup loaded. Image via LANL

The first of these was known as Honeycomb, due to its use of square grids made out of aluminum (which is mostly transparent to neutrons), held in large aluminum frames. Prisms of nuclear fuel, reflectors, neutron absorbers, moderator, and other materials were assembled carefully (to prevent accidental criticality, something that the Pajarito Test Site had seen early in its’ existence in the Demon Core experiments and subsequent accident) to ensure that the modeled behavior of possible core configurations matched closely enough to predicted behavior to justify going through the effort and expense of going on to the next steps of refining and testing fuel elements in an operating reactor core. Especially for cold and warm criticality tests, this test stand was invaluable, but with the cancellation of Project Rover, there was no need to continue using the test stand, and so it was largely mothballed.

PARKA
PARKA, image courtesy LANL

The second was a modified KIWI-A reactor, which used a low-pressure, heavy water moderated island in the center of the reactor to reduce the amount of fissile fuel necessary for the reactor to achieve criticality. This reactor, known as Zepo-A (for zero-power, or cold criticality), was the first of an experiment that was carried out with each successive design in the Rover program, supporting Westinghouse Astronuclear Laboratory and the NNTS design and testing operations. As each reactor went through its’ zero-power neutronic testing, the design was refined, and problems corrected. This sort of testing was completed late in 2017 and early in 2018 at the NCERC in support of the KRUSTY series of tests, which culminated in March with the first full-power test of a new nuclear reactor in the US for more than 40 years, and remain a crucial testing phase for all nuclear reactor and fuel element development. An early, KIWI-type critical assembly test ended up being re-purposed into a test stand called PARKA, which was used to test liquid metal fast breeder reactor (LMFBR, now known as “Integral Fast Reactor or IFR, under development at Idaho National Labs) fuel pins in a low-power, epithermal neutron environment for startup and shutdown transient behavior testing, as well as being a well-understood general radiation source.

Hot Gas Furnace
Hot gas furnace at LASL, image courtesy LANL

Finally, there was a pair of hot gas furnaces (one at LASL, one at WANL) for electrical heating of fuel elements in an H2 environment that used resistive heating to bring the fuel element up to temperature. This became more and more important as the project continued, since development of the clad on the fuel element was a major undertaking. As the fuel elements became more complex, or as materials that were used in the fuel element changed, the thermal properties (and chemical properties at temperature) of these new designs needed to be tested before irradiation testing to ensure the changes didn’t have unintended consequences. This was not just for the clad, the graphite matrix composition changed over time as well, transitioning from using graphite flour with thermoset resin to a mix of flour and flakes, and the fuel particles themselves changed from uranium oxide to uranium carbide, and the particles themselves were coated as well by the end of the program. The gas furnace was invaluable in these tests, and can be considered the grandfather of today’s NTREES and CFEET test stands.

KIWI A Zepo A and Honeycomb shot
KIWI-A, Zepo-A, and Honeycomb mockup in Kiva 3. Image courtesy LANL

An excellent example of the importance of these tests, and the careful checkout that each of the Rover reactors received, can be seen with the KIWI-B4 reactor. Initial mockups, both on Honeycomb and in more rigorous Zepo mockups of the reactor, showed that the design had good reactivity and control capability, but while the team at Los Alamos was assembling the actual test reactor, it was discovered that there was so much reactivity the core couldn’t be assembled! Inert material was used in place of some of the fuel elements, and neutron poisons were added to the core, to counteract this excess reactivity. Careful testing showed that the uranium carbide fuel particles that were suspended in the graphite matrix underwent hydrolysis, moderating the neutrons and therefore increasing the reactivity of the core. Later versions of the fuel used larger particles of UC2, which was then individually coated before being distributed through the graphite matrix, to prevent this absorption of hydrogen. Careful testing and assembly of these experimental reactors by the team at Los Alamos ensured the safe testing and operation of these reactors once they reached the Nevada test site, and supported Westinghouse’s design work, Oak Ridge National Lab’s manufacturing efforts, and the ultimate full-power testing carried out at Jackass Flats.

NTR Core Development Process
NTR Core Design Process, image courtesy IAEA

Once this series of mockup crude criticality testing, zero-power testing, assembly, and checkout was completed, the reactors were loaded onto a special rail car that would also act as a test stand with the nozzle up, and – accompanied by a team of scientists and engineers from both New Mexico and Nevada – transported by train to the test site at Jackass Flats, adjacent to Nellis Air Force Base and the Nevada Test Site, where nuclear weapons testing was done. Once there, a final series of checks was done on the reactors to ensure that nothing untoward had happened during transport, and the reactors were hooked up to test instrumentation and the coolant supply of hydrogen for testing.

Problems at Jackass Flats: Fission is the Easy Part!

The testing challenges that the Nevada team faced extended far beyond the nuclear testing that was the primary goal of this test series. Hydrogen is a notoriously difficult material to handle due to its’ incredibly tiny size and mass. It seeps through solid metal, valves have to be made with incredibly tight clearances, and when it’s exposed to the atmosphere it is a major explosion hazard. To add to the problems, these were the first days of cryogenic H2 experimentation. Even today, handling of cryogenic H2 is far from a routine procedure, and the often unavoidable problems with using hydrogen as a propellant can be seen in many areas – perhaps the most spectacular can be seen during the launch of a Delta-IV Heavy rocket, which is a hydrolox (H2/O2) rocket. Upon ignition of the rocket engines, it appears that the rocket isn’t launching from the pad, but exploding on it, due to the outgassing of H2 not only from the pressure relief valves in the tanks, but seepage from valves, welds, and through the body of the tanks themselves – the rocket catching itself on fire is actually standard operating procedure!

PB K Site Cryo Tank Test
Plu Brook Cryo Tank Pressure Test, image courtesy NASA

In the late 1950s, these problems were just being discovered – the hard way. NASA’s Plum Brook Research Station in Ohio was a key facility for exploring techniques for handling gaseous and liquid hydrogen safely. Not only did they experiment with cryogenic equipment, hydrogen densification methods, and liquid H2 transport and handling, they did materials and mechanical testing on valves, sensors, tanks, and other components, as well as developed welding techniques and testing and verification capabilities to improve the ability to handle this extremely difficult, potentially explosive, but also incredible valuable (due to its’ low atomic mass – the exact same property that caused the problems in the first place!) propellant, coolant, and nuclear moderator. The other options available for NTR propellant (basically anything that’s a gas at reactor operating temperatures and won’t leave excessive residue) weren’t nearly as good of an option due to the lower exhaust velocity – and therefore lower specific impulse.

Plum Brook is another often-overlooked facility that was critical to the success of not just NERVA, but all current liquid hydrogen fueled systems. I plan on doing another post (this one’s already VERY long) looking into the history of the various facilities involved with the Rover and NERVA program.

Indeed, all the KIWI-A tests and the KIWI-B1A used gaseous hydrogen instead of liquid hydrogen, because the equipment that was planned to be used (and would be used in subsequent tests) was delayed due to construction problems, welding issues, valve failures, and fires during checkout of the new systems. These teething troubles with the propellant caused major problems at Jackass Flats, and caused many of the flashiest accidents that occurred during the testing program. Hydrogen fires were commonplace, and an accident during the installation of propellant lines in one reactor ended up causing major damage to the test car, the shed it was contained in, and exposed instrumentation, but only minor apparent damage to the reactor itself, delaying the test of the reactor for a full month while repairs were made (this test also saw two hydrogen fires during testing, a common problem that improved as the program continued and the methods for handling the H2 were improved).

While the H2 coolant was the source of many problems at Jackass Flats, other issues arose due to the fact that these NTRs were using technology that was well beyond bleeding-edge at the time. New construction methods doesn’t begin to describe the level of technological innovation in virtually every area that these engines required. Materials that were theoretical chemical engineering possibilities only a few years before (sometimes even months!) were being utilized to build innovative, very high temperature, chemically and neutronically complex reactors – that also functioned as rocket engines. New metal alloys were developed, new forms of graphite were employed, experimental methods of coating the fuel elements to prevent hydrogen from attacking the carbon of the fuel element matrix (as seen in the KIWI-A reactor, which used unclad graphite plates for fuel, this was a major concern) were constantly being adjusted – indeed, the clad material experimentation continues to this day, but with advanced micro-imaging capabilities and a half century of materials science and manufacturing experience since then, the results now are light-years ahead of what was available for the scientists and engineers in the 50s and 60s. Hydrodynamic principles that were only poorly understood, stress and vibrational patterns that weren’t able to be predicted, and material interactions at temperatures higher than are experienced in the vast majority of situations caused many problems for the Rover reactors.

One common problem in many of these reactors was transverse fuel element cracking, where a fuel element would split across the narrow axis, disrupting coolant flow through the interior channels, exposing the graphite matrix to the hot H2 (which it then would ferociously eat away, exposing both fission products and unburned fuel to the H2 stream and carry it elsewhere – mostly out of the nozzle, but it turned out the uranium would congregate at the hottest points in the reactor – even against the H2 stream – which could have terrifying implications for accidental fission power hot spots. Sometimes, large sections of the fuel elements would be ejected out of the nozzle, spraying partially burned nuclear fuel into the air – sometimes as large chunks, but almost always some of the fuel was aerosolized. Today, this would definitely be unacceptable, but at the time the US government was testing nuclear weapons literally next door to this facility, so it wasn’t considered a cause of major concern.

If this sounds like there were major challenges and significant accidents that were happening at Jackass Flats, well in the beginning of the program that was certainly correct. These early problems were also cited in Congress’ decision to not continue to fund the program (although, without a manned Mars mission, there was really no reason to use the expensive and difficult to build systems, anyway). The thing to remember, though, is that they were EARLY tests, with materials that had been a concept in a material engineer’s imagination only a few years (or sometimes months) beforehand, mechanical and thermal stresses that no-one had ever dealt with, and a technology that seemed the only way to send humans to another planet. The moon was hard enough, Mars was millions of miles further away.

Hot Fire Testing: What Did a Test Look Like?

Nuclear testing is far more complex than just hooking up the test reactor to coolant and instrumentation lines, turning the control drums and hydrogen valves, and watching the dials. Not only are there many challenges associated with just deciding what instrumentation is possible, and where it would be placed, but the installation of these instruments and collecting data from them was often a challenge as well early in the program.

Axial flow path
NRX A2 Flow Diagram, image via NASA (Finseth, 1991)

To get an idea of what a successful hot fire test looks like, let’s look at a single reactor’s test series from later in the program: the NRX A2 technology demonstration test. This was the first NERVA reactor design to be tested at full power by Westinghouse ANL, the others, including KIWI and PHOEBUS, were not technology demonstration tests, but proof-of-concept and design development tests leading up to NERVA, and were tested by LASL. The core itself consisted of 1626 hexagonal prismatic fuel elements This reactor was significantly different from the XE-PRIME reactor that would be tested five years later. One way that it was different was the hydrogen flow path: after going through the nozzle, it would enter a chamber beside the nozzle and above the axial reflector (the engine was tested nozzle-up, in flight configuration this would be below the reflector), then pass through the reflector to cool it, before being diverted again by the shield, through the support plate, and into the propellant channels in the core before exiting the nozzle

Two power tests were conducted, on September 24 and October 15, 1964.

With two major goals and 22 lesser goals, the September 24 test packed a lot into the six minutes of half-to-full power operation (the reactor was only at full power for 40 seconds). The major goals were: 1. Provide significant information for verifying steady-state design analysis for powered operation, and 2. Provide significant information to aid in assessing the reactor’s suitability for operation at steady-state power and temperature levels that were required if it was to be a component in an experimental engine system. In addition to these major, but not very specific, test goals, a number of more specific goals were laid out, including top priority goals of evaluating environmental conditions on the structural integrity of the reactor and its’ components, core assembly performance evaluation, lateral support and seal performance analysis, core axial support system analysis, outer reflector assembly evaluation, control drum system evaluation, and overall reactivity assessment. The less urgent goals were also more extensive, and included nozzle assembly performance, pressure vessel performance, shield design assessment, instrumentation analysis, propellant feed and control system analysis, nucleonic and advanced power control system analysis, radiological environment and radiation hazard evaluation, thermal environment around the reactor, in-core and nozzle chamber temperature control system evaluation, reactivity and thermal transient analysis, and test car evaluation.

Test plot
Image via NASA (Finseth, 1991)

Several power holds were conducted during the test, at 51%, 84%, and 93-98%, all of which were slightly above the power that the holds were planned at. This was due to compressability of the hydrogen gas (leading to more moderation than planned) and issues with the venturi flowmeters used to measure H2 flow rates, as well as issues with the in-core thermocouples used for instrumentation (a common problem in the program), and provides a good example of the sorts of unanticipated challenges that these tests are meant to evaluate. The test length was limited by the availability of hydrogen to drive the turbopump, but despite this being a short test, it was a sweet one: all of the objectives of the test were met, and an ideal specific impulse in a vacuum equivalent of 811 s was determined (low for an NTR, but still over twice as good as any chemical engine at the time).

Low Power Test plot
Image via NASA (Finseth, 1991)

The October 15th test was a low-power, low flow test meant to evaluate the reactor’s operation when not operating in a high power, steady state of operation, focusing on reactor behavior at startup and cool-down. The relevant part of the test lasted for about 20 minutes, and operated at 21-53 MW of power and a flow rate of 2.27-5.9 kg/s of LH2. As with any system, operating at the state that the reactor was designed to operate in was easier to evaluate and model than at startup and shutdown, two conditions that every engine has to go through but are far outside the “ideal” conditions for the system, and operating with liquid hydrogen just made the questions greater. Only four specific objectives were set for this test: demonstration of stability at low LH2 flow (using dewar pressure as a gauge), demonstration of suitability at constant power but with H2 flow variation, demonstration of stability with fixed control drums but variable H2 flow to effect a change in reactor power, and getting a reactivity feedback value associated with LH2 at the core entrance. Many of these tests hinge on the fact that the LH2 isn’t just a coolant, but a major source of neutron moderation, so the flow rate (and associated changes in temperature and pressure) of the propellant have impacts extending beyond just the temperature of the exhaust. This test showed that there were no power or flow instabilities in the low-power, low-flow conditions that would be seen even during reactor startup (when the H2 entering the core was at its’ densest, and therefore most moderating). The predicted behavior and the test results showed good correlation, especially considering the instrumentation used (like the reactor itself) really wasn’t designed for these conditions, and the majority of the transducers used were operating at the extreme low range of their scale.

After the October test, the reactor was wheeled down a shunt track to radiologically cool down (allow the short-lived fission products to decay, reducing the gamma radiation flux coming off the reactor), and then was disassembled in the NRDC hot cell. These post-mortem examinations were an incredibly important tool for evaluating a number of variables, including how much power was generated during the test (based on the distribution of fission products, which would change depending on a number of factors, but mainly due to the power produced and the neutron spectrum that the reactor was operating in when they were produced), chemical reactivity issues, mechanical problems in the reactor itself, and several other factors. Unfortunately, disassembling even a simple system without accidentally breaking something is difficult, and this was far from a simple system. A challenge became “did the reactor break that itself, or did we?” This is especially true of fuel elements, which often broke due to inadequate lateral support along their length, but also would often break due to the way they were joined to the cold end of the core (which usually involved high-temperature, reasonably neutronically stable adhesives).

This issue was illustrated in the A2 test, when there were multiple broken fuel elements that did not have erosion at the break. This is a strong indicator that they broke during disassembly, not during the test itself: hot H2 tends to heavily erode the carbon in the graphite matrix – and the carbide fuel pellets – and is a very good indicator if the fuel rods broke during a power test. Broken fuel elements were a persistent problem in the entire Rover and NERVA programs (sometimes leading to ejection of the hot end portion of the fuel elements), and the fact that all of the fueled ones seem to have not broken was a major victory for the fuel fabricators.

This doesn’t mean that the fuel elements weren’t without their problems. Each generation of reactors used different fuel elements, sometimes multiple different types in a single core, and in this case the propellant channels, fuel element ends, and the tips of the exterior of the elements were clad in NbC, but the full length of the outside of the elements was not, to attempt to save mass and not overly complicate the neutronic environment of the reactor itself. Unfortunately, this means that the small amount of gas that slipped between the filler strips and pyro-tiles placed to prevent this problem could eat away at the middle of the outside of the fuel element (toward the hot end), something known as mid-band corrosion. This occurred mostly on the periphery of the core, and had a characteristic pattern of striations on the fuel elements. A change was made, to ensure that all of the peripheral fuel elements were fully clad with NbC, since the areas that had this clad were unaffected. Once again, the core became more complex, and more difficult to model and build, but a particular problem was addressed due to empirical data gathered during the test. A number of unfueled, instrumented fuel elements in the core were found to have broken in such a way that it wasn’t possible to conclusively rule out handling during disassembly, however, so the integrity of the fuel elements was still in doubt.

The problems associated with these graphite composite fuel elements never really went away during ROVER or NERVA, with a number of broken fuel elements (which were known to have been broken during the test) were found in the PEWEE reactor, the last test of this sort of fuel element matrix (NF-1 used either CERMET – then called composite – or carbide fuel elements, no GC fuel elements were used). The follow-on A3 reactor exhibited a form of fuel erosion known as pin-hole erosion, which the NbC clad was unable to address, forcing the NERVA team to other alternatives. This was another area where long-term use of the GC fuel elements was shown to be unsustainable for long-duration use past the specific mission parameters, and a large part of why the entire NERVA engine was discarded during staging, rather than just the propellant tanks as in modern designs. New clad materials and application techniques show a lot of promise, and GC is able to be used in a carefully designed LEU reactor, but this is something that isn’t really being explored in any depth in most cases (both the LANTR and NTER concepts still use GC fuel elements, with the NTER specifying them exclusively due to fuel swelling issues, but that seems to be the only time it’s actually required).

Worse Than Worst Case: KIWI-TNT

One question that often is asked by those unfamiliar with NTRs is “what happens if it blows up?” The short answer is that they can’t, for a number of reasons. There is only so much reactivity in a nuclear reactor, and only so fast that it can be utilized. The amount of reactivity is carefully managed through fuel loading in the fuel elements and strategically placed neutron poisons. Also, the control systems used for the nuclear reactors (in this case, control drums placed around the reactor in the radial reflector) can only be turned so fast. I recommend checking out the report on Safety Neutronics in Rover Reactors liked at the end of this post if this is something you’d like to look at more closely.

However, during the Rover testing at NRDS one WAS blown up, after significant modification that would not ever be done to a flight reactor. This is the KIWI-TNT test (TNT is short for Transient Nuclear Test). The behavior of a nuclear reactor as it approaches a runaway reaction, or a failure of some sort, is something that is studied in all types of reactors, usually in specially constructed types of reactors. This is required, since the production design of every reactor is highly optimized to prevent this sort of failure from occurring. This was also true of the Rover reactors. However, knowing what a fast excursion reaction would do to the reactor was an important question early in the program, and so a test was designed to discover exactly how bad things could be, and characterize what happened in a worse-than-worst-case scenario. It yielded valuable data for the possibility of an abort during launch that resulted in the reactor falling into the ocean (water being an excellent moderator, making it more likely that accidental criticality would occur), if the launch vehicle exploded on the pad, and also tested the option of destroying the reactor in space after it had been exhausted of its’ propellant (something that ended up not being planned for in the final mission profiles).

B4A Cutaway
KIWI B4A reactor, which KIWI-TNT was based on, image via LANL

What was the KIWI-TNT reactor? The last of the KIWI series of reactors, its’ design was very similar to the KIWI-B4A reactor (the predecessor for the NERVA-1 series of reactors), which was originally designed as a 1000 MW reactor with an exhaust exit chamber temperature of 2000 C. However, a number of things prevented a fast excursion from happening in this reactor: first, the shims used for the fuel elements were made of tantalum, a neutron poison, to prevent excess reactivity; second, the control drums used stepping motors that were slow enough that a runaway reaction wasn’t possible; finally, this experiment would be done without coolant, which also acted as moderator, so much more reactivity was needed than the B4A design allowed. With the shims removed, excess reactivity added to the point that the reactor was less than 1 sub-critical (with control drums fully inserted) and $6 of excess reactivity available relative to prompt critical, and the drum rotation rate increased by a factor of 89(!!), from 45 deg/s to 4000 deg/s, the stage was set for this rapid scheduled disassembly on January 12, 1965. This degree of modification shows how difficult it would be to have an accidental criticality accident in a standard NTR design.

Test Vehicle Schematic
KIWI-TNT Test Stand Schematic, image via LANL

The test had six specific goals: 1. Measure reaction history and total fissions produced under a known reactivity and compare to theoretical predictions in order to improve calculations for accident predictions, 2. to determine distribution of fission energy between core heating and vaporization, and kinetic energies, 3. determination of the nature of the core breakup, including the degree of vaporization and particle sizes produced, to test a possible nuclear destruct system, 4. measure the release into the atmosphere of fission debris under known conditions to better calculate other possible accident scenarios, 5. measure the radiation environment during and after the power transient, and 6. to evaluate launch site damage and clean-up techniques for a similar accident, should it occur (although the degree of modification required to the reactor core shows that this is a highly unlikely event, and should an explosive accident occur on the pad, it would have been chemical in nature with the reactor never going critical, so fission products would not be present in any meaningful quantities).

There were 11 measurements taken during the test: reactivity time history, fission rate time history, total fissions, core temperatures, core and reflector motion, external pressures, radiation effects, cloud formation and composition, fragmentation and particle studies, and geographic distribution of debris. An angled mirror above the reactor core (where the nozzle would be if there was propellant being fed into the reactor) was used in conjunction with high-speed cameras at the North bunker to take images of the hot end of the core during the test, and a number of thermocouples placed in the core.

KIWI-TNT via Pinterest
KIWI-TNT test, AEC image via SomethingAwful

As can be expected, this was a very short test, with a total of 3.1×10^20 fissions achieved after only 12.4 milliseconds. This was a highly unusual explosion, not consistent with either a chemical or nuclear explosion. The core temperature exceeded 17.5000 C in some locations, vaporizing approximately 5-15% of the core (the majority of the rest either burned in the air or was aerosolized into the cloud of effluent), and produced 150 MW/sec of kinetic energy about the same amount of kinetic energy as approximately 100 pounds of high explosive (although due to the nature of this explosion, caused by rapid overheating rather than chemical combustion, in order to get the same effect from chemical explosives would take considerably more HE). Material in the core was observed to be moving at 7300 m/sec before it came into contact with the pressure vessel, and flung the largest intact piece of the pressure vessel (a 0.9 sq. m, 67 kg piece of the pressure vessel) 229 m away from the test location. There were some issues with instrumentation in this test, namely with the pressure transducers used to measure the shock wave. All of these instruments but two (placed 100 ft away) didn’t record the pressure wave, but rather an electromagnetic signal at the time of peak power (those two recorded a 3-5 psi overpressure).

KIWI-TNT residue
KIWI-TNT remains, image via LANL

Radioactive Release during Rover Testing Prequel: Radiation is Complicated

Radiation is a major source of fear for many people, and is the source of a huge amount of confusion in the general population. To be completely honest, when I look into the nitty gritty of health physics (the study of radiation’s effects on living tissue), I spend a lot of time re-reading most of the documents because it is easy to get confused by the terms that are used. To make matters worse, especially for the Rover documentation, everything is in the old, outdated measures of radioactivity. Sorry, SI users out there, all the AEC and NASA documentation uses Ci, rad, and rem, and converting all of it would be a major headache. If someone would like to volunteer helping me convert everything to common sense units, please contact me, I’d love the help! However, the natural environment is radioactive, and the Sun emits a prodigious amount of radiation, only some of which is absorbed by the atmosphere. Indeed, there is evidence that the human body REQUIRES a certain amount of radiation to maintain health, based on a number of studies done in the Soviet Union using completely non-radioactive, specially prepared caves and diets.

Exactly how much is healthy and not is a matter of intense debate, and not much study, though, and three main competing theories have arisen. The first, the linear-no-threshold model, is the law of the land, and states that there’s a maximum amount of radiation that is allowable to a person over the course of a year, no matter if it’s in one incident (which usually is a bad thing), or evenly spaced throughout the whole year. Each rad (or gray, we’ll get to that below) of radiation increases a person’s chance of getting cancer by a certain percentage in a linear fashion, and so effectively the LNT model (as it’s known) determines a minimum acceptable increase in the chance of a person getting cancer in a given timeframe (usually quarters and years). This doesn’t take into account the human body’s natural repair mechanisms, though, which can replace damaged cells (no matter how they’re damaged), which leads most health physicists to see issues with the model, even as they work within the model for their professions.

The second model is known as the linear-threshold model, which states that low level radiation (under the threshold of the body’s repair mechanisms) doesn’t make sense to count toward the likelihood of getting cancer. After all, if you replace your Formica counter top in your kitchen with a granite one, the natural radioactivity in the granite is going to expose you to more radiation, but there’s no difference in the likelihood that you’re going to get cancer from the change. Ramsar, Iran (which has the highest natural background radiation of any inhabited place on Earth) doesn’t have higher cancer rates, in fact they’re slightly lower, so why not set the threshold to where the normal human body’s repair mechanisms can control any damage, and THEN start using the linear model of increase in likelihood of cancer?

The third model, hormesis, takes this one step further. In a number of cases, such as Ramsar, and an apartment building in Taiwan which was built with steel contaminated with radioactive cobalt (causing the residents to be exposed to a MUCH higher than average chronic, or over time, dose of gamma radiation), people have not only been exposed to higher than typical doses of radiation, but had lower cancer rates when other known carcinogenic factors were accounted for. This is evidence that having an increased exposure to radiation may in fact stimulate the immune system and make a person more healthy, and reduce the chance of that person getting cancer! A number of places in the world actually use radioactive sources as places of healing, including radium springs in Japan, Europe, and the US, and the black monazite sands in Brazil. There has been very little research done in this area, since the standard model of radiation exposure says that this is effectively giving someone a much higher risk for cancer, though.

I am not a health physicist. It has become something of a hobby for me in the last year, but this is a field that is far more complex than astronuclear engineering. As such, I’m not going to weigh in on the debate as to which of these three theories is right, and would appreciate it if the comments section on the blog didn’t become a health physics flame war. Talking to friends of mine that ARE health physicists (and whom I consult when this subject comes up), I tend to lean somewhere between the linear threshold and hormesis theories of radiation exposure, but as I noted before, LNT is the law of the land, and so that’s what this blog is going to mostly work within.

Radiation (in the context of nuclear power, especially) starts with the emission of either a particle or ray from a radioisotope, or unstable nucleus of an atom. This is measured with the Curie (Cu) which is a measure of how much radioactivity IN GENERAL is released, or 3.7X10^10 emissions (either alpha, beta, neutron, or gamma) per second. SI uses the term Becquerels (Bq), which is simple: one decay = 1 Bq. So 1 Cu = 3.7X10^10 Bq. Because it’s so small, megaBequerels (Mbq) is often used, because unless you’re looking at highly sensitive laboratory experiments, even a dozen Bq is effectively nothing.

Each different type of radiation affects both materials and biological systems differently, though, so there’s another unit used to describe energy produced by radiation being deposited onto a material, the absorbed dose: this is the rad, and SI unit is the gray (Gy). The rad is defined as 100 ergs of energy deposited in one gram of material, and the gray is defined as 1 joule of radiation absorbed by one kilogram of matter. This means that 1 rad = 0.01 Gy. This is mostly seen for inert materials, such as reactor components, shielding materials, etc. If it’s being used for living tissue, that’s generally a VERY bad sign, since it’s pretty much only used that way in the case of a nuclear explosion or major reactor accident. It is used in the case of an acute – or sudden – dose of radiation, but not for longer term exposures.

This is because there’s many things that go into how bad a particular radiation dose is: if you’ve got a gamma beam that goes through your hand, for instance, it’s far less damaging than if it goes through your brain, or your stomach. This is where the final measurement comes into play: in NASA and AEC documentation, they use the term rem (or radiation equivalent man), but in SI units it’s known as the Sievert. This is the dose equivalent, or normalizing all the different radiation types’ effects on the various tissues of the body, by applying a quality factor to each type of radiation for each part of a human body that is exposed to that type of radiation. If you’ve ever wondered what health physicists do, it’s all the hidden work that goes on when that quality factor is applied.

The upshot of all of this is the way that radiation dose is assessed. There are a number of variables that were assessed at the time (and currently are assessed, with this as an effective starting point for ground testing, which has a minuscule but needing to be assessed consideration as far as release of radioactivity to the general public). The exposure was broadly divided into three types of exposure: full-body (5 rem/yr for an occupational worker, 0.5 rem/yr for the public); skin, bone, and thyroid exposure (30 rem/yr occupational, 3 rem/yr for the public); and other organs (15 rem/yr occupational, 1.5 rem/yr for the public). In 1971, the guidelines for the public were changed to 0.5 rem/yr full body and 1.5 rem/yr for the general population, but as has been noted (including in the NRDS Effluent Final Report) this was more an administrative convenience rather than biomedical need.

Capture
1974 Occupational Radiological Release Standards, Image via EPA

Additional considerations were made for discrete fuel element particles ejected from the core – a less than one in ten thousand chance that a person would come in contact with one, and a number of factors were considered in determining this probability. The biggest concern is skin contact can result in a lesion, at an exposure of above 750 rads (this is an energy deposition measure, not an expressly medical one, because it is only one type of tissue that is being assessed).

Finally, and perhaps the most complex to address, is the aerosolized effluent from the exhaust plume, which could be both gaseous fission products (which were not captured by the clad materials used) and from small enough particles to float through the atmosphere for a longer duration – and possibly be able to be inhaled. The relevant limits of radiation exposure for these tests for off-site populations were 170 mrem/yr whole body gamma dose, and a thyroid exposure dose of 500 mrem/yr. The highest full body dose recorded in the program was in 1966, of 20 mrem, and the highest thyroid dose recorded was from 1965 of 72 mrem.

The Health and Environmental Impact of Nuclear Propulsion Testing Development at Jackass Flats

So how bad were these tests about releasing radioactive material, exactly? Considering the sparsely populated area few people – if any – that weren’t directly associated with the program received any dose of radiation from aerosolized (inhalable, fine particulate) radioactive material. By the regulations of the day, no dose of greater than 15% of the allowable AEC/FRC (Federal Radiation Council, an early federal health physics advisory board) dose for the general public was ever estimated or recorded. The actual release of fission products in the atmosphere (with the exception of Cadmium-115) was never more than 10%, and often less than 1% (115Cd release was 50%). The vast majority of these fission products are very short lived, decaying in minutes or days, so there was not much – if any – change for migration of fallout (fission products bound to atmospheric dust that then fell along the exhaust plume of the engine) off the test site. According to a 1995 study by the Department of Energy, the total radiation release from all Rover and Tory-II nuclear propulsion tests was approximately 843,000 Curies. To put this in perspective, a nuclear explosive produces 30,300,000 Curies per kiloton (depending on the size and efficiency of the explosive), so the total radiation release was the equivalent of a 30 ton TNT equivalent explosion.

Test Release Summary NRDS
Summary of Radiological Release, image via DOE

This release came from either migration of the fission products through the metal clad and into the hydrogen coolant, or due to cladding or fuel element failure, which resulted in the hot hydrogen aggressively attacking the graphite fuel elements and carbide fuel particles.

The amount of fission product released is highly dependent on the temperature and power level the reactors were operated at, the duration of the test, how quickly the reactors were brought to full power, and a number of other factors. The actual sampling of the reactor effluent occurred three ways: sampling by aircraft fitted with special sensors for both radiation and particulate matter, the “Elephant gun” effluent sampler placed in the exhaust stream of the engine, and by postmortem chemical analysis of the fuel elements to determine fuel burnup, migration, and fission product inventory. One thing to note is that for the KIWI tests, effluent release was not nearly as well characterized as for the later Phoebus, NRX, Pewee, and Nuclear Furnace tests, so the data for these tests is not only more accurate, but far more complete as well.

cy67 offsite dose map
Offsite Dose Map, 1967 (a year with higher-than-average release, and the first to employ better sampling techniques) Image via EPA

Two sets of aircraft data were collected: one (by LASL/WANL) was from fixed heights and transects in the six miles surrounding the effluent plume, collecting particulate effluent which would be used (combined with known release rates of 115Cd and post-mortem analysis of the reactor) to determine the total fission product inventory release at those altitudes and vectors, and was discontinued in 1967; the second (NERC) method used a fixed coordinate system to measure cloud size and density, utilizing a mass particulate sampler, charcoal bed, cryogenic sampler, external radiation sensor, and other equipment, but due to the fact that these samples were taken more than ten miles from the reactor tests, it’s quite likely that more of the fission products had either decayed or come down to the ground as fallout, so depletion of much of the fission product inventory could easily have occurred by the time the cloud reached the plane’s locations. This technique was used after 1967.

The next sampling method also came online in 1967 – the Elephant Gun. This was a probe that was stuck directly into the hot hydrogen coming out of the nozzle, and collected several moles of the hot hydrogen from the exhaust stream at several points throughout the test, which were then stored in sampling tanks. Combined with hydrogen temperature and pressure data, acid leaching analysis of fission products, and gas sample data, this provided a more close-to-hand estimate of the fission product release, as well as getting a better view of the gaseous fission products that were released by the engine.

EMAD wikimedia
Engine Maintenance and Disassembly Building at NRDC under construction, image via Wikimedia Commons

Finally, after testing and cool-down, each engine was put through a rigorous post-mortem inspection. Here, the amount of reactivity lost compared to the amount of uranium present, power levels and test duration, and chemical and radiological analysis were used to determine which fission products were present (and in which ratios) compared to what SHOULD have been present. This technique enhanced understanding of reactor behavior, neutronic profile, and actual power achieved during the test as well as the radiological release in the exhaust stream.

Radioactive release from these engine tests varied widely, as can be seen in the table above, however the total amount released by the “dirtiest” of the reactor tests, the Phoebus 1B second test, was only 240,000 Curies, and the majority of the tests released less than 2000 Curies. Another thing that varied widely was HOW the radiation was released. The immediate area (within a few meters) of the reactor would be exposed to radiation during operation, in the form of both neutron and gamma radiation. The exhaust plume would contain not only the hydrogen propellant (which wasn’t in the reactor for long enough to accumulate additional neutrons and turn into deuterium, much less tritium, in any meaningful quantities), but the gaseous fission products (most of which the human body isn’t able to absorb, such as 135Xe) and – if fuel element erosion or breakage occurred – a certain quantity of particles that may either have become irradiated or contain burned or unburned fission fuel.

Isotope Release Distribution 25 mi arc
Image via EPA

These particles, and the cloud of effluent created by the propellant stream during the test, were the primary concern for both humans and the environment from these tests. The reason for this is that the radiation is able to spread much further this way (once emitted, and all other things being equal, radiation goes in a straight line), and most especially it can be absorbed by the body, through inhalation or ingestion, and some of these elements are not just radioactive, but chemically toxic as well. As an additional complication, while alpha and beta radiation are generally not a problem for the human body (your skin stops both particles easily), when they’re IN the human body it’s a whole different ballgame. This is especially true of the thyroid, which is more sensitive than most to radiation, and soaks up iodine (131I is a fairly active radioisotope) like nobody’s business. This is why, after a major nuclear accident (or a theoretical nuclear strike), iodine tablets, containing a radio-inert isotope, are distributed: once the thyroid is full, the excess radioactive iodine passes through the body since nothing else in the body can take it up and store it.

There are quite a few factors that go into how far this particulate will spread, including particle mass, temperature, velocity, altitude, wind (at various altitudes), moisture content of the air (particles could be absorbed into water droplets), plume height, and a host of other factors. The NRDS Effluent Program Final Report goes into great depth on the modeling used, and the data collection methods used to collect data to refine these estimates.

Another thing to consider in the context of Rover in particular is that open-air testing of nuclear weapons was taking place in the area immediately surrounding the Rover tests, which released FAR more fallout (by dozens of orders of magnitude), so it contributed a very minor amount to the amount of radionucleides released at the time.

The offsite radiation monitoring program, which included sampling of milk from cows to estimate thyroid exposure, collected data through 1972, and all exposures measured were well below the exposure limits set on the program.

Since we looked at the KIWI-TNT test earlier, let’s look at the environmental effects of this particular test. After all, a nuclear rocket blowing up has to be the most harmful test, right? Surprisingly, ten other tests released more radioactivity than KIWI-TNT. The discrete particles didn’t travel more than 600 feet from the explosion. The effluent cloud was recorded from 4000 feet to 50 miles downwind of the test site, and aircraft monitoring the cloud were able to track it until it went out over the Pacific ocean (although at that point, it was far less radioactive). By the time the cloud had moved 16,000 feet from the test site, the highest whole body dose from the cloud measured was 1.27X10^-3 rad (at station 16-210), and the same station registered an inhalation thyroid dose of 4.55X10^-3 rads. This shows that even the worst credible accident possible with a NERVA-type reactor has only a negligible environmental and biological impact due to either the radiation released or the explosion of the reactor itself, further attesting to the safety of this engine type.

KIWI-TNT Particle Map
Map of discrete particle distribution, image via LANL

If you’re curious about more in-depth information about the radiological and environmental effects of the KIWI-TNT tests, I’ve linked the (incredibly detailed) reports on the experiment at the end of this post.

KIWI-TNT Rad readings
Radiological distribution from particle monitors, image via LANL

The Results of the Rover Test Program

Throughout the Rover testing program, the fuel elements were the source of most of the non-H2 related issues. While other issues, such as instrumentation, were also encountered, the main headache was the fuel elements themselves.

A lot of the problems came down to the mechanical and chemical properties of the graphite fuel matrix. Graphite is easily attacked by the hot H2, leading to massive fuel element erosion, and a number of solutions were experimented with throughout the test series. With the exception of the KIWI-A reactor (which used unclad fuel plates, and was heavily affected by the propellant), each of the reactors featured FEs that were clad to a greater or lesser extent, using a variety of methods and materials. Often, niobium carbide (NbC) was the favored clad material, but other options, such as tungsten, exist.

CVD coating
CVD Coating device, image courtesy LANL

Chemical vapor deposition was an early option, but unfortunately it was not feasible to consistently and securely coat the interior of the propellant tubes, and differential thermal expansion was a major challenge. As the fuel elements heated, they expanded, but at a different rate than the coating did. This led to cracking, and in some cases, flaking off, of the clad material, leading to the graphite being exposed to the propellant and being eroded away. Machined inserts were a more reliable clad form, but required more complexity to install.

The exterior of the fuel elements originally wasn’t clad, but as time went on it was obvious that this would need to be addressed as well. Some propellant would leak between the prisms, leading to erosion of the outside of the fuel elements. This changed the fission geometry of the reactor, led to fission product and fuel release through erosion, and weakened the already somewhat fragile fuel elements. Usually, though, vapor deposition of NbC was sufficient to eliminate this problem

Fortunately, these issues are exactly the sort of thing that CFEET and NTREES are able to test, and these systems are far more economical to operate than a hot-fired NTR is. It is likely that by the time a hot-fire test is being conducted, the fuel elements will be completely chemically and thermally characterized, so these issues shouldn’t arise.

The other issue with the fuel elements was mechanical failure due to a number of problems. The pressure across the system changes dramatically, which leads to differential stress along the length of the fuel elements. The original, minimally-supported fuel elements, would often undergo transverse cracking, leading to blockage of propellant and erosion. In a number of cases, after the fuel element broke this way, the hot end of the fuel element would be ejected from the core.

Tie Tube
Rover tie tube image courtesy NASA

This led to the development of a structure that is still found in many NTR designs today: the tie tube. This is a hexagonal prism, the same size as the fuel elements, which supports the adjacent fuel elements along their length. In addition to being a means of support, these are also a major source of neutron moderation, due to the fact that they’re cooled by hydrogen propellant from the regeneratively cooled nozzle. The hydrogen would make two passes through the tie tube, one in each direction, before being injected into the reactor’s cold end to be fed through the fuel elements.

The tie tubes didn’t eliminate all of the mechanical issues that the fuel element faced. Indeed, even in the NF-1 test, extensive fuel element failure was observed, although none of the fuel elements were ejected from the core. However, new types of fuel elements were being tested (uranium carbide-zirconium carbide carbon composite, and (U,Zr)C carbide), which offered better mechanical properties as well as higher thermal tolerances.

Current NTR designs still usually incorporate tie tubes, especially because the low-enriched uranium that is the main notable difference in NASA’s latest design requires a much more moderated neutron spectrum than a HEU reactor does. However, the ability to support the fuel element mechanically along its entire length (rather than just at the cold end, as was common in NERVA designs) does also increase the mechanical stability of the reactor, and helps maintain the integrity of the fuel elements.

The KIWI-B and Phoebus reactors were successful enough designs to use as starting points for the NERVA engines. NERVA is an acronym for the Nuclear Energy for Rocket Vehicle Applications, and took place in two parts: NERVA-1, or NERVA-NRX, developed the KIWI-B4D reactor into a more flight-prototypic design, including balance of plant optimization, enhanced documentation of the workings of the reactor, and coolant flow studies. The second group of engines, NERVA-2, were based on the Phoebus 2 type of reactor from Rover, and ended up finally being developed into the NERVA-XE, which was meant to be the engine that would power the manned mission to Mars. The NERVA-XE PRIME test was of the engine in flight configuration, with all the turbopumps, coolant tanks, instrumentation, and even the reactor’s orientation (nozzle down, instead of up) were all the way that it would have been configured during the mission.

The first ground experimental nuclear rocket engine (XE) assembl
NERVA XE-PRIME pre-fire installation and verification, image via Westinghouse Engineer (1974)

The XE-PRIME test series lasted for nine months, from December 1968 to September 1969, and involved 24 startups and shutdowns of the reactor. Using a 1140 MW reactor, operating at 2272 K exhaust temperature, and produced 247 kN of thrust at 710 seconds of specific impulse. This included using new startup techniques from cold-start conditions, verification of reactor control systems – including using different subsystems to be able to manipulate the power and operating temperature of the reactor – and demonstrated that the NERVA program had successfully produced a flight-ready nuclear thermal rocket.

Ending an Era: Post-Flight Design Testing

Toward the end of the Rover program, the engine design itself had been largely finalized, with the NERVA XE-Prime test demonstrating an engine tested in flight configuration (with all the relevant support hardware in place, and the nozzle pointing down), however, some challenges remained for the fuel elements themselves. In order to have a more cost-effective testing program for fuel elements, two new reactors were constructed.

NERVAPewee2, AEC 1971
PEWEE Test Stand, image courtesy LANL

The first, Pewee, was a smaller (75 klbf, the same size as NASA’s new NTR) nuclear rocket engine, which was able to have the core replaced for multiple rounds of testing, but was only used once before the cancellation of the program – but not before achieving the highest specific impulse of any of the Rover engines. This reactor was never tested outside of a breadboard configuration, because it was never meant to be used in flight. Instead, it was a cost-saving measure for NASA and the AEC: due to its’ smaller size, it was much cheaper to built, and due to its’ lower propellant flow rate, it was also much easier to test. This meant that experimental fuel elements that had undergone thermal and irradiation testing would be able to be tested in a fission-powered, full flow environment at lower cost.

Transverse view, Finseth
NF-1 Transverse view, image courtesy NASA

The second was the Nuclear Furnace, which mimicked the neutronic environment and propellant flow rates of the larger NTRs, but was not configured as an engine. This reactor also was the first to incorporate an effluent scrubber, capturing the majority of the non-gaseous fission products and significantly reducing the radiological release into the environment. It also achieved the highest operating temperatures of any of the reactors tested in Nevada, meaning that the thermal stresses on the fuel elements would be higher than would be experienced in a full-power burn of an actual NTR. Again, this was designed to be able to be repeatedly reused in order to maximize the financial benefit of the reactor’s construction, but was only used once before the cancellation of the program. The fuel elements were tested in separate cans, and none of them were the graphite composite fuel form: instead, CERMET (then known as composite) and carbide fuel elements, which had been under development but not extensively used in Rover or NERVA reactors, were tested. This system also used an effluent cleanup system, but that’s something that we’re going to look at more in depth on the next post, as it remains a theoretically possible method of doing hot-fire testing for a modern NTR.

A-type reactor
NRX A reactor, which PAX was based on, image courtesy NASA

Westinghouse ANL also proposed a design based on the NERVA XE, called the PAX reactor, which would be designed to have its’ core replaced, but this never left the drawing boards. Again, the focus had shifted toward lower cost, more easily maintained experimental NTR test stands, although this one was much closer to flight configuration. This would have been very useful, because not only would the fuel be subjected to a very similar radiological and chemical environment, but the mechanical linkages, hydrogen flow paths, and resultant harmonic and gas-dynamic issues would have been able to be evaluated in a near-prototypic environment. However, this reactor was never tested.

Conclusion

As we’ve seen, hot-fire testing was something that the engineers involved in the Rover and NERVA programs were exceptionally concerned about. Yes, there were radiological releases into the environment that are well above and beyond what would be considered today, but when compared to the releases from the open-air nuclear weapons tests that were occurring in the immediate vicinity, they were miniscule.

Today, though, these releases would be unacceptable. So, in the next blog post we’re going to look at the options, and restrictions, for a modern testing facility for NTR hot firing, including a look at the proposals over the years and NASA’s current plan for NTR testing. This will include the exhaust filtration system on the Nuclear Furnace, a more complex (but also more effective) filtering system proposed for the SNTP pebblebed reactor (TimberWind), a geological filtration concept called SAFE, and a full exhaust capture and combustion system that could be installed at NASA’s current rocket test facility at Stennis Space Center.

This post is already started, and I hope to have it out in the next few weeks. I look forward to hearing all your feedback, and if there are any more resources on this subject that I’ve missed, please share them in the comments below!

 

References

Los Alamos Pajarito Site

Los Alamos Critical Assemblies Facility, LA-8762-MS, by R. E. Malenfant,
https://www.osti.gov/servlets/purl/6463833

Thirty-Five Years at Pajarito Canyon Site, LA-7121-H, Rev., by Hugh Paxton
https://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00197750.pdf

A History of Critical Experiments at Pajarito Site, LA-9685-H, by R.E. Malenfant, 1983

https://www.atomictraveler.com/LANLPajarito.pdf

Environmental Impacts and Radiological Release Reports

NRDS Nuclear Rocket Effluent Program, 1959-1970; NERC-LV-539-6, by Bernhardt et al, 1974

https://nepis.epa.gov/Exe/ZyPURL.cgi?Dockey=9100F0RX.TXT

Offsite Monitoring Report for NRX-A2; 1965

https://nepis.epa.gov/Exe/ZyNET.exe/9100FFRU.txt?ZyActionD=ZyDocument&Client=EPA&Index=Prior%20to%201976&Docs=&Query=&Time=&EndTime=&SearchMethod=1&TocRestrict=n&Toc=&TocEntry=&QField=&QFieldYear=&QFieldMonth=&QFieldDay=&UseQField=&IntQFieldOp=0&ExtQFieldOp=0&XmlQuery=&File=D%3A%5CZYFILES%5CINDEX%20DATA%5C70THRU75%5CTXT%5C00000011%5C9100FFRU.txt&User=ANONYMOUS&Password=anonymous&SortMethod=h%7C-&MaximumDocuments=1&FuzzyDegree=0&ImageQuality=r75g8/r75g8/x150y150g16/i425&Display=hpfr&DefSeekPage=x&SearchBack=ZyActionL&Back=ZyActionS&BackDesc=Results%20page&MaximumPages=1&ZyEntry=2

Radiation Measurements of the Effluent from the Kiwi-TNT Experiment; LA-3395-MS, by Henderson et al, 1966

https://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00420212.pdf

Environmental Effects of the KIWI-TNT Effluent: A Review and Evaluation; LA-3449, by R.V.Fultyn, 1968

https://fas.org/sgp/othergov/doe/lanl/docs1/la-3449.pdf

Technological Development and Non-Nuclear Testing

A Review of Fuel Element Development for Nuclear Rocket Engines; LA-5931, by J.M. Taub

http://permalink.lanl.gov/object/tr?what=info:lanl-repo/lareport/LA-05931

Hot Fire Testing

Rover Nuclear Rocket Engine Program: Overview of Rover Engine Tests; N92-15117, by J.L. Finseth, 1992

https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920005899.pdf

Nuclear Furnace 1 Test Report; LA-5189-MS, W.L. Kirk, 1973

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

KIWI-TNT Testing

KIWI-Transient Nuclear Test; LA-3325-MS, 1965

http://permalink.lanl.gov/object/tr?what=info:lanl-repo/lareport/LA-03325-MS

Kiwi-TNT Explosion; LA-3551, by Roy Reider, 1965

https://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00424689.pdf

An Analysis of the KIWI-TNT Experiment with MARS Code; Journal of Nuclear Science and Technology, Hirakawa et al. 1968

https://doi.org/10.1080/18811248.1970.9734632

Miscellaneous Resources

Safety Neutronics for Rover Reactors; LA-3558-MS, Los Alamos Scientific Laboratory, 1965

https://www.osti.gov/servlets/purl/4227707

The Behavior of Fission Products During Nuclear Rocket Reactor Tests; LA-UR-90-3544, by Bokor et al, 1996

https://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00275758.pdf