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Timber Wind: America’s Return to Nuclear Thermal Rockets

 Hello, and welcome to Beyond NERVA! Today, we’re continuing to look at the pebble bed nuclear thermal rocket (check out the most recent blog post on the origins of the PBR nuclear thermal rocket here)!

Sorry it took so long to get this out… between the huge amount of time it took just to find the minimal references I was able to get my hands on, the ongoing COVID pandemic, and several IRL challenges, this took me far longer than I wanted – but now it’s here!

Today is special because it is covering one of the cult classics of astronuclear engineering, Project Timber Wind, part of the Strategic Defense Initiative (better known colloquially as “Star Wars”). This was the first time since Project Rover that the US put significant resources into developing a nuclear thermal rocket (NTR). For a number of reasons, Timber Wind has a legendary status among people familiar with NTRs, but isn’t well reported on, and a lot of confusion has built up around the project. It’s also interesting in that it was an incredibly (and according to the US Office of the Inspector General, overly) classified program, which means that there’s still a lot we don’t know about this program 30 years later. However, as one of the most requested topics I hear about, I’m looking forward to sharing what I’ve discovered with you… and honestly I’m kinda blown away with this concept.

Timber Wind was an effort to build a second stage for a booster rocket, to replace the second (and sometimes third) stage of anything from an MX ballistic missile to an Atlas or Delta booster. This could be used for a couple of different purposes: it could be used similarly to an advanced upper stage, increasing the payload capacity of the rocket and the range of orbits that the payload could be placed in; alternatively it could be used to accelerate a kinetic kill vehicle (basically a self-guided orbital bullet) to intercept an incoming enemy intercontinental ballistic missile before it deploys its warheads. Both options were explored, with much of the initial funding coming from the second concept, before the kill vehicle concept was dropped and the slightly more traditional upper stage took precedence.

Initially, I planned on covering both Timber Wind and the Space Nuclear Thermal Propulsion program (which it morphed into) in a single post, but the mission requirements, and even architectures, were too different to incorporate into a single blog post. So, this will end up being a two-parter, with this post focusing on the three early mission proposals for the Department of Defense (DOD) and Strategic Defense Initiative Organization (SDIO): a second stage of an ICBM to launch an anti-booster kinetic kill vehicle, an orbital transfer vehicle (basically a fancy, restartable second stage for a booster), and a multi-megawatt orbital nuclear power plant. The next post will cover when the program became more open, testing became more prevalent, and grander plans were laid out – and some key restrictions on operating parameters eliminated the first and third missions on this list.

Ah, Nomenclature, Let’s Deal with That

So, there’s a couple things to get out of the way before we begin.

The first is the name. If you do a Google/Yandex/etc search for “Timber Wind,” you aren’t going to find much compared to “Timberwind,” but from what I’ve seen in official reporting it should be the other way around. The official name of this program is Project Timber Wind (two words), which according to the information I’ve been able to find is not unusual. The anecdotal evidence I have (and if you know more, please feel free to leave a comment below!) is that for programs classified Top Secret: Special Access (as this was) had a name assigned based on picking two random words via computer, whereas other Top Secret (or Q, or equivalent) programs didn’t necessarily follow this protocol.

However, when I look for information about this program, I constantly see “Timberwind.” not the original “Timber Wind.” I don’t know when this shift happened – it didn’t ever happen with rare exceptions in official documentation, even in the post-cancellation reporting, but somehow public reporting always uses the single word variation. I kinda attribute it to reading typewritten reports when the reader is used to digitally written documents as personal head-canon, but that’s all that explanation is – my guess which makes sense to me.

So there’s a disconnect between what most easily accessible sources use (single word), and the official reporting (two words). I’m going to use the original, because the only reason I’ve gotten as far as I have by being weird about minor details in esoteric reports, so I’m not planning on stopping now (I will tag the single word in the blog, just so people can find this, but that’s as far as I’m going)!

The second is in nuclear reactor geometry definitions.

Having discrete, generally small fuel elements generally falls into two categories: particle beds and pebble beds. Particles are small, pebbles are big, and where the line falls seems to be fuzzy. In modern contexts, the line seems to fall around the 1 cm diameter mark, although finding a formal definition has so far eluded me. However, pebble beds are also a more colloquial term than particle beds in use: a particle bed is a type of pebble bed in common use, but not vice versa.

In this context, both the RBR and Timber Wind are both particle bed reactors, and I’ll call them such, but if a source calls the reactor a pebble bed (which many do), I may end up slipping up and using the term.

OK, nomenclature lesson done. Back to the reactor!

Project Timber Wind: Back to the Future

For those in the know, Timber Wind is legendary. This was the first time after Project Rover that the US put its economic and industrial might behind an NTR program. While there had been programs in nuclear electric propulsion (poorly funded, admittedly, and mostly carried through creative accounting in NASA and the DOE), nuclear thermal propulsion had taken a back seat since 1972, when Project Rover’s continued funding was canceled, along with plans for a crewed Mars mission, a crewed base on the Moon, and a whole lot of other dreams that the Apollo generation grew up on.

There was another difference, as well. Timber Wind wasn’t a NASA program. Despite all the conspiracy theories, the assumptions based on the number of astronauts with military service records, and the number of classified government payloads that NASA has handled, it remains a civilian organization, with the goal of peacefully exploring the solar system in an open and transparent manner. The Department of Defense, on the other hand, is a much more secretive organization, and as such many of the design details of this reactor were more highly classified than is typical in astronuclear engineering as they deal with military systems. However, in recent years, many details have become available on this system, which we’ll cover in brief today – and I will be linking not only my direct sources but all the other information I’ve found below.

Also unlike NTR designs since the earliest days of Rover, Timber Wind was meant to act as a rocket stage during booster flight. Most NTR designs are in-space only: the reactor is launched into a stable, “nuclear-safe” (i.e. a long-term stable orbit with minimal collision risk with other satellites and debris) orbit, then after being mated to the spacecraft is brought to criticality and used for in-space orbital transfers, interplanetary trajectories, and the like. (Interesting aside, this program’s successor seems to be the first time that now-common term was used in American literature on the subject.)

Timber Wind was meant to support the Strategic Defense Initiative (SDI), popularly known as Star Wars. Started in 1983, this extensive program was meant to provide a ballistic missile shield, among other things, for the US, and was given a high priority and funding level for a number of programs. One of these programs, the Boost Phase Intercept vehicle, meant to destroy an intercontinental ballistic missile during the boost phase of the vehicle using a kinetic impactor which would be launched either from the ground or be pre-deployed in space. A kinetic kill vehicle is basically a set of reaction control thrusters designed to guide a small autonomous spacecraft into its target at high velocity and destroy it. They are typically small, very nimble, and limited only by the sensors and autonomous guidance software available for them.

In order to do this, the NTR would need to function as the second stage of a rocket, meaning that while the engine would be fired only after it had entered the lower reaches of space or the upper reaches of the atmosphere (minimizing the radiation risk from the launch), it would still very much be in a sub-orbital flight path at the time, and would have much higher thrust-to-weight ratio requirements as a result.

The engine that was selected was based on a design by James Powell at Brookhaven National Laboratory (BNL) in the late 1970s. He presented the design to Grumman in 1982, and from there it came to the attention of the Strategic Defense Initiative Organization (SDIO), the organization responsible for all SDI activities.

Haslett 1994

SDIO proceeded to break the program up into three phases:

  • Phase I (November 1987 to September 1989): verify that the pebblebed reactor concept would meet the requirements of the upper stage of the Boost Phase Intercept vehicle, including the Preliminary Design Review of both the stage and the whole vehicle (an MX first stage, with the PBR second stage /exceeding Earth escape velocity after being ignited outside the atmosphere)
  • Phase II (September 1989-October 1991 under SDIO, October 1991-January 1994 when it was canceled under the US Air Force, scheduled completion 1999): Perform all tests to support the ground test of a full PBR NTR system in preparation for a flight test, including fuel testing, final design of the reactor, design and construction of testing facilities, etc. Phase II would be completed with the successful ground hot fire test of the PBR NTR, however the program was canceled before the ground test could be conducted.
    • Once the program was transferred to the US Air Force (USAF), the mission envelope expanded from an impactor’s upper stage to a more flexible, on-orbit multi-mission purpose, requiring a design optimization redesign. This is also when NASA became involved in the program.
    • Another change was that the program name shifted from Timber Wind to the Space Nuclear Thermal Propulsion program (SNTP), reflecting both the change in management as well as the change in the mission design requirements.
  • Phase III (never conducted, planned for 2000): Flight test of the SNTP upper stage using an Atlas II launch vehicle to place the NTR into a nuclear-safe orbit. Once on orbit, a number of on-orbit tests would be conducted on the engine, but those were not specified to any degree due to the relatively early cancellation of the program.

While the program offered promise, many factors combined to ensure the program would not be completed. First, the hot fire testing facilities required (two were proposed, one at San Tan and one at the National Nuclear Security Site) would be incredibly expensive to construct, second the Space Exploration Initiative was heavily criticized for reasons of cost (a common problem with early 90’s programs), and third the Clinton administration cut many nuclear research programs in all US federal departments in a very short period of time (the Integral Fast Reactor at Argonne National Laboratory was another program to be cut at about the same time).

The program would be transferred into a combined USAF and NASA program in 1991, and end in 1994 under those auspices, with many successful hurdles overcome, and it remains an attractive design, one which has become a benchmark for pebble bed nuclear thermal rockets, and a favorite of the astronuclear community to speculate what would be possible with this incredible engine.

To understand why it was so attractive, we need to go back to the beginning, in the late 1970s at Brookhaven National Laboratory in the aftermath of the Rotating Fluidized Bed Reactor (RBR, covered in our last post here).

The Beginning of Timber Wind

When we last left particle bed NTRs, the Rotating Fluidized Bed Reactor program had made a lot of progress on many of the fundamental challenges with the concept of a particle bed reactor, but still faced many challenges. However, the team, including Dr. John Powell, were still very enthusiastic about the promise it offered – and conscious of the limitations of the system.

Dr. Powell continued to search for funding for a particle bed reactor (PBR) NTR program, and interest in NTR was growing again in both industry and government circles, but there were no major programs and funding was scarce. In 1982, eight years after the conclusion of the RBR, he had a meeting with executives in the Grumman Corporation, where he made a pitch for the PBR NTR concept. They were interested in the promise of higher specific impulse and greater thrust to weight ratios compared to what had become the legacy NERVA architecture, but there wasn’t really a currently funded niche for the project. However, they remained interested enough to start building a team of contractors willing to work on the concept, in case the US government revived its NTR program. The companies included other major aerospace companies (such as Garrett Corp and Aerojet) and nuclear contractors (such as Babcock and Wilcox), as well as subcontractors for many components.

At the same time, they tried to sell the concept of astronuclear PBR designs to potentially interested organizations: a 1985 briefing to the Air Force Space Division on the possibility of using the PBR as a boost phase interceptor was an early, but major, presentation that would end up being a major part of the initial Timber Wind architecture, and the next year an Air Force Astronautics Laboratory issues a design study contract for a PBR-based Orbital Transfer Vehicle (OTV, a kind of advanced upper stage for an already-existing booster). While neither of these contracts was big enough to do a complete development program, they WERE enough money to continue to advance the design of the PBR, which by now was showing two distinct parts: the boost phase interceptor, and the OTV. There was also a brief flirtation with using the PBR architecture from Timber Wind as a nuclear electric power source, which we’ll examine as well, but this was never particularly well focused on or funded, so remains a footnote in the program.

Reactor Geometry

From Atomic Power in Space, INL 2015

Timber Wind was a static particle bed reactor, in the general form of a cylinder 50 cm long by 50 cm in diameter, using 19 fuel elements to heat the propellant in a folded flow path. Each fuel element was roughly cylindrical with a 6.4 cm diameter, consisting of a cold frit (a perforated cylinder) made of stainless steel and a hot frit made out of zirconium carbide (ZrC, although rhenium – Rh – clad would also meet thermal and reactivity requirements) coated carbon-carbon composite, which held a total of 125 kg (15 liters) of 500 micron diameter spheres of uranium/zirconium carbide fueled fuel particles which were clad in two layers of different carbon compositions followed by ZrC cladding. These would be held in place through mechanical means, rather than centrifugal force like in the RBR, reducing the mass of the system at the (quite significant materially) cost of developing a hot frit to mechanically contain the fuel. This is something we’ll cover more in depth in the next post.

From Atomic Power in Space, INL 2015

The propellant would then pass into a central, truncated cone central void, becoming wider from the spacecraft to the nozzle end. This is called orificing. An interesting challenge with nuclear reactors is the fact that the distribution of energy generation changes based on location within the reactor, called radial/axial power peaking (something that occurs in individual fuel elements both in isolation and in terms of their location in a core as well, part of why refueling a nuclear reactor is an incredibly complex process), and in this case it was dealt with in a number of ways, but one of the primary ones was individually changing the orificing of each fuel element to accommodate the power generation and propellant flow rate of each fuel element.

Along these lines, another advantage of this type of core is the ability to precisely control the amount of fissile fuel in each fuel element along the length of the reactor, and along the radius of the fuel element. Since the fuel particles are so small, and the manufacturing of each would be a small-batch process (even fueling a hundred of these things would only take 1500 liters of volume, with the fissile component of that volume being a small percentage of that), a variety of fuel loading options were inherently available, and adjustments to power distribution were reasonably easy to achieve from reactor to reactor. This homogenizes power distribution in some reactors, and increases local power in other, more specialized reactors (like some types of NTRs), but here an even power distribution along the length of the fuel element is desired. This power leveling is done in virtually every fuel element in every reactor, but is a difficult and complex process with large fuel elements due to the need to change how much U is in each portion of the fuel elements. With a particle bed reactor, on the other hand, the U content doesn’t need to vary inside each individual fuel paritcles, and both fueled and unfueled particles can be added in specific regions of the fuel element to achieve the desired power balance within the element. There was actually a region of unfueled particles on the last cm of the particle bed in each fuel element to maximize the efficiency of power distribution into the propellant, and the level of enrichment for the 235U fuel was varied from 70% to 93.5% throughout the fueled portions. This resulted in an incredibly flat power profile, with a ratio of only 1.01:1 from the peak power density to the average power density.

Since the propellant would pass from the outside of each fuel element to the inside, cooling the reactor was far easier, and lower-mass (or higher efficiency) options for things such as moderator were an option. This is a benefit of what’s called a folded-flow­ propellant path, something that we’ve discussed before in some depth in our post on Dumbo, the first folded flow NTR concept [insert link]. In short, instead of heating the propellant as it passes down the length of the reactor such as in Rover, a folded flow injects the cold propellant laterally into the fuel element, heating it in a very short distance, and ejecting it through a central void in the fuel element. This has the advantage of keeping the vast majority of the reactor very cool, eliminating many of the thermal structural problems that Rover experienced, at the cost of a more complex gasdynamic behavior system. This also allowed for lighter-weight materials to be used in the construction, such as aluminum structural members and pressure vessel, to further reduce the mass of the reactor.

Interestingly, many of these lower-mass options, such as Li7H moderator, were never explored, since the mass of the reactor came in at only about 0.6 tons, a very small number compared to the 10 ton payload, so it just wasn’t seen as a big enough issue to continue working on at that point.

Finally, because of the low (~1 hr) operating time of the reactor, radiation problems were minimized. With a reactor only shielded by propellant, tankage, and the structures of the NTR itself, it’s estimated that the NOTV would subject its payload to a total of 100 Gy of gamma radiation and a neutron fluence of less than 10^14 n/cm^2. Obviously, reducing this for a crewed mission would be necessary, but depending on the robotic mission payload, additional shielding may not be necessary. The residual radiation would also be minimal due to the short burn time, although if the reactor was reused this would grow over time.

In 1987, the estimated cost per unit (not including development and testing was about $4 million, a surprisingly low number, due to the ease of construction, low thermal stresses requiring fewer exotic materials and solutions, and low uranium load requirements.

This design would continue to evolve throughout Timber Wind and into SNTP as mission requirements changed (this description is based on a 1987 paper linked below), and we’ll look at the final design in the next post.

For now, let’s move on to how this reactor would be used.

Nuclear Thermal Kinetic Kill Vehicle

The true break for the project came in the same year: 1987. This is when the SDIO picked the Brookhaven (and now Grumman) concept as their best option for a nuclear-enhanced booster for their proposed ground deployed boost phase interceptor.

I don’t do nuclear weapons coverage, in fact that’s a large part of why I’ve never covered systems like Pluto here, but it is something that I’ve gained some knowledge of through osmosis through interactions with many intelligent and well-educated people on social media and in real life… but this time I’m going to make a slight exception for strategic ballistic missile shield technology, because an NTR powered booster is… extremely rare. I can think of four American proposals that continued to be pursued after the 1950s, one early (apocryphal) Soviet design in the early 1950s, one modern Chinese concept, and that’s it! I get asked about it relatively frequently, and my answer is basically always the same: unless something significant changes, it’s not a great idea, but in certain contexts it may work. I leave it up to the reader to decide if this is a good context. (The list I can think of is the Reactor In-Flight Test, or RIFT, which was the first major casualty of Rover/NERVA cutbacks; Timber Wind; and for private proposals the Liberty Ship nuclear lightbulb booster and the Nuclear Thermal Turbo Rocket single stage to orbit concept).

So, the idea behind boost stage interception is that it targets an intercontinental ballistic missile and destroys the vehicle while it’s still gaining velocity – the earlier the interception that can destroy the vehicle, the better. There were many ideas on how to do this, including high powered lasers, but the simplest idea (in theory, not in execution) was the kinetic impactor: basically a self-guided projectile would hit the very thin fuel or oxidizer tanks of the ICBM, and… boom, no more ICBM. This was especially attractive since, by this time, missiles could carry over a dozen warheads, and this would take care of all of them at once, rather than a terminal phase interceptor, which would have to deal with each warhead individually.

The general idea behind Timber Wind was that a three-stage weapon would be used to deliver a boost-phase kinetic kill vehicle. The original first stage was based on the LGM-118 Peacekeeper (“MX,” or Missile – Experimental) first stage, which had just deployed two years earlier. This solid fueled ICBM first stage normally used a 500,000 lbf (2.2 MN) SR118 solid rocket motor, although it’s not clear if this engine was modified in any way for Timber Wind. The second stage would be the PBNTR Timber Wind stage, which would achieve Earth escape velocity to prevent reactor re-entry, and the third stage was the kinetic kill vehicle (which I have not been able to find information about).

Here’s a recent Lockheed Martin KKV undergoing testing, so you can get an idea of what this “bullet” looks and behaves like: https://www.youtube.com/watch?v=KBMU6l6GsdM

Needless to say, this would be a very interesting launch profile, and one that I have not seen detailed anywhere online. It would also be incredibly challenging to

  1. detect the launch of an ICBM;
  2. counter-launch even as rapid-fire-capable a missile as a Peacekeeper;
  3. provide sufficient guidance to the missile in real-time to guide the entire stack to interception;
  4. go through three staging events (two of which were greater than Earth escape velocity!);
  5. guide the kinetic kill vehicle to the target with sufficient maneuvering capability to intercept the target;
  6. and finally have a reasonably high chance of mission success, which required both the reactor to go flying off into a heliocentric orbit and have the kinetic kill vehicle impact the target booster

all before the second (or third) staging event for the target ICBM (i.e. before warhead deployment).

This presents a number of challenges to the designers: thrust-to-weight ratio is key to a booster stage, something that to this point (and even today) NTRs struggle with – mostly due to shielding requirements for the payload.

There simply isn’t a way to mitigate gamma radiation in particular without high atomic number nuclei to absorb and re-emit these high energy photons enough times that a lighter shielding material can be used to either stop or deflect the great-great-great-great-…-great grand-daughter photons from sensitive payloads, whether crew or electronics. However, electronics are far less sensitive than humans to this sort of irradiation, so right off the bat this program had an advantage over Rover: there weren’t any people on board, so shielding mass could be minimized.

Ed. Note: figuring out shielded T/W ratio in this field is… interesting to say the least. It’s an open question whether reported T/W includes anything but the thrust structure (i.e. no turbopumps and associated hardware, generally called the “power pack” in rocket engineering), much less whether it includes shielding – and the amount of necessary shielding is another complex question which changes with time. Considering the age of many of these studies, and the advances in computational capability to model not only the radiation being emitted from the reactor vessel but the shielding ability of many different materials, every estimate of required shielding must be taken with 2-3 dump trucks of salt!!! Given that shielding is an integral part of the reactor system, this makes pretty much every T/W estimate questionable.

One of the major challenges of the program, apparently, was to ensure that the reactor would not re-enter the atmosphere, meaning that it had to achieve Earth orbit escape velocity, while still able to deploy the third stage kinetic kill vehicle. I’ve been trying to figure out this staging event for a while now, and have come to the conclusion that my orbital mechanics capabilities simply aren’t good enough to assess how difficult this is beyond “exceptionally difficult.”

However, details of this portion of the program were more highly classified than even the already-highly-classified program, and incredibly few details are available about this portion in specific. We do know that by 1991, the beginning of Phase II of Timber Wind, this portion of the program had been de-emphasized, so apparently the program managers also found it either impractical or no longer necessary, focusing instead on the Nuclear Orbital Transfer Vehicle, or NOTV.

PBR-NOTV: Advanced Upper Stage Flexibility

NOTV Mockup, Powell 1987

At the same time as Timber Wind was gaining steam, the OTV concept was going through a major evolution into the PBR-NOTV (Particle Bed Reactor – Nuclear Orbital Transfer Vehicle). This was another interesting concept, and one which played around with many concepts that are often discussed in the astronuclear field (some related to pebble bed reactors, some related to NTRs), but are almost never realized.

The goals were… modest…

  1. ~1000 s isp
  2. multi-meganewton thrust
  3. ~50% payload mass fraction from LEO to GEO
  4. LEO to GEO transfer time measured in hours, burn time measured in minutes
  5. Customizable propellant usage to change thrust level from same reactor (H2, NH3, and mixtures of the two)

These NOTVs were designed to be the second stage of a booster, similar to the KKV concept we discussed above, but rather than deliver a small kinetic impactor and then leave the cislunar system, these would be designed to place payloads into specific orbits (low Earth orbit, or LEO, mid-Earth orbit, or MEO, and geostationary orbit, GEO, as well as polar and retrograde orbits) using rockets which would normally be far too small to achieve these mission goals. Since the reactor and nozzle were quite small, it was envisioned that a variety of launch vehicles could be used as a first stage, and the tanks for the NTR could be adjusted in size to meet both mission requirements and launch vehicle dimensions. By 1987, there was even discussion of launching it in the Space Shuttle cargo bay, since (until it was taken critical) the level of danger to the crew was negligible due to the lack of oxidizer on board (a major problem facing the Shuttle-launched Centaur with its chemical engine).

There were a variety of missions that the NOTV was designed around, including single-use missions which would go to LEO/MEO/GEO, drop off the payload, and then go into a graveyard orbit for disposal, as well as two way space tug missions. The possibility of on-orbit propellant reloading was also discussed, with propellant being stored in an orbiting depot, for longer term missions. While it wasn’t discussed (since there was no military need) the stage could easily have handled interplanetary missions, but those proposals would come only after NASA got involved.

Multiple Propellants: a Novel Solution to Novel Challenges with Novel Complications

In order to achieve these different orbits, and account for many of the orbital mechanical considerations of launching satellites into particular orbits, a novel scheme for adjusting both thrust and specific impulse was devised: use a more flexible propellant scheme than just cryogenic H2. In this case, the proposal was to use NH3, H2, or a combination of the two. It was observed that the most efficient method of using the two-propellant mode was to use the NH3 first, followed by the H2, since thrust is more important earlier in the booster flight model. One paper observed that in a Hohman transfer orbit, the first part of the perigee burn would use ammonia, followed by the hydrogen to finish the burn (and I presume to circularize the orbit at the end).

When pure ammonia was used, the specific impulse of the stage was reduced to only 500 s isp (compared to the 200-300 s for most second stages), but the thrust would double from 10,000 lbs to 20,000 lbs. By the time the gas had passed into the nozzle, it would have effectively completely dissociated into 3H2 + N2.

One of the main advantages of the composite system is that it significantly reduced the propellant volume needed for the NTR, a key consideration for some of the boosters that were being investigated. In both the Shuttle and Titan rockets, center of gravity and NTR+payload length were a concern, as was volume.

Sadly, there was also a significant (5,000 lb) decrease in payload advantage over the Centaur using NH3 instead of pure H2, but the overall thrust budget could be maintained.

There’s quite a few complications to consider in this design: first, hydrogen behaves very differently than ammonia in a turbopump, not only due to density but also due to compressability: while NH3 is minimally compressible, meaning that it can be considered to have a constant volume for a given pressure and temperature while being accelerated by the turbopump, hydrogen is INCREDIBLY compressible, leading to a lot of the difficulties in designing the power pack (turbopumps, turbines, and supporting hardware of a rocket) for a hydrogen system. It is likely (although not explicitly stated) that at least two turbopumps and two turbines would be needed for this scheme, meaning increased system mass.

Next is chemical sensitivities and complications from the different propellants: while NH3 is far less reactive than H2 at the temperatures an NTR operates at, it nevertheless has its own set of behaviors which have to be accounted for in both chemical reactions and thermal behavior. Ammonia is far more opaque to radiation than hydrogen, for instance, so it’ll pick up a lot more energy from the reactor. This in turn will change the thermal reactivity behavior, which might require the reactor to run at a higher power level with NH3 than it would with H2 to maintain reactor equilibrium.

This leads us neatly into the next behavioral difference: NH3 will expand less than H2 when heated to the same temperature, but at these higher temps the molecule itself may (or will) start to dissociate, as the thermal energy in the molecule exceeds the bonding strength between the covalent bonds in the propellant. This means you’ve now got monatomic hydrogen and various partially-deconstructed nitrogen complexes with different masses and densities to deal with – although this dissociation does decrease propellant mass, increasing specific impulse, and none of the constituent atoms are solids so plating material into your reactor won’t be a concern. These gasdynamic differences have many knock-on effects though, including engine orificing.

See how the top end of the fuel element’s central void is so much narrower than the bottom? One of the reasons for this is that the propellant is hotter – and therefore less dense – at the bottom (it’s also because as you travel down the fuel element more and more propellant is being added). This is something you see in prismatic fuel elements as well, but it’s not something I’ve seen depicted well anywhere so I don’t have as handy a diagram to use.

This taper is called “orificing,” and is used to balance the propellant pressure within an NTR. It depends on the thermal capacity of the propellant, how much it expands, and how much pressure is desired at that particular portion of the reactor – and the result of these calculations is different for NH3 and H2! So some compromises would have to be reached in this cases as well.

Finally, the tankage for the propellant is another complex question. The H2 has to be stored at such a lower temperature compared to the NH3 that a common bulkhead between the tanks simply wouldn’t be possible – the hydrogen would freeze the ammonia. This could lead to a failure mode similar to what happened to SpaceX’s Falcon 9 in September 2016, when the helium tanks became super-chilled and then ruptured on the pad leading to the loss of the vehicle. Of course, the details would be different, but the danger is the same. This leads to the necessity for a complex tankage system in addition to the problems with the power pack that we discussed earlier.

All of this leads to a heavier and heavier system, with more compromises overall, and with a variety of reactor architectures being discussed it was time to consolidate the program.

Multi-Megawatt Power: Electricity Generation

While all these studies were going on, other portions of SDIO were also undergoing studies in astronuclear power systems. The primary electric power system was the SP-100, a multi-hundred kilowatt power supply using technology that had evolved out of the SNAP reactor program in the 60s and 70s. While this program was far along in its development, it was over budget, delayed, and simply couldn’t provide enough power for some of the more ambitious projects within SDIO. Because of this, SDIO (briefly) investigated higher power reactors for their more ambitious – and power-hungry – on-orbit systems.

Power generation was something that was often discussed for pebble bed reactors – the same reasons that make the concept phenomenal for nuclear thermal rockets makes a very attractive high temperature gas cooled reactors (HTGR): the high thermal transfer rates reduce the size of the needed reactor, while the pebble bed allows for very high gas flow rates (necessary due to the low thermal capacity of the coolant in an HTGR). To do this, the gas doesn’t go through a nozzle, but instead through a gas turbine – known as the Brayton cycle. This has huge efficiency advantages over thermoelectric generators, the design being used in SP-100, meaning that the same size reactor can generate much more electricity – but this would definitely not be the same size reactor!

The team behind Timber Wind (including the BNL, SNL and B&W teams) started discussing both electric generation and bimodal nuclear thermal and nuclear electric reactor geometry as early as 1986, before SDIO picked up the program. Let’s take a look at the two proposals by the team, starting with the bimodal proposal.

Particle Bed BNTR: A Hybrid of a Hybrid

Powell et al 1987

The bimodal NTR (BNTR) system never gained any traction, despite it being a potentially valuable addition to the NOTV concept. It is likely that the combination of the increased complexity and mass of the BNTR compared to the design that was finally decided on for Timber Wind, but it was interesting to the team, and they figured someone may be interested as well. This design used the same coolant channels for both the propellant and coolant, which in this case was He. This allowed for similar thermal expansion characteristics and ass flow in the coolant compared to the propellant, while minimizing both corrosion and gas escape challenges.

Horn et al 1987

A total of 37 fuel elements, similar to those used on Timber Wind, were placed in a triangular configuration, with zirconium hydride moderator surrounding them, with twelve control rods for reactivity control. Unusually for many power generation systems, this concept used a conbination of low power, closed loop coolant (using He) and a high power open loop system using H2, which would then be vented out into space through a nozzle (this second option was limited to about 30 mins of high power operation before exhausting H2 reserves). A pair of He Brayton turbines and a radiator was integrated into the BNTR structure. The low power system was designed to operate for “years at a time,” producing 555 kWe of power, while the high power system was rated to 100 Mwe in either rapid ramp or burst mode.

Horn et al 1987

However, due to the very preliminary nature of this design very few things are completely fleshed out in the only report on the concept that I’ve been able to find. The images, such as they are, are also disappointingly poor in quality, but provide at least a vague idea of the geometry and layout of the reactor:

Horn et al 1987

Multi-Megawatt Steady State and Burst Reactor Proposal

By 1989, two years into Timber Wind, SDIO wanted a powerful nuclear reactor to provide two different power configurations: a steady state, 10 Mwe reactor with a 1 year full power lifetime, which was also able to provide bursts of up to 500 MW for long enough to power neutral particle beams and free electron lasers. A variety of proposals were made, including an adaptation of Timber Wind’s reactor core, an adaptation of a NERVA A6 type core (the same family of NERVA reactors used in XE-Prime), a Project Pluto-based core, a hybrid NERVA/Pluto core, a larger, pellet fueled reactor, and two rarer types of fuel: a wire core reactor and a foam fueled reactor. This is in addition to both thermionic and metal Rankine power systems.

The designs for a PBR-based reactor, though, were very different than the Timber Wind reactor. While using the same TRISO-type fuel, they bear little resemblance to the initial reactor proposal. Both the open and closed cycle concepts were explored.

However, this concept, while considered promising, was passed over in preference for more mature fuel forms (under different reactor configurations, namely a NERVA-derived gas reactor.

Finding information about this system is… a major challenge, and one that I’m continuing to work on, but considering this is the best summary I’ve been able to find based on over a week’s searching for source material which as far as I can tel is still classified or has never been digitally documented, as unsatisfying a summary as this is I’m going to leave it here for now.

When I come back to nuclear electric concepts. we’ll come back to this study. I’ve got… words… about it, but at the present moment it’s not something I’m comfortable enough to comment on (within my very limited expertise).

Phase I Experiments

The initial portion of Timber Wind, Phase I, wasn’t just a paper study. Due to the lack of experience with PBR reactors, fuel elements, and integrating them into an NTR, a pair of experiments were run to verify that this architecture was actually workable, with more experiments being devised for Phase II.

Sandia NL ACCR, image DOE

The first of these tests was PIPE (Pulse Irradiation of a Particle Bed Fuel Element), a test of the irradiation behavior of the PBR fuel element which was divided into two testing regimes in 1988 and 1989 at Sandia National Laboratory’s Annular Core Research Reactor using fuel elements manufactured by Babcock and Wilcox. While the ACCR prevented the power density of the fuel elements to achieve what was desired for the full PBR, the data indicated that the optimism about the potential power densities was justified. Exhaust temperatures were close to that needed for an NTR, so the program continued to move forward. Sadly, there were some manufacturing and corrosion issues with the fuel elements in PIPE-II, leading to some carbon contamination in the test loop, but this didn’t impact the ability to gather the necessary data or reduce the promise of the system (just created more work for the team at SNL).

A later test, PIPET (Particle Bed Reactor Integral Performance Tester) began undergoing preliminary design reviews at the same time, which would end up consuming a huge amount of time and money while growing more and more important to the later program (more on that in the next post).

The other major test to occur at this time was CX1, or Critical Experiment 1.

Carried out at Sandia National Laboratory, CX1 was a novel configuration of prototypic fuel elements and a non-prototypical moderator to verify the nuclear worth of fuel elements in a reactor environment and then conduct post-irradiation testing. This sort of testing is vitally important to any new fuel element, since the computer modeling used to estimate reactor designs requires experimental data to confirm the required assumptions used in the calculations.

This novel architecture looked nothing like an NTR, since it was a research test-bed. In fact, because it was a low power system there wasn’t much need for many of the support structures a nuclear reactor generally uses. Instead, it used 19 fuel elements placed within polyethylene moderator plugs, which were surrounded by a tank of water for both neutron reflection and moderation. This was used to analyze a host of different characteristics, from prompt neutron production (since the delayed neutron behavior would be dependent on other materials, this wasn’t a major focus of the testing), as well as initial criticality and excess reactivity produced by the fuel elements in this configuration.

CX-1 was the first of two critical experiments carried out using the same facilities in Sandia, and led to further testing configurations, but we’ll discuss those more in the next post.

Phase II: Moving Forward, Moving Up

With the success of the programmatic, computational and basic experiments in Phase I, it was time for the program to focus on a particular mission type, prepare for ground (and eventual flight) testing, and move forward.

This began Phase II of the program, which would continue from the foundation of Phase I until a flight test was able to be flown. By this point, ground testing would be completed, and the program would be in a roughly similar position to NERVA after the XE-Prime test.

Phase II began in 1990 under the SDIO, and would continue under their auspices until October 1991. The design was narrowed further, focusing on the NOTV concept, which was renamed the Orbital Maneuvering Vehicle.

Many decisions were made at this point which I’ll go into more in the next post, but some of the major decisions were:

  1. 40,000 lbf (~175 kN) thrust level
  2. 1000 MWt power level
  3. Hot bleed cycle power pack configuration
  4. T/W of 20:1
  5. Initial isp est of 930 s

While this is a less ambitious reactor, it could be improved as the program matured and certain challenges, especially in materials and reactor dynamics uncertainties, were overcome.

Another critical experiment (CX2) was conducted at Sandia, not only further refining the nuclear properties of the fuel but also demonstrating a unique control system, called a “Peek-A-Boo” scheme. Here, revolving rings made up of aluminum and gadolinium surrounded the central fuel element, and would be rotated to either absorb neutrons or allow them to interact with the other fuel elements. While the test was promising (the worth of the system was $1.81 closed and $5.02 open, both close to calculated values), but this system would not end up being used in the final design.

Changing of the Guard: Timber Wind Falls to Space Nuclear Thermal Propulsion

Even as Timber Wind was being proposed, tensions with the USSR had been falling. By the time it got going in 1987, tensions were at an all-time low, reducing the priority of the SDIO mission. Finally, the Soviet Union fell, eliminating the need for the KKV concept.

At the same time, the program was meeting its goals (for the most part), and showed promise not just for SDIO but for the US Air Force (who were responsible for launching satellites for DOD and intelligence agencies) as well as NASA.

1990 was a major threshold year for the program. After a number of Senate-requested assessments by the Defense Science Board, as well as assessment by NASA, the program was looking like it was finding a new home, one with a less (but still significantly) military-oriented focus, and with a civilian component as well.

The end of Timber Wind would come in 1991. Control of the program would transfer from SDIO to the US Air Force, which would locate the programmatic center of the project at the Phillips Research Laboratory in Albuquerque, NM – a logical choice due to the close proximity of Sandia National Lab where much of the nuclear analysis was taking place, as well as being a major hub of astronuclear research (the TOPAZ International program was being conducted there as well). Additional stakes in the program were given to NASA, which saw the potential of the system for both uncrewed and crewed missions from LEO to the Moon and beyond.

With this, Timber Wind stopped being a thing, and the Space Nuclear Thermal Propulsion program picked up basically exactly where it left off.

The Promise of SNTP

With the demise of Timber Wind, the Space Nuclear Thermal Propulsion program gained steam. Being a wider collaboration between different portions of the US government, both civil and military, gave a lot of advantages, wider funding, and more mission options, but also brought its’ own problems.

In the next post, we’ll look at this program, what its plans, results, and complications were, and what the legacy of this program was.

References and Further Reading

Timber Wind/SNTP General References

Haslett, E. A. “SPACE NUCLEAR THERMAL PROPULSION
PROGRAM FINAL REPORT
https://apps.dtic.mil/dtic/tr/fulltext/u2/a305996.pdf

Orbital Transfer Vehicle

Powell et al, “NUCLEAR PROPULSION SYSTEMS FOR ORBIT TRANSFER BASED ON THE
PARTICLE BED REACTOR” Brookhaven NL 1987 https://www.osti.gov/servlets/purl/6383303

Araj et al, “ULTRA-HIGH TEMPERATURE DIRECT PROPULSION”” Brookhaven NL 1987 https://www.osti.gov/servlets/purl/6430200

Horn et al, “The Use of Nuclear Power for Bimodal Applications in Space,” Brookhaven NL 1987 https://www.osti.gov/servlets/purl/5555461

Multi-Megawatt Power Plant

Powell et al “HIGH POWER DENSITY REACTORS BASED ON
DIRECT COOLED PARTICLE BEDS” Brookhaven NL 1987 https://inis.iaea.org/collection/NCLCollectionStore/_Public/17/078/17078909.pdf

Marshall, A.C “A Review of Gas-Cooled Reactor
Concepts for SDI Applications” Sandia NL 1987 https://www.osti.gov/servlets/purl/5619371

“Atomic Power in Space: a History, chapter 15” https://inl.gov/wp-content/uploads/2017/08/AtomicPowerInSpaceII-AHistory_2015_chapters6-10.pdf

Categories
Development and Testing Forgotten Reactors History Non-nuclear Testing Nuclear Thermal Systems Test Stands

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!

If you would like to support my work, consider becoming a Patreon supporter at https://www.patreon.com/beyondnerva . Not only do you get early access to blog posts, but I post extra blogs, images from the 3d models I’m working on of both spacecraft and reactors, and more! Every bit helps.

You can also follow me on Twitter (https://twitter.com/BeyondNerva) for more content and conversation!

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
History Nuclear Thermal Systems

Radiator LNTR: The Last of the Line

Hello, and welcome back to Beyond NERVA! Today, we’re finishing (for now) our in-depth look at liquid fueled nuclear thermal rockets, by looking at the second major type of liquid NTR (LNTR): the radiator-type LNTR. If you’re just joining us, make sure to check out the introduction (available here) and the bubbler post (available here) for some important context to understand how this design got here.

Rather than passing the propellant directly through the molten fuel, in this system the propellant would pass through the central void of the fuel element, becoming heated primarily through radiation (although some convection within the propellant flow would occur, overall it was a minor effect), hence the name.

This concept had been mentioned in previous works on bubbler-type LNTRs, and initial studies on the thermalization behavior of the propellant, and conversely fuel cooling behavior, were conducted during the early 1960s, but the first major study wouldn’t occur until 1966. However, it would also extend into the 1990s in its development, meaning that it was a far longer-lived design.

Let’s begin by looking at the differences between the bubbler and radiator designs, and why the radiator offers an attractive trade-off compared to the bubbler.

The Vapor Problem, or Is Homogenization of Propellant/Fuel Temp Worth It?

Liquid fuels offer many advantages for an NTR, including the fact that the fuel will distribute its heat evenly across the volume of the fuel element, the fact that the effective temperature of the fuel can be quite high, and that the fuel is able to be reasonably well contained with minimal theoretical challenges.

The bubbler design had an additional advantage: by passing the propellant directly through the fuel, in discrete enough bundles (the bubbles themselves) that the fuel and the propellant would have the same temperature.

Maximum specific impulse due to vapor pressure, Barrett Jr.

Sadly, there are significant challenges to making this sort of nuclear reactor into a rocket, the biggest one being propellant mass. These types of NTRs still use hydrogen propellant, the problem occurs in the fuel mass itself. As the bubbles move through the zirconium/niobium-uranium carbide fuel, it heats up rapidly, and the pressure drops significantly during this process. This means that all of the components of the fuel (the Zr/Nb, C, and U) end up vaporizing into the bubbles, to the point that the bubble is completely saturated by a mix of these elements in vapor form by the time it exits the fuel body. This is called vapor entrainment.

This is a major problem, because it means that the propellant leaving the nozzle has a far higher mass than the hydrogen that was originally input into the system. While there’s the possibility that a different propellant could be used which would not entrain as much of the fuel mass, but would also be higher molecular mass to start – to the point that the gains might likely outweigh the losses (if you feel like exploring this trade-off on a more technical footing, please let me know! I’d love to explore this more), and it wouldn’t eliminate the entrainment problem.

This led people to wonder if you have to pass the propellant through the fuel in the first place. After all, while there is a thermodynamically appealing symmetry to homogenizing your fuel and propellant temperatures, this isn’t actually necessary, is it? The fuel elements are already annular in shape, after all, so why not use them as a more traditional fuel element for an NTR? The lower surface area would mean that there’s less chance for the inconveniently high vapor pressure of the fuel would be mitigated by the fact that the majority of the propellant wouldn’t come in contact with the fuel (or even the layer of propellant that does interact with the fuel), meaning that the overall propellant molecular mass would be kept low… right?

The problem is that this means that the only method of heating the propellant becomes radiation (there’s a small amount of convection, but it’s so negligible that it can be ignored)… which isn’t that great in hydrogen, especially in the UV spectrum that most of the photons from the nuclear reaction are emitted in. The possibility of using either microparticles or vapors which would absorb the UV and re-emit it in a lower wavelength, which would be more easily absorbed by the hydrogen, was already being investigated in relation to gas core NTRs (which have the same problem, but in a completely different order of magnitude), and offered promise, but was also a compromise: this deliberately increases the molar mass of the propellant one way to minimize the molar mass a different way. This was a design possibility that needed to be carefully studied before it could be considered more feasible than the bubbler LNTR.

The leader of the effort to study this trade-off was one of the best-known fluid fueled NTR designers on the NASA side: Robert Ragsdale at Lewis Research Center (LRC, and we’ll come back to Ragsdale in gas core NTR design as well). There were a collection of studies around a particular design, beginning with a study looking at reactor geometry and fuel element size optimization to not only maximize the thrust and specific impulse, but also to minimize the uranium loss rates of the reactor.

This study concluded that there were many advantages to the radiator-type LNTR over the bubbler-type. The first consideration, minimizing the vapor entrainment problem that was laguing the bubbler, was minimized, but not completely eliminated, in the radiator design. The next conclusion is that the specific impulse of the negine could be maintained, or increased, to 1400 s isp or more. Finally, one of th emost striking was in thrust-to-core-weight ratio, which went from about 1:1 in the Nelson/Princeton design that we discussed in the last post all the way up to 19:1 (potentially)! This is because the propellant flow rate isn’t limited y the bubble velocity moving through the fuel (for more detail on this problem, and the other related constraints, check out the last blog post, here).

These conclusions led to NASA gathering a team of researchers, including Ragsdale, Kasack, Donovan, Putre, and others ti develop the Lewis LNTR reactor.

Lewis LNTR: The First of the Line

Lweis Radiator LNTR, Ragsdale 1967

Once the basic feasibility of the radiator LNTR was demonstrated, a number of studies were conducted to determine the performance characteristics, as well as the basic engineering challenges, facing this type of NTR. They were conducted in 1967/68, and showed distinct promise, for the desired 2000 to 5000 MWt power range (similar to the Phoebus 2 reactor’s power goal, which remains the most powerful nuclear reactor ever tested at 3500 MWt).

Fuel tube cross-section, Putre 1968

As with any design, the first question was the basics of reactor configuration. The LRC team never looked at a single-tube LNTR, for a variety of reasons, and instead focused their efforts on a multi-tube design, but the number and diameter of the tubes was one of the major questions to be addressed in initial studies. Because of this, and the particular characteristics of the heat transfer required, the reactor would have many fuel elements with a diameter of between 1 and 4 inches, but which diameter was best would be a matter of quite some study.

Another question for the study team was what the fuel element temperature would be. As in every NTR design, the hotter the propellant, the higher the isp (all other things being equal), but as we saw in the bubbler design, higher temperatures also mean higher vapor pressure, meaning that mass is lost more easily into the propellant – which increases the propellant mass and reduces the isp, and at some point even cost more specific impulse due to higher mass than is gained with the higher temperature. Because the propellant and the fuel would only interact on the surface of the fuel element, the surface temperature of the fuel was the overriding consideration, and was also explored, in the range of 5000 to 6100 K.

Effect of Reactor Pressure on T/W Ratio and U mass loss ratio in H, Ragsdale 1967

The final consideration which was optimized in this design was engine operating pressures. Because this design wasn’t fundamentally limited by the bubble velocity and void fraction of the molten fuel, the chamber pressure could be increased significantly, leading to both more thrust and a higher thrust-to-weight ratio. However, the trade-off here is that at some point the propellant isn’t being completely thermalized, resulting in a lower specific impulse. This final consideration was explored in the range of 200 to 1000 atm (2020-10100 N/cm2).

The three primary goals were: to maximize specific impulse, maximize thrust-to-weight ratio, and minimize uranium mass loss. They quickly discovered that they couldn’t have their cake and eat it, too: higher temperatures, and therefore higher isp, led to faster U mass loss rates, increasing T/W ratio reduced the specific impulse, and minimizing the U loss rate hurt both T/W and isp. They could improve any one (or often two) of these characteristics, but always at the cost of the third characteristic.

Four potential LNTR configurations, note the tradeoffs between isp, T/W, and fuel loss rates. Ragsdale 1967

We’ll look at many of the design characteristics and engineering considerations of the LRC work in the next section on general design challenges and considerations for the radiator LNTR, but for now we’ll look at their final compromise reactor.

The reactor itself would be made up of several (oddly, never specified) fuel elements, in a beryllium structure, with each fuel element being made up of Be as well. These would be cooled by cryogenic hydrogen moving from the nozzle end to the spacecraft end of the reactor, before flowing back into the central void of the fuel element. As it was fed through the central annulus, it would be seeded with tungsten microparticles to increase the amount of heat the propellant would absorb. Finally, it would be exhausted through a standard De Laval nozzle to provide thrust.

Reference LRC LNTR design characteristics, Putre 1968

The final fuel that they settled on was a liquid ternary carbide design, with the majority of the fuel being niobium carbide (although ZrC was also considered), with a molar mass fraction of 0.02 being UC2. This compromise offered good power density for the reactor while minimizing the vaporization rate of the fuel mass. This would be held in 2 inch diameter, 5 foot long fuel element tubes, with a fuel surface temperature of 5060 K. The propellant would be pressurized to 200 atm in the reactor.

Final LRC LNTR Fuel Characteristics, Putre 1968

This led to a design that struck a compromise between isp, T/W, and U mass loss which was not only acceptable, but impressive: 1400 s isp (on par with some forms of electric propulsion), a T/W ratio (of the core alone) of 4, and a hydrogen-to-uranium flow rate ratio of 50.

They did observe that none of these characteristics were as high as they could be, especially in terms of T/W ratio (which they calculated could go as high as 19!), or isp (with a theoretical maximum of 1660), and the uranium loss was twice the theoretical minimum, but sadly the cost of maximizing any of these characteristics was so high from an engineering point of view that it wasn’t feasible.

Sadly, I haven’t been able to find any documentation on this reactor design – and very few references to it – after February 1968. The exact time of the cancellation, and the reasons why, are a mystery to me. If someone is able to help me find that information it would be greatly appreciated.

LARS: The Brookhaven Design

LARS cross section,

The radiator LNTR would remain dormant for decades, as astronuclear funding was scarce and focused on particular, well-characterized systems (most of which were electric powerplant concepts), until the start of the Space Exploration Initiative. In 1991, a conference was held to explore the use of various types of NTR in future crewed space missions. This led to many proposals, including one from the Department of Energy’s Brookhaven National Laboratory in New York. This was the Liquid Annular Reactor System, or LARS.

A team of physicists and engineers, including Powell, Ludewig, Lazareth, and Maise decided to revisit the radiator LNTR design, but as far as I can tell didn’t use any of the research done by the LRC team. Due to the different design philosophies, lack of references, and also the general compartmentalization of knowledge within the different parts of the astronuclear community, I can only conclude that they began this design from scratch (if this is incorrect, and anyone has knowledge of this program, please get in contact with me!).

LARS was a very different design than the LRC concept, and seems to have gone through two distinct iterations. Rather than the high-pressure system that the LRC team investigated, this was a low-pressure, low-thrust design, which optimized a different characteristic: hydrogen dissociation. This maximizes the specific impulse of the NTR by reducing the mass of the propellant to the lowest theoretically possible mass while maintaining the temperature of the propellant (up to 1600 s, according to the BNL team). The other main distinction from the LRC design was the power level: rather than having a very powerful reactor (3000 to 5000 MWt), this was a modest reactor of only 200 MWt. This leads to a very different set of design tradeoffs, but many of the engineering and materials challenges remain the same.

LARS would continue to us NbC diluted with UC2, but the fuel would not completely melt in the fuel element, leaving a solid layer against the walls of the beryllium fuel element tube. This in turn would be regeneratively cooled with hydrogen flowing through a number of channels in the durm, as well as a gap surrounding the body of the fuel element which would also be filled with cold hydrogen. A drive system would be attached on the cold end of the tube to spin it at an appropriate rate (which was not detailed in the papers). The main changes were in the fuel element configuration, size, and number.

The first iteration of LARS was an interesting concept, using a folded-flow system. This used many small fuel element tubes, arranged in a similar manner to the flow channels in the Dumbo reactor, with the propellant moving from the center of the reactor to the outer circumference, before being ejected out of the nozzle of the reactor. Each layer of fuel elements contained eleven individual tubes, with between 1 and 10 layers of fuel elements in the reactor. As the number of layers increased, the length and radius of the fuel elements decreased.

One of the important notes that was made by the team at this early design date was that the perpendicular fuel element orientation would minimize the amount of fission products that would be ejected from the rocket. I’m unable to determine how this was done, other than if they were solids which would stick to the outside of the propellant flue, however.

Unfortunately, I haven’t been able to discover exactly why this design was abandoned for a more traditional LNTR architecture, but the need to cool the entire exterior of the reactor to keep it from melting seems to possibly be a concern. Reversing the flow, with the hot propellant being in the center of the reactor rather than the external circumference, seems like an easy fix if this was the primary concern, but the discussions of reactor architecture after this seem to pretty much ignore this early iteration. Another complication would be the complexity of the reactor architecture. Whether with dedicated motors, or with a geared system allowing one motor to spin multiple fuel elements, a complex system is needed to spin the fuel elements, which would not only be something which would potentially break down more, but also require far more mass than a simpler system.

The second version of LARS kept the same type of fuel, power output, and low pressure operation, but rather than using the folded flow concept it went with seven fuel elements in a beryllium body. The propellant would be used to cool first the nozzle of the rocket, then the rotating beryllium drum which contained the fuel element, before entering the main propellant channel. The final thermalization of the propellant would be facilitated by the use of tungsten microparticles in the H2, necessary due to the low partial pressure and high transparency of pure H2 (while the vapor pressure issues of any LNTR were acknowledged, the effect that this would have on the thermalization seems to have not been considered a significant factor in the seeding necessity) Two versions, defined by the emissivity of the fuel element, were proposed.

Final two LARS options, f is fuel emissivity, Maise 1999

This design was targeted to reach up to 2000 s isp, but due to uncertainties in U loss rates (as well as C and Nb), the overall mass of the propellant upon exiting the reactor was uncertain, so the authors used a range of 1600-2000 s. The thrust of the engine was approximately 20,000 N, which would result in a T/W ratio of about 1;1 when including a shadow shield (one author points out that without the shield the ratio would be about 3-4/1.

I have been unable to find the research reports themselves for this program (unlike the LRC design), so the specifics of the reactor physics tradeoffs, engineering compromises, actual years of research and the like aren’t something that I’m able to discuss. The majority of my sources are conference papers and journal articles, which occurred in 1991 and 1992, but there was one paper from 1999, so it was at least under discussion through the 1990s (interestingly, that paper discussed using LARS for the 550 AU mission concept, which later got remade into the FOCAL gravitational lens mission: https://www.centauri-dreams.org/2010/11/15/a-focal-mission-into-the-oort-cloud/ ).

This seems to be the last time that LARS has been mentioned in the technical literature, so while it is mentioned as the “baseline” liquid core concept in places such as Atomic Rockets (http://www.projectrho.com/public_html/rocket/enginelist2.php#id–Nuclear_Thermal–Liquid_Core–LARS) it has not been explored in depth since.

Lessons Learned, Lessons to Learn: The Challenges of LNTR

In many ways, the apparent dual genesis of radiator LNTRs offers the ability to look into two particular thought processes in what the challenges are with radiator-type LNTRs. One example of this is what’s discussed more in the “fundamental challenges” sections of the introductory section of the reports: for the LRC team they focus on vapor entrainment minimization, whereas in the BNL presentations it seems quite important to point out that “yes, containing a refractory in a spinning, gas cooled drum is relatively trivial.” This juxtaposition of foci is interesting to me, as an examination of the different engineering philosophies of the two teams, and the priorities of the times.

Wall Construction

Both the LRC and LARS LNTRs ended up with similar fuel element configurations: a high temperature material, with coolant tubes traveling the length of the fuel element walls to regeneratively cool the walls. This material would have to withstand not only the temperature of the fuel element, but also resist chemical attack by the hydrogen used for the regenerative cooling, as well as being able to withstand the mechanical strain of not only the spinning fuel, but also the torque from whatever drive system is used to spin the fuel element to maintain the centripetal force used to contain the fuel element.

Another constant concern is the temperature of the wall. While high temperature loadings can be handled using regenerative cooling, the more heat is removed from the fuel during the regenerative cooling step, but it reduces the specific impulse of the engine. Here’s a table from the LRC study that examines the implications of wall cooling ratio vs specific impulse in that design, which will also apply as a general rule of thumb for LARS:

However, from there, the two designs differed significantly. The LARS design is far simpler: a can of beryllium, with a total of 20% of the volume being the regenerative cooling channel. As mentioned previously, the fuel didn’t become completely molten, but remained solid (and mostly containing the ZrC/NbC component, with very little U). Surrounding the outside of the fuel element can itself was another coolant gap. This would then be mounted to the reactor body with a drive system at the ship end, and a bearing at the hot end. This would then be mounted in the stationary moderator which made up the majority of the internal volume of the reactor core, which was shielded from the heat in the fuel element in a very heterogeneous temperature profile.

The LRC concept on the other hand, was far more complex in some ways. Rather than using a metal can, the LRC design used graphite, which maintains its strength far better than many metals at high temperatures. A number of options were considered to maintain the wall of the can, due not only to the fuel mixture potentially attacking the graphite (as the carbon could be dissolved into the carbide of the fuel element), as well as attack from the hydrogen in the coolant channels (which would be able to be addressed in a similar way to how NERVA fuel elements used refractory metal coatings to prevent the same erosive effects).

The LRC design, since the fuel would be completely molten across the entire volume of the fuel element, was a more complex challenge. A number of options were considered to minimize the wall heating of the fuel element, including:

  • Selective fuel loading
    • A common strategy in solid fuel elements, this creates hotter and cooler zones in the fuel element
      • Neutron heating will distribute the radiative heating past U distribution
    • Convection and fuel mixing will end up distributing the fuel over time
    • May be able to be limited by affecting the temperature and viscosity of the fuel for the life of the reactor
  • Multiple fluids in fuel
    • Step beyond selective loading, a different material may be used as the outer layer of the fuel body, resisting mixing and reducing thermal load on the wall
  • Vapor insulation along exterior of fuel body
    • Using thermally opaque vapor to insulate the fuel element wall from the fuel body
    • Significantly reduces the heating on the outer wall
    • Two options for maintaining vapor wall:
      • Ablative coating on inner wall of fuel element can
      • Porous wall in can (similar to a low-flow version of a bubbler fuel element) pumping vapor into gap between fuel and can
    • Maximum stable vapor-layer thickness based on vapor bubble force balance vs centripetal force of liquid fuel
      • Two phase flow dynamics needed to maintain the vapor layer would be complex

This set of options offer a trade-off: either a simpler option, which sets hard limits on the fuel element temperature in order to ensure the phase gradient in the fuel element (the LARS concept), or the fully liquid, more complex-behaving LRC design which has better power distribution, and a higher theoretical fuel element temperature – only limited by the vapor pressure increase and fuel loss rates in the fuel element, rather than the wall heating temperature limits of the LARS design.

Anyone designing a new radiator LNTR has much work that they can draw from, but other than the dynamics of the actual fuel behavior (which have never gone through a criticality test), the fuel element can design will be perhaps the largest set of engineering challenges in this type of system (although simpler than the bubbler-type LNTR).

Propellant Thermalization

The major change between the bubbler and radiator-type LNTRs is the difference in the thermalization behavior of the propellant: in a bubbler-type LNTR, assuming the propellant can be fed through the fuel, the two components reach thermal equilibrium, so the only thing needed is to direct it out of the nozzle; a radiator on the other hand has a similar flow path to the Rover-type NTRs, once through from nozzle to ship side for regenerative cooling, then a final thermalization pass through the central void of the fuel element.

This is a problem for hydrogen propellant, which is largely transparent to the EM radiation coming off the reactor. This thermal transfer accounted for all but 10% of the thermalization effects in the LARS design, and in many of the LRC studies this was completely ignored as negligible, with the convective effects in the propellant mainly being a concern in terms of fuel mass loss and propellant mass increase.

While the fuel mass loss would increase the opacity of the gas (making it absorb more heat), a far better option was available: adding a material in microparticle form to the propellant flow as it goes through the final thermalization cycle. The preferred material for the vast majority of these applications, which we’ll see in the gas cycle NTRs as well, is microparticles of tungsten.

This has been studied in a host of different applications, and will be something that I’ll discuss in depth on a section of the propellant webpage in the future (which I’ll link to here once it’s done), but for the LRC design the target goal for increasing the opacity of the H2 was to achieve between 10,000 and 20,000 cm^2/gm, for a reduction in single-digit percentage of specific impulse due to the higher mass. They pointed out that the simplified calculations used for the fuel mass loss behavior could lead to an error that they were unable to properly address, and which could either increase or decrease the amount of additive used.

The LARS concept used tungsten microparticles as well, and their absorption actually was the defining factor in the two designs they proposed: the emissivity and reflectivity of the fuel in terms of the absorption of the wall and the propellant.

Two other options are available for increasing the opacity of the hydrogen gas.

The first is to use a metal vapor deliberately, as was the paradigm in Soviet gas core design. Here, they used either NaK or Li vapor, both of which have small neutron absorption cross-sections and high thermal capacity. This has the advantage of being more easily mixed with the turbulent propellant stream, as well as being far lower mass than the W that is often used in US designs, but may be less opaque to the EM frequencies being emitted by the fuel’s surface in an LNTR design. I’m still trying to track down a more thorough writeup of the use of these vapors in NTR design at the moment (a common problem in both Soviet and Russian astronuclear literature is a lack of translations), but when I do I’ll discuss it in far more depth, since it’s an idea that doesn’t seem to have translated into the American NTR design paradigm.

As I said, this is a concept that I’m going to cover more in depth in both the gas core and general propellant pages, so with one final – and fascinating – note, we’ll move on to the conclusion.

An Interesting Proposal

The final option is something that Cavan Stone mentioned to me on Facebook a while ago: the use of lithium deuteride (LiD) as a propellant or additive in this design. This is an interesting concept, since Li7 is a fissile material, and is reasonably opaque to the frequencies being discussed in these reactors. The use of deuterium rather than protium also increases the neutron moderation of the propellant, which can in turn increase fissile efficiency of the reactor. The Li will harden the neutron spectrum overall, while the D and Be (in the fuel element can/reactor body) will thermalize the spectrum.

There was a discussion of using LiD as a propellant in NTRs in the 1960s [https://www.osti.gov/biblio/4764043-nuclear-effect-using-lithium-hydride-propellant-nuclear-rocket-reactor-thesis], but sadly I can’t find it anywhere online. If someone is able to help me find it, please let me know. This is a fascinating concept, and one that I’m very glad Cavan brought up to me, but also one that is complex enough that I really need to see an in-depth study by someone far more knowledgeable than me to be able to intelligently discuss the implications of.

Conclusion, or The Future of the Forgotten Reactor

While often referenced in passing in any general presentation on nuclear thermal rockets, the liquid core NTR seems to be the least studied of the different NTR types, and also the least covered. While the bubbler offers distinct advantages from a purely thermodynamic point of view, the radiator offers far more promise from a functional perspective.

Sadly, while both solid and gas core NTRs have been studied into the 20th century, the liquid core has been largely forgotten, and the radiator in particular seems to have gone through a reinvention of the wheel, as it were, between the 1960s NASA design and the 1990s DOE design, with few of the lessons learned from the LRC concept being applied to the BNL design as far as vapor dynamics, thermal transfer, and the like.

This doesn’t mean that the design is without promise, though, or that the challenges that the reactor faces are insurmountable. A number of hurdles in testing need to be overcome for this design to work – but many of the problems that there simply isn’t any data for can be solved with a simple set of criticality and reactor physics tests, something well within the capabilities of most research nuclear programs with the capability to test NTRs.

With the advances in nuclear and two-phase flow modeling, a body of research that doesn’t seem to have been examined in depth for over two decades, and the possibility of a high-isp, moderate-to-high thrust engine without the complications of a gas core NTR (a subject that we’ll be covering soon), the LNTR, and the radiator in particular, offer a combination of promise and ability to develop advanced NTRs as low hanging fruit that few systems are able to offer.

Final Note

With that, we’re leaving the realm of liquid fueled NTRs for now. This is a fascinating field, and one that I haven’t seen much discussion of outside the original technical papers, so I hope you enjoyed it! I’m going to work on getting these posts into a more easily-referenced form on the website proper, and will make a note of that in my blog (and on my social media) when I do! If anyone is aware of any additional references pertaining to the LNTR, as well as its thermophysical behavior, fuel materials options, or anything else relating to these desgins, please let me know, either in the comments or by sending me a message to beyondnerva at gmail dot com.

Our next blog post will be on droplet and vapor core NTRs, and will be covered by a good friend of mine and fellow astronuclear enthusiast: Calixto Lopez. These reactors have fascinated him since he was in school many moons ago, and he’s taught me the majority of what I know about them, so I asked him if he was willing to write that post.

After that, we’re going to move on to the closed cycle gas core NTR, which I’ve already begun research on. There’s lots of fascinating tidbits about this reactor type that I’ve already uncovered, so this may end up being another multiple part blog series.

Finally, to wrap up our discussion of advanced NTRs, we’re going to do a series on the open cycle gas core NTR types. This is going to be a long, complex series on not only the basic physics challenges, but the design evolution of the engine type, as well as discussion on various engineering methods to mitigate the major fuel loss and energy waste issues involved in this type of engine. There may be a delay between the closed and open cycle NTR posts due to the sheer amount of research necessary to do open cycles justice, but rest assured I’m already doing research on them.

As you can guess, this blog takes a lot of time, and a lot of research, to write. If you would like to support me in my efforts to bring the wide and complex history of astronuclear engineering to light, consider supporting me on Patreon: https://www.patreon.com/beyondnerva . Every dollar helps, and you get access to not only early releases of every blog post and webpage, but at the higher donation amounts you also get access to the various 3d models that I’m working on, videos, and eventually the completed 3d models themselves for your own projects (with credit for the model construction, of course!).

I’m also always looking for new or forgotten research in astronuclear engineering, especially that done by the Soviet Union, Russia, China, India, and European countries. If you run across anything interesting, please consider sending it to beyondnerva at gmail dot com.

You can find me on Twitter ( @beyondnerva), as well as join my facebook group (insert group link) for more astronuclear goodness.

References

General References

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

ANALYSIS OF VAPORIZATION OF LIQUID URANIUM, METAL, AND CARBON SYSTEMS AT 9000” AND 10,000” R Kaufman and Peters 1966 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19660025363.pdf

Fundamental Material Limitations in Heat-Exchanger Nuclear Rockets, Kane and Wells, Jr. 1965 https://www.osti.gov/servlets/purl/4610034/

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

Radiator-Specific LNTR References

Lewis Research Center LNTR

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 IN A RADIATING LIQUID-CORE NUCLEAR ROCKET CONCEPT, Putre and Kasack 1968 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19680007624.pdf

Liquid Annular Reactor System (LARS)

[Paywall] Conceptual Design of a LARS Based Propulsion System, Ludewig et al 1991 https://arc.aiaa.org/doi/abs/10.2514/6.1991-3515

The Liquid Annular Reactor System (LARS) Propulsion, Powell et al 1991 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19910012832.pdf

LIQUID ANNULUS, Ludewig 1992 https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19920001886.pdf

[Paywall] The liquid annular reactor system (LARS) for deep space exploration, Maise et al 1999 https://www.sciencedirect.com/science/article/abs/pii/S0094576599000442

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.

If you would like early access to these, as well as all future blog posts and websites, consider becoming a Patron of the page! My Patrons help me be able to devote the time that I need to the website, and provide strong encouragement for me to put out more material as well! You can sign up here: https://www.patreon.com/beyondnerva

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
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
Fission Power Systems Low Enriched Uranium Nuclear Thermal Systems Spacecraft Concepts

LEU NTP Part Three: Spacecraft Overview

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

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

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

Advances in LEU Nuclear Propulsion

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

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

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

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

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

The Beginnings of the Modern Astronuclear Thermal Era

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

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

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

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

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

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

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

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

 

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

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

Radiation Shielding

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

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

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

What is Radiation, and How is it Shielded?

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

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

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

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

Neutron Shielding

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

Gamma and X-Ray Shielding

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

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

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

NTP-Specific Shielding

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

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

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

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

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

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

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

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

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

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

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

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

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

The Propellant Tanks

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

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

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

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

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

 

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

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

 

Nuclear + Rockets: Always Complicated

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

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

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

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

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

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

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

Enter Active Cooling: The Zero Boil-Off Tank

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

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

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

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

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

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

 

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

GTA at MSFC
Test Article at MSFC, Image Courtesy NASA

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

LH2 Shielding Goes Away Through the Mission

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

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

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

LEU NTP Mars Vehicle

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

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

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

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

Core Propulsion Stage

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

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

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

In-Line Propellant Tank

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

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

Saddle Truss with Drop Tanks

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

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

Mission Payload

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

The Integrated Stack

MTV Copernicus (NCPS Config), NASA

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

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

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

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

Stack Construction

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

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

SLS Sensitivity Chart

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

LEU NTP: The Latest Plan to Get to Mars

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

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

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

Additional Reading

Nuclear Thermal Propulsion

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

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

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

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

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

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

Radiation and Shielding

Types of Radiation video, FermiLab

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

In-Space Radiation Environment and Crew Quarters Shielding

Neutron Astronomy; Casadei, University of Birmingham, 2017

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

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

Homepage, Rutherford Appleton Laboratory Mini Magnetosphere Project

In-Space Reactor Shielding

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

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

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

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

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

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

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

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

Auxiliary Support Systems for NTR

 

Propellant Tanks and Zero Boil-Off

Passive Thermal Protection

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

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

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

Zero Boil-Off

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

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

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

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

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

Nuclear Cryogenic Propulsion Stage

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

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

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

Launch Vehicles

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

Ares V Wikipedia Page

Ares V Fact Sheet, NASA 2006

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

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

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

NASA Space Launch System Main Page

NASA SLS Overview Page

NASA’s Mars Design Reference Mission 5.0 and Associated Considerations

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

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

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

Blog Updates

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

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

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

Categories
Development and Testing Low Enriched Uranium Nuclear Thermal Systems

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

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

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

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

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

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

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

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

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

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

CERMET Fuel Elements

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

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

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

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

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

CERMET Composition and Manufacture

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

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

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

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

Angular UO2 Microparticles
Angular UO2 microparticles, image courtesy NASA

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

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

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

HIP process
Image courtesy NASA

 

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

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

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

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

The Beginnings of CERMET Fuels

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

ANL 200 MW Reactor
Image courtesy DOE

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

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

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

Non-Spherical Microparticles
UO2 particles

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

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

Can component fit check pic

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

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

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

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

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

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

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

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

NTREES

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

Houts NTREES Facility 2013

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

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

Chamber installation
Pressure Chamber during upgrade, image courtesy NASA MSFC

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

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

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

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

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

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

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

 

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

 

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

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

CFEET

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

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

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

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

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

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

More info on CFEET is available on the CFEET page!

What Have They Taught Us?

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

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

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

Development Challenges for LEU NTP, and a New Direction

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

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

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

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

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

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

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

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

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

How This Changes the Core

BWXT Core
BWXT Core, image via BWXT

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

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

More to Come!

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

Sources and Additional Reading

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

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

CERMET Fueled Reactors, Cowan et al 1987

A CERMET Fueled Reactor for Nuclear Propulsion, Kruger 1991

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

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

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

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

Low Enriched Nuclear Thermal Propulsion Systems, Houts et al 2017

NTP CERMET Fuel Development Status, Barnes 2017

2017 Game Changing Development program Mid-year Review Slides

Channel update:

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

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

 

Cpoyright 2018 Beyond NERVA. Contact for reprint permission.