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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!

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References

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

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

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

Categories
Development and Testing Forgotten Reactors Nuclear Thermal Systems

The Bubbler: Liquid NTRs Without Barriers

Hello, and welcome back to Beyond NERVA! Today, we continue our look at liquid fueled nuclear thermal rockets (LNTRs), with a deep dive into the first of the two main types: what I call the bubbler LNTR.

This potentially attractive form of advanced NTR is a design that has been largely forgotten in the history of NTR designs outside some minor footnotes. Because of this, I felt that it was a great subject for the blog! All of the sources that I can find on the designs are linked at the end of this post, including a couple that are not available digitally, so if you’re interested in a more technical analysis of the concept please check that out!

What is a Bubbler LNTR?

Every NTR has to heat the (usually hydrogen) propellant in some way, which is usually done through (usually thermal) radiation from the fuel’s surface into the propellant.

Bubbles passing through fuel, Nelson 1963

This design, though, changes that paradigm by passing the propellant through the liquid fuel (usually a mix of uranium carbide (UC2) and some other carbide – either zirconium (ZrC) or niobium (NbC). This is done by having a porous outer wall which the propellant is injected through. This is known as a “folded flow propellant path,” and is seen in other NTRs as well, notably the Dumbo reactor from the early days of Project Rover.

In order to keep the fuel in place, each fuel element is spun at a high enough rate to keep the fuel in place using centrifugal force. The number of fuel elements is one of the design choices that varies from design to design, and the overall diameter, as well as the thickness of the fuel layer, is a matter of some design flexibility as well, but on average the individual fuel elements range from about 2 to about 6 inches in diameter, with the ratio between the thickness of the fuel layer and the thickness of the central void where the now-hot propellant passes through to the nozzle being roughly 1:1.

This was the first type of LNTR to be proposed, and was a subject of study for over a decade, but seems to have fallen out of favor with NTR designers in the late 1960s/early 1970s due to fuel/propellant interaction complications and engineering challenges related to the physical structures for injecting the propellant (more on that later).

Let’s look at the history of bubbler LNTR in more depth, and see how the proposals have evolved over time.

History of the Bubbler-type LNTR: The First of its Kind

McCarthy, 1954

Image from Barrett, Jr 1964

The first proposal for a liquid fueled NTR was in 1954, by J McCarthy in “Nuclear Reactors for Rockets” [ed. Note I have been unable to locate this report in digital form, if anyone is able to help me get ahold of it I would greatly appreciate your assistance; the following summary is based on references to this study in later works]: This design was the first to suggest the centrifugal containment of liquid fuel, and was also the first of the bubbler designs. It used a single fuel element as the entire reactor, with a large central void in the center of the fuel body as the propellant flow channel once it left the fuel itself.

This design was fundamentally limited by three factors:

  1. A torus is a terrible neutronic structure, and while the hydrogen propellant in the central void of the fuel would provide some neutron moderation, McCarthy found upon running the MCNP calculations that the difference was so negligible that it could be assumed to be a vacuum; and
  2. Only a certain amount of heat could be removed from the fuel by the propellant based on assumed fuel element geometry, and that cooling the reactor could pose a major challenge at higher reactor powers; and
  3. The behavior of the hydrogen as it passes through, and also out of, the liquid fuel was not well understood in practice, and
  4. the vapor pressure of the fuel’s constituent components could lead to fuel being absorbed in the gas as vapor in both the bubbles and exhausting propellant flow, causing both a loss of specific impulse and fissile fuel. This process is called “entrainment,” and is a (if not the) major issue for this type of reactor.

However, despite these problems this design jump started the design of LNTRs, defined the beginnings of the design envelope for this type of engine, and introduced the concept of the bubbler LNTR for the first time.

The Princeton LNTR, 1963

Princeton LNTR, Nelson et al 1963

The next major design step was undertaken by Nelson et al at Princeton’s Dept. of Aeronautical Engineering in 1963, under contract by NASA. This was a far more in-depth study than the proposal by McCarthy, and looked to address many of the challenges that the original design faced.

Perhaps the most notable change was the shift from a single large fuel element to multiple smaller ones, arranged in a hexagonal matrix for maximum fuel element packing. This does a couple of things:

  1. It homogenizes the reactor more. While heterogeneous (mixed-region) reactors work well, for a variety of reasons it’s beneficial to have a more consistent distribution of materials through the core – mainly for neutronic properties and ease of modeling (this is 1963, MCNP in a heterogeneous core using a slide rule sounds… agonizing).
  2. Given a materially limited, fixed specific impulse (see the Fuel Materials Constraints section for more in-depth discussion on this) NTR, the thrust is proportional to the total surface area of the fuel/propellant interface. By using multiple fuel elements (which they call vortices), the total available surface area increases in the same volume, increasing the thrust without compromising isp (this also implies a greater specific power, another good thing in an NTR).

This was a thermal (0.37 eV) neutron spectrum reactor, fueled by a mix of UC2 and ZrC, varying the dilution level for greater moderation and increased thermal limits. It was surrounded by a 21 cm reflector of beryllium (a “standard reflector”).

From there, the basic geometry of the reactor, from the number of fuel elements and their fueled thickness, to the core diameter and volume (the length was at a fixed ratio compared to the radius), to the shape, velocity, and number of bubbles (as well as vapor entrainment losses of the fuel material) were studied.

This was a fairly limited study, despite its length, due to the limitations of the resources available. Transients and reactor kinetics were specifically excluded from this study, the hydrogen was again replaced with vacuum in calculations, the temperature was assumed to be higher than possible due to vapor entrainment problems (4300 K, instead of 3600 K at 10 atm, 3800 at 30 atm) the chamber pressure was limited to only >1 atm, and age-diffusion theory calculations only give results within an order of magnitude… but it’s still one of the most thorough study of LNTRs I’ve found, and the most researched bubbler architecture. They pointed out the potential benefits of the use of 233U, or a larger but neutronically equivalent volume of 232Th (turning the reactor into a thermal breeder), in order to improve the overall vaporization characteristics, but this was not included in the study.

Barrett LNTR, 1964

The next year, W. Louis Barrett presented a variation of the Princeton LNTR at the AIAA Conference. The main distinction between the two designs was the addition of zirconium hydride in the areas between the fuel elements and the outer reflector, and presented the first results from a study being conducted on the bubble behavior in the fuel (being conducted at Princeton at the time). The UC2/ZrC fuel was the same, as were the number of fuel elements and reactor dimensions. The author concluded that a specific impulse of 1500-1550 seconds was possible, with a T/W of 1 at 100 atm, with thrust not being limited by heat transfer but by available flow area.

Below are the two relevant graphs from his findings: the first is the point at which the fissile fuel itself would end up becoming captured by the passing gas, and the second looks at the maximum specific impulse any particular fissile fuel could theoretically offer. The image for the McCarthy reactor above was from the same paper.

Final Work: Bubbles are Annoying

For this reactor to work, the heat must be adequately transferred from the fuel element to the propellant as it bubbles through the fuel mass radially. The amount of heat that needs to be removed, and the time and distance that it can be removed in, is a function of both the fuel and the bubbles of H2.

Sadly, the most comprehensive study of this has never been digitized, but for anyone who’s able to get documents digitized at Princeton University and would like to help make the mechanics of bubbler-type LNTRs more accessible, here’s the study: Liebherr, J.F., Williams, P.M., and Grey, J., “Bubble Motion Studies for the Liquid Core Nuclear Rocket,” Princeton University Aeronautical Engineering Report No. 673, December 1963. Apparently you can check it out after you can convince the librarians to excavate it, based on their website: https://catalog.princeton.edu/catalog/1534764.

McGuirk 1972

Here, a clear plastic housing was constructed which consisted of two main layers: an outer, solid casing which formed the outer body of the apparatus, and a perforated, inner cylinder, which simulated the fuel element canister. Water was used as the fuel element analog, and the entire apparatus was spun along its long axis to apply centrifugal acceleration to the water at various rotation rates. Pressurized air (again, at various pressures) was used in place of the hydrogen coolant. Stroboscopic photography was used to document bubble size, shape, and behavior, and these behaviors were then used to calculate the potential thermal exchange, vapor entrainment, and other characteristics of the behavior of this system.

One significant finding, based on Gray’s reporting, though, is that there’s a complex relationship between the dimensions, shape, velocity, and transverse momentum of the bubbles and their thermal uptake capacity, as well as their vapor entrainment of fuel element components. However, without being able to read this work, I can only hope someone can make this work accessible to the world at large (and if you’ve got technical knowledge and interest in the subject, and feel like writing about it, let me know: I’m more than happy to have you write a blog post on here on this INSANELY complex topic).

The last reference to a bubbler LNTR I can find is from AIAA’s Engineering Notes from May 1972 by McGuirk and Park, “Propellant Flow Rate through Simulated Liquid-Core Nuclear Rocket Fuel Bed.” This paper brings up a fundamental problem that heretofore had not been addressed in the literature on bubblers, and quite possibly spelled their death knell.

Every study until this point greatly simplified, or ignored, two phase flow thermodynamic interactions. If you’re familiar with thermodynamics, this is… kinda astounding, to be honest. It also leads me to a diversion that could be far longer than the two pages that this report covers, but I won’t indulge myself. In short, two phase flow is used to model the thermal transfer, hydro/gasdynamic properties, and other interactions between (in this case) a liquid and a gas, or a melting or boiling liquid going through a phase change.

This is… a problem, to say the least. Based on the simplified modeling, the fundamental thermal limitation for this sort of reactor was vapor entrainment of the fuel matrix, reducing the specific impulse and changing he proportions of elements in the matrix, causing potential phase change and neutronics complications.

This remains a problem, but is unfortunately not the main thermal limitation of this reactor, rather it was discovered that the amount of thermal rejection available through the bubbling of the propellant through the fuel is not nearly as high as was expected at lower propellant flow rates, and higher flow rates led to splattering of the bubbles bursting, as well as unstable flow in the system. We’ll look at the consequences of this later, but needless to say this was a major hiccup in the development of the bubbler type LNTR.

While there may be further experimentation on the bubbler type LNTR, this paper came out shortly before the cancellation of the vast majority of astronuclear funding in the US, and when research was restarted it appears that the focus had shifted to radiator-type LNTRs, so let’s move on to looking at them.

Bubbler-Specific Constraints

Fuel Element Thickness and Heat Transfer

One of the biggest considerations in a bubbler LNTR is the thickness of the fuel within each fuel canister. The fundamental trade-off is one of mechanical vs thermodynamic requirements: the smaller the internal radius at the fuel element’s interior surface, the higher the angular velocity has to be to maintain sufficient centrifugal force to contain the fuel, btu also the greater time and distance the bubbles are able to collect heat from the fuel.

In the Princeton study, the total volume within the fuel canister was roughly equally divided between fuel and propellant to achieve a comfortable trade-off between fuel mass, reactor volume, and thermal uptake in the propellant. In this case, they included the volume of the propellant as it passed through the fuel to be part of the central annulus’ volume, which eases the neutronic calculations, but also induces a complication in the actual diameter of the central void: as propellant flow increases, the void diameter decreases, requiring more angular momentum to maintain sufficient centrifugal force.

A thinner fuel element, on the other hand, runs into the challenge of requiring a greater volume of propellant to pass through it to remove the same amount of energy, but an overall lower temperature of the propellant that is used. This, in turn, reduces the propellant’s final velocity, resulting in lower specific impulse but higher thrust. However, another problem is that the fluid mixture of the propellant/fuel can only contain so much gas before major problems develop in the behavior of the mixture. In an unpublished memorandum from 1963 (“Some Considerations on the Liquid Core Reactor Concept,” Mar 23), Bussard speculated that the maximum ratio of gas to fuel would be around 0.3 to 0.4; at this point the walls of the bubbles are likely to merge, converting the fuel into a very liquidy droplet core reactor (a concept that we’ll discuss in a future blog post), as well as leading to excess splattering of the fuel into the central void of the fuel element. While some sort of recapture system may be possible to prevent fuel loss, in a classic bubbler LNTR this is an unacceptable situation, and therefore this type of limitation (which may or may not actually be 0.3-0.4, something for future research to examine) intrinsically ties fuel element thickness to maximum propellant flow rates based on volume.

There are some additional limits here, as well, but we’ll discuss those in the next section. While the propellant will gain some additional power through its passage out of the fuel element and toward the nozzle, as in the radiator type LNTR, this will not be as significant as the propellant is entering along the entire length fuel element.

Bubble Dynamics

This is probably the single largest problem that a bubbler faces: the behavior of the bubbles themselves. As this is the primary means of cooling the fuel, as well as thermalizing the propellant, the behavior of these bubbles, and the ability of the propellant stream to control the entirety of the heat generated in the fuel, is of absolutely critical importance. We looked briefly in the last section at the impacts of the thickness of the fuel, but what occurs within that distance is a far more complex topic than it may appear at first glance. With advances in two phase flow modeling (which I’m unable to accurately assess), this problem may not be nearly as daunting as it was when this reactor was being researched, but in all likelihood this set of challenges is perhaps the single largest reason that the bubbler LNTR disappeared from the design literature when it did.

The other effect that the bubbles have on the fuel is that they are the main source of vapor entrainment of fuel element materials in a bubbler, since they are the liquid/gas interface that occurs for the longest, and have the largest relative surface area. We aren’t going to discuss this particular dynamic to any great degree, but the behavior of this interaction compared to inner surface interactions will potentially be significant, both due to the fact that these bubbles are the longest-lived liquid/gas interaction by surface area and are completely encircled by the fuel itself while undergoing heating (and therefore expansion, exacerbated by the decreasing pressure from the centrifugal acceleration gradient). One final note on this behavior: it may be possible that the bubbles may become saturated with vapor during their thermalization, preventing uptake of more material while also increasing the thermal uptake of energy from the fuel (metal vapors were suggested by Soviet NTR designers, including Li and NaK, to deal with the thermal transparency of H2 in advanced NTR designs).

The behavior of the bubbles depends on a number of characteristics:

  1. Size: The smaller the bubble, the greater the surface area to volume ratio, increasing the amount of heat the can be absorbed in a given time relative to the volume, but also the less thermal energy that can be transported by each bubble. The size of the bubbles will increase as they move through the fuel element, gaining energy though heat, and therefore expanding and becoming less dense.
  2. Shape: Partially a function of size, shape can have several impacts on the behavior and usefulness of the bubbles. Only the smallest bubbles (how “small” depends on the fluids under consideration) can retain a spherical shape. The other two main shape classifications of bubbles in the LNTR literature are oblate spheroid and spherical cap. In practice, the higher propellant flow rates result in the largest, spherical cap-type bubbles in the fuel, which complicate both thermal transfer and motion modeling. One consequence of this is that the bubbles tend to have a high Reynolds number, leading to more turbulent behavior as they move through the fuel mass. Most standard two-phase modeling equations at the time had a difficult time adequately predicting the behavior of these sorts of bubbles. Another important consideration is that the bubbles will change shape to a certain degree as they pass through the fuel element, due to the higher temperature and lower centrifugal force being experienced on them as they move into the central void of the fuel element.
  3. Velocity: A function of centrifugal force, viscosity of the fuel, initial injection pressure of the propellant, density of the constituent gas/vapor mix, and other factors, the velocity of a bubble through the fuel element determines how much heat – and vapor – can be absorbed by a bubble of a given size and shape. An increase in velocity also changes the bubble shape, for instance from an oblate spheroid to a spherical cap. One thing to note is that the bubbles don’t move directly along the radius of the fuel element, both oscillation laterally and radially occur as the shape deforms and as centrifugal, convective, and other forces interact with the bubble; whether this effect is significant enough to change the necessary modeling of the system will depend on a number of factors including fuel element thickness, convective and Coriolis behavior in the fuel mass, bubble Reynolds number, and angular velocity of the fuel element,
  4. Distribution: One concern in a bubbler LNTR is ensuring that the bubbles passing through the fuel mass don’t combine into larger conglomerations, or that the density of bubbles results in a lack of overall cohesion in the fuel mass. This means that the distribution system for the bobbles must balance propellant flow rate, bubble size, velocity, and shape, non-vertical behavior of the bubbles, and the overall gas fraction of the fuel element based on the fuel element design being used.

As mentioned previously, the final paper on the bubbler I was able to find looked at the challenges of bubble dynamics in a simulated LNTR fuel element; in this case using water and compressed air. Several compromises had to be made, leading to unpredictable behavior of the propellant stream and the simulated fuel behavior, which could be due to the challenges of using water to simulate ZrC/UC2, including insufficient propellant pressure, bubble behavior irregularities, and other problems. Perhaps the most major challenge faced in this study is that there were three distinct behavioral regimes in the two phase system: orderly (low prop pressure), disordered (medium prop pressure), and violent (high prop pressure), each of which was a function of the relationship between propellant flow and centrifugal force being applied. As suspected, having too high a void fraction within the fuel mass led to splattering, and therefore fuel mass loss rates that were unacceptably high, but the point that this violent disorder occurred was low enough that it was not assured that the propellant might not be able to completely remove all the thermal energy from the fuel element itself. If the energy level of each fuel element is reduced (by reducing the fissile component of the fuel while maintaining a critical mass, for instance), this can be compensated for, but only by losing power density and engine performance. The alternative, increasing the centrifugal force on the system, leads to greater material and mechanical challenges for the system.

Adequately modeling these characteristics was a major challenge at the time these studies were being conducted, and the number of unique circumstances involved in this type of reactor makes realistic modeling remain non-trivial; advances in both computational and modeling techniques make this set of challenges more accessible than in the 1960s and 70s, though, which may make this sort of LNTR more feasible than it once was, and restarting interest in this unique architecture.

These constraints define many things in a bubbler LNTR, as they form the single largest thermodynamic constraint on the engine. Increasing centrifugal force increases the stringency for both the fuel element canister (with incorporated propellant distribution system), mechanical systems to maintain angular velocity for fuel containment, maximum thrust and isp for a given design, and other considerations.

Suffice to say, until the bubble behavior, and its interactions with the fuel mass, can be adequately modeled and balanced, the bubbler LNTR would require significant basic empirical testing to be able to be developed, and this limitation was probably a significant contributor to the reason that it hasn’t been re-examined since the early-to-mid 1970s.

The “Restart Problem”

The last major issue in a bubbler-type design is the “restart problem”: when the reactor is powered down, there will be a period of time when the fuel is still molten, requiring centrifugal containment, but the reactor being powered down allows for the fuel to be pressed into the pores of the fuel element canister, blocking the propellant passages.

One potential solution for the single fuel element design was proposed by L. Crocco, who suggested that the fuel material is used for the bubbling structure itself. When powered up, the fuel would be completely solid, and would radiate heat in all directions until the fuel becomes molten [ed. Note: according to Crocco, this would occur from the inner surface to the outer one, but I can’t find backup for that assumption of edge power peaking behavior, or how it would translate to a multi-fuel-element design], and propellant would be able to pass through the inner layers of the fuel element once the liquid/solid interface reached the pre-drilled propellant channels in the fuel element.

Another would be to continue to pass the hydrogen propellant through the fuel element until the pressure to continue pumping the H2 reaches a certain threshold pressure, then use a relief valve to vent the system elsewhere while continuing to reject the final waste heat until a suitable wall temperature has been reached. This is going to both make the fuel element less dense, and also result in a lower fuel element density near the wall than at the inner surface of the fuel element. While this could maybe [ed. Note: speculation on my part] make it so that the fuel is more likely to melt from the inner surface to the outer one, the trapped H2 may also be just enough to cause power peaking around the bubbles, allow chemical reactions to occur during startup with unknown consequences, and other complications that I couldn’t even begin to guess at – but the tubes would be kept clear.

Wall Material Constraints

Other than the “restart problem,” additional constraints apply to the wall material. It needs to be able to handle the rotational stresses of the spinning fuel element, be permeable to the propellant, and able to withstand rather extreme thermal gradients: on one side, gaseous hydrogen at near-cryogenic temperatures (the propellant would have already absorbed some heat from the reactor body) to about 6000 K on the inside, where it comes in contact with the molten fuel.

Also, the bearings holding the fuel element will need to be designed with care. Not only do they need to handle the rather large amount of thermal expansion that will occur in all directions during reactor startup, they have to be able to deal with high rotation rates throughout the temperature range.

The Paths Not (Yet?) Taken

Perhaps due to the early time period in which the LNTR was explored, a number of design options don’t seem to have been explored in this sort of reactor.

One option is neutron moderator. Due to the high thermal gradients in this reactor, ZrH and other thermally sensitive moderators could be used to further thermalize the neutron spectrum. While this might not be explicitly required, it may help reduce the fissile requirements of the reactor, and would not be likely to significantly increase reactor mass.

A host of other options are possible as well, if you can think of one, comment below!

Diffuser LNTR

The other option was brought up by Michael Turner at Project Persephone, in regards to the vapor entrainment and restart problem issues: what if you get rid of the holes in the walls of the fuel element, and the bubbles through the fuel element, altogether? As we saw when discussing Project Rover, hydrogen gets through EVERYTHING, especially hot metals. This diffusion process is done through individual molecules, not through bubbles, meaning that the possibility of vapor entrainment is eliminated. The down side is that the propellant mass flow will be extremely reduced, resulting in a higher-isp (due to the ability to increase fuel temp because the vapor losses are minimized), much-lower-thrust reactor than those designed before. As he points out, this may be able to be mixed with bubbles for a high-thrust, lower-isp mode, if “shutters” on the fuel element outer frit were able to be engineered. Another possible requirement would be to reduce the fissile component density of the fuel to match the power output to the hydrogen flow rates, or to create a hybrid diffuser/radiator LNTR to balance the propellant flow and thermal output of the reactor.

I have not been able to calculate if this would be feasible or not, and am reasonably skeptical, but found it an intriguing possibility.

Conclusion

The bubbler liquid nuclear thermal rocket is a fascinating concept which has not been explored nearly as much as many other advanced NTR designs. The advantage of being able to fully thermalize the propellant to the highest fuel element temperature while maintaining cryogenic temperatures outside the fuel element is a rarity in NTR design, and offers many options for structures outside the fuel elements themselves. After over a decade of research at Princeton (and other centers), the basic research on the dynamics of this type of reactor has been established, and with the computational and modeling capabilities that were unavailable at the time of these studies, new and promising versions of this concept may come to light if anyone chooses to study the design.

The problems of vapor entrainment, fissile fuel loss, and restarting the reactor are significant, however, and impact many area of the reactor design which have not been addressed in previous studies. Nevertheless, the possibility remains that this drive may one day indeed make a useful stepping stone from the solid-fueled NTRs of tomorrow to the advanced NTRs of the decades ahead.

References

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

The Liquid Core Nuclear Rocket, Grey 1965 (pg 92) https://permalink.lanl.gov/object/tr?what=info:lanl-repo/lareport/LA-03229-MS

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

Propellant Flow Rate through Simulated Liquid-Core Nuclear Rocket Fuel Bed, McGuirk and Park 1972 https://arc.aiaa.org/doi/abs/10.2514/3.61690?journalCode=jsr

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
Development and Testing RHU RTGs Test Stands

ESA’s RTG Program: Breaking New Ground

Hello, and welcome back to Beyond NERVA! I would like to start by apologizing for the infrequency in posting recently. My wife is currently finishing her thesis in wildlife biology (multispecies occupancy, or the combination of statistical normalization of detection likelihood with the interactions between various species in an ecosystem to determine inter-species interactions and sensitivities), which has definitely made our household more hectic in the last few months. She should defend her thesis soon, and things will return to somewhat normal afterward, including a return to more frequent posting.

Today, we return to radioisotope thermoelectric generators (RTGs), which have once again been in the news. This time, the new is on a happier note than the passing of one of the pioneers in the field: the refining of a different type of fuel for RTGs: Americium 241. This work was done at the United Kingdom’s National Nuclear Laboratory Cumbria Lab by a team including personnel from the University of Leicester, and announced on May 3rd. This material was isolated out of the UK’s nuclear weapons stockpile of plutonium for nuclear warheads, which is something that we’ll look at more in the post itself.

A quick note on nomenclature and notation, since this is something that varies a bit: the way I learned to notate isotopes of an element (or nuclei of a given element with different numbers or neutrons) is as follows, and is what I generally use. If the element is spelled out, the result is [Element name][atomic mass of the nucleus](isomer state – if applicable); if it’s abbreviated it’s [atomic mass](isomer)[element symbol]. An isomer shows the energetic state of the nucleus: if gamma rays are absorbed by the nucleus, then a nucleon can jump to a higher energy state, just as electrons do for lower-wavelength (and lower energy) photons, and for this post it may not matter, but will come up with this element in the future. This means that Plutonium 238 will become 238Pu, and Americium 242(m) becomes 242(m)Am – I chose that because it’s a long-lived isomer that we will return to at some point due to its incredible usefulness in astronuclear design and reactor geometry advantages.

In this post, we’ll take a brief look at the fuels we currently use, as well as this “new” fuel, and one or two more that have been used in the past.

Radioisotope Power Source Fuels: What Are They, and How Do they Work?

Radioisotopes release energy in proportion to the instability of the isotope of whatever element is being used. This level of instability is a very complex issue, and how they decay largely is determined by where they fall in relation to the Valley of Stability, or the region in the Chart of the Isotopes where the number of protons and the number of neutrons balances the strong nuclear and electroweak forces in the nucleus. We aren’t going to go into this in too much detail, but the CEA did an excellent video on this topic which is highly recommended viewing if you want more information on the mechanics of radioactive decay and its different manifestations:


For our purposes, the important things to consider about radioactive decay are: how much energy is released in a given amount of time, and how easy is that energy to harness into kinetic energy, and therefore heat, while minimizing the amount of unharnessable radiation that could potentially damage sensitive components in the payload of the spacecraft. The amount of energy released in a given time is inversely proportional to the half-life of the isotope: the shorter the half-life, the more energy is released in a given time, but as a consequence the energy is being released at a faster rate and therefore the fuel will produce useful energy levels for a shorter period of time. How much of that energy is useful is determined by the decay chain of the isotope in question, which shows the potential decay mechanisms that the isotope will go through, how much energy is released in each decay, and how what the subsequent “daughter” isotopes are – with their decay mechanisms as well. The overall energy release from a particular isotope has to take all of these daughters, granddaughters, etc into consideration in order to calculate the total energy release, and how useful that energy is. If it’s mostly (or purely) alpha radiation, that’s ideal since it’s easily converted to kinetic energy, and each decay releases a proportionally larger amount of energy on average due to the high relative mass of the alpha particle. Beta radiation is not as good, but still easily shieldable (which converts the energy into kinetic energy and therefore heat), and so while not as ideal for heating is still acceptable.

For an RTG designer, the biggest problem (in this area at least) is gamma radiation. Not only is this very hard to convert into useful energy, but it can damage the payload and components of a spacecraft. This is because of the way gamma radiation interacts with matter: the wavelength is so short that it is absorbed most efficiently by very dense elements with a high number of nucleons (protons and neutrons), after which that nucleon jumps to a higher energy state (an isomer), and usually just spits it back out at a lower energy state. This process is repeated until there isn’t enough energy in the short wavelength photon to actually do anything meaningful.

Diverted particle (blue) emitting a photon (green) through brehmsstrahlung. Image Wikipedia

Unfortunately, radioactive decay isn’t the only way to produce gamma rays, you also have to deal with one of the key concepts in radiation shielding which always ties my fingers in knots whenever I try and type it (my tongue just says “nah, not even going to try”): Brehmsstrahlung. This is also known by the name “braking radiation,” and is close to the concept of “cyclotron radiation.” Basically, when a massive, high energy particle is diverted from its original course, two things happen: it slows down, and that slowing down means that energy has to be conserved somehow. In the case of quantum mechanics, the elastic collisions between the particle and whatever force acts on it (in practice the electromagnetic field it interacts with) creates kinetic energy (heat) if it hits a particle, and in every case (including magnetic containment of the particle) a photon is produced. This photon’s wavelength is proportional to the energy the particle originally has, the degree that its direction is changed, and how much energy is lost through mechanisms other than elastic collisions with other nucleons – and it’s never able to be fully slowed by elastic collisions for very complex quantum mechanical reasons. The amount converted to brehmstrahlung is proportional to the initial velocity of the particle in the case we’re talking about (magnetic containment and diversion of radiation isn’t the subject of this blog post after all). This means that if you have a high energy alpha particle emitter, and a low energy alpha emitter, the high energy alpha emitter will create more gamma radiation, meaning that it’s more problematic in terms of energy efficiency of a radioisotope for heat production.

Another concern is what chemical form the radioisotope is used in the fuel element itself. This defines a number of things, including how well any heat produced is transmitted to where it’s useful (or if the heat isn’t transferred efficiently enough, the thermal failure point of the fuel), what kind of nuclear interactions will occur within the fuel, the chemical impact of the elemental composition of the fuel changing through the radioactive decay that is the entire point of using these materials, how much the particles are slowed within the fuel itself (and any subsequent brehmsstrahlung), how much of the radiation that comes from the radioisotope decay is shielded by the fuel itself, and a whole host of other characteristics that are critical to the real-world design of a radioisotope power source.

Now that we’ve (briefly) looked at how radioisotopes are used as power sources, let’s look at what options are available for fueling these systems, which ones have been used in the past, and which ones may be used in the future.

How Are Radioisotopes Used in Space?

The best known use of radioisotopes in space is the use of radioisotope thermoelectric generators, RTGs. RTGs are devices that use the thermoelectric effect, paired with a material containing an isotope going through radioactive decay, to produce electricity. While these are the most well known of the radioisotope power sources, they aren’t the only ones, and in fact they are made up of an even more fundamental type of power source: a radioisotope heating unit, or RHU.

These RHUs are actually more common than RTGs, since often the challenge isn’t getting power to your systems but keeping them from freezing in the extreme cold of space (when there isn’t direct sunlight, in which case it’s extreme heat which is the problem). In that case, a small pellet of radioisotope is connected to a heat transfer mechanism, which could be simple thermal conduction through metal components or the use of heat pipes, which use a wick to transfer a working fluid from a heat source to a cold sink through the boiling and then condensation of the working fluid. Heat pipes have become well known in the astronuclear community thanks to the Kilopower reactor, but are common components in electronics and other systems, and have been used in many flight systems for decades.

RTGs use RHUs as well, as the source of their heat to produce electricity. In fact, the design of a common RHU for both RTG and spacecraft thermal management requirements was a focus of NASA for years, resulting in the General Purpose Heating Unit (GPHU) and its successor variations by the same name. This is important to efficiently and reliably manufacture the radioisotope fuel needed for the wide variety of systems that NASA has used in its spacecraft over the decades. While RTGs are the focus of this technology, we aren’t interested in either the power conversion system or the heat rejection system that make up two of the four main systems in an RTG, so we won’t delve into the details of these systems in particular. Rather, our focus is on the RHU radioisotope fuel itself, and the shielding requirements that this fuel mandates for both spacecraft and payload functionality (the other two major systems of an RTG). Because of this, for the rest of the post we will be discussing RHUs rather than RTGs for the most part (although mentions of RTG implications will be liberally scattered throughout).

Another potential use for radioisotopes in space seems to be something that is rare in discussions of space systems, but is common on Earth: as a source of well-characterized and -understood radiation for both analysis and manufacture of materials and systems. Radioisotope tracking is common in everything from medicine to agriculture, and radioactive analysis is used on everything from ancient artifacts to modern currency. The ISS has had experiments that use mildly radioactive isotopes to analyze the growth of living beings and microbes in microgravity environments, for instance, a very common use in agriculture to analyze nutrient uptake in crops, and a variation of a technique used in medicine to analyze everything from circulatory flow to tumor growth in nuclear medicine. X-ray analysis of materials is also a common method used in high-end manufacturing, and as groups such as Made in Space, Relativity Space and Tethers Unlimited continue exploring 3d printing and ISRU in microgravity and low gravity environments, this will be an invaluable tool. However, this is a VERY different subject, and so this will be where we leave this biological analysis and technological development technology and move to the most common use of radioisotopes: to provide heat to do some sort of work.

RHU Fuel: The Choices Available, and the Choices Made

Most RHUs, for space applications at least, are made from 238Pu, which is an isotope of plutonium that is not only not able to undergo fission, but in fairly minute quantities will render Pu meant for nuclear weapons completely unusable. In the early days of the American (and possibly Soviet) nuclear weapons program, small amounts of this isotope were isolated from material meant for nuclear weapons, but as time went on and the irradiation process became more efficient for producing weapons (which is very different from producing power), the percentage of 238Pu dropped from the single digits to insignificant. By this time, though, the Mound Laboratories in Miamisburg, Ohio had become very interested in the material as a source of radioactive decay heat for a variety of uses. These uses ranged from spacecraft to pacemakers (yes, pacemakers… they were absolutely safe unless the person was cremated, and the fact that the removal of said pacemakers couldn’t be guaranteed is what killed the program). This doesn’t mean that they didn’t investigate the rest of what we’ll be looking at as well, but much of their later focus was on 238Pu.

The advantages of 238Pu are significant: it only undergoes alpha decay with an exceptionally small chance of spontaneous fission (which occurs in the vast majority, if not all, isotopes of uranium and above on the periodic table), it has a good middle-of the road half-life (87.84 years) for long term use as a power source for missions lasting decades (as the majority of outer solar system missions require just to get tot he target destination and provide useful science returns), it has a good amount of energy released during alpha decay (5.59 MeV), and its daughter – 234U – is incredibly stable, with a half-life so long that it basically won’t undergo any significant decay until long after the fuel is useless and the spacecraft has fulfilled its mission centuries ago (245,500 years). The overall radiation power for 238Pu is 0.57 W/g, which is one of the best power densities available for long lived radioisotopes.

Additionally, the fuel used PuO2, is incredibly chemically stable, and when 238Pu becomes 234U, those two oxygen atoms are easily able to reattach to the newly formed U to form UO2 – the same fissile fuel form used in most nuclear reactors (although since it’s 234 rather than 235, it would be useless in a reactor), which is also incredibly chemically stable. Finally, the fuel itself is largely self-shielding once encased in the usual iridium clad to protect the fuel during handling and transport, meaning that the minimal amount of gamma radiation coming off the power source is only a major consideration for instruments looking in the x-ray and gamma bands of the EM spectrum causing noise, rather than significant material or electronic degradation.

238Pu is generated by first making a target of 237Np, usually through irradiation of a precursor material in a nuclear reactor. This target is then exposed to the neutrons produced by a nuclear reactor for a set length of time, and then chemically separated into 238Pu. Due to the irradiation time and energy level, the final result is almost pure 238Pu (close enough for the DOE and NASA’s purposes, anyway), which can then be turned into the ceramic pellet and encased in the clad material. This is then mated to the system that the spacecraft will use the power source for, usually just before spacecraft integration into the launch vehicle. Due to the rigid and strict interpretations of radiation protection, this is an incredibly complex and challenging process – but one that is done on a fairly regular basis. The supply of 238Pu has been a challenge from a bureaucratic perspective, but a recent shakeup of the American astronuclear industry due to a GAO report released last year offers hope for sufficient supply for American flight systems.

This is far from the only radioisotope source used for RHUs – in fact, it wasn’t even the first. Many early US designs used strontium 90, as did many Soviet designs. The first RTG to fly, SNAP-3, used this power source, as did multiple nautical navigation buoys, Soviet lighthouses along their northern coast, and many other systems. This isotope has a half-life of 28.79 years, which means that it’s more useful for shorter-lived systems than 238Pu, but is still a long enough half-life that it’s still a useful radioisotope source. The disadvantage is that it decays via beta decay (at 546 KeV), to 90Y, which is less efficient in converting the radioactive decay into heat. However, this isotope goes through another decay as well. 90Y only has a 2.66 day (!) half-life, ejecting a 2.28 MeV beta particle, resulting in a final decay product of 90Zr, which is radiologically stable. This means that the total energy release is 2.826 MeV, through two beta emissions. The overall energy release in the initial decay is attractive, at 0.9 W/g, however, so either sufficient shielding (of sufficiently high thermal conductivity) to convert this beta radiation into kinetic energy, or a different power conversion system than one that is thermally based is a potential way to increase the efficiency of these systems.

Finally, we come to the one that scares many, and has a horrible reputation in international politics, polonium 210. This is most famous for being used as a poison in the case of Alexander Litvinenko, a Russian defector who was poisoned with Po in his tea (and a whole lot of other people being exposed), but much of the effectiveness is due to chemical toxicity, not radiotoxicity. 210Po has an incredibly short half-life, of only 138.38 days. This is only acceptable for short mission times, but the massive amount of heat generated is still incredibly attractive for designers looking for very high temperature applications. Designs such as radioisotope thermal rockets (where the decay heats hydrogen or other propellant, much like an NTR driven by decay rather than a reactor) that have their efficiency defined purely by temperature can gain significant advantages thanks to this high decay energy. 210Po decays via alpha decay into lead 206, a stable isotope, so there are no complexities from daughter products emitting additional radiation, and a 5.4 MeV alpha particle carries quite a lot of energy.

Other isotopes are also available, and there’s a fascinating table in one of my sources that shows the ESA decision process when it comes to radioisotope selection from 2012:


241Am: The Fuel in the News

Americium-241 is the big news, however, and the main focus of this post. 241Am is produced during irradiation of uranium, which goes through neutron capture during irradiation to 241Pu, which is one of the isotopes that degrades the usefulness of weapons-grade 239Pu. It decays in a very short time (14 day half-life) into 241Am, which is not useful as weapons-grade material, but is still useful in nuclear reactors. According to Ed Pheil, the CEO of Elysium Industries, the Monju sodium cooled fast reactor in Japan faced a restart problem because the 241Pu in the reactor’s fuel decayed before restarting the reactor, causing a reactivity deficit and delaying startup.

241Am decays through a 5.64 MeV alpha emission into 237Np, which in turn goes through a 4.95 MeV alpha decay with a half-life of 2.14 x 10^6 years. This means that the daughter is effectively radiologically stable. With its longer half-life of 433 years (compared to 88 years for 238Pu), 241Am won’t put out as much energy at any given time compared to 238Pu fuel, but the difference in power output as the mission continues will allow for a more steady power supply for the spacecraft. A comparison of beginning of life power to 20 year end of life power for a 238Pu RTG shows a reduction in power output of about 15%, compared to only about 3.5% for 241Am. This allows for more consistent power availability, and for extremely long range or long duration missions a greater amount of power available. However, in the ESA design documentation, this extreme longevity is not something that is examined, a curious omission on first inspection. This could be explained by two factors, however: mission component lifetime, which could be influenced by multiple factors independent of power supply, and the continuing high cost of maintaining a mission control team and science complement to support the probe.

Depending on what power level is needed (more on that in the next section about the ESA RTG design), and how long the mission is, the longer half-life could make 241Am superior in terms of useful energy released compared to 238Pu, and is one of the reasons ESA started looking at 241Am as the main focus of their RTG efforts.

Why the focus on weapons material in the description of production methods? Because that’s where the UK’s National Nuclear Laboratory gained their 241Am in the latest announcement. The NNL is responsible for production of all 241Am for Europe’s RTG programs, but doesn’t HAVE to produce the material from weapons stockpiles.

Fuel cycle using civilian power reactors, Summerer 2012

The EU reprocesses their fuel, unlike the US, and use the Pu to create mixed-oxide (MOX) fuel. If the Pu is chemically separated from the irradiated fuel pellets, then allowed to decay, the much shorter half-life of 241Pu compared to all of the others will lead to the ability to chemically separate the 241Am from the Pu fuel for the MOX. This could, in theory, allow for a steady supply of 241Am for European space missions. As to how much 241Am that would be available through reprocessing, this is a complex question, and one that I have not been able to explore sufficiently to give a good answer to how much 241Am would be available through reprocessing. Jaro Franta was kind enough to provide a pressurized water reactor spent fuel composition table, which provides a vague baseline:

However, MOX fuel generally undergoes higher burnup, and according to several experts the Pu is quickly integrated into fuel as part of the reprocessing of spent fuel. This could be to ensure weapons material is not lying around in Le Hague, but also prevents enough of a decay time to separate the 241Am – plus, as we see in 238Pu production, where the materials are fabricated in one place, separated in another, and made into fuel in a third, Le Hague and the Cumbria Laboratory are not only in different locations but different countries, and after this process the Pu is even more useful for weapons, this bureaucratic requirement makes the process of using spent nuclear fuel for 241Am production an iffy proposition at best. However, according to Summerer and Stephenson (referenced in one of the papers, but theirs is behind a paywall) the economical separation of 241Am from spent civilian fuel can be economical (I’m assuming due to the short half-life of 241Pu), so it seems like the problem is systemic, not technical.

241Am is used as RHU fuel in the form of Am2O3. This allows for very good chemical stability, as well as reasonable thermal transfer properties (for an oxide). This fuel is encased in a “multilayer containment structure similar to that of the general-purpose heat source (GPHS) system,” with thermal and structural trade-offs made to account for the different thermal profile and power level of the ESA RTG (which, as far as I can tell, doesn’t have a catchy name like “GPHS RTG” or “MMRTG” yet). Neptunium oxide most often takes the form of NpO2, meaning that a deficit of oxygen will occur over time in the fuel pellet. The implications of this are something that I am completely unable to answer, and something that is definitely distinct from the use of 238Pu, which then becomes 234U, both of which have two oxygen atoms in their most common oxide state. However, considering there’s a stoichiometric mismatch between the initial material, the partially-decayed material, and the final, fully-degenerate state of the fuel element. I know just enough to know that this is far from ideal, and will change a whole host of properties, from thermal conductivity to chemical reactivity with the clad, so there will be (potentially insignificant) other factors that have impacts on fuel element life from the chemical point of view rather than the nuclear one.

ESA’s RTG: the 241Am RTG

ESA has been interested in RTGs for basically its entire existence, but for reasons that I haven’t been able to determine for certainty, they have not investigated 238Pu production for their RTG fuel to any significant degree for that entire time. Rather, they have focused on 241Am. This comes with a trade-off: due to the longer half-life, at any given time less energy is available per unit mass (0.57 W/g for 238Pu vs 0.1 W/g for 241Am), but as previously noted there are a couple of threshold points at which this longer half-life is an advantage.

Mass to power design envelope, Ambrose et al 2013

The first advantage is in mission lifetime (assuming the radiochemistry situation is of minimal concern): the centuries-long half-life could allow for unprecedented mission durations, where overall mass budget for the power source is of less concern in mission design than longevity of the mission. The second comes to when a very small amount of power is used, and this is the focus of ESA’s RTG program. The GPHS used in American systems produces 250 Wt at the time of manufacture in no more than 1.44 kg. Sadly, it’s very difficult to determine what the ESA fuel element’s mass and specific composition, so a direct comparison is currently impossible. Based on a 2012 presentation, there were two parallel options being explored, with different specific power characteristics, but also different thermal conductivity – and more efficient thermal transport. One was the use of CERMET fuel, where the oxide fuel is encapsulated in a refractory metal (which was unspecified), manufactured by spark plasma sintering. This is a technology that we’ve examined in terms of fissile fuel rather than radioisotope fuel, but many of the advantages apply in both cases. In addition, the refractory metal provides internal brehmstrahlung shielding, potentially offsetting the need for robust clad materials, but this is potentially offset by the need to contain the radioisotope in the case of a launch failure, which requires a mechanical strength that may ensure that the reduction in radiation shielding requirements are rendered effectively moot. The second was a parallel of American RHU design, using a multi-layered clad structure, using refractory metals, carbon-based insulators, and carbon-carbon insulators. This seems to be the architecture which was ultimately chosen, after a contract with Areva TA in France and other partners (which I have not been able to find documentation on, if you have that info please post it in the comments below!).

Cross-section of 1st generation ESA RTG, Ambrosi et al 2013

ESA has two RTG designs that they’ve been discussing over the last two decades: an incredibly small one, of only 1 W of electricity, and a larger one, in the 10-50 We range. These systems are similar in that they are both starting from the same design philosophy, but at the same time different materials and design tradeoffs were required for the two systems, so they have evolved in different directions.

BiTe thermocouple unit, Summerer 2012

The 10-50 We RTG combines the 241Am RHU with a composite clad, surrounded by a bismuth telluride thermocouple structure. This is very similar to lead telluride TE converters in many ways, but is more efficient in lower operating temperatures. PbTe is also under investigation, developed and manufactured by Fraunhofer IPM in Germany, but BiTe seems to be the current technological forerunner in ESA RTG development. This seems to be commercially available, but sadly based on the large number of thermopiles (another name for TE converters) available, it is not clear which is being used.

The 1-1.5 We RTG is one that doesn’t seem to be explored in depth in the currently published literature. This power output is useful for certain applications, such as powering an individual sensor, but is a much more niche application. Details on this design are very thin on the ground, though, so we will leave this design at that and move on to the experiment performed – again, very little is available on the specifics, but the experiment WAS described in previous papers.

Gen 2 ESA RTG, Ambrosi 2015

This design went through an evolution into the model seen in the public announcement video. This is called the Gen 2 Flight System Design, and likely was introduced sometime in the 2015 timeframe. At the same time, the Radioisotope Heating Unit itself went to TRL 3.

On the fuel element fabrication side, after 241Am was selected in 2010 two phases of isotope production occurred. The first was from 2011 to 2013, and the second from 2013 to 2016.

This was the first test of that batch of RTG fuel to produce electricity, which is a major achievement, for which the entire team should be congratulated.

The Experiment in the News: What is All the Fuss Actually About?

Sadly, there’s no publicly available information about the Am-fueled test in the news. The promotional video and press release from NNL/UL provided only two pieces of information: 241Am fuel was used, and they lit a light bulb. No details about how much 241Am, its fuel form or clad, the thermocouples used, the power requirements of the light bulb… this was meant for maximum viral distribution, not for conveying technical information.

The best way to look at the test is by looking at preceeding tests. RTG design from the ground up takes time, and in the case of ESA and the NNL, this process was only funded for the last ten years. They have had to pick a fuel type, continue to go through selection on thermocouple type, and will be working to finalize the design of their flight system.

Electrically heated breadboard experiment, Ambrosi 2012

In 2012, an electrically heated breadboard experiment was conducted. The test used a cuboid form factor, rather than the more common cylindrical form, and was conducted in a liquid nitrogen cooled vacuum chamber. It was designed around a theoretical fuel element of Am2O3 which would provide 83 Wt of power to the thermocouple. This thermocouple, in the proposal phase, was either a commercially available BiTe or bespoke PbTe thermocouple, contained in a cover gas of argon. The maximum output of electrical power was 5 We, which is less than the minimum size of the “normal” 10 We RTG design. Depending on the fuel element used, it’s possible that the experiment was carried out at 83 Wt/ 5We to allow for maximum comparison between the electrically heated version and the radioisotope powered version, but the lack of RI-powered experiment data prevents us from knowing if this was the case, or whether the experiment was performed at 10 We (166 Wt?) due to manufacturing and fuel element design constraints.

The electrically heated design in the breadboard experiment demonstrated that in the 5-50 We power output range, a specific power of 2 We/kg is feasible. There are possibilities that this could be improved somewhat, but it provides a good baseline for the power range and the specific power that these units will provide. This paper (linked below, Development and Testing of Americium-241 Radioisotope Thermoelectric Generator: Concept Designs and Breadboard System, Ambrosi et al) notes that only at small power outputs can 241Am compete with 238Pu systems, but extended mission lifetime considerations were not addressed.


ESA and the University of Leicester continue to look at expanding the use of RTGs in the future. The focus for the thermocouples in the first flight design was in bismuth telluride TEGs as of 2015. They are also looking into Stirling convertors as well, continuing in the current drive to move away from the

It takes time to do the things that they’re attempting, there are few specialists in these areas (although the fundamental tasks aren’t difficult if you know what you’re doing, it takes a while to know the ins and outs of any sort of large-scale chemistry), and it requires a lot of research to verify that each step of the way is both safe and reliable without compromising efficiency.

I have reached out to Dr. Ambrosi at University of Leicester for additional information about this test. If I hear anything from him, I will add the information about this particular test to the page on this NTR system, which should release soon.

Conclusions: 241Am, Is It the RTG Fuel of the Future?

As with most things in astronuclear engineering, the choice of an RHU fuel is a very complex question, and one which has no simple answer. 238Pu remains the preferred long-duration RTG fuel for space missions in terms of specific power, but its expense, and the requirements as far as infrastructure for fabrication and manufacture provide a high cost barrier for new entrants into the use of RHUs. For Europe, this barrier to entry is considered unacceptable, and that has kept them out of the RTG-flying world community for the entirety of their history.

241Am, on the other hand, is available to nations that conduct reprocessing of spent nuclear fuel, such as signatories to the Euratom treaty (as well as the UK, who are withdrawing voluntarily from the treaty as part of Brexit… don’t ask me why, it’s not required), where reprocessing of spent nuclear fuel is normal practice. Similar challenges to bureaucratic roadblocks to significant production of 238Pu by the US can be seen in European production of 241Am, but the existence of significant reprocessing capabilities make it theoretically far more available. 241Am is also available commercially in the US, meaning that at least in one country the regulatory barriers to possession, and therefore cost, are significantly lower than 238Pu.

This choice, however, comes at a cost of halving the specific power available to RTG systems due to the lower specific power of the fuel, at least as the design is historically described. Optimization calculations by ESA and its partners, primarily the University of Leicester, show that in the 5-50 We range of electric output the impact on mission mass is minimal, and for very low power applications (1-1.5 We) it is superior. The increased availability, and lower cost of acquisition of sufficiently pure 241Am, offset the advantages of 238Pu from the European perspective, as it integrates into current industrial capabilities, which outweigh the engineering advantages of 238Pu for the organizations involved. Even so, this is an exciting development for deep space exploration nerds, and one that can’t be overstated.

While this was a fascinating experiment, and I will be trying to find more information, the significance of this experiment boils down to one thing: this is the first time, outside the US or Russia, that an RTG designed for spacecraft use produced electrical power. It opens up new mission opportunities for ESA, who have been hampered in deep space exploration by their lack of suitable RHU fuel, and offers hope for more missions, more science, and more discoveries in the future.

References

Isotope information for 241Am, Periodictable.com https://periodictable.com/Isotopes/095.241/index.html

Isotope information for 238Pu, Periodictable.com https://periodictable.com/Isotopes/094.238/index2.full.dm.prod.html

Isotope information for 90Sr, Periodictable.com https://periodictable.com/Isotopes/038.90/index2.full.dm.prod.html

Isotope information for 210Po, Periodictable.com https://periodictable.com/Isotopes/084.210/index2.full.dm.prod.html

241Am ESA RTG Design

Development and Testing of Americium-241 Radioisotope Thermoelectric Generator: Concept Designs and Breadboard System; Ambrosi et al 2012 https://www.lpi.usra.edu/meetings/nets2012/pdf/3043.pdf

Americium-241 Radioisotope Thermoelectric Generator Development for Space Applications, Ambrosi et al 2013 https://inis.iaea.org/collection/NCLCollectionStore/_Public/45/066/45066049.pdf

Nuclear Power Sources for Space Applications – a key enabling technology (slideshow), Summerer et al, ESA 2012 https://www.euronuclear.org/events/enc/enc2012/presentations/L-Summerer.pdf

Space Nuclear Power Systems: Update on Activities and Programmes in the UK, Ambrosi (University of Leicester) and Tinsley (National Nuclear Laboratory), 2015 http://www.unoosa.org/pdf/pres/stsc2015/tech-15E.pdf

Categories
Development and Testing Fission Power Systems Forgotten Reactors Nuclear Electric Propulsion Test Stands

Topaz International part II: The Transition to Collaboration


Hello, and welcome back to Beyond NERVA! Before we begin, I would like to announce that our Patreon page, at https://www.patreon.com/beyondnerva, is live! This blog consumes a considerable amount of my time, and being able to pay my bills is of critical importance to me. If you are able to support me, please consider doing so. The reward tiers are still very much up for discussion with my Patrons due to the early stage of this part of the Beyond NERVA ecosystem, but I can only promise that I will do everything I can to make it worth your support! Every dollar counts, both in terms of the financial and motivational support!

Today, we continue our look at the collaboration between the US and the USSR/Russia involving the Enisy reactor: Topaz International. Today, we’ll focus on the transfer from the USSR (which became Russia during this process) to the US, which was far more drama-ridden than I ever realized, as well as the management and bureaucratic challenges and amusements that occurred during the testing. Our next post will look at the testing program that occurred in the US, and the changes to the design once the US got involved. The final post will overview the plans for missions involving the reactors, and the aftermath of the Topaz International Program, as well as the recent history of the Enisy reactor.

For clarification: In this blog post (and the next one), the reactor will mostly be referred to as Topaz-II, however it’s the same as the Enisy (Yenisey is another common spelling) reactor discussed in the last post. Some modifications were made by the Americans over the course of the program, which will be covered in the next post, but the basic reactor architecture is the same.

When we left off, we had looked at the testing history within the USSR. The entry of the US into the list of customers for the Enisy reactor has some conflicting information: according to one document (Topaz-II Design History, Voss, linked in the references), the USSR approached a private (unnamed) US company in 1980, but the company did not purchase the reactor, instead forwarding the offer up the chain in the US, but this account has very few details other than that; according to another paper (US-Russian Cooperation… TIP, Dabrowski 2013, also linked), the exchange built out of frustration within the Department of Defense over the development of the SP-100 reactor for the Strategic Defense Initiative. We’ll look at the second, more fleshed out narrative of the start of the Topaz International Program, as the beginning of the official exchange of technology between the USSR (and soon after, Russia) and the US.

The Topaz International Program (TIP) was the final name for a number of programs that ended up coming under the same umbrella: the Thermionic System Evaluation Test (TSET) program, the Nuclear Electric Propulsion Space Test Program (NEPSTP), and some additional materials testing as part of the Thermionic Fuel Element Verification Program (TFEVP). We’ll look at the beginnings of the overall collaboration in this post, with the details of TSET, NEPSTP, TFEVP, the potential lunar base applications, and the aftermath of the Topaz International Program, in the next post.

Let’s start, though, with the official beginnings of the TIP, and the challenges involved in bringing the test articles, reactors, and test stands to the US in one of the most politically complex times in modern history. One thing to note here: this was most decidedly not the US just buying a set of test beds, reactor prototypes, and flight units (all unfueled), this was a true international technical exchange. Both the American and Soviet (later Russian) organizations involved on all levels were true collaborators in this program, with the Russian head of the program, Academician Nikolay Nikolayvich Ponomarev-Stepnoy, still being highly appreciative of the effort put into the program by his American counterparts as late as this decade, when he was still working to launch the reactor that resulted from the TIP – because it’s still not only an engineering masterpiece, but could perform a very useful role in space exploration even today.

The Beginnings of the Topaz International Program

While the US had invested in the development of thermionic power conversion systems in the 1960s, the funding cuts in the 1970s that affected so many astronuclear programs also bit into the thermionic power conversion programs, leading to their cancellation or diminution to the point of being insignificant. There were several programs run investigating this technology, but we won’t address them in this post, which is already going to run longer than typical even for this blog! An excellent resource for these programs, though, is Thermionics Quo Vadis by the Defense Threat Reduction Agency, available in PDF here: https://www.nap.edu/catalog/10254/thermionics-quo-vadis-an-assessment-of-the-dtras-advanced-thermionics (paywall warning).

Our story begins in detail in 1988. The US was at the time heavily invested in the Strategic Defense Initiative (SDI), which as its main in-space nuclear power supply was focused on the SP-100 reactor system (another reactor that we’ll be covering in a Forgotten Reactors post or two). However, certain key players in the decision making process, including Richard Verga of the Strategic Defense Initiative Organization (SDIO), the organizational lynchpin on the SDI. The SP-100 was growing in both cost and time to develop, leading him to decide to look elsewhere to either meet the specific power needs of SDI, or to find a fission power source that was able to operate as a test-bed for the SDI’s technologies.

Investigations into the technological development of all other nations’ astronuclear capabilities led Dr. Verga to realize that the most advanced designs were those of the USSR, who had just launched the two TOPOL-powered Plasma-A satellites. This led him to invite a team of Soviet space nuclear power program personnel to the Eighth Albuquerque Space Nuclear Power Symposium (the predecessor to today’s Nuclear and Emerging Technologies for Space, or NETS, conference, which just wrapped up recently at the time of this writing) in January of 1991. The invitation was accepted, and they brought a mockup of the TOPAZ. The night after their presentation, Academician Nikolay Nicolayvich Ponomarev-Stepnoy, the Russian head of the Topol program, along with his team of visiting academicians, met with Joe Wetch, the head of Space Power Incorporated (SPI, a company made up mostly of SNAP veterans working to make space fission power plants a reality), and they came to a general understanding: the US should buy this reactor from the USSR – assuming they could get both governments to agree to the sale. The terms of this “sale” would take significant political and bureaucratic wrangling, as we’ll see, and sadly the problems started less than a week later, thanks to their generosity in bringing a mockup of the Topaz reactor with them. While the researchers were warmly welcomed, and they themselves seemed to enjoy their time at the conference, when it came time to leave a significant bureaucratic hurdle was placed in their path.

Soviet researchers at Space Nuclear Power Symposium, 1991, image Dabrowski

This mockup, and the headaches surrounding being able to take it back with the researchers, were a harbinger of things to come. While this mockup was non-functional, but the Nuclear Regulatory Commission claimed that, since it could theoretically be modified to be functional (a claim which I haven’t found any evidence for, but is theoretically possible), and as such was considered a “nuclear utilization facility” which could not be shipped outside the US. Five months later, and with the direct intervention of numerous elected officials, including US Senator Pete Domenici, the mockup was finally returned to Russia. This decision by the NRC led to a different approach to importing further reactors from the USSR and Russia, when the time came to do this. The mockup was returned, however, and whatever damage the incident caused to the newly-minted (hopeful) partnership was largely weathered thanks to the interpersonal relationships that were developed in Albuquerque.

Teams of US researchers (including Susan Voss, who was the major source for the last post) traveled to the USSR, to inspect the facilities used to build the Enisy (Yenisey is another common spelling, the reactor was named after the river in Siberia). These visits started in Moscow, with Drs Wetch and Britt of SPI, when a revelation came to the American astronuclear establishment: there wasn’t one thermionic reactor in the USSR, but two, and the most promising one was available for potential export and sale!

These visits continued, and personal relationships between the team members from both sides of the Iron Curtain grew. Due to headaches and bureaucratic difficulties in getting technical documentation translated effectively in the timeframe that the program required, often it was these interpersonal relationships that allowed the US team to understand the necessary technical details of the reactor and its components. The US team also visited many of the testing and manufacturing locations used in the production and development of the Enisy reactor (if you haven’t read it yet, check out the first blog post on the Enisy for an overview of how closely these were linked), as well as observing testing in Russia of these systems. This is also the time when the term “Topaz-II” was coined by one of the American team members, to differentiate the reactor from the original Topol (known in the west as Topaz, and covered in our first blog post on Soviet astronuclear history) in the minds of the largely uninformed Western academic circles.

The seeds of the first cross-Iron Curtain technical collaboration on astronuclear systems development, planted in Albuquerque, were germinating in Russian soil.

The Business of Intergovernmental Astronuclear Development

During this time, due to the headaches involved in both the US and the USSR from a bureaucratic point of view (I’ve never found any information that showed that the two teams ever felt that there were problems in the technological exchange, rather they all seem to be political and bureaucratic in nature, and exclusively from outside the framework of what would become known as the Topaz International Program), two companies were founded to provide an administrative touchstone for various points in the technological transfer program.

The first was International Scientific Products, which from the beginning (in 1989) was made specifically to facilitate the purchase of the reactors for the US, and worked closely with the SDIO Dr. Verga was still intimately involved, and briefed after every visit to Russia on progress in the technical exchange and eventual purchase of the reactors. This company was the private lubricant for the US government to be able to purchase these reactor systems (for reasons too complex to get into in this blog post). The two main players in ISP were Drs Wetch and Britt, who also appear to be the main administrative driving force in the visits. The company gave a legal means to transmit non-classified data from the USSR to the US, and vice versa. After each visit, these three would meet, and Dr. Verga kept his management at SDIO consistently briefed on the progress of the program.

The second was the International Nuclear Energy Research and Technology corporation, known as INERTEK. This was a joint US-USSR company, involving the staff of ISP, as well as individuals from all of the Soviet team of design bureaus, manufacturing centers (except possibly in Talinn, but I haven’t been able to confirm this, it’s mainly due to the extreme loss of documentation from that facility following the collapse of the USSR), and research institutions that we saw in the last post. These included the Kurchatov Institute of Atomic Energy (headed by Academician and Director Ponomarev-Stepnoy, the head of the Russian portion of the Topaz International Program), the Scientific Industrial Association “LUCH” (represented by Deputy Director Yuri Nikolayev), the Central Design Bureau for Machine Building (represented by Director Vladmir Nikitin), and the Keldysh Institute of Rocket Research (represented by Director Academician Anatoli Koreteev). INERTEK was the vehicle by which the technology, and more importantly to the bureaucrats the hardware, would be exported from the USSR to the US. Academician Ponomarev-Stepnoy was the director of the company, and Dr Wetch was his deputy. Due to the sensitive nature of the company’s focus, the company required approval from the Ministry of Atomic Energy (Minatom) in Moscow, which was finally achieved in December 1990.

In order to gain this approval, the US had to agree to a number of demands from Minatom. This included: the Topaz-II reactors had to be returned to Russia after the testing and that the reactors could not be used for military purposes. Dr. Verga insisted on additional international cooperation, including staff from the UK and France. This not only was a cost-saving measure, but reinforced the international and transparent nature of the program, and made military use more challenging.

While this was occurring, the Americans were insistent that the non-nuclear testing of the reactors had to be duplicated in the US, to ensure they met American safety and design criteria. This was a major sticking point for Minatom, and delayed the approval of the export for months, but the Americans did not slow in their preparations for building a test facility. Due to the concentration of space nuclear power research resources in New Mexico (with Los Alamos and Sandia National Laboratories, the US Air Force Philips Laboratory, and the University of New Mexico’s New Mexico Engineering Research Institute (NMERI), as well as the presence of the powerful Republican senator Pete Domenici to smooth political feathers in Washington, DC (all of the labs were within his Senatorial district in the north of the state), it was decided to test the reactors in Albuquerque, NM. The USAF purchased an empty building from the NMERI, and hired personnel from UNM to handle the human resources side of things. The selection of UNM emphasized the transparent, exploratory nature of the program, an absolute requirement for Minatom, and the university had considerable organizational flexibility when compared to either the USAF or the DOE. According to the contract manager, Tim Stepetic:

The University was very cooperative and accommodating… UNM allowed me to open checking accounts to provide responsive payments for the support requirements of the INTERTEK and LUCH contracts – I don’t think they’ve ever permitted such checkbook arrangements either before or since…”

These freedoms were necessary to work with the Russian team members, who were in culture shock and dealing with very different organizational restrictions than their American counterparts. As has been observed both before and since, the Russian scientists and technicians preferred to save as much of their (generous in their terms) per diem for after the project and the money would go further. They also covered local travel expenses as well. One of the technicians had to leave the US for Russia for his son’s brain tumor operation, and was asked by the surgeon to bring back some Tylenol, a request that was rapidly acquiesced to with bemusement from his American colleagues. In addition, personal calls (of a limited nature due to international calling rates at the time) were allowed for the scientists and technicians to keep in touch with their families and reduce their homesickness.

As should be surprising to no-one, the highly unusual nature of this financial arrangement, as well as the large amount of money involved (which ended up coming out to about $400,000 in 1990s money), a routine audit led to the Government Accounting Office being called in to investigate the arrangement later. Fortunately, no significant irregularities in the financial dealings of the NMERI were found, and the program continued. Additionally, the reuse of over $500,000 in equipment scrounged from SNL and LANL’s junk yards allowed for incredible cost savings in the program.

With the business side of the testing underway, it was time to begin preparations for the testing of the reactors in the US, beginning with the conversion of an empty building into a non-nuclear test facility. The building’s conversion, under the head of Frank Thome on the facilities modification side, and Scott Wold as the TSET training manager, began in April of 1991, only four months after Minatom’s approval of INTERTEK. Over the course of the next year, the facility would be prepared for testing, and would be completed just before the delivery of the first shipment of reactors and equipment from Russia.

By this point, the test program had grown to include two programs. The first was the Thermionic Systems Evaluation Test (TSET), which would study mechanical, thermophysical, and chemical properties of the reactors to verify the data collected in Russia. This was to flight-qualify the reactors for American space mission use, and establish the collaboration of the various international participants in the Topaz International Program.

The second program was the Nuclear Electric Propulsion Space Test Program (NEPSTP); run by the Johns Hopkins Applied Physics Laboratory, and funded by the SDIP Ballistic Missile Defense Organization, it proposed an experimental spacecraft that would use a set of six different electric thrusters, as well as equipment to monitor the environmental effects of both the thrusters and the reactor during operation. Design work for the spacecraft began almost immediately after the TSET program began, and the program was of interest to both the American and Russian parts of the team.

Later, one final program would be added: the Thermionic Fuel Element Verification Program (TFEVP). This program, which had predated TIP, is where many of the UK and French researchers were involved, and focused of increasing the lifetime of the thermionic fuel elements from one year (the best US estimate before the TSET) to at least three, and preferably seven, years. This would be achieved through better knowledge of materials properties, as well as improved manufacturing methods.

Finally, there were smaller programs that were attached to the big three, looking at materials effets in intense radiation and plasma environments, as well as long-term contact with cesium vapor, chemcal reactions within the hardware itself, and the surface electrical properties of various ceramics. These tests, while not the primary focus of the program, WOULD contribute to the understanding of the environment an astronuclear spacecraft would experience, and would significantly affect future spacecraft designs. These tests would occur in the same building as the TSET testing, and the teams involved would frequently collaborate on all projects, leading to a very well-integrated and collegial atmosphere.

Reactor Shipment: A Funny Little Thing Occurred in Russia

While all of this was going on in the Topaz International Program, major changes were happening thoughout the USSR: it was falling apart. From the uprisings in Latvia and Lithuania (violently put down by the Soviet military), to the fall of the Berlin Wall, to the ultimate lowering of the hammer and sickle from the Kremlin in December 1991 and its replacement with the tricolor of the Russian Federation, the fall of the Iron Curtain was accelerating. The TIP teams were continuing to work at their program, knowing that it offered hope for the Topaz-II project as well as a vehicle to form closer technological collaborations with their former adversaries, but the complications would rear their heads in this small group as well.

The American purchase of the Topaz reactors was approved by President George H.W. Bush on 27 March, 1992 during a meeting with his Secretary of State, James Barker, and Secretary of Defense Richard Cheney. This freed the American side of the collaboration to do what needed to be done to make the program happen, as well as begin bringing in Russian specialists to begin test facility preparations.

Trinity site obelisk

The first group of 14 Russian scientists and technicians to arrive in the US for the TSET program arrived on April 3, 1992, but only got to sleep for a few hours before being woken up by their guests (who also brought their families) for a long van journey. This was something that the Russians greatly appreciated, because April 4 is a special day in one small part of the world: it’s one of only two days of the year that the Trinity Site, the location of the first nuclear explosion in history, is open to the public. According to one of them, Georgiy Kompaniets:

It was like for a picnic! And at the entrance to the site there were souvenir vendors selling t-shirts with bombs and rocks supposedly at the epicenter of the blast…” (note: no trinitite is allowed to be collected at the Trinity site anymore, and according to some interpretations of federal law is considered low-level radioactive waste from weapons production)

The Russians were a hit at the Trinity site, being the center of attention from those there, and were interviewed for television. They even got to tour the McDonald ranch house, where the Gadget was assembled and the blast was initiated. This made a huge impression on the visiting Russians, and did wonders in cementing the team’s culture.

Hot air balloon in New Mexico, open source

Another cultural exchange that occurred later (exactly when I’m not sure) was the chance to ride in a hot air balloon. Albuquerque’s International Balloon Fiesta is the largest hot air ballooning event in the world, and whenever atmospheric conditions are right a half dozen or more balloons can be seen floating over the city. A local ballooning club, having heard about the Russian scientists and technicians (they had become minor local celebrities at this point) offered them a free hot air balloon ride. This is something that the Russians universally accepted, since none of them had ever experienced this.

According to Boris Steppenov:

The greatest difficulty, it seemed, was landing. And it was absolutely forbidden to touch down on the reservations belonging to the Native Americans, as this would be seen as an attack on their land and an affront to their ancestors…

[after the flight] there were speeches, there were oaths, there was baptism with champagne, and many other rituals. A memory for an entire life!”

The balloon that Steppenov flew in did indeed land on the Sandia Pueblo Reservation, but before touchdown the tribal police were notified, and they showed up to the landing site, issued a ticket to the ballooning company, and allowed them to pack up and leave.

These events, as well as other uniquely New Mexican experiences, cemented the TIP team into a group of lifelong friends, and would reinforce the willingness of everyone to work together as much as possible to make TIP as much of a success as it could be.

C-141 taking off, image DOD

In late April, 1992, a team of US military personnel (led by Army Major Fred Tarantino of SDIO, with AF Major Dan Mulder in charge of logistics), including a USAF Airlift Control Element Team, landed in St. Petersburg on a C-141 and C-130, carrying the equipment needed to properly secure the test equipment and reactors that would be flown to the US. Overflight permissions were secured, and special packing cases, especially for the very delicate tungsten TISA heaters, were prepared. These preparations were complicated by the lack of effective packing materials for these heaters, until Dr. Britt of both ISP and INTERTEK had the idea of using foam bedding pads from a furniture store. Due to the large size and weight of the equipment, though, the C-141 and C-130 aircraft were not sufficient for airlifting the equipment, so the teams had to wait on the larger C-5 Galaxy transports intended for this task, which were en route from the US at the time.

Sadly, when the time came for the export licenses to be given to the customs officer, he refused to honor them – because they were Soviet documents, and the Soviet Union no longer existed. This led Academician Ponomarev-Stepnoy and INTERTEK’s director, Benjamin Usov, to travel to Moscow on April 27 to meet with the Chairman of the Government, Alexander Shokhin, to get new export licenses. After consulting with the Minister of Foreign Economic Relations, Sergei Glazev, a one-time, urgent export license was issued for the shipment to the US. This was then sent via fast courier to St. Petersburg on May 1.

C-5 Galaxy, image USAF

The C-5s, though, weren’t in Russia yet. Once they did land, though, a complex paperwork ballet needed to be carried out to get the reactors and test equipment to America. First, the reactors were purchased by INTERTEK from the Russian bureaus responsible for the various components. Then, INTERTEK would sell the reactors and equipment to Dr. Britt of ISP once the equipment was loaded onto the C-5. Dr. Britt then immediately resold the equipment to the US government. This then avoided the import issues that would have occurred on the US side if the equipment had been imported by ISP, a private company, or INTERTEK, a Russian-led international consortium.

One of them landed in St. Petersburg on May 6, was loaded with the two Topaz-II reactors (V-71 and Ya-21U) and as much equipment as could be fit in the aircraft, and left the same day. It would arrive in Albuquerque on May 7. The other developed maintenance problems, and was forced to wait in England for five days, finally arriving on May 8. The rest of the equipment was loaded up (including the Baikal vacuum chamber), and the plane left later that day. Sadly, it ran into difficulties again upon reaching England, as was forced to wait two more days for it to be repaired, arriving in Albuquerque on May 12.

Preparations for Testing: Two Worlds Coming Together

Unpacking and beryllium checks at TSET Facility in Albuquerque, Image DOE/NASA

Once the equipment was in the US, detailed examination of the payload was required due to the beryllium used in the reflectors and control drums of the reactor. Berylliosis, or the breathing in of beryllium dust, is a serious health issue, and one that the DOE takes incredibly seriously (they’ll evacuate an entire building at the slightest possibility that beryllium dust could be present, at the cost of millions of dollars on occasion). Detailed checks, both before the equipment was removed from the aircraft and during the unpackaging of the reactors. However, no detectable levels of beryllium dust were detected, and the program continued with minimal disruption.

Then it came time to unbox the equipment, but another problem arose: this required the approval of the director of the Central Design Bureau of Heavy Machine Building, Vladmir Nikitin, who was in Moscow. Rather than just call him for approval, Dr Britt called and got approval for Valery Sinkevych, the Albuquerque representative for INTERTEK, to have discretional control over these sorts of decisions. The approval was given, greatly smoothing the process of both setup and testing during TIP.

Sinkevych, Scott Wold and Glen Schmidt worked closely together in the management of the project. Both were on hand to answer questions, smooth out difficulties, and other challenges in the testing process, to the point that the Russians began calling Schmidt “The Walking Stick.” His response was classic: that’s my style, “Management by Walking Around.”

Soviet technicians at TSET Test Facility, image Dabrowski

Every day, Schmidt would hold a lab-wide meeting, ensuring everyone was present, before walking everyone through the procedures that needed to be completed for the day, as well as ensuring that everyone had the resources that they needed to complete their tasks. He also made sure that he was aware of any upcoming issues, and worked to resolve them (mostly through Wetch and Britt) before they became an issue for the facility preparations. This was a revelation to the Russian team, who despite working on the program (in Russia) for years, often didn’t know anything other than the component that they worked on. This synthesis of knowledge would continue throughout the program, leading to a far

Initial estimates for the time that it would take to prepare the facility and equipment for testing of the reactors were supposed to be 9 months. Due to both the well-integrated team, as well as the more relaxed management structure of the American effort, this was completed in only 6 ½ months. According to Sinkevych:

The trust that was formed between the Russian and American side allowed us in an unusually short time to complete the assembly of the complex and demonstrate its capabilities.”

This was so incredible to Schmidt that he went to Wetch and Britt, asking for a bonus for the Russians due to their exceptional work. This was approved, and paid proportional to technical assignment, duration, and quality of workmanship. This was yet another culture shock for the Russian team, who had never received a bonus before. The response was twofold: greatly appreciative, and also “if we continue to save time, do we get another bonus?” The answer to this was a qualified “perhaps,” and indeed one more, smaller bonus was paid due to later time savings.

Installation of Topaz-II reactor at TSET Facility, image DOE/NASA

Mid-Testing Drama, and the Second Shipment

Both in the US and Russia, there were many questions about whether this program was even possible. The reason for its success, though, is unequivocally that it was a true partnership between the American and Russian parts of TIP. This was the first Russian-US government-to-government cooperative program after the fall of the USSR. Unlike the Nunn-Lugar agreement afterward, TIP was always intended to be a true technological exchange, not an assistance program, which is one of the main reasons why the participants of TIP still look fondly and respectfully at the project, while most Russian (and other former Soviet states) participants in N-L consider it to be demeaning, condescending, and not something to ever be repeated again. More than this, though, the Russian design philosophy that allowed full-system, non-nuclear testing of the Topaz-II permanently changed American astronuclear design philosophy, and left its mark on every subsequent astronuclear design.

However, not all organizations in the US saw it this way. Drs. Thorne and Mulder provided excellent bureaucratic cover for the testing program, preventing the majority of the politics of government work from trickling down to the management of the test itself. However, as Scott Wold, the TSET training manager pointed out, they would still get letters from outside organizations stating:

[after careful consideration] they had concluded that an experiment we proposed to do wouldn’t be possible and that we should just stop all work on the project as it was obviously a waste of time. Our typical response was to provide them with the results of the experiment we had just wrapped up.”

As mentioned, this was not uncommon, but was also a minor annoyance. In fact, if anything it cemented the practicality of collaborations of this nature, and over time reduced the friction the program faced through proof of capabilities. Other headaches would arise, but overall they were relatively minor.

Sadly, one of the programs, NEPSTP, was canceled out from under the team near the completion of the spacecraft. The new Clinton administration was not nearly as open to the use of nuclear power as the Bush administration had been (to put it mildly), and as such the program ended in 1993.

One type of drama that was avoided was the second shipment of four more Topaz-II reactors from Russia to the US. These were the Eh-40, Eh-41, Eh-43, and Eh-44 reactors. The use of these terms directly contradicts the earlier-specified prefixes for Soviet determinations of capabilities (the systems were built, then assessed for suitability for mechanical, thermal, and nuclear capabilities after construction, for more on this see our first Enisy post here). These units were for: Eh-40 thermal-hydraulic mockup, with a functioning NaK heat rejection system, for “cold-test” testing of thermal covers during integration, launch, and orbital injection; Eh-41 structural mockup for mechanical testing, and demonstration of the mechanical integrity of the anticriticality device (more on that in the next post), modified thermal cover, and American launch vehicle integration; Eh-43 and -44 were potential flight systems, which would undergo modal testing, charging of the NaK coolant system, fuel loading and criticality testing, mechanical vibration, shock, and acoustic tests, 1000 hour thermal vacuum steady-state stability and NaK system integrity tests, and others before launch.

An-124, image Wikimedia

How was drama avoided in this case? The previous shipment was done by the US Air Force, which has many regulations involved in the transport of any cargo, much less flight-capable nuclear reactors containing several toxic substances. This led to delays in approval the first time this shipment method was used. The second time, in 1994, INTERTEK and ISP contracted a private cargo company, Russian Volga Dnepr Airlines, to transport these four reactors. In order to do this, Volga Dnepr Airlines used their An-124 to fly these reactors from St. Petersburg to Albuquerque.

For me personally, this was a very special event, because I was there. My dad got me out of school (I wasn’t even a teenager yet), drove me out to the landing strip fence at Kirtland AFB, and we watched with about 40 other people as this incredible aircraft landed. He told me about the shipment, and why they were bringing it in, and the seed of my astronuclear obsession was planted.

No beryllium dust was found in this shipment, and the reactors were prepared for testing. Additional thermophysical testing, as well as design work for modifications needed to get the reactors flight-qualified and able to be integrated with the American launchers, were conducted on these reactors. These tests and changes will be the subject of the next blog post, as well as the missions that were proposed for the reactors.

These tests would continue until 1995, and the end of testing in Albuquerque. All reactors were packed up, and returned to Russia per the agreement between INTERTEK and Minatom. The Enisy would continue to be developed in Russia until at least 2007.

More Coming Soon!

The story of the Topaz International Program is far from over. The testing in the US, as well as the programs that the US/Russian team had planned have not even been touched on yet besides very cursory mentions. These programs, as well as the end of the Topaz International Program and the possible future of the Enisy reactor, are the focus of our next blog post, the final one in this series.

This program provided a foundation, as well as a harbinger of challenges to come, in international astronuclear collaboration. As such, I feel that it is a very valuable subject to spend a significant amount of time on.

I hope to have the next post out in about a week and a half to two weeks, but the amount of research necessary for this series has definitely surprised me. The few documents available that fill in the gaps are, sadly, behind paywalls that I can’t afford to breach at my current funding availability.

As such, I ask, once again, that you support me on Patreon. You can find my page at https://www.patreon.com/beyondnerva every dollar counts.

References:

US-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program, Dabrowski 2013 https://www.researchgate.net/profile/Richard_Dabrowski/publication/266516447_US-Russian_Cooperation_in_Science_and_Technology_A_Case_Study_of_the_TOPAZ_Space-Based_Nuclear_Reactor_International_Program/links/5433d1e80cf2bf1f1f2634b8/US-Russian-Cooperation-in-Science-and-Technology-A-Case-Study-of-the-TOPAZ-Space-Based-Nuclear-Reactor-International-Program.pdf

Topaz-II Design Evolution, Voss 1994 http://gnnallc.com/pdfs/NPP%2014%20Voss%20Topaz%20II%20Design%20Evolution%201994.pdf

Categories
Development and Testing Fission Power Systems Forgotten Reactors History Test Stands

Topaz International part 1: ENISY, the Soviet Years

Hello, and welcome back to Beyond NERVA! Today, we’re going to return to our discussion of fission power plants, and look at a program that was unique in the history of astronuclear engineering: a Soviet-designed and -built reactor design that was purchased and mostly flight-qualified by the US for an American lunar base. This was the Enisy, known in the West as Topaz-II, and the Topaz International program.

This will be a series of three posts on the system: this post focuses on the history of the reactor in the Soviet Union, including the testing history – which as we’ll see, heavily influenced the final design of the reactor. The next will look at the Topaz International program, which began as early as 1980, while the Soviet Union still appeared strong. Finally, we’ll look at two American uses for the reactor: as a test-bed reactor system for a nuclear electric test satellite, and as a power supply for a crewed lunar base. This fascinating system, and the programs associated with it, definitely deserve a deep dive – so let’s jump right in!

We’ve looked at the history of Soviet astronuclear engineering, and their extensive mission history. The last two of these reactors were the Topaz (Topol) reactors, on the Plasma-A satellites. These reactors used a very interesting type of power conversion system: an in-core thermionic system. Thermionic power conversion takes advantage of the fact that certain materials, when heated, eject electrons, gaining a positive static charge as whatever the electrons impact gain a negative charge. Because the materials required for a thermionic system can be made incredibly neutronically robust, they can be placed inside the core of the reactor itself! This is a concept that I’ve loved since I first heard of it, and remains as cool today as it did back then.

Diagram of multi-cell thermionic fuel element concept, Bennett 1989

The original Topaz reactor used a multi-cell thermionic element concept, where fuel elements were stacked in individual thermionic conversion elements, and several of these were placed end-to-end to form the length of the core. While this is a perfectly acceptable way to set up one of these systems, there are also inefficiencies and complexities associated with so many individual fuel elements. An alternative would be to make a single, full-length thermionic cell, and use either one or several fuel rods inside the thermionic element. This is the – wait for it – single cell thermionic element design, and is the one that was chosen for the Enisy/Topaz-II reactor (which we’ll call Enisy in this post, since it’s focusing on the Soviet history of the reactor). While started in 1967, and tested thoroughly in the 70s, it wasn’t flight-qualified until the 80s… and then the Soviet Union collapsed, and the program died.

After the fall of the USSR, there was a concerted effort by the US to keep the specialist engineers and scientists of the former Soviet republics employed (to ensure they didn’t find work for international bad actors such as North Korea), and to see what technology had been developed behind the Iron Curtain that could be purchased for use by the US. This is where the RD-180 rocket engine, still in use by the United Launch Alliance Atlas rockets, came from. Another part of this program, though, focused on the extensive experience that the Soviets had in astronuclear missions, and in paricular the most advanced – but as yet unflown – design of the renowned NPO Luch design bureau, attached to the Ministry of Medium Industry: the Enisy reactor (which had the US designation of Topaz-II due to early confusion about the design by American observers).

Enisy power supply, image Department of Defense

The Enisy, in its final iteration, was designed to have a thermal output of 115 kWt (at the beginning of life), with a mission requirement of at least 6 kWe at the electrical outlet terminals for at least three years. Additional requirements included a ten year shelf life after construction (without fissile fuel, coolant, or other volatiles loaded), a maximum mass of 1061 kg, and prevention of criticality before achieving orbit (which was complicated from an American point of view, more on that below). The coolant for the reactor remained NaK-78, a common coolant in most reactors we’ve looked at so far. Cesium was stored in a reservoir at the “bottom” (away from the spacecraft) end of the reactor vessel, to ensure the proper partial pressure between the cathode and anode of the fuel elements, which would leak out over time (about 0.5 g/day during operation). This was meant to be the next upgrade in the Soviet astronuclear fleet, and as such was definitely a step above the Topaz-I reactor.

Perhaps the most interesting part of the design is that it was designed to be able to be tested as a complete system without the use of fissile fuels in the reactor. Instead, electrical resistance heaters could be inserted in the thermionic fuel elements to simulate the fission process, allowing for far more complete testing of the system in flight configuration before launch. This design decision heavily influenced US nuclear power plant design and testing procedures, and continues to influence designs today (the induction heating testing of the KRUSTY thermal simulator is a good recent example of this concept, even if it’s been heavily modified for the different reactor geometry), however, the fact that the reactor used cylindrical fuel elements made this process much easier.

So what did the Enisy look like? This changed over time, but we will look at the basics of the power plant’s design in its final Soviet iteration in this post, and the examine the changes that the Americans made during the collaboration in the next post. We’ll also look at why the design changed as it did.

First, though, we need to look at how the system worked, since compared to every system that we’ve looked at in depth, the physics behind the power conversion system are quite novel.

Thermionics: How to Keep Your Power Conversion System in the Core

We haven’t looked at power conversion systems much in this blog yet, but this is a good place to discuss the first kind as it’s so integral to this reactor. If the details of how the power conversion system actually worked don’t interest you, feel free to skip to the next section, but for many people interested in astronuclear design this power conversion system offers the promise to potentially be the most efficient and reliable option available for in-space nuclear reactors geared towards electricity production.

In short, thermionic reactions are those that occur when a material is heated and gives off charged particles. This is something that has been known since ancient times, even though the physical mechanism was completely unknown until after the discovery of the electron. The name comes from the term “thermions,” or “thermal ions.” One of the first to describe this effect used a hot anode in a vacuum: the modern incandescent lightbulb: Thomas Edison, who observed a static charge building up on the glass of his bulbs while they were turned on. However, today this has expanded to include the use of anodes, as well as solid-state systems and systems that don’t have a vacuum.

The efficiency of these systems depends on the temperature difference between the anode and cathode, the work function (or minimum thermodynamic work needed to remove an electron from a solid to a vacuum immediately outside the solid surface) of the emitter used, and the Boltzmann Constant (which relates to the average kinetic energy of particles in a gas), as well as a number of other factors. In modern systems, however, the structure of a thermionic convertor which isn’t completely solid state is fairly standard: a hot cathode is separated from a cold anode, with cesium vapor in between. For nuclear systems, the anode is often tungsten, the cathode seems to vary depending on the system, and the gap between – called the inter-electrode gap – is system specific.

The cesium exists in an interesting state of matter. Solid, liquid, gas, and plasma are familiar to pretty much everyone at this point, but other states exist under unusual circumstances; perhaps the best known is a supercritical fluid, which exhibits the properties of both a liquid and a gas (although this is a range of possibilities, with some having more liquid properties and some more gaseous). The one that concerns us today is something called Rydberg matter, one of the more exotic forms of matter – although it has been observed in many places across the universe. In its simplest form, Rydberg matter can be seen as small clusters of interconnected molecules within a gas (the largest number of atoms observed in a laboratory is 91, according to Wikipedia, although there’s evidence for far larger numbers in interstellar gas clouds). These clumps end up affecting the electron clouds of those atoms in the clusters, causing them to orbit across the nuclei of those atoms, causing a new lowest-energy state for the entire cluster to occur. These structures don’t degrade any faster under radioactive bombardment due to a number of quantum mechanical properties, which brought them to the attention of the Los Alamos Scientific Laboratory staff in the 1950s, and a short time later Soviet nuclear physicists as well.

This sounds complex, and it is, but the key point is this: because the clumps act as a unit within Rydberg matter, their ability to transmit electricity is enhanced compared to other gasses. In particular, cesium seems to be a very good vehicle for creating Rydberg matter, and cesium vapor seems to be the best available for the gap between the cathode and anode of a thermionic convertor. The density of the cesium vapor is variable and dependent on many factors, including the materials properties of the cathode and anode, the temperature of the cathode, the inter-electrode gap distance, and a number of other factors. Tuning the amount of cesium in the inter-electrode gap is something that must occur in any thermionic power conversion system; in fact the original version of the Enisy had the ability to vary the inter-electrode gap pressure (this was later dropped when it was discovered to be superfluous to the efficient function of the reactor).

This type of system comes in two varieties: in-core and out-of-core. The out-of-core variant is very similar to the power conversion systems we saw (briefly) on the SNAP systems: the coolant from the reactor passes around or through the radiation shield of the system, heats the anode, which then emits electrons into the gap, collected by the cathode, and then the electricity goes through the power conditioning unit and into the electrical system of the spacecraft. Because thermionic conversion is theoretically more efficient, and in practice is more flexible in temperature range, than thermoelectric conversion, even keeping the configuration of the power conversion system’s relationship to the rest of the power plant offers some advantages.

The in-core variant, on the other hand, wraps the power conversion system directly around the fissile fuel in the core, with electrical power being conducted out of the core itself and through the shield. The coolant runs across the outside of the thermionic unit, providing the thermal gradient for the system to work, and then exits the reactor. While this increases the volume of the core (admittedly, not by much), it also eliminates the need for more complex plumbing for the primary coolant loop. Additionally, it allows for less heat loss from the coolant having to travel a farther difference. Finally, there’s far less chance of a stray meteor hitting your power conversion system and causing problems – if a thermionic fuel element is damaged by a foreign object, you’re going to have far bigger problems with the system as a whole, since it means that it damaged your control systems and pressure vessel on the way to damaging your power conversion unit!

The in-core thermionic power conversion system, while originally proposed by the US, was seen as a curiosity on their side of the Iron Curtain. Some designs were proposed, but none were significantly researched to the level of being able to be serious contenders in the struggle to gain the significant funding needed to develop as complex a system as an astronuclear fission power plant, and the low conversion efficiency available in practice prevents its application in terrestrial power plants, which to this day continue to use steam turbine generators.

On the other side of the Iron Curtain, however, this was seen as the ideal solution for a power conversion system: the only systems needed for the system to work could be solid-state, with no moving parts: heaters to vaporize the cesium, and electromagnetic pumps to move it through the reactor. Greater radiation resistance and more flexible operating temperatures, as well as greater conversion efficiency, all offered more promise to Soviet astronuclear systems designers than the thermoelectric path that the US ended up following. The first Soviet reactor designed for in-space use, the Romashka, used a thermionic power conversion system, but the challenges involved in the system itself led the Krasnya Zvezda design bureau (who were responsible for the Romasha, Bouk, and Topol reactors) to initially choose to use thermoelectric convertors in their first flight system: the BES-5 Bouk, which we’ve seen before.

Now that we’ve looked at the physics behind how you can place your power conversion system within the reactor vessel of your power plant (and as far as I’ve been able to determine, if you’re looking to generate electricity beyond what a simple sensor needs, this is the only option without going to something very exotic), let’s look at the reactor itself.

Enisy: The Design of the TOPAZ-II Reactor

The Enisy was a uranium oxide fueled, zirconium hydride moderated, sodium-potassium eutectic cooled reactor, which used a single-element thermionic fuel element design for in-core power conversion. The multi-cell version was used in the Topol reactor, where each fuel pellet was wrapped in its own thermionic convertor. This is sometimes called a “flashlight” configuration, since it looks a bit like the batteries in a large flashlight, but this comes at the cost of complexity, mass, and increased inefficiencies. To offset this, many issues are easier to deal with in this configuration, especially as your fuel reaches higher burnup percentages and your fuel swells. The ultimate goal was single-unit thermionic fuel elements, which were realized in the Enisy reactor. While more challenging in terms of materials requirements, the greater simplicity, lower mass, and greater efficiency of the system offered more promise.

The power plant was required to provide 6 kWe of electrical power at the reactor terminals (before the power conditioning unit) at 27 volts. It had to have an operational life of three years, and a storage life if not immediately used in a mission of at least ten years. It also had to have an operational reliability of >95%, and could not under any circumstances achieve criticality before reaching orbit, nor could the coolant freeze at any time during operation. Finally, it had to do all of this in less than 1061 kg (excluding the automatic control system).

TFE Full Length, image DOD

Thirty-seven fuel elements were used in the core, which was contained in a stainless steel reactor vessel. These contained uranium oxide fuel pellets, with a central fission gas void about 22% of the diameter of the fuel pellets to prevent swelling as fission products built up. The emitters were made out of molybdenum, a fairly common choice for in-core applications. Al2O3 (sapphire) insulators were used to electrically isolate the fuel elements from the rest of the core. Three of these would be used to power the cesium heater and pump directly, while another (unknown) number powered the NaK coolant pump (my suspicion is that it’s about the same number). The rest would output power directly from the element into the power conditioning unit on the far side of the power plant.

Enisy Core Cross-section, image DOD

Nine control drums, made mostly out of beryllium but with a neutron poison along one portion of the outer surface (Boron carbide/silicon carbide) surrounded the core. Three of these drums were safety drums, with two positions: in, with the neutron poison facing the center of the core, and out, where the beryllium acted as a neutron reflector. The rest of the drums could be rotated in or out as needed to maintain reactivity at the appropriate level in the core. These had actuators mounted outside the pressure vessel to control the rotation of the drums, and were connected to an automatic control system to ensure autonomous stable function of the reactor within the mission profile that the reactor would be required to support.

Image DOD

The NaK coolant would flow around the fuel elements, driven by an electromagnetic pump, and then pass through a radiator, in an annular flow path immediately surrounding the TFEs. Two inlet and two outlet pipes were used to connect the core to the radiator. In between the radiator and the core was a radiation shield, made up of stainless steel and lithium hydride (more on this seemingly odd choice when we look at the testing history).

The coolant tubes were embedded in a zirconium hydride moderator, which was contained in stainless steel casings.

Finally, a reservoir of cesium was at the opposite end of the reactor from the radiator. This was necessary for the proper functioning of the thermionic fuel elements, and underwent many changes throughout the design history of the reactor, including a significant expansion as the design life requirements increased.

Once the Topaz International program began, additional – and quite significant – changes were made to the reactor’s design, including a new automated control system and an anti-criticality system that actually removed some of the fuel from the core until the start-up commands were sent, but that’s a discussion for the next post.

TISA Heater Installation During Topaz International, image NASA

I saved the coolest part of this system for last: the TISA, or “Thermal Simulators of Apparatus Cores” (the acronym was from the original Russian), heaters. These units were placed in the active section of the thermionic fuel elements to simulate the heat of fission occurring in the thermionic fuel elements, with the rest of the systems and subsystems being in flight configuration. This led to unprecedented levels of testing capability, but at the same time would lead to a couple of problems later in testing – which would be addressed as needed.

How did this design end up this way? In order to understand that, the development and testing process of the Soviet design team must be looked at.

The History of Enisy’s Design

The Enisy reactor started with the development of the thermionic fuel element by the Sukhumi Institute in the early 1960s, which had two options: the single cell and multiple cell variants. In 1967, these two options were split into two different programs: the Topol (Topaz), which we looked at in the Soviet Astronuclear History post, led by the Krasnaya Zvezda design bureau in Moscow, and Enisy, which was headed by the Central Design Bureau of Machine Building in Leningrad (now St. Petersburg). Aside from the lead bureau, in charge of the overall program and system management, a number of other organizations were involved with the fabrication and testing of the reactor system: the design and modeling team consisted of: the Kurchatov Institute of Atomic Energy was responsible for nuclear design and analytics, the Scientific Industrial Association Lutch was responsible for the thermionic fuel elements, the Sukhumi Institute remained involved in the reactor’s automatic control systems design; fabrication and testing was the responsibility of: the Research Institute of Chemical Machine Building for thermal vacuum testing, the Scientific Institute for Instrument Building’s Turaevo nuclear test facility, Kraznoyarsk Spacecraft Designer for mechanical testing and spacecraft integration, Prometheus Laboratory for materials development (including liquid metal eutectic development for the cooling system and materials testing) and welding, and the Enisy manufacturing facility was located in Talinn, Estonia (a decision that would cause later headaches during the collaboration).

The Enisy originally had three customers (the identities of which I am not aware of, simply that at least one was military), and each had different requirements for the reactor. Originally designed to operate at 6 kWe for one year with a >95% success rate, but customer requirements changed both of these characteristics significantly. As an example, one customer needed a one year system life, with a 6 kWe power output, while another only needed 5 kWe – but needed a three year mission lifetime. This longer lifetime ended up becoming the baseline requirement of the system, although the 6 kWe requirement and >95% mission success rate remained unchanged. This led to numerous changes, especially to the cesium reservoir needed for the thermionic convertors, as well as insulators, sensors, and other key components in the reactor itself. As the cherry on top, the manufacture of the system was moved from Moscow to Talinn, Estonia, resulting in a new set of technicians needing to be trained to the specific requirements of the system, changes in documentation, and at the fall of the Soviet Union loss of significant program documentation which could have assisted the Russia/US collaboration on the system.

The nuclear design side of things changed throughout the design life as well. An increase in the number of thermionic fuel elements (TFEs) occurred in 1974, from 31 to 37 in the reactor core, an increase in the height of the “active” section of the TFE, although whether the overall TFE length (and therefore the core length) changed is information I have not been able to find. Additional space in the TFEs was added to account for greater fuel swelling as fission products built up in the fuel pellets, and the bellows used to ensure proper fitting of the TFEs with reactor components were modified as well. The moderator blocks in the core, made out of zirconium hydride, were modified at least twice, including changing the material that the moderator was kept in. Manufacturing changes in the stainless steel reactor vessel were also required, as were changes to the gamma shielding design for the shadow shield. All in all, the reactor went through significant changes from the first model tested to theend of its design life.

Another area with significantly changing requirements was the systems integration side of things. The reactor was initially meant to be launched in a reactor-up position, but this was changed in 1979 to a reactor-down launch configuration, necessitating changes to several systems in what ended up being a significant effort. Another change in the launch integration requirements was an increase in the acceleration levels required during dynamic testing by a factor of almost two, resulting in failures in testing – and resultant redesigns of many of the structures used in the system. Another thing that changed was the boom that mounted the power plant to the spacecraft – three different designs were used through the lifetime of the system on the Russian side of things, and doubtless another two (at least) were needed for the American spacecraft integration.

Perhaps the most changed design was the coolant loop, due to significant problems during testing and manufacturing of the system.

Design Driven by (Expected) Failure: The USSR Testing Program

Flight qualification for nuclear reactors in the USSR at the time was very different from the way that the US did flight qualification, something that we’ll look at a bit more later in this post. The Soviet method of flight qualification was to heavily test a number of test-beds, using both nuclear and non-nuclear techniques, to validate the design parameters. However, the actual flight articles themselves weren’t subjected to nearly the same level of testing that the American systems would be, instead going through a relatively “basic” (according to US sources) workmanship examination before any theoretical launch.

In the US, extensive systems modeling is a routine part of nuclear design of any sort, as well as astronautical design. Failures are not unexpected, but at the same time the ideal is that the system has been studied and modeled mathematically thoroughly enough that it’s not unreasonable to predict that the system will function correctly the first time… and the second… and so on. This takes not only a large amount of skilled intellectual and manual labor to achieve, but also significant computational capabilities.

In the Soviet Union, however, the preferred method of astronautical – and astronuclear – development was to build what seemed to be a well-designed system and then test it, expecting failure. Once this happened, the causes of the failure were analyzed, the problem corrected, and then the newly upgraded design would be tested again… and again, for as many times as were needed to develop a robust system. Failure was literally built into the development process, and while it could be frustrating to correct the problems that occurred, the design team knew that the way their system could fail had been thoroughly examined, leading to a more reliable end result.

This design philosophy leads to a large number of each system needing to be built. Each reactor that was built underwent a post-manufacturing examination to determine the quality of the fabrication in the system, and from this the appropriate use of the reactor. These systems had four prefixes: SM, V, Ya, and Eh. Each system in this order was able to do everything that the previous reactor would be able to do, in addition to having superior capabilities to the previous type. The SM, or static mockup, articles were never built for anything but mechanical testing, and as such were stripped down, “boilerplate” versions of the system. The V reactors were the next step up, which were used for thermophysical (heat transfer, vibration testing, etc) or mechanical testing, but were not of sufficient quality to undergo nuclear testing. The Ya reactors were suitable for use in nuclear testing as well, and in a pinch would be able to be used in flight. The Eh reactors were the highest quality, and were designated potential flight systems.

In addition to this designation, there were four distinct generations of reactor: the first generation was from V-11 to Ya-22. This core used 31 thermionic fuel elements, with a one year design life. They were intended to be launched upright, and had a lightweight radiation shield. The next generation, V-15 to Ya-26, the operational lifetime was increased to a year and a half.

The third generation, V-71 to Eh-42 had a number of changes. The number of TFEs was increased from 31 to 37, in large part to accommodate another increase in design life, to above 3 years. The emitters on the TFEs were changed to the monocrystaline Mo emitters, and the later ones had Nb added to the Mo (more on this below). The ground testing thermal power level was reduced, to address thermal damage from the heating units in earlier non-nuclear tests. This is also when the launch configuration was changed from upright to inverted, necessitating changes in the freeze-prevention thermal shield, integration boom, and radiator mounting brackets. The last two of this generation, Eh -41 and Eh-42, had the heavier radiation shield installed, while the rest used the earlier, lighter gamma shield.

The final generation, Ya-21u to Eh-44, had the longest core lifetime requirement of three years at 5.5 kWe power output. These included all of the other changes above, as well as many smaller changes to the reactor vessel, mounting brackets, and other mechanical components. Most of these systems ended up becoming either Ya or Eh units due to lessons learned in the previous three generations, and all of the units which would later be purchased by the US as flight units came from this final generation.

A total of 29 articles were built by 1992, when the US became involved in the program. As of 1992, two of the units were not completed, and one was never assembled into its completed configuration.

Sixteen of the 21 units were tested between 1970 and 1989, providing an extensive experimental record of the reactor type. Of these tests, thirteen underwent thermal, mechanical, and integration non-nuclear testing. Nuclear testing occurred six times at the Baikal nuclear facility. As of 1992, there were two built, but untested, flight units available: the E-43 and E-44, with the E-45 still under construction.

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Unit Name

Generation

Series #

Core Life

# of TFEs

TFE Generation

ACS Unit

Launch configuration

Manufacturing location

Test type

Test stand

Testing begin

Testing end

Testing duration

System notes

SM-0

0

Static Model 1

n/a

n/a

n/a

Upright

CDBMB

Static

01/01/76

01/01/76

Original mockup, with three main load bearing systems.

SM-1

0

Static Model 2

n/a

n/a

n/a

Inverted

CDBMB

Static

Krasnoyarsk

01/01/83

01/01/84

Inverted launch configuration static test model.

SM-2

0

Static Model 3

n/a

n/a

n/a

Inverted

CDBMB

Static

Krasnoyarsk

01/01/83

01/01/84

Inverted launch configuration static test model.

V-11

1

Prototype 1

1

1

Upright

CDBMB

Electric heat

Baikal

07/23/71

02/03/72

3200

Development of system test methods and operations. Incomplete set of TFEs

V-12

1

Prototype 2

1

31

1

Upright

CDBMB

Electrical

Baikal

06/21/72

04/18/73

850

Development of technology for prelaunch operations and system testing

V-13

1

Prototype 3

1

31

1

Upright

Talinn

Mechanical

Baikal, Mechanical

08/01/72

05/01/73

?

Transportation, dynamic, shock, cold temperature testing. Reliability at freezing and heating.

Ya(?)-20

1

Specimen 1

1

31

1

Upright

Talinn

Nuclear

Romashka

10/01/72

03/01/74

2500

Zero power testing. Neutron physical characteristics, radiation field characterization. Development of nuclear tests methods.

Ya-21

1

Specimen 2

1

31

1

Upright

Talinn

Nuclear

Baikal, Romashka

?

?

?

Nuclear test methods and test stand trials. Prelaunch operations. Neutron plysical characteristics

Ya-22

1

Specimen 3

1

31

1

Upright

Talinn

n/a

n/a

n/a

n/a

n/a

Unfabricated, was intended to use Ya-21 design documents

Unit Name

Generation

Series #

Core Life

# of TFEs

TFE Generation

ACS Unit

Launch configuration

Manufacturing location

Test type

Test stand

Testing begin

Testing end

Testing duration

System notes

V-15

2

Serial 1

1-1.5

31

2

Upright

Talinn

Cold temp

Baikal, Cold Temp Testing

02/12/80

?

Operation and functioning tests at freezing and heating.

V-16

2

Serial 2

1-1.5

31

2

Upright

Talinn

Mechanical, Electrical

Mechanical

08/01/79

12/01/79

2300

Transportation, vibration, shock. Post-mechanical electirc serviceability testing.

Ya-23

2

Serial 3

1-1.5

31

2

SAU-35

Upright

Talinn

Nuclear

Romashka

03/10/75

06/30/76

5000

Nuclear testing revision and development, including fuel loading, radiation and nuclear safety. Studied unstable nuclear conditions and stainless steel material properties, disassembly and inspection. LiH moderator hydrogen loss in test.

Eh-31

2

Serial 4

1-1.5

31

2

SAU-105

Upright

Talinn

Nuclear

Romashka

02/01/76

09/01/78

4600

Nuclear ground test. ACS startup, steady-state functioning, post-operation disassembly and inspection. TFE lifetime limited to ~2 months due to fuel swelling

Ya-24

2

Serial 5

1-1.5

31

2

SAU-105

Upright

Talinn

Nuclear

Tureavo

12/01/78

04/01/81

14000

Steady state nuclear testing. Significant TFE shortening post-irradiation.

(??)-33

2

Serial 6

1-1.5

31

2

Upright

Talinn

Spacecraft integration

Tureavo

n/a

n/a

n/a

TFE needed redesign, no systems testing. Installed at Turaevo as mockup. Used to establish transport and handling procedures

V(?)-25

2

Serial 7

1-1.5

31

2

Upright

Talinn

Spacecraft integration

Krasnoyarsk

n/a

n/a

?

System incomplete. Used as spacecraft mockup, did not undergo physical testing.

(??)-35

2

Serial 8

1-1.5

31

2

Upright

Talinn

Test stand preparation

Baikal

?

?

?

Second fabrication stage not completed. Used for some experiments with Baikal test stand. Disassembled in Sosnovivord.

V(?)-26

2

Serial 9

1-1.5

31

2

Upright

Talinn, CDBMB

n/a

n/a

n/a

n/a

n/a

Refabricated at CDBMB. TFE burnt and damaged during second fadrication. Notch between TISA and emitter

Unit Name

Generation

Series #

Core Life

# of TFEs

TFE Generation

ACS Unit

Launch configuration

Manufacturing location

Test type

Test stand

Testing begin

Testing end

Testing duration

System notes

V-71

3

Serial 10

1.5

37

3

Upright, Inverted

Talinn

Mechanical, Electrical, Spacecraft integration

Baikal, Krasnoyarsk, Cold Temp Testing

01/01/81

01/01/87

1300

Converted from upright to inverted launch configuration, spacecraft integration heavily modified. First to use 37 TFE core configuration. Transport testing (railroad vibration and shock), cold temperature testing. Electrical testing post-mechanical. Zero power testing at Krasnoyarsk.

Ya-81

3

Serial 11

1.5

37

3

Ground control (no ACS)

Inverted

Talinn

Nuclear

Romashka

09/01/80

01/01/83

12500

Nuclear ground test, steady state operation. Leaks observed in two cooling pipes 120 hrs into test; leaks plugged and test continued. Disassembly and inspection.

Ya-82

3

Serial 12

1.5

37

3

Prototype Sukhumi ACS

Inverted

Talinn

Nuclear

Tureavo

09/01/83

11/01/84

8300

Nuclear ground test, startup using ACS, steady state. Initial leak in EM pump led to large leak later in test. Test ended in loss of coolant accident. Reactor disassembled and inspected post-test to determine leak cause.

Eh(?)-37

3

Serial 13

1.5

37

3

Inverted

Talinn

Static

?

?

?

?

Quality not sufficient for flight (despite Eh “flight” designation). Static and torsion tests conducted.

Eh-38

3

Serial 14

1.5

37

3

Factory #1

Inverted

Talinn

Nuclear

Romashka

02/01/86

05/01/86

4700

Nuclear ground test, pre-launch simulation. ACS startup and operation. Steady state test. Post-operation disassembly and examination.

(??)-39

3

Serial 15

1.5

37

3

Inverted

Talinn

special

special

special

special

special

Fabrication begin in Estonia, with some changed components. After changes, system name changed to Eh-41, and serial number changed to 17. Significant reactor changes.

Eh-40

3

Serial 16

1.5

37

3

Inverted

Talinn

Cold temp, coolant flow

?

01/03/88

12/31/88

?

Cold temperature testing. No electrical testing. Filled with NaK during second stage of fabrication.

Eh-41

3

Serial 17

1.5

37

3

Inverted

Talinn

Mechanical, Leak

Baikal, Mechanical

01/01/88

?

?

Began life as Eh(?)-39, post-retrofit designation. Transportation (railroad) dynamic, and impact testing. Leak testing done post-mechanical testing. First use of increased shield mass.

Eh-42

3

Serial 18

1.5

37

3

Inverted

Talinn

n/a

n/a

Critical component welding failure during fabrication. Unit never used.

Unit Name

Generation

Series #

Core Life

# of TFEs

TFE Generation

ACS Unit

Launch configuration

Manufacturing location

Test type

Test stand

Testing begin

Testing end

Testing duration

System notes

Ya-21u

4

Serial 19

3

37

4

Inverted

Talinn

Electrical

Baikal

12/01/87

12/01/89

?

First Gen 4 reactor using modified TFEs. Electrical testing on TFEs conducted. New end-cap insulation on TFEs tested.

Eh-43

4

Serial 20

3

37

4

Inverted

Talinn

n/a

6/30/88 (? Unclear what testing is indicated)

n/a

n/a

n/a

Flight unit. First fabrication phase in Talinn completed, second incomplete as of 1994

Eh-44

4

Serial 21

3

37

4

Inverted

Talinn

n/a

n/a

n/a

n/a

n/a

Flight unit. First fabrication phase in Talinn completed, second incomplete as of 1994

Eh(?)-45

4

Serial 22

3

37

4

Inverted

Talinn

n/a

n/a

n/a

n/a

n/a

Partially fabricated unit with missing components.

Not many fine details are known about the testing of these systems, but we do have some information about the tests that led to significant design changes. These changes are best broken down by power plant subsystem, because while there’s significant interplay between these various subsystems their functionality can change in minor ways quite easily without affecting the plant as a whole. Those systems are: the thermionic fuel elements, the moderator, the pressure vessel, the shield, the coolant loop (which includes the radiator piping), the radiator coatings, the launch configuration, the cesium unit, and the automatic control system (including the sensors for the system and the drum drive units). While this seems like a lot of systems to cover, many of them have very little information about their design history to pass on, so it’s less daunting than it initially appears.

Thermionic Fuel Elements

It should come as no surprise that the thermionic fuel elements (TFEs) were extensively modified throughout the testing program. One of the big problems was short circuiting across the inter-electrode gap due to fuel swelling, although other problems occurred to cause short circuits as well.

Perhaps the biggest change was the change from 31 to 37 TFEs in the core, one of the major changes to minimize fuel swelling. The active core length (where the pellets were) was increased by up to 40 mm (from 335 mm to 375 mm), the inter-electrode gap was widened by 0.05 mm (from 0.45 to 0.5 mm). In addition, the hole through the center of the fuel element was increased in diameter to allow for greater internal swelling, reducing the mechanical stress on the emitter.

The method of attaching the bellows for thermal expansion were modified (the temperature was dropped 10 K) to prevent crystalization of the palladium braze and increase bellows thermal cycling capability after failures on the Ya-24 system (1977-1981).

Perhaps the biggest change was to the materials used in the TFE. The emitter started off as a polycrystaline molybdenum in the first two generations of reactors, but the grain boundaries between the Mo crystals caused brittleness over time. Because of this, they developed the capability to use monocrystalline Mo, which improved performance in the early third generation of reactors – just not enough. In the final version seen in later 3rd generation and fourth generation systems, the Mo was doped with 3% niobium, which created the best available material for the emitter.

There were many other changes during the development of the thermionic fuel elements, including the addition of coatings on some materials for corrosion resistance, changes in electrical insulation type, and others, but these were the most significant in terms of functionality of the TFEs, and their impact on the overall systems design.

ZrH Moderator

The zirconium hydride neutron moderator was placed around the outside of the core. Failures were observed several times in testing, including the Ya-23 test, which resulted in loss of hydrogen in the core and the permanent shutdown of that reactor. Overpower issues, combined with a loss of coolant, led to moderator failure in Ya-82 as well, but in this case the improved H barriers used in the stainless steel “cans” holding the ZrH prevented a loss of hydrogen accident despite the ZrH breaking up (the failure was due to the ZrH being spread more thinly across the reactor, not the loss of H due to ZrH damage).

This development process was one of the least well documented areas of the Soviet program.

Reactor Vessel

Again, this subsystem’s development seems poorly documented. The biggest change, though, seems to be changing the way the triple coating (of chrome, then nickel, then enamel) was applied to the stainless steel of the reactor vessel. This was due to the failure of the Ya-23 unit, which failed at the join between the tube and the end of the tube on one of the TFEs. The crack self-sealed, but for future units the coatings didn’t go all the way to the weld, and the hot CO2 used as a cover gas was allowed to carbonize the steel to prevent fatigue cracking.

Radiation Shield

The LiH component of the radiation shield (for neutron shielding) seems to not have changed much throughout the development of the reactor. The LiH was contained in a 1.5 mm thick stainless steel casing, polished on the ends for reflectivity and coated black on the outside face.

However, the design of the stainless steel casing was changed in the early 1980s to meet more stringent payload gamma radiation doses. Rather than add a new material such as tungsten or depleted uranium as is typical, the designers decided to just thicken the reactor and spacecraft sides of the LiH can to 65 mm and 60 mm respectively. While this was definitely less mass-efficient than using W or U, the manufacturing change was fairly trivial to do with stainless steel, and this was considered the most effective way to ensure the required flux rates with the minimum of engineering challenges.

The first unit to use this was the E-41, fabricated in 1985, which was also the first unit to be tested in the inverted flight configuration. The heavier shield, combined with the new position, led to the failure of one of the shield-to-reactor brackets, as well as the attachment clips for the radiator piping. These components were changed, and no further challenges occurred with the shield in the rest of the test program.

Coolant Loop

The NaK coolant loop was the biggest source of headaches dueing the development of the Enisy. A brief list of failures, and actions taken to correct them, is here:

V-11 (July 1971-February 1972): A weld failed at the join between the radiator tubing and collector during thermophysical testing. The double weld was changed to a triple weld to correct the failure mode.

Ya-21 (1971): This reactor seemed to have everything go wrong with it. Another leak at the same tube-to-collector interface led to the welding on of a small sleeve to repair the crack. This fix seemed to solve the problem of failures in that location.

Ya-23 (March 1975-June 1976): Coolant leak between coolant tube and moderator cavity. Both coating changes and power ramp-up limits eliminated issues.

V-71 (January 1981-1994?): NaK leak in radiator tube after 290 hours of testing. Plugged, testing continued. New leak occurred 210 test hours later, radiator examined under x-ray. Two additional poorly-manufactured tubes replaced with structural supports. One of test reactors sent to US under Topaz International.

Ya-81 (September 1980-January 1983): Two radiator pipe leaks 180 hours into nuclear testing (no pre-nuclear thermophysical testing of unit). Piping determined to be of lower quality after switching manufacturers. Post-repair, the unit ran for 12,500 hours in nuclear power operation.

Ya-82 (September 1983 to November 1984): Slow leak led to coolant pump voiding and oscillations, then one of six pump inlet lines being split. There were two additional contributions to this failure: the square surfaces were pressed into shape from square pipes, which can cause stress microfractures at the corners, and second the inlet pump was forced into place, causing stress fracturing at the joint. This failure led to reactor overheating due to a loss-of-coolant condition, and led to the failure of the ZrH moderator blocks. This led to increased manufacturing controls on the pump assembly, and no further major pump failures were noted in the remainder of the testing.

Eh-38 (February 1986-August 1986): This failure is a source of some debate among the Russian specialists. Some believe it was a slow leak that began shortly after startup, while others believe that it was a larger leak that started at some point toward the end of the 4700 hour nuclear test. The exact location of the leak was never located, however it’s known that it was in the upper collector of the radiator assembly.

Ya-21u (December 1987-December 1989): Caustic stress-corrosion cracking occurred about a month and a half into thermophysical testing in the lower collector assembly, likely caused by a coating flaw growing during thermal cycling. This means that subsurface residual stresses existed within the collector itself. Due to the higher-than-typical (by U.S. standards) carbon content in the stainless steel (the specification allowed for 0.08%-0.12% carbon, rather than the less than 0.8% carbon content in the U.S. SS-321), the steel was less ductile than was ideal, which could have been a source of the flaw growing as it did. Additionally, increased oxygen levels in the NaK coolant could have exacerbated the problem more as well. A combination of ensuring that heat treatments had occurred post-forming, as well as ensuring a more oxygen-poor environment, were essential to reducing the chances of this failure happening again.

Radiator

Pen and ink diagram of radiator, image DOD

The only known data poing on the radiator development was during the Ya-23 test, where the radiator coating changed the nuclear properties of the system at elevated temperature (how is unknown). This was changed to something that would be less affected by the radiation environment. The final radiator configuration was a chrome and polymer substrate with an emissivity of 0.85 at beginning of life.

Launch configuration

As we saw, the orientation that the reactor was to be launched in was changed from upright to inverted, with the boom to connect the reactor to the spacecraft being side by side inside the payload fairing. This required the thermal cover used to prevent the NaK from freezing to be redesigned, and modified after the V-13 test, when it was discovered to not be able to prevent freezing of the coolant. The new cover was verified on the V-15 tests, and remained largely unchanged after this.

Some of the load-bearing brackets needed to be changed or reinforced as well, and the clips used to secure the radiator pipes to the structural components of the radiator.

Cesium Supply Unit

For the TFEs to work properly, it was critical that the Cs vapor pressure was within the right pressure range relative tot he temperature of the reactor core. This system was designed from first physical principles, leading to a novel structure that used temperature and pressure gradients to operate. While initially throttleable, but there were issues with this functionality during the Ya-24 nuclear test. This changed when it was discovered that there was an ideal pressure setting for all power levels, so the feed pressure was fixed. Sadly, on the Ya-81 test the throttle was set too high, leading to the need to cool the Cs as it returned to the reservoir.

Additional issues were found in the startup subsystem (a single-use puncture valve) used to vent the inert He gas from the interelectrode gap (this was used during launch and before startup to prevent Cs from liquefying or freezing in the system), as well as to balance the Cs pressure by venting it into space at a rate of about 0.4 g/day. The Ya-23 test saw a sensor not register the release of the He, leading to an upgraded spring for the valve.

Finally, the mission lifetime extension during the 1985/86 timeframe tripled the required lifetime of the system, necessitating a much larger Cs reservoir to account for Cs venting. This went from having 0.455 g to 1 kg. These were tested on Ya-21u and Eh-44, despite one (military) customer objecting due to insufficient testing of the upgraded system. This system would later be tested and found to be acceptable as part of the Topaz International program.

Automatic Control System

The automatic control system, or ACS, was used for automatic startup and autonomous reactor power management, and went through more significant changes than any other system, save perhaps the thermionic fuel elements. The first ACS, called the SAU-35, was used for the Ya-23 ground test, followed by the SAU-105 for the Eh-31 and Ya-24 tests. Problems arose, however, because these systems were manufactured by the Institute for Instrument Building of the Ministry of Aviation Construction, while the Enisy program was under the purview of the Ministry of Atomic Energy, and bureaucratic problems reared their heads.

This led the Enisy program to look to the Sukhumi Institute (who, if you remember, were the institute that started both the Topol and Enisy programs in the 1960s before control was transferred elsewhere) for the next generation of ACS. During this transition, the Ya-81 ground nuclear test occurred, but due to the bureaucratic wrangling, manufacturer change, and ACS certification tests there was no unit available for the test. This led the Ya-81 reactor to be controlled from the ground station. The Ya-82 test was the first to use a prototype Sukhumi-built ACS, with nine startups being successfully performed by this unit.

The loss-of-cooling accident potentially led to the final major change to the ACS for the Eh-38 test: the establishment of an upper temperature limit. After this, the dead-band was increased to allow greater power drift in the reactor (reducing the necessary control drum movement), as well as some minor modifications rerouting the wires to ensure proper thermocouple sensor readings, were the final significant modifications before Topaz International started.

Sensors

The sensors on the Enisy seem to have been regularly problematic, but rather than replace them, they were either removed or left as instrumentation sensors rather than control sensors. These included the volume accumulator sensors on the stainless steel bellows for the thermionic fuel elements (which were removed), and the set of sensors used to monitor the He gas in the TFE gas gap (for fission product buildup), the volume accumulator (which also contained Ar), and the radiation shield. This second set of sensors was kept in place, but was only able to measure absolute changes, not precise measurements, so was not useful for the ACS.

Control Drive Unit

The control drive unit was responsible for the positioning of the control drums, both on startup as well as throughout the life of the reactor to maintain appropriate reactivity and power levels. Like in the SNAP program, these drive systems were a source of engineering headaches.

Perhaps the most recurring problem during the mid-1970s was the failure of the position sensor for the drive system, which was used to monitor the rotational position of the drum relative to the core. This failed in the Ya-20, Ya-21, and Ya-23, after which it was replaced with a sensor of a new design and the problem isn’t reported again. The Ya-81 test saw the loss of the Ar gas used as the initial lubricant in the drive system, and later seizing of the bearing the drive system connected to, leading to its replacement with a graphite-based lubricant.

The news wasn’t all bad, however. The Eh-40 test demonstrated greater control of drum position by reducing the backlash in the kinematic circuit, for instance, and improvements to the materials and coatings used eliminated problems of coating delamination, improving the system’s resistance to thermal cycling and vibrational stresses, and radiator coating issues.

The Eh-44 drive unit was replaced against the advice of one of the Russian customers due to a lack of mandatory testing on the advanced drive system. This system remained installed at the time of Topaz International, and is something that we’ll look at in the next blog post.

A New Customer Enters the Fold

During this testing, an American company (which is not named) was approached about possibly purchasing nearly complete Enisy reactors: the only thing that the Soviets wouldn’t sell was the fissile fuel itself, and that they would help with the manufacturing on. This was in addition to the three Russian customers (at least one of which was military, but again all remain unnamed). This company did not purchase any units, but did go to the US government with this offer.

This led to the Topaz International program, funded by the US Department of Defense’s Ballistic Missile Defense Organization. The majority of the personnel involved were employees of Los Alamos and Sandia National Laboratories, and the testing occurred at Kirtland Air Force Base in Albuquerque, NM.

As a personal note, I was just outside the perimeter fence when the aircraft carrying the test stand and reactors landed, and it remains one of the formational events in my childhood, even though I had only the vaguest understanding of what was actually happening, or that some day, more than 20 years, later, I would be writing about this very program, which I saw reach a major inflection point.

The Topaz International program will be the subject of our next blog post. It’s likely to be a longer one (as this was), so it may take me a little longer than a week to get out, but the ability to compare and contrast Soviet and American testing standards on the same system is too golden an opportunity to pass up.

Stay tuned! More is coming soon!

References:

Topaz II Design Evolution, Voss 1994 https://www.researchgate.net/publication/234517721_TOPAZ_II_Design_Evolution

Russian Topaz II Test Program, Voss 1993 http://gnnallc.com/pdfs_r/SD%2006%20LA-UR-93-3398.pdf

Overview of the Nuclear Electric Propulsion Space Test Program, Voss 1994 https://www.osti.gov/servlets/purl/10157573

Thermionic System Evaluation Test: Ya-21U System, Topaz International Program, Schmidt et al 1996 http://www.dtic.mil/dtic/tr/fulltext/u2/b222940.pdf

Categories
Development and Testing Fission Power Systems History Nuclear Electric Propulsion

SNAP-50: The Last of the SNAP Reactors

Hello, and welcome to Beyond NERVA, for our first blog post of the year! Today, we reach the end of the reactor portion of the SNAP program. A combination of the holidays and personal circumstances prevented me from finishing this post as early as I would have liked to, but it’s finally here! Check the end of the blog post for information on an upcoming blog format change. [Author’s note: somehow the references section didn’t attach to the original post, that issue is now corrected, and I apologize, references are everything in as technical a field as this.]

The SNAP-50 was the last, and most powerful, of the SNAP series of reactors, and had a very different start when compared to the other three reactors that we’ve looked at. A fifth reactor, SNAP-4, also underwent some testing, but was meant for undersea applications for the Navy. The SNAP-50 reactor started life in the Aircraft Nuclear Propulsion program for the US Air Force, and ended its life with NASA, as a power plant for the future modular space station that NASA was planning before the budget cuts of the mid to late 1970s took hold.

Because it came from a different program originally, it also uses different technology than the reactors we’ve looked at on the blog so far: uranium nitride fuel, and higher-temperature, lithium coolant made this reactor a very different beast than the other reactors in SNAP. However, these changes also allowed for a more powerful reactor, and a less massive power plant overall, thanks to the advantages of the higher-temperature design. It was also the first major project to move the space reactor development process away from SNAP-2/10A legacy designs.

The SNAP-50 would permanently alter the way that astronuclear reactors were designed, and would change the course of in-space reactor development for over 20 years. By the time of its cancellation in 1973, it had approached flight readiness to the point that funding and time allowed, but changes in launch vehicle configuration rang the death knell of the SNAP-50.

The Birth of the SNAP-50

Mockup of SNAP-50, image DOE

Up until now, the SNAP program had focused on a particular subset of nuclear reactor designs. They were all fueled with uranium-zirconium hydride fuel (within a small range of uranium content, all HEU), cooled with NaK-78, and fed either mercury Rankine generators or thermoelectric power conversion systems. This had a lot of advantages for the program: fuel element development improvements for one reactor could be implemented in all of them, challenges in one reactor system that weren’t present in another allowed for distinct data points to figure out what was going on, and the engineers and reactor developers were able to look at each others’ work for ideas on how to improve reliability, efficiency, and other design questions.

Tory IIA reactor inlet end, image DOE

However, there was another program that was going on at about the same time which had a very different purpose, but similar enough design constraints that it could be very useful for an in-space fission power plant: the Aircraft Nuclear Propulsion program (ANP), which was primarily run out of Oak Ridge National Laboratory. Perhaps the most famous part of the ANP program was the series of direct cycle ramjets for Project PLUTO: the TORY series. These ramjets were nuclear fission engines using the atmosphere itself as the working fluid. There were significant challenges to this approach, because the clad for the fuel elements must not fail, or else the fission products from the fuel elements would be released as what would be virtually identical to nuclear fallout, only different due to the method that it was generated. The fuel elements themselves would be heavily eroded by the hot air moving through the reactor (which turned out to be a much smaller problem than was initially anticipated). The advantage to this system, though, is that it was simple, and could be made to be relatively lightweight.

Another option was what was known as the semi-indirect cycle, where the reactor would heat a working fluid in a closed loop, which would then heat the air through a heat exchanger built into the engine pod. While this was marginally safer from a fission product release point of view, there were a number of issues with the design. The reactor would have to run at a higher temperature than the direct cycle, because there are always losses whenever you transfer heat from one working fluid to another, and the increased mass of the system also required greater thrust to maintain the desired flight characteristics. The primary coolant loop would become irradiated when going through the reactor, leading to potential irradiation of the air as it passed through the heat exchanger. Another concern was that the heat exchanger could fail, leading to the working fluid (usually a liquid metal) being exposed at high temperature to the superheated air, where it could easily explode. Finally, if a clad failure occurred in the fuel elements, fission products could migrate into the working fluid, making the primary loop even more radioactive, increasing the irradiation of the air as it passed through the engine – and releasing fission products into the atmosphere if the heat exchanger failed.

The alternative to these approaches was an indirect cycle, where the reactor heated a working fluid in a closed loop, transferred this to another working fluid, which then heated the air. The main difference between these systems is that, rather than having the possibly radioactive primary coolant come in close proximity with the air and therefore transferring ionizing radiation, there is an additional coolant loop to minimize this concern, at the cost of both mass and thermal efficiency. This setup allowed for far greater assurances that the air passing through the engine would not be irradiated, because the irradiation of the secondary coolant loop would be so low as to be functionally nonexistent. However, if the semi-indirect cycle was more massive, this indirect cycle would be the heaviest of all of the designs, meaning far higher power outputs and temperatures were needed in order to get the necessary thrust-to-weight ratios for the aircraft. Nevertheless, from the point of view of the people responsible for the ANP program, this was the most attractive design for a crewed aircraft.

Both SNAP and ANP needed many of the same things out of a nuclear reactor: it had to be compact, it had to be lightweight, it had to have a VERY high power density and it needed to be able to operate virtually maintenance-free in a variety of high-power conditions. These requirements are in stark contrast to terrestrial, stationary nuclear reactors which can afford heavy weight, voluminous construction and can thus benefit of low power density. As a general rule of thumb, an increase in power density, will also intensify the engineering, materials, and maintenance challenges. The fact that the ANP program needed high outlet temperatures to run a jet engine also bore the potential of having a large thermal gradient across a power conversion system – meaning that high-conversion-efficiency electrical generation was possible. That led SNAP program leaders to see about adapting an aircraft system into a spacecraft system.

Image DOE

The selected design was under development at the Connecticut Advanced Nuclear Engine Laboratory (CANEL) in Middletown, Connecticut. The prime contractor was Pratt and Whitney. Originally part of the indirect-cycle program, the challenges of heat exchanger design, adequate thrust, and a host of other problems continually set back the indirect cycle program, and when the ANP program was canceled in 1961, Pratt and Whitney no longer had a customer for their reactor, despite doing extensive testing and even fabricating novel alloys to deal with certain challenges that their reactor design presented. This led them to look for another customer for the reactor, and they discovered that both NASA and the US Air Force were both interested in high-power-density, high temperature reactors for in-space use. Both were interested in this high powered reactor, and the SNAP-50 was born.

PWAR-20 cross-section and elevation, image DOE

This reactor was an evolution of a series of test reactors, the PWAR series of test reactors. Three reactors (the PWAR-2, -4, and -8, for 2, 4, and 8 MW of thermal power per reactor core) had already been run for initial design of an aircraft reactor, focused on testing not only the critical geometry of the reactor, but the materials needed to contain its unique (at the time) coolant: liquid lithium. This is because lithium has an excellent specific heat capacity, or the amount of energy that can be contained as heat per unit mass at a given temperature: 3.558 J/kg-C, compared to the 1.124 J/kg-C of NaK78, the coolant of the other SNAP reactors. This means that less coolant would be needed to transport the energy away from the reactor and into the engine in the ANP program, and for SNAP this meant that less working fluid mass would be needed transferring from the reactor to the power conversion system. The facts that Li is much less massive than NaK, and that less of it would be needed, makes lithium a highly coveted option for an astronuclear reactor design. However, this design decision also led to needing novel concepts for how to contain liquid lithium. Even compared to NaK, lithium is highly toxic, highly corrosive in most materials and led, during the ANP program, to Pratt and Whitney investigating novel elemental compositions for their containment structures. We’ll look at just what they did later.

SNAP-50: Designing the Reactor Core

This reactor ended up using a form of fuel element that we have yet to look at in this blog: uranium nitride, UN. While both UC (you can read more about carbide fuels here) and UN were considered at the beginning of the program, the reactor designers ended up settling on UN because of a unique capacity that this fuel form offers: it has the highest fissile fuel density of any type of fuel element. This is offset by the fact that UN isn’t the most heat tolerant of fuel elements, requiring a lower core operating temperature. Other options were considered as well, including CERMET fuels using oxides, carbides, and nitrides suspended in a tungsten metal matrix to increase thermal conductivity and reduce the temperature of the fissile fuel itself. The decision between UN, with its higher mass efficiency (due to its higher fissile density), and uranium carbide (UC), with the highest operating temperature of any solid fuel element, was a difficult decision, and a lot of fuel element testing occurred at CANEL before a decision was reached. After a lot of study, it was determined that UN in a tungsten CERMET fuel was the best balance of high fissile fuel density, high thermal conductivity, and the ability to manage low fuel burnup over the course of the reactor’s life.

From SNAP-50/SPUR Design Summary

Perhaps the most important design consideration for the fuel elements after the type of fuel was how dense the fuel would be, and how to increase the density if this was desired in the final design. While higher density fuel is generally speaking a better idea when it comes to specific power, it was discovered that the higher density the fuel was, the lower the amount of burnup would be possible before the fuel would fail due to fission product gas buildup within the fuel itself. Initial calculations showed that there was an effectively unlimited fuel burnup potential of UN at 80% of its theoretical density since a lot of the gasses could diffuse out of the fuel element. However, once the fuel reached 95% density, this was limited to 1% fuel burnup. Additional work was done to determine that this low burnup was in fact not a project killer for a 10,000 hour reactor lifetime, as was specified by NASA, and the program moved ahead.

These fuel pellets needed a cladding material, as most fuel does, and this led to some additional unique materials challenges. With the decision to use lithium coolant, and the need for both elasticity and strength in the fuel element cladding (to deal with both structural loads and fuel swelling), it was necessary to do extensive experimentation on the metal that would be used for the clad. Eventually, a columbium-zirconium alloy with a small amount of carbon (CB-1ZR-0.6C) was decided on as a barrier between the Cb-Zr alloy of the clad (which resisted the high-temperature lithium erosion on the pressure vessel side of the clad) and the UN-W CERMET fuel (which would react strongly without the carburized layer).

This decisions led to an interesting reactor design, but not necessarily one that is unique from a non-materials point of view. The fuel would be formed into high-density pellets, which would then be loaded into a clad, with a spring to keep the fuel to the bottom (spacecraft end) of the reactor. The gap between the top of the fuel elements and the top of the clad was for the release of fission product gasses produced during operation of the reactor. These rods would be loaded in a hexagonal prism pattern into a larger collection of fuel elements, called a can. Seven of these cans, placed side by side (one regular hexagon, surrounded by six slightly truncated hexagons), would form the fueled portion of the reactor core. Shims of beryllium would shape the core into a cylinder, which was surrounded by a pressure vessel and lateral reflectors. Six poison-backed control drums mounted within the reflector would rotate to provide reactor control. Should the reactor need to be scrammed, a spring mechanism would return all the drums to a position with the neutron poison facing the reactor, stopping fission from occurring.

SNAP-50 flow diagram, image DOE

The lithium, after being heated to a temperature of 2000°F (1093°C), would feed into a potassium boiler, before being returned to the core at an inlet temperature of 1900 F (1037°C). From the boiler, the potassium vapor, which is 1850°F (1010°C), would enter a Rankine turbine which would produce electricity. The potassium vapor would cool down to 1118°F (603°C) in the process and return – condensed to its liquid form – to the boiler, thus closing the circulation. Several secondary coolant loops were used in this reactor: the main one was for the neutron reflectors, shield actuators, control drums, and other radiation hardened equipment, and used NaK as a coolant; this coolant was also used as a lubricant for the condensate pump in the potassium system. Another, lower temperature organic coolant was used for other systems that weren’t in as high a radiation flux. The radiators that were used to reject heat also used NaK as a working fluid, and were split into a primary and secondary radiator array. The primary array pulled heat from the condenser, and reduced it from 1246°F (674°C) to 1096°F (591°C), while the secondary array took the lower-temperature coolant from 730°F (388°C) to 490°F (254°C). This design was designed to operate in both single and dual loop situations, with the second (identical) loop used for high powered operation and to increase redundancy in the power plant.

These design decisions led to a flexible reactor core size, and the ability to adapt to changing requirements from either NASA or the USAF, both of which were continuing to show interest in the SNAP-50 for powering the new, larger space stations that were becoming a major focus of both organizations.

The Power Plant: Getting the Juice Flowing

By 1973, the SNAP 2/10A program had ended, and the SNAP-8/ZrHR program was winding down. These systems simply didn’t provide enough power for the new, larger space station designs that were being envisaged by NASA, and the smaller reactor sizes (the 10B advanced designs that we looked at a couple blog posts back, and the 5 kWe Thermoelectric Reactor) didn’t provide capabilities that were needed at the time. This left the SNAP-50 as the sole reactor design that was practical to take on a range of mission types… but there was a need to have different reactor power outputs, so the program ended up developing two reactor sizes. The first was a 35 kWe reactor design, meant for smaller space stations and lunar bases, although this particular part of the 35 kWe design seems to have never been fully fleshed out. A larger, 300 kWe type was designed for NASA’s proposed modular space station, a project which would eventually evolve into the actual ISS.

Unlike in the SNAP-2 and SNAP-8 programs, the SNAP-50 kept its Rankine turbine design, which had potassium vapor as its working fluid. This meant that the power plant was able to meet its electrical power output requirements far more easily than the lower efficiency demanded by thermoelectric conversion systems. The CRU system meant for the SNAP-2 ended up reaching its design requirements for reliability and life by this time, but sadly the overall program had been canceled, so there was no reactor to pair to this ingenious design (sadly, it’s so highly toxic that testing would be nearly impossible on Earth). The boiler, pumps, and radiators for the secondary loop were tested past the 10,000 hour design lifetime of the power plant, and all major complications discovered during the testing process were addressed, proving that the power conversion system was ready for the next stage of testing in a flight configuration.

One concern that was studied in depth was the secondary coolant loop’s tendency to become irradiated in the neutron flux coming off the reactor. Potassium has a propensity for absorbing neutrons, and in particular 41K (6% of unrefined K) can capture a neutron and become 42K. This is a problem, because 42K goes through gamma decay, so anywhere that the secondary coolant goes needs to have gamma radiation shielding to prevent the radiation from reaching the crew. This limited where the power conversion system could be mounted, to keep it inside the gamma shielding of the temporary, reactor-mounted shield, however the compact nature of both the reactor core and the power conversion system meant that this was a reasonably small concern, but one worthy of in-depth examination by the design team.

The power conversion system and auxiliary equipment, including the actuators for the control drums, power conditioning equipment, and other necessary equipment was cooled by a third coolant loop, which used an organic coolant (basically the oil needed for the moving parts to be lubricated), which ran through its own set of pumps and radiators. This tertiary loop was kept isolated from the vast majority of the radiation flux coming off the reactor, and as such wasn’t a major concern for irradiation damage of the coolant/lubricant.

Some Will Stay, Some Will Go: Mounting SNAP-50 To A Space Station

SNAP50 mounted to early NASA modular space station concept, image DOE

Each design used a 4-pi (a fully enclosing) shield with a secondary shadow shield pointing to the space station in order to reduce radiation exposure for crews of spacecraft rendezvousing or undocking from the space station. This primary shield was made out of a layer of beryllium to reflect neutrons back into the core, and boron carbide (B4C, enriched in boron-10) to absorb the neutrons that weren’t reflected back into the core. These structures needed to be cooled to ensure that the shield wouldn’t degrade, so a NaK shield coolant system (using technology adapted from the SNAP-8 program) was used to keep the shield at an acceptable temperature.

The shadow shield was built in two parts: the entire structure would be launched at the same time for the initial reactor installation for the space station, and then when the reactor needed to be replaced only a portion of the shield would be jettisoned with the reactor. The remainder, as well as the radiators for the reactor’s various coolant systems, would be kept mounted to the space station in order to reduce the amount of mass that needed to be launched for the station resupply. The shadow shield was made out of layers of tungsten and LiH, for gamma and neutron shielding respectively.

Image DOE

When it came time to replace the core of the reactor at the end of its 10,000 hour design life (which was a serious constraint on the UN fuels that they were working with due to fuel burnup issues), everything from the separation plane back would be jettisoned. This could theoretically have been dragged to a graveyard orbit by an automated mission, but the more likely scenario at the time would have been to leave it in a slowly degrading orbit to give the majority of the short-lived isotopes time to decay, and then design it to burn up in the atmosphere at a high enough altitude that diffusion would dilute the impact of any radioisotopes from the reactor. This was, of course, before the problems that the USSR ran into with their US-A program [insert link], which eliminated this lower cost decommissioning option.

Image DOE

After the old reactor core was discarded, the new core, together with the small forward shield and power conversion system, could be put in place using a combination of off-the-shelf hardware, which at the time was expected to be common enough: either Titan-III or Saturn 1B rockets, with appropriate upper stages to handle the docking procedure with the space station. The reactor would then be attached to the radiator, the docking would be completed, and within 8 hours the reactor would reach steady-state operations for another 10,000 hours of normal use. The longest that the station would be running on backup power would be four days. Unfortunately, information on the exact docking mechanism used is thin, so the details on how they planned this stage are still somewhat hazy, but there’s nothing preventing this from being done.

A number of secondary systems, including accumulators, pumps, and other equipment are mounted along with the radiator in the permanent section of the power supply installation. Many other systems, especially anything that has been exposed to a large radiation flux or high temperatures during operation (LiH, the primary shielding material, loses hydrogen through outgassing at a known rate depending on temperature, and can almost be said to have a half-life), will be separated with the core, but everything that was practicable to leave in place was kept.

This basic design principle for reloadable (which in astronuclear often just means “replaceable core”) reactors will be revisited time and again for orbital installations. Variations on the concept abound, although surface power units seem to favor “abandon in place” far more. In the case of large future installations, it’s not unreasonable to suspect that refueling of a reactor core would be possible, but at this point in astronuclear mission utilization, even having this level of reusability was an impressive feat.

35 kWe SNAP-50: The Starter Model

In the 1960s, having 35 kWe of power for a space station was considered significant enough to supply the vast majority of mission needs. Because of this, a smaller version of the SNAP-50 was designed to fit this mission design niche. While the initial power plant would require the use of a Saturn 1B to launch it into orbit, the replacement reactors could be launched on either an Atlas-Centaur or Titan IIIA-Centaur launch vehicle. This was billed as a low cost option, as a proof of concept for the far larger – and at this point, far less fully tested – 300 kWe version to come.

NASA was still thinking of very large space stations at this time. The baseline crew requirements alone were incredible: 24-36 crew, with rotations lasting from 3 months to a year, and a station life of five years. While 35 kWe wouldn’t be sufficient for the full station, it would be an attractive option. Other programs had looked at nuclear power plants for space stations as well, like we saw with the Manned Orbiting Laboratory and the Orbital Workshop (later Skylab), and facilities of that size would be good candidates for the 35 kWe system.

The core itself measured 8.3 inches (0.211 m) across, 11.2 inches (0.284 m) long, and used 236 fuel elements arranged into seven fuel element cans within the pressure vessel of the core. Six poison-backed control drums were used for primary reactor control. The core would produce up to 400 kW of thermal power. The pressure vessel, control drums, and all other control and reflective materials together measured just 19.6 inches (4.98 m) by 27.9 inches (7.09 m), and the replaceable portion of the reactor was between four and five feet (1.2 m and 1.5 m) tall, and five and six feet (1.5 m and 1.8 m) across – including shielding.

SNAP-50 powered probe concept, image DOE

This reactor could also have been a good prototype reactor for a nuclear electric probe, a concept that will be revisited later, although there’s little evidence that this path was ever seriously explored. Like many smaller reactor designs, this one did not get the amount of attention that its larger brother offered, but at the time this was considered a good, solid space station power supply.

300 kWe SNAP-50: The Most Powerful Space Reactor to Date

While there were sketches for more powerful reactors than the 300 kWe SNAP-50 variant, they never really developed the reactors to any great extent, and certainly not to the point of experimental verification that SNAP-50 had achieved. This was considered to be a good starting point for possibly a crewed nuclear electric spacecraft, as well as being able to power a truly huge space station.

The 300 kWe variant of the reactor was slightly different in more than size when compared to its smaller brother. Despite using the same fuel, clad, and coolant as the 35 kWe system, the 300 kWe system could achieve over four times the fuel burnup of the smaller reactor (0.32% vs 1.3%), and had a higher maximum fuel power density as well, both of which have a huge impact on core lifetimes and dynamics. This was partially achieved by making the fuel elements almost half as narrow, and increasing the number of fuel elements to 1093, held in 19 cans within the core. This led to a core that was 10.2 inches (0.259 m) wide, and 14.28 inches (0.363 m) long (keeping the same 1:1.4 gore geometry between the reactors), and a pressure vessel that was 12” (0.305 m) in diameter by 43” (1.092 m) in length. It also increased the thermal output of the reactor to 2200 kWt. The number of control drums was increased from six to eight longer control drums to fit the longer core, and some rearrangement of lithium pumps and other equipment for the power conversion system occurred within the larger 4 pi shield structure. The entire reactor assembly that would undergo replacement was five to six feet high, and six to seven feet in diameter (1.5 m; 1.8 m; 2.1 m).

Lander-based SNAP-50 concept, image DOE

Sadly, even the ambitious NASA space station wasn’t big enough to need even the smaller 35 kWe version of the reactor, much less the 300 kWe variants. Plans had been made for a fleet of nuclear electric tugs that would ferry equipment back and forth to a permanent Moon base, but cancellation of that program occurred at the same time as the death of the moon base itself.

Mass Tradeoffs: Why Nuclear Instead of Solar?

By the middle of the 1960s, photovoltaic solar panels had become efficient and reliable enough for use in spacecraft on a regular basis. Because of this, it was a genuine question for the first time ever whether to go with solar panels or a nuclear reactor, whereas in the 1950s and early 60s nuclear was pretty much the only option. However, solar panels have a downside: drag. Even in orbit, there is a very thin atmosphere, and so for lower orbits a satellite has to regularly raise itself up or it will burn up in the atmosphere. Another down side comes from MM/OD: micro meteorites and orbital debris. Since solar panels are large, flat, and all pointing at the sun all the time, there’s a greater chance that something will strike one of those panels, damaging or possibly even destroying it. Managing these two issues is the primary concern of using solar panels as a power supply in terms of orbital behavior, and determines the majority of the refueling mass needed for a solar powered space station.

Image DOE, from SNAP-50 Design Summary

On the nuclear side, by 1965, there were two power plant options on the table: the SNAP-8 (pre-ZrHR redesign) and the SNAP-50, and solar photovoltaics had developed to the point that they could be deployed in space. Because of this, a comparison was done by Pratt and Whitney of the three systems to determine the mass efficiency of each system, not only in initial deployment but also in yearly fueling and tankage requirements. Each of the systems was compared at a 35 kWe power level to the space station in order to allow for a level playing field.

One thing that stands out about the solar system (based on a pair of Lockheed and General Electric studies) is that it’s marginally the lightest of all the systems at launch, but within a year the total system maintenance mass required far outstrips the mass of the nuclear power plants, especially the SNAP-50. This is because the solar panels have a large sail area, which catches the very thin atmosphere at the station’s orbital altitude and drags the station down into the thicker atmosphere, so thrust is needed to re-boost the space station. This is something that has to be done on a regular basis for the ISS. The mass of the fuel, tankage, and structure to allow for this reboost is extensive. Even back in 1965 there were discussions on using electric propulsion for the reboosting of the space station, in order to significantly reduce the mass needed for this procedure. That discussion is still happening casually with the ISS, and Ad Astra still hopes to use VASIMR for this purpose – a concept that’s been floated for the last ten or so years.

Overall, the mass difference between the SNAP-50 and the optimistic Lockheed proposal of the time was significant: the original deployment was only about 70 lbs (31.75 kg) different, but the yearly maintenance mass requirements would be 5,280 lbs (2395 kg) different – quite a large amount of mass.

Because the SNAP-50 and SNAP-8 don’t have these large sail areas, and the radiators needed can be made aerodynamically enough to greatly reduce the drag on the station, the reboost requirements are significantly lower than for the solar panels. The SNAP-50 weighs significantly less than the SNAP-8, and has significantly less surface area, because the reactor operates at a far higher temperature, and therefore needs a smaller radiator. Another difference between the reactors is volume: the SNAP-50 is physically smaller than the SNAP-8 because of that same higher temperature, and also due to the fact that the UN fuel is far more dense than its U-ZrH fueled counterpart.

These reactors were designed to be replaced once a year, with the initial launch being significantly more massive than the follow-up launches, benefitting of the sectioned architecture with a separation plane just at the small of the shadow shield as described above. Only the smaller section of shield remained with the reactor when it was separated. The larger, heavier section, on the other hand, would remain with the space station, as well as the radiators, and serve as the mounting point for the new reactor core and power conversion system, which would be sent via an automated refueling launch to the space station.

Solar panels, on the other hand, require both reboost to compensate for drag as well as equipment to repair or replace the panels, batteries, and associated components as they wear out. This in turn requires a somewhat robust repair capability for ongoing maintenance – a requirement for any large, long term space station, but the more area you have to get hit by space debris, which means more time and mass spent on repairs rather than doing science.

Of course, today solar panels are far lighter, and electric thrusters are also far more mature than they were at that time. This, in addition to widespread radiophobia, make solar the most widespread occurrence in most satellites, and all space stations, to date. However, the savings available in overall lifetime mass and a sail area that is both smaller and more physically robust, remain key advantages for a nuclear powered space station in the future

The End of an Era: Changing Priorities, Changing Funding

The SNAP-50, even the small 35 kWe version, offered more power, more efficiency, and less mass and volume than the most advanced of SNAP-8’s children: the A-ZrHR [Link]. This was the end of the zirconium hydride fueled reactor era for the Atomic Energy Commission, and while this type of fuel continues to be used in reactors all over the world in TRIGA research and training reactors (a common type of small reactor for colleges and research organizations), its time as the preferred fuel for astronuclear designs was over.

In fact, by the end of the study period, the SNAP-50 was extended to 1.5 MWe in some designs, the most powerful design to be proposed until the 1980s, and one of the most powerful ever proposed… but this ended up going nowhere, as did much of the mission planning surrounding the SNAP program.

At the same time as these higher-powered reactor designs were coming to maturity, funding for both civilian and military space programs virtually disappeared. National priorities, and perceptions of nuclear power, were shifting. Technological advances eliminated many future military crewed missions in favor of uncrewed ones with longer lifetimes, less mass, less cost – and far smaller power requirements. NASA funding began falling under the axe even as we were landing on the Moon for the first time, and from then on funding became very scarce on the ground.

The transition from the Atomic Energy Commission to the Department of Energy wasn’t without its hiccups, or reductions in funding, either, and where once every single AEC lab seemed to have its own family of reactor designs, the field narrowed greatly. As we’ll see, even at the start of Star Wars the reactor design was not too different from the SNAP-50.

Finally, the changes in launch system had their impact as well. NASA was heavily investing in the Space Transport System (the Space Shuttle), which was assumed to be the way that most or all payloads would be launched, so the nuclear reactor had to be able to be flown up – and in some cases returned – by the Shuttle. This placed a whole different set of constraints on the reactor, requiring a large rewrite of the basic design. The follow-on design, the SP-100, used the same UN fuel and Li coolant as the SNAP-50, but was designed to be launched and retrieved by the Shuttle. The fact that the STS never lived up to its promise in launch frequency or cost (and that other launchers were available continuously) means that this was ultimately a diversion, but at the time it was a serious consideration.

All of this spelled the death of the SNAP-50 program, as well as the end of dedicated research into a single reactor design until 1983, with the SP-100 nuclear reactor system, a reactor we’ll look at another time.

While I would love to go into many of the reactors that were developed up to this time, including heat pipe cooled reactors (SABRE at Los Alamos), thermionic power conversion systems (5 kWe Thermionic Reactor), and other ideas, there simply isn’t time to go into them here. As we look at different reactor components they’ll come up, and we’ll mention them there. Sadly, while some labs were able to continue funding some limited research with the help of NASA and sometimes the Department of Defense or the Defense Nuclear Safety Agency. The days of big astronuclear programs, though, were fading into a thing of the past. Both space and nuclear power would refocus, and then fade in the rankings of budgetary requirements over the years. We will be looking at these reactors more as time goes on, in our new “Forgotten Reactors” column (more on that below).

The Blog is Changing!

With the new year, I’ve been thinking a lot about the format of both the website and the blog, and where I hope to go in the next year. I’ve had several organizational projects on the back burner, and some of them are going to be started here soon. The biggest part is going to be the relationship between the blog and the website, and what I write more about where.

Expect another blog post shortly (it’s already written, just not edited yet) about our plans for the next year!

I’ve got big plans for Beyond NERVA this year, and there are a LOT of things that are slowly getting started in the background which will greatly improve the quality of the blog and the website, and this is just the start!

References

SNAP-50/SPUR Program Summary, Pratt and Whitney staff, 1964 https://www.osti.gov/servlets/purl/4307107

35 and 300 kWe SNAP-50/SPUR Power Plants for the Manned Orbiting Space Station Application, Pratt and Whitney staff, 1965 https://www.osti.gov/servlets/purl/4307103

Uranium Nitride Fuel Development SNAP-50, Pratt and Whitney staff, 1965
https://www.osti.gov/servlets/purl/4324037

SNAP Program Summary Report, Voss 1984 https://apps.dtic.mil/dtic/tr/fulltext/u2/a146831.pdf

Categories
Development and Testing Fission Power Systems History Spacecraft Concepts

US Astro-nuclear History part 2: SNAP-8, NASA’s Space Station Power Supply

Hello, and welcome back to Beyond NERVA! As some of you may have noticed, the website has moved! Yes, we’re now at beyondnerva.com! I’m working on updating the webpage, and am getting the pieces together for a major website redesign (still a ways off, but lots of the pieces are starting to fall into place) to make the site easier to navigate and more user friendly. Make sure to update your bookmarks with this new address! With that brief administrative announcement out of the way, let’s get back to our look at in-space fission power plants.

Today, we’re going to continue our look at the SNAP program, America’s first major attempt to provide electric power in space using nuclear energy, and finishing up our look at the zirconium hydride fueled reactors that defined the early SNAP reactors by looking at the SNAP-8, and its two children – the 5 kW Thermoelectric Reactor and the Advanced Zirconium Hydride Reactor.

SNAP 8 was the first reactor designed with these space stations in mind in mind. While SNAP-10A was a low-power system (at 500 watts when flown, later upgraded to 1 kW), and SNAP-2 was significantly larger (3 kW), there was a potential need for far more power. Crewed space stations take a lot of power (the ISS uses close to 100 kWe, as an example), and neither the SNAP-10 or the SNAP-2 were capable of powering the space stations that NASA was in the beginning stages of planning.

Initially designed to be far higher powered, with 30-60 kilowatts of electrical power, this was an electric supply that could power a truly impressive outpost for humanity in orbit. However, the Atomic Energy Commission and NASA (which was just coming into existence at the time this program was started) didn’t want to throw the baby out with the bath water, as it were. While the reactor was far higher powered than the SNAP 2 reactor that we looked at last time, many of the power system’s components are shared: both use the same fuel (with minor exceptions), both use similar control drum structures for reactor control, both use mercury Rankine cycle power conversion systems, and perhaps most attractively both were able to evolve with lessons learned from the other part of the program.

While SNAP 8 never flew, it was developed to a very high degree of technical understanding, so that if the need for the reactor arose, it would be available. One design modification late in the SNAP 8 program (when the reactor wasn’t even called SNAP 8 anymore, but the Advanced Zirconium Hydride Reactor) had a very rare attribute in astronuclear designs: it was shielded on all sides for use on a space station, providing more than twice the electrical power available to the International Space Station without any of the headaches normally associated with approach and docking with a nuclear powered facility.

Let’s start back in 1959, though, with the SNAP 8, the first nuclear electric propulsion reactor system.

SNAP 8: NASA Gets Involved Directly

The SNAP 2 and SNAP 10A reactors were both collaborations between the Atomic Energy Commission (AEC), who were responsible for the research, development, and funding of the reactor core and primary coolant portions of the system, and the US Air Force, who developed the secondary coolant system, the power conversion system, the heat rejection system, the power conditioning unit, and the rest of the components. Each organization had a contractor that they used: the AEC used Atomics International (AI), one of the movers and shakers of the advanced reactor industry, while the US Air Force went to Thompson Ramo Wooldridge (better known by their acronym, TRW) for the SNAP-2 mercury (Hg) Rankine turbine and Westinghouse Electric Corporation for the SNAP-10’s thermoelectric conversion unit.

S8ER Slightly Disassembled
SNAP-8 with reflector halves slightly separated, image DOE

1959 brought NASA directly into the program on the reactor side of things, when they requested a fission reactor in the 30-60 kWe range for up to one year; one year later the SNAP-8 Reactor Development Program was born. It would use a similar Hg-based Rankine cycle as the SNAP-2 reactor, which was already under development, but the increased power requirements and unique environment that the power conversion system necessitated significant redesign work, which was carried out by Aerojet General as the prime contractor. This led to a 600 kWe rector core, with a 700 C outlet temperature As with the SNAP-2 and SNAP-10 programs, the SNAP 8’s reactor core was funded by the AEC, but in this case the power conversion system was the funding responsibility of NASA.

The fuel itself was similar to that in the SNAP-2 and -10A reactors, but the fuel elements were far longer and thinner than those of the -2 and -10A. Because the fuel element geometry was different, and the power level of the reactor was so much higher than the SNAP-2 reactor, the SNAP-8 program required its own experimental and developmental reactor program to run in parallel to the initial SNAP Experimental and Development reactors, although the materials testing undertaken by the SNAP-2 reactor program, and especially the SCA4 tests, were very helpful in refining the final design of the SNAP-8 reactor.

The power conversion system for this reactor was split in two: identical Hg turbines would be used, with either one or both running at any given time depending on the power needs of the mission. This allows for more flexibility in operation, and also simplifies the design challenges involved in the turbines themselves: it’s easier to design a turbine with a smaller power output range than a larger one. If the reactor was at full power, and both turbines were used, the design was supposed to produce up to 60 kW of electrical power, while the minimum power output of a single turbine would be in the 30 kWe range. Another advantage was that if one was damaged, the reactor would continue to be able to produce power.

S8ER Assembly 1962
SNAP-8 Experimental Reactor being assembled, 1962

Due to the much higher power levels, an extensive core redesign was called for, meaning that different test reactors would need to be used to verify this design. While the fuel elements were very similar, and the overall design philosophy was operating in parallel to the SNAP-2/10A program, there was only so much that the tests done for the USAF system would be able to help the new program. This led to the SNAP-8 development program, which began in 1960, and had its first reactor, the SNAP-8 Experimental Reactor, come online in 1963.

SNAP-8 Experimental Reactor: The First of the Line

S8ER Cutaway
Image DOE

The first reactor in this series, the SNAP 8 Experimental Reactor (S8ER), went critical in May 1963, and operated until 1965. it operated for 2522 hours at above 600 kWt, and over 8000 hours at lower power levels. The fuel elements for the reactor were 14 inches in length, and 0.532 inches in diameter, with uranium-zirconium hydride (U-ZrH, the same basic fuel type as the SNAP-2/10A system that we looked at last time) enriched to 93.15% 235U, with 6 X 10^22 atoms of hydrogen per cubic centimeter.

S8ER Cutaway Drawing
S8ER Fuel Element

The biggest chemical change in this reactor’s fuel elements compared to the SNAP-2/10A system was the hydrogen barrier inside the metal clad: instead of using gadolinium as a burnable poison (which would absorb neutrons, then decay into a neutron-transparent element as the reactor underwent fission over time), the S8ER used samarium. The reasons for the change are rather esoteric, relating to the neutron spectrum of the reactor, the particular fission products and their ratios, thermal and chemical characteristics of the fuel elements, and other factors. However, the change was so advantageous that eventually the different burnable poison would be used in the SNAP-2/10A system as well.

S8ER Cross SectionThe fuel elements were still loaded in a triangle array, but makes more of a cylinder than a hexagon like in the -2/10A, with small internal reflectors to fill out the smooth cylinder of the pressure vessel. The base and head plates that hold the fuel elements are very similar to the smaller design, but obviously have more holes to hold the increased number of fuel elements. The NaK-78 coolant (identical to the SNAP-2/10A system) entered in the bottom of the reactor into a space in the pressure vessel (a plenum), flowed through the base plate and up the reactor, then exits the top of the pressure vessel through an upper plenum. A small neutron source used as a startup neutron source (sort of like a spark plug for a reactor) was mounted to the top of the pressure vessel, by the upper coolant plenum. The pressure vessel itself was made out of 316 stainless steel.

Instead of four control drums, the S8ER used six void-backed control drums. These were directly derived from the SNAP-2/10A control system. Two of the drums were used for gross reactivity control – either fully rotated in or out, depending on if the reactor is under power or not. Two were used for finer control, but at least under nominal operation would be pretty much fixed in their location over longer periods of time. As the reactor approached end of life, these drums would rotate in to maintain the reactivity of the system. The final two were used for fine control, to adjust the reactivity for both reactor stability and power demand adjustment. The drums used the same type of bearings as the -2/10A system.

S8ER Facility Setup
SNAP-8 Experimental Reactor test facility, Santa Susana Field Site, image DOE

The S8ER first underwent criticality benchmark tests (pre-dry critical testing) from September to December 1962 to establish the reactor’s precise control parameters. Before filling the reactor with the NaK coolant, water immersion experiments for failure-to-orbit safety testing (as an additional set of tests to the SCA-4 testing which also supported SNAP-8) was carried out between January and March of 1963. After a couple months of modifications and refurbishment, dry criticality tests were once again conducted on May 19, 1963, followed in the next month with the reactor reaching wet critical power levels on June 23. Months of low-power testing followed, to establish the precise reactor control element characteristics, thermal transfer characteristics, and a host of other technical details before the reactor was increased in power to full design characteristics.

S8ER Core Containment StructureThe reactor was shut down from early August to late October, because some of the water coolant channels used for the containment vessel failed, necessitating the entire structure to be dug up, repaired, and reinstalled, with significant reworking of the facility being required to complete this intensive repair process. Further modifications and upgrades to the facility continued into November, but by November 22, the reactor underwent its first “significant” power level testing. Sadly, this revealed that there were problems with the control drum actuators, requiring the reactor to be shut down again.

After more modifications and repairs, lower power testing resumed to verify the repairs, study reactor transient behavior, and other considerations. The day finally came for the SNAP-8 Experimental Reactor achieved its first full power, at temperature testing on December 11, 1963. Shortly after, the reactor had to be shut down again to repair a NaK leak in one of the primary coolant loop pumps, but the reactor was up and operating again shortly after. Lower power tests were conducted to evaluate the samarium burnable poisons in the fuel elements, measure xenon buildup, and measure hydrogen migration in the core until April 28, interrupted briefly by another NaK pump failure and a number of instrumentation malfunctions in the automatic scram system (which was designed to automatically shut down the reactor in the case of an accident or certain types of reactor behaviors). However, despite these problems, April 28 marked 60 days of continuous operation at 450 kWt and 1300 F (design temperature, but less-than-nominal power levels).

S8ER Drive Mechanism
S8ER Control drum actuator and scram mechanism, image DOE

After a shutdown to repair the control drive mechanisms (again), the reactor went into near-continuous operation, either at 450 or 600 kWt of power output and 1300 F outlet temperature until April 15, 1965, when the reactor was shut down for the last time. By September 2 of 1964, the S8ER had operated at design power and temperature levels for 1000 continuous hours, and went on in that same test to exceed the maximum continuous operation time of any SNAP reactor to date on November 5 (1152 hours). January 18 of 1965 it achieved 10,000 hours of total operations, and in February of that year reached 100 days of continuous operation at design power and temperature conditions. Just 8 days later, on February 12, it exceeded the longest continuous operation of any reactor to that point (147 days, beating the Yankee reactor). March 5 marked the one year anniversary of the core outlet temperature being continuously at over 1200 F. By April 15, when the reactor was shut down for the last time it achieved an impressive set of accomplishments:

  1. 5016.5 continuous operations immediately preceeding the shutdown (most at 450 kWt, all at 1200 F or greater)
  2. 12,080 hours of total operations
  3. A total of 5,154,332 kilowatt-hours of thermal energy produced
  4. 91.09% Time Operated Efficiency (percentage of time that the reactor was critical) from November 22, 1963 (the day of first significant power operations of the reactor), and 97.91% efficiency in the last year of operations.

Once the tests were concluded, the reactor was disassembled, inspected, and fuel elements were examined. These tests took place at the Atomics International Hot Laboratory (also at Santa Susana) starting on July 28, 1965. For about 6 weeks, this was all that the facility focused on; the core was disassembled and cleaned, and the fuel elements were each examined, with many of them being disassembled and run through a significant testing regime to determine everything from fuel burnup to fission product percentages to hydrogen migration. The fuel element tests were the most significant, because to put it mildly there were problems.

S8ER FE Damage Map
Core cross section with location and type of damage to fuel elements, image DOE

Of the 211 fuel elements in the core, only 44 were intact. Many of the fuel elements also underwent dimensional changes, either swelling (with a very small number actually decreasing) across the diameter or the length, becoming oblong, dishing, or other changes in geometry. The clad on most elements was damaged in one way or another, leading to a large amount of hydrogen migrating out of the fuel elements, mostly into the coolant and then out of the reactor. This means that much of the neutron moderation needed for the reactor to operate properly migrated out of the core, reducing the overall available reactivity even as the amount of fission poisons in the form of fission products was increasing. For a flight system, this is a major problem, and one that definitely needs to be addressed. However, this is exactly the sort of problem that an experimental reactor is meant to discover and assess, so in this way as well the reactor was a complete success, if not as smooth a development as the designers would likely have preferred.

S8ER FE Damage Photo

It was also discovered that, while the cracks in the clad would indicate that the hydrogen would be migrating out of the cracks in the hydrogen diffusion barrier, far less hydrogen was lost than was expected based on the amount of damage the fuel elements underwent. In fact, the hydrogen migration in these tests was low enough that the core would most likely be able to carry out its 10,000 hour operational lifetime requirement as-is; without knowing what the mechanism that was preventing the hydrogen migration was, though, it would be difficult if not impossible to verify this without extensive additional testing, when changes in the fuel element design could result in a more satisfactory fuel clad lifetime, reduced damage, and greater insurance that the hydrogen migration would not become an issue.

S8ER Post irradiation fuel characteristics

The SNAP-8 Experimental Reactor was an important stepping stone to nuclear development in high-temperature ZrH nuclear fuel development, and greatly changed the direction of the whole SNAP-8 program in some ways. The large number of failures in cladding, the hydrogen migration from the fuel elements, and the phase changes within the crystalline structure of the U-ZrH itself were a huge wake-up call to the reactor developers. With the SNAP-2/10A reactor, these issues were minor at best, but that was a far lower-powered reactor, with very different geometry. The large number of fuel elements, the flow of the coolant through the reactor, and numerous other factors made the S8ER reactor far more complex to deal with on a practical level than most, if any, anticipated. Plating of the elements associated with Hastelloy on the stainless steel elements caused concern about the materials that had been selected causing blockages in flow channels, further exacerbating the problems of local hot spots in the fuel elements that caused many of the problems in the first place. The cladding material could (and would) be changed relatively easily to account for the problems with the metal’s ductility (the ability to undergo significant plastic deformation before rupture, in other words to endure fuel swelling without the metal splitting, cracking, fracturing or other ways that the clad could be breached) under high temperature and radiation fluxes over time. A number of changes were proposed to the reactor’s design, which strongly encouraged – or required – changes in the SNAP-8 Development Reactor that was currently being designed and fabricated. Those changes would alter what the SNAP-8 reactor would become, and what missions it would be proposed for, until the program was finally put to rest.

After the S8ER test, a mockup reactor, the SNAP-8 Development Mockup, was built based on the 1962 version of the design. This mockup never underwent nuclear testing, but was used for extensive non-nuclear testing of the designs components. Basically, every component that could be tested under non-nuclear conditions (but otherwise identical, including temperature, stress loading, vacuum, etc.) was tested and refined with this mockup. The tweaks to the design that this mockup suggested are far more minute than we have time to cover here, but it was an absolutely critical step to preparing the SNAP-8 reactor’s systems for flight test.

SNAP-8 Development Reactor: Facing Challenges with the Design

S8DR with installed reflector assembly
SNAP 8 Development Reactor post-reflector installation, before being lowered into containment structure. Image DOE

The final reactor in the series, the SNAP-8 Development Reactor, was a shorter-lived reactor, in part because many of the questions that needed to be answered about the geometry had been answered by the S8ER, and partly because the unanswered materials questions were able to be answered with the SCA4 reactor. This reactor underwent dry critical testing in June 1968, and power testing began at the beginning of the next year. From January 1969 to December 1969, when the reactor was shut down for the final time, the reactor operated at nominal (600 kWt) power for 668 hours, and operated at 1000 kWt for 429 hours.

S8DR Cutaway Drawing in test vault
S8DR in Test Vault, image DOE

The SNAP-8 Development Reactor (S8DR) was installed in the same facility as the S8ER, although it operated under different conditions than the S8ER. Instead of having a cover gas, the S8DR was tested in a vacuum, and a flight-type radiation shield was mounted below it to facilitate shielding design and materials choices. Fuel loading began on June 18, 1968, and criticality was achieved on June 22, with 169 out of the 211 fuel elements containing the U-ZrH fuel (the rest of the fuel elements were stainless steel “dummy” elements) installed in the core. Reactivity experiments for the control mechanisms were carried out before the remainder of the dummy fuel elements were replaced with actual fuel in order to better calibrate the system.

Finally, on June 28, all the fuel was loaded and the final calibration experiments were carried out. These tests then led to automatic startup testing of the reactor, beginning on December 13, 1968, as well as transient analysis, flow oscillation, and temperature reactivity coefficient testing on the reactor. From January 10 to 15, 1969, the reactor was started using the proposed automated startup process a total of five times, proving the design concept.

1969 saw the beginning of full-power testing, with the ramp up to full design power occurring on January 17. Beginning at 25% power, the reactor was stepped up to 50% after 8 hours, then another 8 hours in it was brought up to full power. The coolant flow rates in both the primary and secondary loops started at full flow, then were reduced to maintain design operating temperatures, even at the lower power setting. Immediately following these tests on January 23, an additional set of testing was done to verify that the power conversion system would start up as well. The biggest challenge was verification that the initial injection of mercury into the boiler would operate as expected, so a series of mercury injection tests were carried out successfully. While they weren’t precisely at design conditions due to test stand limitations, the tests were close enough that it was possible to verify that the design would work as planned.

Control RoomAfter these tests, the endurance testing of the reactor began. From January 25 to February 24 was the 500-hour test at design conditions (600 kWt and 1300 F), although there were two scram incidents that led to short interruptions. Starting on March 20, the 9000 hour endurance run at design conditions lasted until April 10. This was followed by a ramp up to the alternate design power of 1 MWt. While this was meant to operate at only 1100 F (to reduce thermal stress on the fuel elements, among other things), the airblast heat exchanger used for heat rejection couldn’t keep up with the power flow at that temperature, so the outlet temperature was increased to 1150 F (the greater the temperature difference between a radiator and its environment, the more efficient it is, something we’ll discuss more in the heat rejection posts). After 18 days of 1 MWt testing, the power was once again reduced to 600 kWt for another 9000 hour test, but on June 1, the reactor scrammed itself again due to a loss of coolant flow. At this point, there was a significant loss of reactivity in the core, which led the team to decide to proceed at a lower temperature to mitigate hydrogen migration in the fuel elements. Sadly, reducing the outlet temperature (to 1200 F) wasn’t enough to prevent this test from ending prematurely due to a severe loss in reactivity, and the reactor scrammed itself again.

The final power test on the S8ER began on November 20, 1969. For the first 11 days, it operated at 300 kWt and 1200 F, when it was then increased in power back to 600 kWt, but the outlet temperature was reduced to 1140F, for an additional 7 days. An increase of outlet temperature back to 1200 F was then dialed in for the final 7 days of the test, and then the reactor was shut down.

This shutdown was an interesting and long process, especially compared to just removing all the reactivity of the control drums by rotating them all fully out. First, the temperature was dropped to 1000 F while the reactor was still at 600 kWt, and then the reactor’s power was reduced to the point that both the outlet and inlet coolant temperatures were 800 F. This was held until December 21 to study the xenon transient behavior, and then the temperatures were further reduced to 400 F to study the decay power level of the reactor. On January 7, the temperature was once again increased to 750 F, and two days later the coolant was removed. The core temperature then dropped steadily before leveling off at 180-200F.

Once again, the reactor was disassembled and examined at the Hot Laboratory, with special attention being paid to the fuel elements. These fuel elements held up much better than the S8ER’s fuel elements, with only 67 of the 211 fuel elements showing cracking. However, quite a few elements, while not cracked, showed significant dimensional changes and higher hydrogen loss rates. Another curiosity was that a thin (less than 0.1 mil thick) metal film, made up of iron, nickel, and chromium, developed fairly quickly on the exterior of the cladding (the exact composition changed based on location, and therefore on local temperature, within the core and along each fuel element).

S8DR FE Damage Map
S8DR Fuel Element Damage map, image DOE

The fuel elements that had intact cladding and little to no deformation showed very low hydrogen migration, an average of 2.4% (this is consistent with modeling showing that the permeation barrier was damaged early in its life, perhaps during the 1 MWt run). However, those with some damage lost between 6.8% and 13.2 percent of their hydrogen. This damage wasn’t limited to just cracked cladding, though – the swelling of the fuel element was a better indication of the amount of hydrogen lost than the clad itself being split. This is likely due to phase changes in the fuel elements, when the UzrH changes crystalline structure, usually due to high temperatures. This changes how well – and at what bond angle – the hydrogen is kept within the fuel element’s crystalline structure, and can lead to more intense hot spots in the fuel element, causing the problem to become worse. The loss of reactivity scrams from the testing in May-July 1969 seem to be consistent with the worst failures in the fuel elements, called Type 3 in the reports: high hydrogen loss, highly oval cross section of the swollen fuel elements (there were a total of 31 of these, 18 of them were intact, 13 were cracked). One interesting note about the clad composition is that where there was a higher copper content due to irregularities in metallography there was far less swelling of the Hastelloy N clad, although the precise mechanism was not understood at the time (and my rather cursory perusal of current literature didn’t show any explanation either). However, at the time testing showed that these problems could be mitigated, to the point of insignificance even, by maintaining a lower core temperature to ensure localized over-temperature failures (like the changes in crystalline structure) would not occur.

S8DR H loss rate table
Image DOE

The best thing that can be said about the reactivity loss rate (partially due to hydrogen losses, and partially due to fission product buildup) is that it was able to be extrapolated given the data available, and that the failure would have occurred after the design’s required lifetime (had S8DR been operated at design temperature and power, the reactor would have lost all excess reactivity – and therefore the ability to maintain criticality – between October and November of 1970).

On this mixed news note, the reactor’s future was somewhat in doubt. NASA was certainly still interested in a nuclear reactor of a similar core power, but this particular configuration was neither the most useful to their needs, nor was it exceptionally hopeful in many of the particulars of its design. While NASA’s reassessment of the program was not solely due to the S8DR’s testing history, this may have been a contributing factor.

One way or the other, NASA was looking for something different out of the reactor system, and this led to many changes. Rather than an electric propulsion system, focus shifted to a crewed space station, which has different design requirements, most especially in shielding. In fact, the reactor was split into three designs, none of which kept the SNAP name (but all kept the fuel element and basic core geometry).

A New Life: the Children of SNAP-8

Even as the SNAP-8 Development Reactor was undergoing tests, the mission for the SNAP-8 system was being changed. This would have major consequences for the design of the reactor, its power conversion system, and what missions it would be used in. These changes would be so extensive that the SNAP-8 reactor name would be completely dropped, and the reactor would be split into four concepts.

The first concept was the Space Power Facility – Plumbrook (SPT) reactor, which would be used to test shielding and other components at NASA’s Plum Brook Research Center outside Cleveland, OH, and could also be used for space missions if needed. The smallest of the designs (at 300 kWt), it was designed to avoid many of the problems associated with the S8ER and S8DR; however, funding was cut before the reactor could be built. In fact, it was cut so early that details on the design are very difficult to find.

The second reactor, the Reactor Core Test, was very similar to the SPF reactor, but it was the same power output as the nominal “full power” reactor, at 600 kWt. Both of these designs increased the number of control drums to eight, and were designed to be used with a traditional shadow shield. Neither of them were developed to any great extent, much less built.

A third design, the 5 kWe Thermoelectric Reactor, was a space system, meant to take many of the lessons from the SNAP-8 program and apply them to a medium-power system which would apply both the lessons of the SNAP-8 ER and DR as well as the SNAP-10A’s experience with thermoelectric power conversion systems to a reactor between the SNAP-10B and Reference Zirconium Hydride reactor in power output.

The final design, the Reference Zirconium Hydride Reactor (ZrHR), was extensively developed, even if geometry-specific testing was never conducted. This was the most direct replacement for the SNAP-8 reactor, and the last of the major U-ZrH fueled space reactors in the SNAP program. Rather than powering a nuclear electric spacecraft, however, this design was meant to power space stations.

The 5 kWe Thermoelectric Reactor: Simpler, Cleaner, and More Reliable

Artists Depiction
5 kWe Thermoelectric Reactor, artist’s concept cutaway drawing. Image DOE

The 5 kWe Thermoelectric Reactor (5 kWe reactor) was a reasonably simple adaptation of the SNAP-8 design, intended to be used with a shadow shield. Unsurprisingly, a lot of the design changes mirrored some of the work done on the SNAP-10B Interim design, which was undergoing work at about the same time. Meant to supply 5 kWe of power for 5 years using lead telluride thermoelectric convertors (derived from the SNAP-10A convertors), this system was meant to provide power for everything from small crewed space stations to large communications satellites. In many ways, this was a very different departure from the SNAP-8 reactor, but at the same time the changes that were proposed were based on evolutionary changes during the S8ER and S8DR experimental runs, as well as advances in the SNAP 2/10 core which was undergoing parallel post-SNAPSHOT design evolution (the SNAP-10A design had been frozen for the SNAPSHOT program at this point, so these changes were either for the followon SNAP-10A Advanced or SNAP-10B reactors). The change from mercury Rankine to thermoelectric power conversion, though, paralleled a change in the SNAP-2/10A origram, where greater efficiency was seen as unnecessary due to the constantly-lower power requirements of the systems.

5 kWe SchematicThe first thing (in the reactor itself, at least) that was different about the design was that the axial reflector was tapered, rather than cylindrical. This was done to keep the exterior profile of the reactor cleaner. While aerodynamic considerations aren’t a big deal (although they do still play a minute part in low Earth orbit) for astronuclear power plants, everything that’s exposed to the unshielded reactor becomes a radiation source itself, due to radiation scattering and material activation under neutron bombardment. If you could get your reactor to be a continuation of the taper of your shadow shield, rather than sticking out from that cone shape, you can make the shadow shield smaller for a given reactor. Since the shield is often many times heavier than the power system itself, especially for crewed applications, the single biggest place a designer can save mass is in the shadow shield.

This tapered profile meant two things: first, there would be a gradient in the amount of neutron moderation between the top and the bottom of the reactor, and second, the control system would have to be reworked. It’s unclear exactly how far the neutronics analysis for the new reflector configuration had proceeded, sadly, but the control systems were adaptations of the design changes that were proposed for the SNAP-10B reactor: instead of having the wide, partial cylinder control drums of the original design, large sections (235 degrees in total) of the reflector would be slid up or down around the core containment vessel to control the amount of reactivity available. This is somewhat similar to the SNAP-10B and BES-5 concepts in its execution, but the mechanism is quite different from a neutronics perspective: rather than capturing the unwanted neutrons using a neutron poison like boron or europium, they’re essentially vented into space.

A few other big changes from the SNAP-8 reference design when it comes to the core itself. The first is in the fuel: instead of having a single long fuel rod in the clad, the U-XrH fuel was split into five different “slugs,” which were held together by the clad. This would create a far more complex thermal distribution situation in the fuel, but would also allow for better thermal stress management within the hydride itself. The number of fuel elements was reduced to 85, and they came in three configurations: one set of 27 had radial fins to control the flow that spiralled around the fuel element in a right-hand direction, another set of 27 had fins in the left-hand direction, and the final 31 were unfinned. This was done to better manage the flow of the NaK coolant through the core, and avoid some of the hydrodynamic problems that were experienced on the S8DR.

Blueprint Layout of System
5kWe Thermoelectric power system. Image DOE

Summary TableSummary Table 2

The U-ZrH Reactor: Power for America’s Latest and Greatest Space Stations.

ZrHR Cutaway Drawing
Cutaway drawing, Image DOE

The Reference ZrH Reactor was begun in 1968, while the S8DR was still under construction. Because of this increased focus on having a crewed space station configuration, and the shielding requirement changes, some redesign of the reactor core was needed. The axial shield would change the reactivity of the core, and the control drums would no longer be able to effectively expose portions of the core to the vacuum of space to get rid of excess reactivity. Because of this, the number of fuel elements in the core were increased from 211 to 295. Another change was that rather than the even spacing of fuel elements used in the S8DR, the fuel elements were spaced in such a way that the amount of coolant around each fuel element was proportional to the amount of power produced by each fuel element. This means that the fuel elements on the interior of the core were wider spaced than the fuel elements around the periphery. This made it far more unlikely that local hot spots will develop which could lead to fuel element failures, but it also meant that the flow of coolant through the reactor core would need to be far more thoroughly studied than was done on the SNAP 8 reactor design. These thermohydrodynamic studies would be a major focus of the ZrHR program.

Reference Design Xsec and ElevtionAnother change was in the control drum configuration, as well as the need to provide coolant to the drums. This was because the drums were now not only fully enclosed solid cylinders, but were surrounded by a layer of molten lead gamma shielding. Each drum would be a solid cylinder in overall cross section; the main body was beryllium, but a crescent of europium alloy was used as a neutron poison (this is one of the more popular alternatives to boron for control mechanisms that operate in a high temperature environment) to absorb neutrons when this portion of the control drum was turned toward the core. These drums would be placed in dry wells, with NaK coolant flowing around them from the spacecraft (bottom) end before entering the upper reactor core plenum to flow through the core itself. The bearings would be identical to those used on the SNAP-8 Development Reactor, and minimal modifications would be needed for the drum motion control and position sensing apparatus. Fixed cylindrical beryllium reflectors, one small one along the interior radius of the control drums and a larger one along the outside of the drums, filled the gaps left by the control drums in this annular reflector structure. These, too, would be kept cool by the NaK coolant flowing around them.

Surrounding this would be an axial gamma shield, with the preferred material being molten lead encased in stainless steel – but tungsten was also considered as an alternative. Why the lead was kept molten is still a mystery to me, but my best guess is that this was due to the thermal conditions of the axial shield, which would have forced the lead to remain above its melting point. This shield would have made it possible to maneuver near the space station without having to remain in the shadow of the directional shield – although obviously dose rates would still be higher than being aboard the station itself.

TE Layout diagramAnother interesting thing about the shielding is that the shadow shield was divided in two, in order to balance thermal transfer and radiation protection for the power conversion system, and also to maximize the effectiveness of the shadow shields. Most designs used a 4 pi shield design, which is basically a frustrum completely surrounding the reactor core with the wide end pointing at the spacecraft. The primary coolant loops wrapped around this structure before entering the thermoelectric conversion units. After this, there’s a small “galley” where the power conversion system is mounted, followed by a slightly larger shadow shield, with the heat rejection system feed loops running across the outside as well. Finally, the radiator – usually cylindrical or conical – completed the main body of the power system. The base of the radiator would meet up with the mounting hardware for attachment to the spacecraft, although the majority of the structural load was an internal spar running from the core all the way to the spacecraft.

While the option for using a pure shadow shield concept was always kept on the table, the complications in docking with a nuclear powered space station which has an unshielded nuclear reactor at one end of the structure were significant. Because of this, the ZrHR was designed with full shielding around the entire core, with supplementary shadow shields between the reactor itself and the power conversion system, and also a second shadow shield after the power conversion system. These shadow shields could be increased to so-called 4-pi shields for more complete shielding area, assuming the mission mass budget allowed, but as a general rule the shielding used was a combination of the liquid lead gamma shield and the combined shadow shield configuration. These shields would change to a fairly large extent depending on the mission that the ZrHR would be used on.

Radiator Design BaselineAnother thing that was highly variable was the radiator configuration. Some designs had a radiator that was fixed in relation to the reactor, even if it was extended on a boom (as was the case of the Saturn V Orbital Workshop, later known as Skylab). Others would telescope out, as was the case for the later Modular Space Station (much later this became the International Space Station). The last option was for the radiators to be hinged, with flexible joints that the NaK coolant would flow through (this was the configuration for the lunar surface mission), and those joints took a lot of careful study, design, and material testing to verify that they would be reliable, seal properly, and not cause too many engineering compromises. We’ll look at the challenges of designing a radiator in the future, when we look at heat rejection systems (at this point, possibly next summer), but suffice to say that designing and executing a hinged radiator is a significant challenge for engineers, especially with a material at hot, and as reactive, as liquid NaK.

The ZrHR was continually being updated, since there was no reason to freeze the majority of the design components (although the fuel element spacing and fin configuration in the core may have indeed been frozen to allow for more detailed hydrodynamic predictability), until the program’s cancellation in 1973. Because of this, many design details were still in flux, and the final reactor configuration wasn’t ever set in stone. Additional modifications for surface use for a crewed lunar base would have required tweaking, as well, so there is a lot of variety in the final configurations.

The Stations: Orbital Missions for SNAP-8 Reactors

OWS with two CSMs, 1966.PNG
Frontispiece for nuclear-powered Saturn V Orbital Workstation (which flew as Skylab), image NASA 

At the time of the redesign, three space stations were being proposed for the near future: the first, the Manned Orbiting Research Laboratory, (later changed to the Manned Orbiting Laboratory, or MOL), was a US Air Force project as part of the Blue Gemini program. Primarily designed as a surveillance platform, advances in photorecoinnasance satellites made this program redundant after just a single flight of an uncrewed, upgraded Gemini capsule.

The second was part of the Apollo Applications Program. Originally known as the Saturn V Orbital Workshop (OWS), this later evolved into Skylab. Three crews visited this space station after it was launched on the final Saturn V, and despite huge amounts of work needed to repair damage caused during a particularly difficult launch, the scientific return in everything from anatomy and physiology to meteorology to heliophysics (the study of the Sun and other stars) fundamentally changed our understanding of the solar system around us, and the challenges associated with continuing our expansion into space.

The final space station that was then under development was the Modular Space Station, which would in the late 1980s and early 1990s evolve into Space Station Freedom, and at the start of its construction in 1998 (exactly 20 years ago as of the day I’m writing this, actually) was known as the International Space Station. While many of the concepts from the MSS were carried over through its later iterations, this design was also quite different from the ISS that we know today.

TE Detail FlowBecause of this change in mission, quite a few of the subsystems for the power plant were changed extensively, starting just outside the reactor core and extending through to shielding, power conversion systems, and heat rejection systems. The power conversion system was changed to four parallel thermoelectric convertors, a more advanced setup than the SNAP-10 series of reactors used. These allowed for partial outages of the PCS without complete power loss. The heat rejection system was one of the most mission-dependent structures, so would vary in size and configuration quite a bit from mission to mission. It, too, would use NaK-78 as the working fluid, and in general would be 1200 (on the OWS) to 1400 (reference mission) sq. ft in surface area. We’ll look more at these concepts in later posts on power conversion and heat rejection systems, but these changes took up a great deal of the work that was done on the ZrHR program.

Radiation Shield ZonesOne of the biggest reasons for this unusual shielding configuration was to allow a compromise between shielding mass and crew radiation dose. In this configuration, there would be three zones of radiation exposure: only shielded by the 4 pi shield during rendezvous and docking (a relatively short time period) called the rendezvous zone; a more significant shielding for the spacecraft but still slightly higher than fully shielded (because the spacecraft would be empty when docked the vast majority of the time) called the scatter shield zone; and the crewed portion of the space station itself, which would be the most shielded, called the primary shielded zone. With the 4 pi shield, the entire system would mass 24,450 pounds, of which 16,500 lbs was radiation shielding, leading to a crew dose of between 20 and 30 rem a year from the reactor.

The mission planning for the OWS was flexible in its launch configuration: it could have launched integral to the OWS on a Saturn V (although, considering the troubles that the Skylab launch actually had, I’m curious how well the system would have performed), or it could have been launched on a separate launcher and had an upper stage to attach it to the OWS. The two options proposed were either a Saturn 1B with a modified Apollo Service Module as a trans-stage, or a Titan IIIF with the Titan Trans-stage for on-orbit delivery (the Titan IIIC was considered unworkable due to mass restrictions).

Deorbit System ConceptAfter the 3-5 years of operational life, the reactor could be disposed of in two ways: either it would be deorbited into a deep ocean area (as with the SNAP-10A, although as we saw during the BES-5’s operational history this ended up not being considered a good option), or it could be boosted into a graveyard orbit. One consideration which is very different from the SNAP-10A is that the reactor would likely be intact due to the 4 pi shield, rather than burning up as the SNAP-10A would have, meaning that a terrestrial impact could lead to civilian population exposures to fission products, and also having highly enriched (although not quite bomb grade) uranium somewhere for someone to be able to relatively easily pick up. This made the deorbiting of the reactor a bit pickier in terms of location, and so an uncontrolled re-entry was not considered. The ideal was to leave it in a parking orbit of at least 400 nautical miles in altitude for a few hundred years to allow the fission products to completely decay away before de-orbiting the reactor over the ocean.

Nuclear Power for the Moon

Lunar configuration cutaway
Lunar landing configuration, image DOE

The final configuration that was examined for the Advanced ZrH Reactor was for the lunar base that was planned as a follow-on to the Apollo Program. While this never came to fruition, it was still studied carefully. Nuclear power on the Moon was nothing new: the SNAP-27 radioisotope thermoelectric generator had been used on every single Apollo surface mission as part of the Apollo Lunar Surface Experiment Package (ALSEP). However, these RTGs would not provide nearly enough power for a permanently crewed lunar base. As an additional complication, all of the power sources available would be severely taxed by the 24 day long, incredibly cold lunar night that the base would have to contend with. Only nuclear fission offered both the power and the heat needed for a permanently staffed lunar base, and the reactor that was considered the best option was the Advanced ZrH Reactor.

The configuration of this form of the reactor was very different. There are three options for a surface power plant: the reactor is offloaded from the lander and buried in the lunar regolith for shielding (which is how the Kilopower reactor is being planned for surface operations); an integral lander and power plant which is assembled in Earth (or lunar) orbit before landing, with a 4 pi shield configuration; finally an integrated lander and reactor with a deployable radiator which is activated once the reactor is on the surface of the moon, again with a 4 pi shield configuration. There are, of course, in-between options between the last two configurations, where part of the radiator is fixed and part deploys. The designers of the ZrHR decided to go with the second option as their best design option, due to the ability to check out the reactor before deployment to the lunar surface but also minimizing the amount of effort needed by the astronauts to prepare the reactor for power operations after landing. This makes sense because, while on-orbit assembly and checkout is a complex and difficult process, it’s still cheaper in terms of manpower to do this work in Earth orbit rather than a lunar EVA due to the value of every minute on the lunar surface. If additional heat rejection was required, a deployable radiator could be used, but this would require flexible joints for the NaK coolant, which would pose a significant materials and design challenge. A heat shield was used when the reactor wasn’t in operation to prevent exessive heat loss from the reactor. This eased startup transient issues, as well as ensuring that the NaK coolant remained liquid even during reactor shutdown (frozen working fluids are never good for a mechanical system, after all). The power conversion system was exactly the same configuration as would be used in the OWS configuration that we discussed earlier, with the upgraded, larger tubes rather than the smaller, more numerous ones (we’ll discuss the tradeoffs here in the power conversion system blog posts).

Surface configuration diagram

This power plant would end up providing a total of 35.5 kWe of conditioned (i.e. usable, reliable power) electricity out of the 962 kWt reactor core, with 22.9 kWe being delivered to the habitat itself, for at least 5 years. The overall power supply system, including radiator, shield, power conditioning unit, and the rest of the ancillary bits and pieces that make a nuclear reactor core into a fission power plant, ended up massing a total of 23,100 lbs, which is comfortably under the 29,475 lb weight limit of the lander design that was selected (unfortunately, finding information on that design is proving difficult). A total dose rate at a half mile for an unshielded astronaut would be 7.55 mrem/hr was considered sufficient for crew radiation safety (this is a small radiation dose compared to the lunar radiation environment, and the astronauts will spend much of their time in the much better shielded habitat.

Sadly, this power supply was not developed to a great extent (although I was unable to find the source document for this particular design: NAA-SR-12374, “Reactor Power Plants for Lunar Base Applications, Atomics International 1967), because the plans for the crewed lunar base were canceled before much work was done on this design. The plans were developed to the point that future lunar base plans would have a significant starting off point, but again the design was never frozen, so there was a lot of flexibility remaining in the design.

The End of the Line

Sadly, these plans never reached fruition. The U-ZrH Reactor had its budget cut by 75% in 1971, with cuts to alternate power conversion systems such as the use of thermionic power conversion (30%) and reactor safety (50%), and the advanced Brayton system (completely canceled) happening at the same time. NERVA, which we covered in a number of earlier posts, also had its funding slashed at the same time. This was due to a reorientation of funds away from many current programs, and instead focusing on the Space Shuttle and a modular space station, whose power requirements were higher than the U-ZrH Reactor would be able to offer.

At this point, the AEC shifted their funding philosophy, moving away from preparing specific designs for flight readiness and instead moving toward a long-term development strategy. In 1973 head of the AEC’s Space Nuclear Systems Division said that, given the lower funding levels that NASA was forced to work within, “…the missions which were likely to require large amounts of energy, now appear to be postponed until around 1990 or later.” This led to the cancellation of all nuclear reactor systems, and a shift in focus to radioisotope thermoelectric generators, which gave enough power for NASA and the DoD’s current mission priorities in a far simpler form.

Funding would continue at a low level all the way to the current day for space fission power systems, but the shift in focus led to a very different program. While new reactor concepts continue to be regularly put forth, both by Department of Energy laboratories and NASA, for decades the focus was more on enhancing the technological capability of many areas, especially materials, which could be used by a wide range of reactor systems. This meant that specific systems wouldn’t be developed to the same level of technological readiness in the US for over 30 years, and in fact it wouldn’t be until 2018 that another fission power system of US design would undergo criticality testing (the KRUSTY test for Kilopower, in early 2018).

More Coming Soon!

Originally, I was hoping to cover another system in this blog post as well, but the design is so different compared to the ZrH fueled reactors that we’ve been discussing so far in this series that it warranted its own post. This reactor is the SNAP-50, which didn’t start out as a space reactor, but rather one of the most serious contenders for the indirect-cycle Aircraft Nuclear Propulsion program. It used uranium carbide/nitride fuel elements, liquid lithium coolant, and was far more powerful than anything that weve discussed yet in terms of electric power plants. Having it in its own post will also allow me to talk a little bit about the ANP program, something that I’ve wanted to cover for a while now, but considering how much more there is to discuss about in-space systems (and my personal aversion to nuclear reactors for atmospheric craft on Earth), hasn’t really been in the cards until now.

This series continues to expand, largely because there’s so much to cover that we haven’t gotten to yet – and no-one else has covered these systems much either! I’m currently planning on doing the SNAP-50/SPUR system as a standalone post, followed by the SP-100 and a selection of other reactor designs. After that, we’ll cover the ENISY reactor program in its own post, followed by the NEP designs from the 90s and early 00s, both in the US and Russia. Finally, we’ll cover the predecessors to Kilopower, and round out our look at fission power plant cores by revisiting Kilopower to have a look at what’s changed, and what’s stayed the same, over the last year since the KRUSTY test. We will then move on to shielding materials and design (probably two or three posts, because there’s a lot to cover there) before moving on to power conversion systems, another long series. We’ll finish up the nuclear systems side of nuclear power supplies by looking at heat sinks, radiators, and other heat rejection systems, followed by a look at nuclear electric spacecraft architecture and design considerations.

A lot of work continues in the background, especially in terms of website planning and design, research on a lot of the lesser-known reactor systems, and planning for the future of the web page. The blog is definitely set for topics for at least another year, probably more like two, just covering the basics and history of astronuclear design, but getting the web page to be more functional is a far more complex, and planning-heavy, task.

I hope you enjoyed this post, and much more is coming next year! Don’t forget to join us on Facebook, or follow me on Twitter!

References

SNAP 8 Summary Report, AEC/Atomics International Staff, 1973 https://www.osti.gov/servlets/purl/4393793

SNAP-8, the First Electric Propulsion Power System, Wood et al 1961 https://www.osti.gov/servlets/purl/4837472

SNAP 8 Reactor Preliminary Design Summary, ed. Rosenberg, 1961 https://www.osti.gov/servlets/purl/4476265

SNAP 8 Reactor and Shield, Johnson and Goetz 1963 https://www.osti.gov/servlets/purl/4875647

SNAP 8 Reactors for Unmanned and Manned Applications, Mason 1965 https://www.osti.gov/servlets/purl/4476766

SNAP 8 Reactor Critical Experiment, ed. Crouter. 1964 https://www.osti.gov/servlets/purl/4471079

Disassembly and Postoperation Examination of the SNAP 8 Experimental Reactor, Dyer 1967 https://www.osti.gov/servlets/purl/4275472

SNAP 8 Experimental Reactor Fuel Element Behavior, Pearlman et al 1966 https://www.osti.gov/servlets/purl/4196260

Summary of SNAP 8 Development Reactor Operations, Felten and May 1973 https://www.osti.gov/servlets/purl/4456300

Structural Analysis of the SNAP 8 Development Reactor Fuel Cladding, Dalcher 1969 https://www.osti.gov/servlets/purl/6315913

Reference Zirconium Hydride Thermoelectric System, AI Staff 1969 https://www.osti.gov/servlets/purl/4004948

Reactor-Thermoelectric System for NASA Space Station, Gylfe and Johnson 1969 https://www.osti.gov/servlets/purl/4773689