Development and Testing Low Enriched Uranium Nuclear Thermal Systems Test Stands

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

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

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

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

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

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

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

Alumina sleeve during test, Bradley

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

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


TRIGA reactor core, image courtesy Wikimedia

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



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

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


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

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

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


High Flux Isotope Reactor (HFIR), Oak Ridge National Lab, image courtesy Wikimedia

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






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

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


Advanced Test Reactor, Idaho NL. Image courtesy DOE

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

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

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

Fuel Properties and Parameters to Test
Image courtesy Oak Ridge NL

NTR Hot Fire Testing For Today’s Regulatory Environment

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

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

The Way We Did It Before: Nuclear Furnace Exhaust Scrubbers

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

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

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

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

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

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

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

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

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

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

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

Above Ground Test Option #1: Exhaust Scrubbing

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

SNTP Test Facility
Image courtesy DOE

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

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

Saddle Mountain facility diagram
Alternative test facility layout

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

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

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

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

The Borehole Option: Subsurface Active Filtration of Exhaust

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

SAFE schematic
SAFE concept, Howe 2012, image courtesy NASA

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

SAFE injector model
SAFE injection system model, Howe 2012

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

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

P-Tunnel setup

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

INL geology 2
INL geological composition, image courtesy DOE

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

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

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

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

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

Above Ground Test Option #2: Complete Capture

Flow Diagram Coote 2017
Image via Coote 2017, courtesy NASA

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

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

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

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

NTP Testing at Stennis Space Center

SSC overview
Stennis SC test facilities, image courtesy NASA

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

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

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

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

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

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

Notional schedule
Notional Development Timeline

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

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

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

Regulatory Burdens, Bureaucratic Concerns, and Other Matters

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

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

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

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

Upcoming Posts!

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

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

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



In Pile Testing

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

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

10 CFR Part 830, Nuclear Safety Management;

High Flux Isotope Reactor homepage:

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

Transient Reactor Test Facility homepage:

Al 6061 Matweb page:

300 Stainless Steel; Pennsylvania Stainless,

Grade 5 Titanium Matweb page:

SIGRATHERM, SGL (manufacturer) website:

Nuclear Furnace ECS

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

DOE Fact Sheet, Appendix 2

Above Ground Effluent Treatment System

Space Nuclear Thermal Propulsion Final Report, R.A. Haslett, Grumman Aerospace Corp, 1995

Space Nuclear Thrmal Propulsion Test Facilities Subpanel Final Report, Allen et al, 1993

Subsurface Active Filtration of Exhaust (SAFE)

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

Subscale Validation of the Subsurface Active Filtration of Exhaust Approach to NTP Ground Testing, Marshall et al, NASA Glenn RC, 2015 (Conference Paper) and (Presentation Slides)

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

Complete Exhaust Capture and NASA Planning

Stennis Space Center Activities and Plans Overview presentation, NASA

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

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

NASA FY18 Budget Estimates:

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

Fission Power Systems Low Enriched Uranium Nuclear Thermal Systems Spacecraft Concepts

LEU NTP Part Three: Spacecraft Overview

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

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

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

Advances in LEU Nuclear Propulsion

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

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

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

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

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

The Beginnings of the Modern Astronuclear Thermal Era

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

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

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

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

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

NERVAPewee2, AEC 1971
Pewee on test stand, 1971, courtesy DOE
SNRE Diagram, Borowski 2010
Small Nuclear Rocket Engine, Borowski NASA GRC

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


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

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

Radiation Shielding

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

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

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

What is Radiation, and How is it Shielded?

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

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

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

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

Neutron Shielding

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

Gamma and X-Ray Shielding

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

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

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

NTP-Specific Shielding

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

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

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

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

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

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

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

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

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

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

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

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

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

The Propellant Tanks

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

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

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

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

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


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

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


Nuclear + Rockets: Always Complicated

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

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

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

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

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

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

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

Enter Active Cooling: The Zero Boil-Off Tank

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

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

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

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

CRYOTE System Cutaway, image via NASA

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


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

Test Article at MSFC, Image Courtesy NASA

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

LH2 Shielding Goes Away Through the Mission

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

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

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

LEU NTP Mars Vehicle

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

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

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

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

Core Propulsion Stage

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

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

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

In-Line Propellant Tank

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

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

Saddle Truss with Drop Tanks

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

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

Mission Payload

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

The Integrated Stack

MTV Copernicus (NCPS Config), NASA

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

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

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

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

Stack Construction

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

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

SLS Sensitivity Chart

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

LEU NTP: The Latest Plan to Get to Mars

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

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

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

Additional Reading

Nuclear Thermal Propulsion

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

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

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

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

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

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

Radiation and Shielding

Types of Radiation video, FermiLab

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

In-Space Radiation Environment and Crew Quarters Shielding

Neutron Astronomy; Casadei, University of Birmingham, 2017

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

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

Homepage, Rutherford Appleton Laboratory Mini Magnetosphere Project

In-Space Reactor Shielding

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

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

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

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

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

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

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

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

Auxiliary Support Systems for NTR


Propellant Tanks and Zero Boil-Off

Passive Thermal Protection

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

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

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

Zero Boil-Off

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

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

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

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

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

Nuclear Cryogenic Propulsion Stage

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

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

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

Launch Vehicles

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

Ares V Wikipedia Page

Ares V Fact Sheet, NASA 2006

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

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

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

NASA Space Launch System Main Page

NASA SLS Overview Page

NASA’s Mars Design Reference Mission 5.0 and Associated Considerations

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

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

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

Blog Updates

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

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

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

Development and Testing Low Enriched Uranium Nuclear Thermal Systems

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

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

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

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

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

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

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

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

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

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

CERMET Fuel Elements

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

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

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

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

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

CERMET Composition and Manufacture

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

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

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

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

Angular UO2 Microparticles
Angular UO2 microparticles, image courtesy NASA

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

W-ZrO2 CVD Coated Particles, image courtesy NASA

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

HIP process
Image courtesy NASA


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

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

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

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

The Beginnings of CERMET Fuels

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

ANL 200 MW Reactor
Image courtesy DOE

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

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

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

Non-Spherical Microparticles
UO2 particles

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

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

Can component fit check pic

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

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

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

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

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

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

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

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


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

Houts NTREES Facility 2013

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

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

Chamber installation
Pressure Chamber during upgrade, image courtesy NASA MSFC

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

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

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

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

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

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

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


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


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

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


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

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

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

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

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

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

More info on CFEET is available on the CFEET page!

What Have They Taught Us?

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

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

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

Development Challenges for LEU NTP, and a New Direction

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

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

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

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

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

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

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

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

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

How This Changes the Core

BWXT Core, image via BWXT

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

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

More to Come!

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

Sources and Additional Reading

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

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

CERMET Fueled Reactors, Cowan et al 1987

A CERMET Fueled Reactor for Nuclear Propulsion, Kruger 1991

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

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

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

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

Low Enriched Nuclear Thermal Propulsion Systems, Houts et al 2017

NTP CERMET Fuel Development Status, Barnes 2017

2017 Game Changing Development program Mid-year Review Slides

Channel update:

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

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


Cpoyright 2018 Beyond NERVA. Contact for reprint permission.