Fission Power Plant Reactor Cores

The geometry of the reactor core is often quite different from that seen in terrestrial reactor designs. The biggest difference is that it will be far smaller; often only a couple meters long and wide, or smaller; as opposed to the dozens or hundreds of meters that large terrestrial fission plants have. This is, of course, to save mass of the spacecraft, and to get more power out of less mass and volume. We’ll examine the different options for core geometry more as we go through this blog series, but the shape ranges from plugs of uranium oxide or carbide arranged in a row, surrounded by a power conversion system that we’ll look at later in this post; to cylindrical, square, or hexagonal fuel elements with coolant channels running through them (as we’ve seen with the Kilopower reactor, although that’s unique in that there’s just one fuel element, not several); to more exotic options, which all depend on a number of factors.

One big difference between these reactor cores and the ones that we’ve been examining in the nuclear thermal rocket posts is that they don’t run nearly as hot. While there are efficiency benefits to having a hotter reactor, as we’ve seen the thermal stresses, increased chemical reaction rates, and limitations on the materials that can be used, often mean that this is simply more trouble than its worth in most cases. An NTR has to be as hot as possible to maximize the specific impulse, or rocket efficiency, of the engine; whereas to produce electricity there are many options that work well enough at cooler temperatures, so engineers generally decide to run at cooler temperatures and save themselves a lot of the concerns and headaches that these high temperatures cause.

How hot is hot enough? Well, that depends on a number of factors, mostly to do with the power conversion system and the heat rejection system. The fact that these systems are linked together (in much the same way as the turbopumps, propellant being used, and other components of an NTR work together) is by now a familiar concept to regulars of this blog. These considerations, no matter what the details are, are called “balance of plant” issues; and if anything they’re the biggest concern for a reactor designer just beginning to design an in-space fission power system. Many concepts, like amount of power that needs to be provided, mass and volume of individual components, maximum and minimum working fluid outlet temperature, radiator requirements, and many other considerations, all work together to define a system; but perhaps the most important considerations are the first and third: power requirements and power conversion system. The power requirements, as well as mass requirements, will be defined by the mission that will use this reactor, so they can be taken as a given: if NASA wants a reactor to provide 100 megawatts of electricity (MWe) for a manned spacecraft using nuclear electric propulsion… well it’s the job of the reactor designer to do that, ideally within the mass budget allotted (or else other systems on the spacecraft have to get lighter, something that’s difficult or impossible to do).

Lets look at some examples of proposed, tested, or flown reactor cores:

Russian Astronuclear Reactor Cores

Romaskha: The First Soviet Astronuclear Reactor

For more general information on Romashka, check out the Romashka page!

Romashka 2A series of disc-shaped uranium carbide (UC2) fuel elements were used in this reactor (90% 235U), with holes drilled through the center, and roughly halfway from the central hole of the disc to the edge of the fuel element. Both of these holes were used to thread the thermionic power conversion system through the core of the reactor. Spacing of the fuel elements was provided by a mixture of beryllium oxide and graphite, which was also used to slightly moderate the neutron spectrum – but the neutron spectrum in the reactor remained in the fast spectrum. Surrounding the reactor core itself, both radially and at the ends of the core, were beryllium reflectors. Boron and boron nitride control rods placed in the radial reflector and base axial reflector were used to maintain reactor control through the use of a hydraulic system, however a large negative thermal reactivity coefficient in the reactor core was also meant to largely control the reactor in the case of normal operations. Finally, the reactor was surrounded by a finned steel casing that provided all heat rejection through passive radiation – no pumps required! The nominal operating temperature of the reactor was meant to be between 1200 C and 1800 C at the center of the core, and about 800 C at the edges of the core at the ends of the cylinder.

BES-5 Buk, The Most Flown Reactor in History

The Bouk (“Beech”) reactor, also known as the “Buk,” or BES-5 reactor, is arguably the most successful astronuclear design in history. Begun in 1960 by the Krasnya Zvesda Scientific and Propulsion Association, this reactor promised greater power output than the Romashka, at the cost of additional complexity, and requiring coolant to operate. From 1963 to 1969, testing of the fuel elements and reactor core was carried out without using the thermoelectric fuel elements (TFE), which were still under development. From 1968 to 1970, three reactor cores with full TFEs were tested at Baikal; and, with successful testing completed, the reactor design was prepared for launch, integrated into the Upravlenniye Sputnik Aktivny (US-A; in the West, RORSAT, for Radar Ocean Reconnaisance SATellite) spacecraft, designed to use radar for naval surveillance.

DOE Sketch based on KOSMOS 954

LLNL sketch of BES-5 based on KOSMOS-954 wreckage, DOE image via Sven Grahn

Rather than having stacked discs of UC2, the BES-5 used 79 fuel rods made out of uranium (90% enriched, total U mass 30 kg) molybdenum alloy metal, encased in high-temperature steel. NaK was used as a coolant for the reactor, pumped using the energy from 19 of the fuel assemblies to run an electromagnetic fuel pump. Producing over 100 kW of thermal energy, after electric conversion using in-core germanium-silicon thermoelectric power conversion elements (which use the difference in charge potential between two different metals along a boundary to create an electrical charge when a temperature gradient is applied across the join; again, more in a later post), a maximum of 5 kW of electrical energy was available for the spacecraft’s instrumentation. The fact that this core used thermoelectric conversion rather than thermionic is a good indicator that the common use of the term, TOPAZ, for this reactor is incorrect. Reactor control was provided by six beryllium reflector drums that would be slowly lowered through holes in the radial reflector over the reactor’s life to increase the local neutron flux to account for the buildup of neutron poisons.

TEU-5 TOPOL

For more information on the TEU-5, Check out our TOPAZ-I page!

The TOPOL program was begun in the 1960s, under the Russian acronym for Thermionic Experimental Converter in the Active Zone, which translates directly into Topaz in English, but ground testing didn’t begin until 1970. This was a multi-cell thermionic fuel element design similar in basic concept to Romashka, however it was a far more complex design. Instead of a single stack of disc-shaped fuel elements, a “garland” of fuel elements were formed into a thermionic fuel element. The fissile fuel element was surrounded by a thimble of tungsten or molybdenum, which formed the cathode of the thermionic converter, while the anode of the converter was a thin niobium tube; as with most thermionic converters the gap between cathode and anode was filled with cesium vapor. The anode was cooled with pumped NaK, although some sources indicate that lithium was also considered as a coolant for higher-powered versions of the reactor.

BES-5 core cross section

TEU-5 core cross-section, DOE

The differences between the BES-5 and TEU-5 were far more than the power conversion system. Instead of being a fast reactor, the Topaz was designed for the thermal neutron spectrum, and as such used zirconium hydride for in-core moderation (also creating a thermal limitation for the materials in the core; however, hydrogen loss mitigation measures were taken throughout the development process). Rather than using the metal fuels that its predecessor had, or the carbides of the Romashka, the Topol used a far more familiar material to nuclear power plant operators: uranium oxide (UO2), enriched to 90% 235U. This, along with reactor core geometry changes, allowed the amount of uranium needed for the core to drop from 30 kg in the BES-5 to 11.5 kg. NaK remained the coolant, due to its low melting temperature, good thermal conductivity, and neutronic transparency. The cathode temperature in the TEU-5 was in the range of 1500-1800C, which resulted in an electrical power output of up to 10 kW.

ENISY/TOPAZ-II

Coming soon!

TEM Power Unit

Coming soon!

American Astronuclear Reactor Cores

Systems for Nuclear Auxiliary Power

SNAP was the main focus of a number of organizations in the US looking to use nuclear power in space and in remote regions on Earth where maintenance of a fission power system wasn’t feasible.

The program extended everywhere from the bottom of the seas (SNAP-4, which we won’t be covering in this post) to deep space travel with electric propulsion. SNAP was divided up into an odd/even numbering scheme, with the odd model numbers (starting with the SNAP-3) being radioisotope thermoelectric generators, and the even numbers (beginning with SNAP-2) being fission reactor electrical power systems.

Due to the sheer scope of the SNAP program, even eliminating systems that aren’t fission-based, this is going to be a two post subject. This post will cover the US Air Force’s portion of the SNAP reactor program: the SNAP-2 and SNAP-10A reactors; their development programs; the SNAPSHOT mission; and a look at the missions that these reactors were designed to support, including satellites, space stations, and other crewed and uncrewed installations. The next post will cover the NASA side of things: SNAP-8 and its successor designs as well as SNAP-50/SPUR. The one after that will cover the SP-100, SABRE, and other designs from the late 1970s through to the early 1990s, and will conclude with looking at a system that we mentioned briefly in the last post: the ENISY/TOPAZ II reactor, the only astronuclear design to be flight qualified by the space agencies and nuclear regulatory bodies of two different nations.

SNAP capabilities 1964

For a detailed overview of the SNAP program, check out our blog posts on the program available here: Part 1 and Part 2 (coming soon!)

SNAP Experimental Reactor Program

SNAP Critical Assembly

S2 Critical Assembly

SNAP Critical Assembly core, image DOE

SNAP-2’s first critical assembly test was in October of 1957, shortly after Sputnik-1’s successful launch. With 93% enriched 235U making up 8% of the weight of the U-ZrH fuel elements, a 1” beryllium inner reflector, and an outer graphite reflector (which could be varied in thickness), separated into two rough hemispheres to control the construction of a critical assembly; this device was able to test many of the reactivity conditions needed for materials testing on a small economic scale, as well as test the behavior of the fuel itself. The primary concerns with testing on this machine were reactivity, activation, and intrinsic steady state behavior of the fuel that would be used for SNAP-2. A number of materials were also tested for reflection and neutron absorbency, both for main core components as well as out-of-core mechanisms. This was followed by the SNAP-2 Experimental Reactor in 1959-1960 and the SNAP 2 Development Reactor in 1961-1962.

SNAP Experimental Reactor

S2ER Cross Section

SNAP-2 Experimental Reactor core cros section diagram, image DOE

The SNAP-2 Experimental Reactor (S2ER or SER) was built to verify the core geometry and basic reactivity controls of the SNAP-2 reactor design, as well as to test the basics of the primary cooling system, materials, and other basic design questions, but was not meant to be a good representation of the eventual flight system. Construction started in June 1958, with construction completed by March 1959. Dry (Sept 15) and wet (Oct 20) critical testing was completed the same year, and power operations started on Nov 5, 1959. Four days later, the reactor reached design power and temperature operations, and by April 23 of 1960, 1000 hours of continuous testing at design conditions were completed. Following transient and other testing, the reactor was shut down for the last time on November 19, 1960, just over one year after it had first achieved full power operations. Between May 19 and June 15, 1961, the reactor was disassembled and decommissioned. Testing on various reactor materials, especially the fuel elements, was conducted, and these test results refined the design for the Development Reactor.

S2ER Schedule and Timeline

SNAP-2 Development Reactor

S2DR Core Xsec

SNAP 2 Development Reactor core cross section, image DOE

The SNAP-2 Development Reactor (S2DR or SDR, also called the SNAP-2 Development System, S2DS) was installed in a new facility at the Atomics International Santa Susana research facility to better manage the increased testing requirements for the more advanced reactor design. While this wasn’t going to be a flight-type system, it was designed to inform the flight system on many of the details that the S2ER wasn’t able to. This, interestingly, is much harder to find information on than the S2ER. This reactor incorporated many changes from the S2ER, and went through several iterations to tweak the design for a flight reactor. Zero power testing occurred over the summer of 1961, and testing at power began shortly after (although at SNAP-10 power and temperature levels. Testing continued until December of 1962, and further refined the SNAP-2 and -10A reactors.

S2DR Development Timeline

SNAP Critical Assembly 4 (SCA-4)

A third set of critical assembly reactors, known as the SNAP Development Assembly series, was constructed at about the same time, meant to provide fuel element testing, criticality benchmarks, reflector and control system worth, and other core dynamic behaviors. These were also built at the Santa Susana facility, and would provide key test capabilities throughout the SNAP program. This water-and-beryllium reflected core assembly allowed for a wide range of testing environments, and would continue to serve the SNAP program through to its cancellation. Going through three iterations, the designs were used more to test fuel element characteristics than the core geometries of individual core concepts. This informed all three major SNAP designs in fuel element material and, to a lesser extent, heat transfer (the SNAP-8 used thinner fuel elements) design.

SNAP-2/10A and -10B

The original request for the SNAP program, which ended up becoming known as SNAP 2, occurred in 1955, from the AEC’s Defense Reactor Development Division and the USAF Wright Air Development Center. It was for possible power sources in the 1 to 10 kWe range that would be able to autonomously operate for one year, and the original proposal was for a zirconium hydride moderated sodium-potassium (NaK) metal cooled reactor with a boiling mercury Rankine power conversion system (similar to a steam turbine in operational principles, but we’ll look at the power conversion systems more in a later post), which is now known as SNAP-2. The design was refined into a 55 kWt, 5 kWe reactor operating at about 650°C outlet temperature, massing about 100 kg unshielded, and was tested for over 10,000 hours. This epithermal neutron spectrum would remain popular throughout much of the US in-space reactor program, both for electrical power and thermal propulsion designs. This design would later be adapted to the SNAP-10A reactor, with some modifications, as well, creating the SNAP-2/10A program.

SNAP-10B

The first -10B reactor, the -10B Basic (B-10B), was a very simple and direct evolution of the U-10A/2, with nothing but the reflector and control structures changed to the -10B configuration. Other than a slight drop in power density (to 0.30 kWt/lb), the rest of the performance characteristics of the B-10B were identical to the U-10A/2. This design would have been a simple evolution of the -10A/2, with a slimmer profile to help with payload integration challenges.

10B Basic Table

Image DOE

The next iteration of the SNAP-10B, the Advanced -10B (A-10B), had options for significant changes to the reactor core and the fuel elements themselves. One thing to keep in mind about these reactors is that they were being designed above and beyond any specific mission needs; and, on top of that, a production schedule hadn’t been laid out for them. This means that many of the design characteristics of these reactors were never “frozen,” which is the point in the design process when the production team of engineers need to have a basic configuration that won’t change in order to proceed with the program, although obviously many minor changes (and possibly some major ones) would continue to be made up until the system was flight qualified.

 

Up until now, every SNAP-10 design used a 37 fuel element core, with the only difference in the design occurring in the Upgraded -10A/2 and Basic -10B reactors (which changed the hydrogen barrier ceramic enamel inside the fuel element clad). However, with the A-10B there were three core size options: the first kept the 37 fuel element core, a medium-sized 55-element core, and a large 85-element core. There were other questions about the final design, as well, looking at two other major core changes (as well as a lot of open minor questions). The first option was to add a “getter,” a sheath of hydrophilic (highly hydrogen-absorbing) metal to the clad outside the steel casing, but still within the active region of the core. While this isn’t as ideal as containing the hydrogen within the U-ZrH itself, the neutron moderation provided by the hydrogen would be lost at a far lower rate. The second option was to change the core geometry itself, as the temperature of the core changed, with devices called “Thermal Coefficient Augmenters” (TCA). There were two options that were suggested: first, there was a bellows system that was driven by NaK core temperature (using ruthenium vapor), which moves a portion of the radial reflector to change the core’s reactivity coefficient; second, the securing grids for the fuel elements themselves would expand as the NaK increased in temperature, and contract as the coolant dropped in temperature.

Between the options available, with core size, fuel element design, and variable core and fuel element configuration all up in the air, the Advanced SNAP-10B was a wide range of reactors, rather than just one. Many of the characteristics of the reactors remained identical, including the fissile fuel itself, the overall core size, maximum outlet temperature, and others. However, the number of fuel elements in the core alone resulted in a wide range of different power outputs; and, which core modification the designers ultimately decided upon (Getter vs TCA, I haven’t seen any indication that the two were combined) would change what the capabilities of the reactor core would actually be. However, both for simplicity’s sake, and due to the very limited documentation available on the SNAP-10B program, other than a general comparison table from the SNAP Systems Capability Study from 1966, we’ll focus on the 85 fuel element core of the two options: the Getter core and the TCA core.

A final note, which isn’t clear from these tables: each of these reactor cores was nominally optimized to a 100 kWt power output, the additional fuel elements reduced the power density required at any time from the core in order to maximize fuel lifetime. Even with the improved hydrogen barriers, and the variable core geometry, while these systems CAN offer higher power, it comes at the cost of a shorter – but still minimum one year – life on the reactor system. Because of this, all reported estimates assumed a 100 kWt power level unless otherwise stated.

Yt Getter FE

Fuel element with yttrium getter, image DOE

The idea of a hydrogen “getter” was not a new one at the time that it was proposed, but it was one that hadn’t been investigated thoroughly at that point (and is a very niche requirement in terrestrial nuclear engineering). The basic concept is to get the second-best option when it comes to hydrogen migration: if you can’t keep the hydrogen in your fuel element itself, then the next best option is keeping it in the active region of the core (where fission is occurring, and neutron moderation is the most directly useful for power production). While this isn’t as good as increasing the chance of neutron capture within the fuel element itself, it’s still far better than hydrogen either dissolving into your coolant, or worse yet, migrating outside your reactor and into space, where it’s completely useless in terms of reactor dynamics. Of course, there’s a trade-off: because of the interplay between the various aspects of reactor physics and design, it wasn’t practical to change the external geometry of the fuel elements themselves – which means that the only way to add a hydrogen “getter” was to displace the fissile fuel itself. There’s definitely an optimization question to be considered; after all, the overall reactivity of the reactor will have to be reduced because the fuel is worth more in terms of reactivity than the hydrogen that would be lost, but the hydrogen containment in the core at end of life means that the system itself would be more predictable and reliable. Especially for a static control system like the A-10B, this increase in behavioral predictability can be worth far more than the reactivity that the additional fuel would offer. Of the materials options that were tested for the “getter” system, yttrium metal was found to be the most effective at the reactor temperatures and radiation flux that would be present in the A-10B core. However, while improvements had been made in the fuel element design to the point that the “getter” program continued until the cancellation of the SNAP-2/10 core experiments, there were many uncertainties left as to whether the concept was worth employing in a flight system.

The second option was to vary the core geometry with temperature, the Thermal Coefficient Augmentation (TCA) variant of the A-10B. This would change the reactivity of the reactor mechanically, but not require active commands from any systems outside the core itself. There were two options investigated: a bellows arrangement, and a design for an expanding grid holding the fuel elements themselves.

A-10B Bellows Diagram

Ruthenium vapor bellows design for TCA, image DOE

The first variant used a bellows to move a portion of the reflector out as the temperature increased. This was done using a ruthenium reservoir within the core itself. As the NaK increased in temperature, the ruthenium would boil, pushing a bellows which would move some of the beryllium shims away from the reactor vessel, reducing the overall worth of the radial reflector. While this sounds simple in theory, gas diffusion from a number of different sources (from fission products migrating through the clad to offgassing of various components) meant that the gas in the bellows would not just be ruthenium vapor. While this could have been accounted for, a lot of study would have needed to have been done with a flight-type system to properly model the behavior.

A-10B Expandable BaseplateThe second option would change the distance between the fuel elements themselves, using base plate with concentric accordion folds for each ring of fuel elements called the “convoluted baseplate.” As the NaK heated beyond optimized design temperature, the base plates would expand radially, separating the fuel elements and reducing the reactivity in the core. This involved a different set of materials tradeoffs, with just getting the device constructed causing major headaches. The design used both 316 stainless steel and Hastelloy C in its construction, and was cold annealed. The alternative, hot annealing, resulted in random cracks, and while explosive manufacture was explored it wasn’t practical to map the shockwave propagation through such a complex structure to ensure reliable construction at the time.

A-10B Expandable Baseplate 2While this is certainly a concept that has caused me to think a lot about the concept of a variable reactor geometry of this nature, there are many problems with this approach (which could have possibly been solved, or proven insurmountable). Major lifetime concerns would include ductility and elasticity changes through the wide range of temperatures that the baseplate would be exposed to; work hardening of the metal, thermal stresses, and neutron bombardment considerations of the base plates would also be a major concern in this concept.

These design options were briefly tested, but most of these ended up not being developed fully. Because the reactor’s design was never frozen, many engineering challenges remained in every option that had been presented. Also, while I know that a report was written on the SNAP-10B reactor’s design (R. J . Gimera, “SNAP 1OB Reactor Conceptual Design,” NAA-SR-10422), I can’t find it… yet. This makes writing about the design difficult, to say the least.

SNAP-8

Coming soon!

Advanced ZrH Reactor

Coming soon!

SNAP-50/SPUR

Coming soon!

SP-100

Coming soon!

Fission Surface Power

Coming soon!

Prometheus

Coming soon!

Kilopower

LANL Kilopower screencap core highlight

Kilopower’s 1 kWe nuclear reactor (the one to be directly tested with KRUSTY) uses metal fuel cast as a single cylinder 11 cm in diameter, with a 4 cm central hole for the test stand and cutouts to accommodate the heat pipes along the periphery. This is made out of uranium-molybdenum alloy, 92% uranium by weight (enriched to 95% 235U), and 8% molybdenum (also by weight). Y12 has extensive experience dealing with this particular fuel form, so it’s well understood both from a reactor physics perspective and from a fabrication and manufacture perspective.

The geometry of the reactor means that there’s a strong negative temperature reactivity coefficient, meaning that this reactor is largely self-regulating and the control rod is only needed for startup, shutdown, or major power level changes. Concepts for increasing the power level of this reactor have heat pipe channels inside the U-Mo fuel as well, but we’ll look at that more later.

There are minor changes to the geometry of the core for the test, but mainly they are to accommodate the reflector being raised around the core rather than a control rod being used to start the fission reaction.

The Kilopower Family of Reactor Cores

Kilopower 4 kWtThis is just the first step for the program, however. The 1 kWe design is the smallest one that is considered practicable for a fission power system, but the basic design concept can be extended up to 40 kWe. While the basic reactor is able to produce up to about 4 kWe, any additional power increase starts to exceed the heat transfer limits of the heat pipes. In order to increase the amount of heat transferred from the core to the power conversion system, changes are needed to the heat pipe system.

12 kWt Kilopower coreThe first option is to move the heat pipes from the periphery of the core to the interior. This is harder than it looks at first glance. The different components of the heat pipe affect the reactivity of the reactor in a number of different ways, which are best assessed “in the wild” with the 1-4 kWt design before moving on to this change.

The advantage to this is that a greater surface area of the heat pipe is exposed to the fuel element generating the heat, so more heat can be transported using the same heat pipes, or even by using smaller pipes (in this case, there’s a reduction in major diameter from 3/8” to 1/2”, and an increase in number to 12). By moving the heat pipes from the periphery to the center of the fuel element, power can be boosted to 13 kWt using only two more kg of 235U, and only 24 kg more in reactor mass.

21 kWt KilopowerThe larger sizes continue this trend, increasing the number of heat pipes in the core to increase the amount of heat removal. Because the reactor has a strong negative thermal reactivity coefficient, it has a corresponding tendency to increase reactivity as heat is more completely removed, increasing the power output of the reactor. The configuration and size of the heat pipes is based on Monte Carlo and thermal conductivity modeling to ensure that the temperature gradient across the fuel is acceptable, even with heat pipe failure.

43 kWt KilopowerWith the testing of KRUSTY, enough information will be gained both in reactor engineering and in fuel element manufacturing to enable the internal heat pipes. Additional expansions of the test area at the Nevada site will allow for this expansion of reactor sizes to allow for ground testing of these larger reactors (to test KRUSTY, a site regulation had to be changed to account for the amount of reactivity being inserted into the reactor).

At this point, the fuel that is used in KRUSTY, the UMo metal fuel, can’t be used anymore. There are issues of critical density of fuel, power levelling across a monolithic fuel element, and other issues mean that metal fuel can’t be used for a larger reactor. Metal fuel is relatively rare in reactor designs on Earth, oxide fuels being more common. This is also an option for a heat pipe cooled reactor, and this is a very attractive small modular reactor in its own right.

Reactor Size Cross section diagram

Image via DOE/LANL

This is the Megapower concept, a concept being explored by the Department of Defense for forward operating bases, disaster relief, and other missions where the lack of a supply chain for electrical power is critical.

2MWe Megapower

Image via Poston, LLNL