Another Nuclear Power Option That Might Make More Sense

By Merwin Brown, Guest Contributor

In the article, “A Nuclear Power Option That Makes Sense,” in the April 9 Smart Energy Portal newsletter, Pete Baston highlighted some of the merits of the Molten Salt Breeder Reactor, more specifically the liquid fluoride thorium reactor (LFTR). Among its many merits, it is especially well suited for breeding uranium-233, a very good thermal nuclear power reactor fuel, from a feedstock of thorium. In general he got much right. There are some alternative nuclear power reactor designs which offer benefits beyond those of the ubiquitous light water reactor (LWR) nuclear power plants. Consider another “nuclear power option that makes sense,” the high temperature gas-cooled reactor (HTGR), which might make even more sense than either the LWR or the LFTR.

The thorium-uranium-233 (Th-U) fuel cycle feature is one where the LFTR probably outshines the other reactor concepts, largely because of the LFTR’s convenient fuel reprocessing configuration. The major reasons one might opt for this fuel cycle are that it expands the nuclear fuel options beyond uranium, doesn’t require a complex enrichment process, and would not produce plutonium, which I believe is the radioactive waste to which Baston referred in his article. However, given the major technical and economic barriers to overcome in switching from today’s fuel cycle based on uranium feedstock, the benefits of the Th-U fuel cycle appear to be many years into the future. I believe there are more compelling near term advantages of the HTGR that warrant a more urgent consideration.

As the name implies, the HTGR uses a gaseous coolant, helium, and can produce high outlet coolant temperatures, up to at least 950 C in advanced designs. These features, plus some other attendant structural aspects that go along with high temperature capabilities, provide a variety of performance, versatility and inherent safety advantages.

The ability to operate at very high temperatures enables highly efficient electrical generation. The structural temperature limits of LWRs keep their maximum outlet coolant temperatures considerably below the maximum efficiency capabilities of even conventional fossil fuel steam cycles. In the HTGR, the reactor is “thermally loafing” while the steam cycle is ideally running at higher temperatures and pressures for maximum efficiency. If one wants to really put the HTGR reactor to work, there is room thermally for a closed-loop helium Brayton gas turbine topping cycle, yielding thermodynamic efficiencies above 50%. These high temperatures are also ideal for process heat applications such as making steel and hydrogen fuels, thereby expanding the use of nuclear as a low greenhouse gas emissions fuel.

To achieve these high temperature operations, the HTGR designers chose helium gas as its coolant, graphite as a main structural element and as its neutron moderator (needed in all thermal nuclear power plants), and a special fuel “element” design in which the fission process takes place and the radioactive fission products are trapped. The figure below, taken from brochure 11-50243-01, of the U.S. Department of Energy’s Idaho National Laboratory, shows the HTGR’s coated particle fuel pellet and two fuel “element” configurations using graphite.

The silicon carbide coated fuel particle used in the HTGR (which by the way has been designed to use the “conventional” U-235 fuel similar to LWRs, but has used U-233 from thorium in some instances) can withstand very high temperatures (>3000 F) without melting or failure. Functionally, this multi-layered fuel design effectively replaces the steel pressure vessel used in liquid-cooled reactors (existing and proposed LWRs and molten salt reactors) with billions of much stronger smaller vessels, i.e., the fuel particles. If one or more of the layers were to fail and degrade the ability of the fuel to retain radionuclides, the fuel particles would fail statistically and in small numbers with minimal releases. Furthermore, under normal operation, these fuel particles release negligible amounts of radioactive fission products into the helium coolant.

This fuel design has implications for working in and around the nuclear plant. I’ve toured both LWR and HTGR reactors. Even with the LWR not operating, to go inside the reactor secondary containment dome, i.e., the reactor building, usually requires suiting up in protective gear and then only staying inside for limited amounts of time to avoid too much exposure to radiation. In the HTGR, I have conducted tours in street clothes everywhere in the plant even while the reactor was operating, because there is essentially no radioactive material leaked into the primary coolant and secondary steam power loops.

Helium, while not perhaps as effective at heat removal as water or molten salt, is good enough at cooling, and brings some extra benefits. It won’t freeze at low temperatures, change phase at high temperatures and cause pressure excursions, nor cause dramatic sharp changes in heat transfer rates, which can stress reactor components. Furthermore, because it is essentially transparent to neutrons, any change in density of the helium doesn’t affect the control of nuclear reaction in the core, and it doesn’t become radioactive from exposure to neutrons produced in the nuclear fission process. Also, because helium is chemically inert, it isn’t a corrosive agent to the power plant structure.

The use of a graphite core structure takes advantage of a number of this material’s properties. Graphite is a very good neutron moderator. It also can withstand very high temperatures and not melt. And because a large mass of graphite is used, it heats up quite slowly should there be a problem with heat removal from the reactor core. The relatively long time it takes for a HTGR core to increase from normal temperatures to damaging levels, i.e., hours and days, has led to informally labeling the HTGR as the “walk away reactor,” meaning that there should be considerable time, if needed, for operators to think and plan a response. This feature, coupled with the fact the reactor can’t totally lose all of its gas coolant, and its core can be cooled by natural convection, or even radiant and conductive heat mechanisms, should the helium circulators that pressurize and pump the helium around the system fail, has also resulted in the label as an “inherently safe reactor.” While that label is perhaps a bit of hyperbole, the HTGR is richly endowed with inherent safety features.

While the HTGR is probably the most inherently safe fission reactor, it does have some failure mechanisms that must be managed. Those usually involve getting water in the reactor, for example, from a steam generator tube failure, causing hydrogen-water shift reactions (but so do the other reactors at high enough temperatures), or the potential for graphite oxidation if considerable amounts of air enter the reactor, for example, from the reactor building. But even in these rare events, the consequences are bounded to a level that offsite emergency sheltering is not required and the operator has considerable time to take remedial action.

Because of the above attributes, I also have heard the HTGR referred to in the past as the “utility-preferred” reactor. So why isn’t the HTGR being used as the preferred reactor in the world? Largely it is because the HTGR’s low power density, one of its safety advantages, made it too large to fit on ships and submarines. So the navy, in the mid-20th century, chose to develop the LWR technology, and the utility industry later leveraged that development and adapted it for use on land. Once the LWR industry was established, the utility industry was reluctant to change to a new design, and over time has relied on incremental improvements in the LWR.

HTGR reactors have been built and operated in 5 countries. This sets the HTGR apart from other new concepts, and today there are some serious proponents for building future HTGRs. Interest is being revived around the world with active programs in the US, Europe, China, and Korea. GA, AREVA, and Pebble Bed Modular Reactor/Westinghouse (PBMR/W) have been active in the DOE Next Generation Nuclear Project (NGNP). AREVA was recently selected as the vendor for the US NGNP program by the NGNP Alliance, a consortium of industrial end users, utilities, and vendors. Given the high efficiency performance, versatility in application, e.g., process heat, and inherent safety features of the HTGR, it is “a nuclear power option that makes sense.


Merwin Brown, Electric Grid Program Director for the California Institute for Energy and Environment (CIEE), manages a team helping develop and commercialize technologies for the modern electric grid for California’s aggressive energy-policy goals. The team develops, administers, and conducts R&D programs for reliable, safe, affordable, and environmentally sound transmission and distribution systems. California’s Public Interest Energy Research Transmission Program at the California Energy Commission largely funds this work.

Brown’s knowledge of electric utilities and new technologies results from 40 years of experience with firms such as Pacific Gas and Electric Company, Arizona Public Service, the Pacific Northwest National Laboratory, and the National Renewable Energy Laboratory. He has managed R&D programs of $50 million per year with groups as large as 100 scientists and engineers, and R&D projects as large as $20 million.

Brown has experience in strategic business planning and provides leadership among stakeholders, researchers, and funding agencies. He has served in advisory positions for many electricity industry organizations, including as an Arizona Solar Energy Commissioner, on the Board of the American Council for an Energy Efficiency Economy, and as advisor to the Electric Power Research Institute.

Brown has numerous technical publications and presentations, and holds B.S. and Ph.D. degrees in nuclear engineering from Kansas State University. Contact him at