Fort St. Vrain in Pictures: 2

by Will Davis

The Fort St. Vrain project was innovative in more than one respect, and while it did not blaze the trail in HTGR (High Temperature Gas-cooled Reactor) commercialization – a feat accomplished by the Peach Bottom Atomic Power Station – it did considerably advance the technology toward full commercial operation and duplication on a wide scale through some significant design changes.  One of the most interesting of these was the use of a prestressed concrete reactor vessel (PCRV) for the reactor itself and primary components instead of a steel vessel.

Fort St Vrain Model 1Above, a scale model of the Fort St. Vrain Nuclear Generating Station as actually constructed.  In this view, what we might call the primary or nuclear portion of the plant is at left, inside the tallest structure; the steam plant is in the structure to the right of it.  The PCRV can be seen in the lower section of the tall reactor building; on top of it is a refueling operations floor that we’ll see more of in a future installment.

Fort St Vrain reactorThe design of nuclear steam supply system (NSSS) selected for Fort St. Vrain is shown above.  The 841 MWt core is seen in the upper end of the system, with the steam generators and helium circulators located below.  Control rod drives entered the core from above and are not shown here.  This system was roughly 31 feet in diameter; the cavity in the PCRV was 75 feet tall.

Fort St Vrain reactor and primary parts

This view gives a perspective of the relative size of the PCRV and the primary system.  The 12 steam generator modules’ and 4 helium circulators’ arrangement can be better appreciated here, as can the skirt area below the PCRV.  Overall this PCRV was 106 feet tall and 61 feet in diameter.

Selection of the prestressed concrete PCRV concept was made as part of the process to compact the NSSS for future plants as compared to Peach Bottom.  Using a unitary concrete structure such as this allowed the entire NSSS to become essentially integral inside of it.  According to General Atomic, integrating the components in this manner “reduces the primary coolant inventory, eliminates external piping, removes the requirements for steam generator pressure shells, provides a single integral radiation shield, and results in minimum building size.”  Later on, when the AEC requested the reactor vendors to study nuclear plants in a fantastic 10,000 MWt range, both Westinghouse and GE informed the AEC that they would have to move to integral, concrete vessels of some sort to make nuclear steam supply systems of that scale practical — but it never happened, and neither did later design generations of gas cooled reactors designed by General Atomic.  Thus, Fort St. Vrain remained unique in the United States in the respect of its use of concrete to contain the primary system.

Fort St Vrain PCRV partsWith the nuclear components removed, we now see the key components of the PCRV itself.  The cavity for the nuclear steam supply system (NSSS) was lined with carbon steel, which was applied to form a gas-tight seal.  The thermal barrier and cooling tubes were necessary to keep the concrete cool; a manway (“access penetration”) was included through the bottom of the PCRV.

Fort St Vrain PCRV prestressingThe concrete of the PCRV structure had to be steel reinforced and prestressed in order to contain the pressure of the primary system gas, helium, which was pressurized to to roughly 700 psia although the maximum design pressure of this PCRV was comfortably higher at 845 psia.  The PCRV itself comprised about 6480 cubic yards of concrete and the 3/4 inch carbon steel liner, as well as 448 steel tendons.  The head and longitudinal tendons were loaded to 1.4 million pounds force, while the circumferential tendons were loaded to 1.25 million pounds force.

There was another significant advantage to incorporating a concrete vessel like this in the plant design – mobility, or that is to say, the ability to site a plant anywhere.  By the time this plant was designed, there were already problems with getting the very large and heavy components required for light water reactors to some prospective nuclear plant sites.  In fact, the size and weight of such components could and did prevent siting a plant of a given type in a given desired location in some instances.  It was for this reason that GE and Chicago Bridge & Iron developed the ability to field assemble large BWR (Boiling Water Reactor) pressure vessels on site at the end of the 1960’s.  This allowed utilities to site nuclear plants in locations that could not take shipments of either a large prefabricated BWR vessel, or else a slightly smaller but heavier PWR vessel and attendant steam generators.  With the PCRV concept, however, the PCRV itself was built right on the spot, there was practically no limit as to where a gas cooled reactor plant could be constructed because anything required to make it could be shipped anywhere it was possible to construct a large power plant in the first place.

As we now know, although the PCRV concept itself was eventually proved out by operation of the Fort St. Vrain Nuclear Generating Station when completed, no other PCRV’s were built in the United States.  It will probably be some time in the future when this highly sensible and practical idea is revived – quite possibly for gas cooled reactors – in the United States.

Prestressed Concrete Reactor Vessel one-fifth scale model being tested at General Atomic, near San Diego.

Prestressed Concrete Reactor Vessel one-fifth scale model being tested at General Atomic, near San Diego.  As can be seen here, it was intended that tendons (both on the model and the actual plant) be accessible for inspection and even for replacement.  Tests of this model assured General Atomic that there was no credible situation that could have lead to a major breach of this structure that then would have led to the rapid loss of the helium coolant – a very important consideration for gas cooled reactors.  Quite aside from the safety and site selection aspects, a PCRV allowed for the use of local materials and labor in construction of a plant as well as avoidance of the cost of shipping large primary components.

In later installments –  Design of plant components, construction of the plant, and more!  Illustrations used in this series come from a Fort St. Vrain Nuclear Generating Station press package and the various materials contained therein, including brochures, papers and photographs.

Will DavisWill Davis is a member of the Board of Directors for the N/S Savannah Association, Inc. He is a consultant to the Global America Business Institute, a contributing author for Fuel Cycle Week, and he writes his own popular blog Atomic Power Review. Davis is also a consultant and writer for the American Nuclear Society, and serves on the ANS Communications Committee and the Book Publishing Committee. He is a former U.S. Navy reactor operator and served on SSBN-641, USS Simon Bolivar.  His popular Twitter account is @atomicnews.





About Will Davis

Will Davis is the Communications Director for the N/S Savannah Association, Inc. where he also serves as historian, newsletter editor and member of the board of directors. Davis has recently been engaged by the Global America Business Institute as a consultant. He is also a consultant to, and writer for, the American Nuclear Society; an active ANS member, he is serving on the ANS Communications Committee 2013–2016. In addition, he is a contributing author for Fuel Cycle Week, and writes his own popular blog Atomic Power Review. Davis is a former US Navy reactor operator, qualified on S8G and S5W plants.

18 thoughts on “Fort St. Vrain in Pictures: 2

  1. Roger Dudley

    There are several films of FSV construction at the Denver Public Library in the Western History & Genealogy Department. They are part of the Public Service Company of Colorado Collection, WH1367 is the number of that collection.

  2. publius

    Incidentally, relevant to Rod’s question, for larger HTGRs, GA proposed a more “square” PCRV layout in which the steam generators would be arranged around the core, rather than below it. This is the same approach taken by the British design, which was based heavily on MAGNOX & AGR experience, as well as the “Dragon” experimental reactor. It’s very unfortunate that UKAEA & CEGB didn’t go ahead, instead choosing a PWR design for Sizewell B.

  3. Ted Quinn

    Thanks Will for publishing this, the great history of FSV and initiating all of the great comments that came in too! I learned a lot about the history of FSV and this technology! I also look forward to your future updates and pictures!

  4. Ted Borst

    Thanks for a fascinating article. As the Radiation Protection Manager for PSCo at FSV, I can attest Jim Joosten’s second “interesting tidbit” is completely untrue. Elevated radiation levels were discovered in the Visitor Center, not the training building, during decommissioning. The area of interest was promptly and easily cleaned. This discovery, although unexpected, did not have any financial impact on the decommissioning program.

  5. Will Davis Post author

    Mr. August, I’d like to incorporate some of your comments here into the piece that’ll cover the Fort St. Vrain NSSS if you’d agree to it!

  6. Greg Skelly

    The “Top Head” of the reactor was accessible by fuel handling equipment. Refueling of the reactor’s core was accomplished by changing out the graphite fuel blocks.

    Steam Generators above the core would of hampered refueling operations.

  7. JK August

    Rod Adams:

    Have you come across any mention of why Ft. St. Vrain’s steam generators were below the reactor instead of above it, as in a PWR, for example?

    Rod, the basic reason was that FSV was a gas-cooled plant. There was no need for natural recirculation flow as is required for the safety case of a PWR plant, for example. In fact, in a similar condition, all the HTGR has to do, within 72 hours, is blow the notorious boron balls that safely shutdown the plant (mentioned elsewhere). OBTW, the reason the boron balls failed was not the fundamental design. It was prolonged dryout operations over the space of about 15 years (11973-1986) that allowed moisture to contribute to boric acid “bridging” of the balls. (Hmmm, where else have we heard about boric acid problems — that hard white glue that binds up stuff all over where it comes out?) IMHO, had the original plant’s helium PC chemistry been maintained as designed, there would not only not have been a boron ball bridging issue, but about ten other problems would have vanished, too. SCC on the control rod drive mechanism cables, for example. I have to give it to the US NRC, after all that; they let the Fort have a good go of it, what with all the problems it had. The oversight on the part of GA to not provide better shutdown PC cleanup capacity for moisture is one redesign effort that could easily correct about 1/2 of everything not right about the original plant. Oh well!!!

  8. JK August

    Okay, the discussion has reached one of my pet peeves — the Helium Circulators with their water bearings and intermediate “buffer-midbuffer” labyrinth seals. In my opinion, the basic design of a circulator with a water bearing and intermediate midbuffer leak off point worked perfectly. That fundamental design was right. Okay, you say, why all the disparaging hype about the Bearing Water System, water injection into the PCRV primary circuit and its related negative content?

    The answer is the complicated two loop design and single loop shutdown logic of the plant’s Loop Shutdown and Reactor Trip Safety Logic. They were just too complex to work reliably, especially in concert with the Emergency Feedwater (EFW) System backup to the Backup Bearing Water (BUBW). We took the BUBW system and its EFW feed out of service for several years in the early 1980’s, and had the best operating results of all time. (!!) Why, oh why then, you ask, did we put them back into service?!! Well, in large part, it was our oversight, the US NRC, and how they interpreted the book. They found two sentences in our FSAR that implied that BUBW — a non-safety system — would always be in service (back in the day, before Standard Tech Specs and FSAR licensing). Net effect: the problematic but massively overkill safety contribution (about 10 layers of Defense in Depth) went back into service, bringing along with it all the problematic historical Primary Coolant water injections that eventually killed the commercial plant.

    A relicensed FSV under current PRA risk-informed initiatives would not have suffered the same fate, IMHO, and could have been operated to technical if not commercial success.* Oh well, it’s just history now. My takeaway is that any new design must be developed and operated as a pilot plant by DOE, first — just like the first LWR plants were, as Navy sub designs. Once all of the bugs are out, including licensing bugs, then you can take them to the mat, commercially. Just the humble opinion of someone familiar with the legacy of the plant.

    *FSV could have never been a commercial success, operating with highly enriched U fuel — 93.6%. It depended on fuel reprocessing for its overall favorable fuel cost structure. Jimmy Carter ended that in 1977, when, IMHO, PSCo should have sued the Federal Government on the part of its ratepayers for investment recovery (~$1 billion over life of the plant) and politely bowed out. All my opinion, by the way, but the economics are a well-known fact. FSV was designed to incorporate spent fuel reprocessing, and without it, it was an economically disabled plant.

  9. John Solakiewicz

    Phil Wagner, I was at FSV from 1974 to 1980. My first assignment as a PSC Nuclear Engineer was to determine the report-ability of our accidental insertion of our backup shutdown system’s B4C “Boron balls” into the reactor. Yes, these balls did freeze up in the Chicago Nuclear Pile 1, but not at FSV. I have a candy jar of them on my shelf. The Results personnel were pressurizing the B4C chamber, but a valving error released the pressure too quickly and it burst the rupture disc, releasing the balls. GA had a great set of engineers supporting us.Great fuel design and reactor design. I calculated a 39% reactor efficiency in our start-up testing.

    The plant’s final issue was with the weld ID markings, using sharp edged dies, which over time was resulting in pipe crack initiation.

    Helium’s short half life was great in leaving a relatively clean radioactive foot print, resulting in low or no contamination issues.

  10. Steve Short

    Thanks Will for the excellent history of the Ft. St. Vrain plant in this and your previous article. Because of the excellent safety (“melt”-safe TRISO fuel) and radiological (low occupational dose) characteristics of HTGRs, their relatively high thermal conversion efficiency, and use of the Brayton thermodynamic cycle (which negates the need for large quantities of water), I believe HTGR technology to be far superior to LWR technology. Hence, I agree with Sam and hope that HTGR technology gets another chance, which is a huge challenge given the significant investment and infrastructure that already exists around LWRs.
    However, I do disagree with your assessment that PCRVs are the way to go. With today’s emphasis on the need to be economically-competitive with other technologies such as natural gas, the on-site construction required for PCRVs only contributes to the already high capital cost of nuclear power. Modularization (i.e., maximizing factory fabrication and minimizing on-site construction), in the absence of incentives such as a tax on greenhouse gas emissions, is almost essential to demonstrating economic competitiveness. I am not aware of any current HTGR designs proposing use of PCRVs.
    I’m looking forward to your additional forthcoming articles on the Ft. St. Vrain plant.

  11. Jim Joosten

    For the analysts, here are a few more picture and a little further technical detail about Ft. St. Vrain’s low radiation levels:

    “During the four-year decommissioning period, and despite the fact that personnel spent 340 percent more time in the radiologically controlled areas than originally forecast, the project
    experienced a total radiation exposure of only 380 person REM (3.80 person sievert). This number, approximately 12 percent under the original radiation exposure estimate, is roughly
    equivalent to the expected person-REM exposure during one year of operation for a light water reactor. In addition, the FSV personnel contamination rates were only 54 percent and
    24 percent of the contamination rates for typical pressurized water reactor and boiling water reactor outages, respectively. Moreover, the project maintained a low (including all
    subcontractors) lost-workday incident rate of 0.70 per 200,000 person-hours. This rate, when compared to the construction industry average incident rate of 5.5, is exemplary.” – M. FISHER, PSCC / IAEA

  12. Jim Joosten

    Good story. Thanks. Let me add another interesting tidbit. Back in the 1980s, I had a chance to tour Ft. St. Vrain with an NRC Commissioner. At the time, we were interested in why the plant had such a horrible record with respect to radiological safety citations. As it turned out, to our surprise, it wasn’t a problem with fuel leaks and radiation levels per se, BUT rather the lack of it. The fuel in this HTGR reactor was so perfect and radiation levels in the plant so low, that the staff had become somewhat lax with their diligence. They were being cited repeatedly for not taking RP controls seriously enough.
    I recall another interesting tidbit, if true – namely that the reactor had a relatively costly decommissioning program for the time. As it turned out, a significant cost component was reportedly not due to plant contamination from fuel leakage but rather due to a human error in the training building, i.e., apparently an employee inadvertently dropped and dispersed the thorium mantle from a camping lantern that was being used to demonstrate detectors. That’s how low the reactor induced radiation levels were.

    Bottomline, Ft. St. Vrain was a unique and very successful reactor design IMHO. The project collapsed and became uneconomic not due to poor engineering but rather poor NSSS support. GA was a brilliant company, but a bit too academic and not sufficiently supportive of the utility. GA lacked the navy nuclear experience that Westinghouse brought to the table in the PWR world.

  13. Will Davis Post author

    Thank you very much, Sam – and thank you for providing this back story! I have the when-and-where of all this but decided to leave it out as I was trying to really limit myself to a photo-montage. That’s very hard to do! So, I’m very glad you’ve given us the background and having this from the first person is really priceless and great to have preserved here as a comment. Wishing you all the best.

  14. Sam Ross

    I am impressed by the way Will Davis has accurately reconstructed historical details about Fort St. Vrain. My first experience with nuclear power occurred when, in 1961, my employer, Public Service Company of Colorado (PSCo), assigned me to spend two years with General Atomic (GA) in San Diego as PSCos representative for High Temperature Reactor Development Associates. HTRDA was the 53-member group mentioned by Will Davis that was formed to lend support for construction of the 40 mW prototype Peach Bottom Atomic Power Station on the Susquehanna River in southeastern Pennsylvania. Peach Bottom incorporated a graphite moderated High Temperature Gas-cooled Reactor (HTGR) using helium as the coolant.
    I returned to PSCo in 1963. At that time PSCo was the lead company in a group of Western utilities known as Advanced Reactor Development Associates (ARDA). Results of the ARDA program were eventually incorporated into the Fort St. Vrain plant. When I returned to PSCo in 1963, I undertook a study of potential power plant sites. After PSCo entered into a contract with GA in 1965 for construction of the plant, we selected a site only about three miles from one of the sites that I had studied. I then became involved in coordinating licensing activities with GA. Licensing activities culminated in scheduling a meeting in Washington in early April 1968 with the AEC Advisory Committee on Reactor Safeguards. Instead, race riots erupted in Washington, central Washington was set on fire, and we were holed up in our rooms in the Watergate Hotel. The contract with GA specified that construction of Fort St. Vrain was contingent upon the successful operation of Peach Bottom. Hence, beginning in 1965, I made frequent visits to Peach Bottom to monitor final construction and start-up. As construction of Fort St. Vrain proceeded, we organized a Nuclear Safety Review Committee, and I became Secretary.
    It was all an interesting, though sometimes hectic, process. I still believe that the HTGR is the best reactor concept that has reached the commercial operation stage, but it suffered the fate of other historical examples of “better” technology that could not overcome the momentum of other well established technologies, in this case particularly the PWR which had the advantage of concentrated development in the Navy submarine program. Most of the problems at Fort St. Vrain related to mechanical problems, particularly in developmental equipment such as the helium circulators. The strength of the HTGR probably lies in the fuel and physics which, in the case of Fort St. Vrain, performed exactly as predicted. Each of the tiny fuel particles of highly enriched uranium encased in layers of pyrolytic carbon provided a miniature containment sphere. The reactor operated on the thorium cycle which in itself offers advantages. I am still hopeful that the HTGR will surface again the future of reactor technology.

  15. Philip Wagner

    There were many details about the poor engineering that went into the design of the FSV facility. While in theory many seemed like an efficient design, the hanging heavy, water driven He circulators vertically and spinning them at 10krpm while lubricating with water and having labyrinth seals to prevent that water from leaking into the PSCRV was a major problem. While good for thermal efficiency, it was ripe for problems that plagued the facility from the start and limited actual operation to single digits over its lifetime. In addition, the ability to test the emergency shutdown system was not provided and when an actual test was conducted the boron carbide balls did not drop through the tube because of moisture causing the individual balls to form a lump. Those are just two of the bigger problems but not the extent of the poor design that essentially destroyed any hope of advancing HTGR technology. Had the follow-on HTGR plants been developed and put into operation with numerous modifications (e.g. motor driven circulators) we would probably still be building HTGRs. FSV is a poor example of a great technology.

  16. doug

    Thanks. Very interesting. When is the US going to get smart enough to generate a long term energy policy in which the basis is nuclear power. Falling behind to Korea, China and Japan is a disgrace to our pioneering science & technology heritage.

  17. Rod Adams


    As usual, your historical photos and technical details are fascinating and awe inspiring.

    As much as I have studied and read about Ft. St. Vrain over the years, I’m not sure that I was ever as aware of the fact that the steam generators were completely below the reactor.

    In most heat transfer loops, the heat sinks are located higher in elevation that the heat sources.

    Have you come across any mention of why Ft. St. Vrain’s steam generators were below the reactor instead of above it, as in a PWR, for example?

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