SAWTOOTH LIFT: Nuclear Power Design
February 22, 2011 • 10:37AM

The crucial inflection point in the 1964 NAWAPA design of the Ralph M. Parsons Co. is the transfer function, where flows from the collection function exiting the Rocky Mountain Trench are transferred from the Columbia Basin up into the Great Basin and on through the distribution function. This transfer function hinges on the Sawtooth Lift, consisting of 6 pump lifts, for a total of 2450 feet, and dealing with a flow rate of 85,500 cubic feet per second flowing through the Sawtooth Lift, would require 26GW of power. Powering this 26 GW pumping system with the most advanced nuclear power considerations, serves as a the trigger for a long awaited nuclear Renaissance in the United States. Below is an educated estimate of the nuclear power options for NAWAPA's pumping necessities and some relevant considerations involved, presented by retired nuclear engineer Dewitt Moss, contributor to the fuel concept of the Washington State Fast Flux Test Facility(FFTF).

By Dewitt Moss


The time/effort involved through communications with you and your contributors plus some recent technical news articles have caused me to reflect on the monumental task confronting this program. They may be surmountable and solved with extensive and expert management—but we should be aware of the issues in order to address them as they relate to the NAWAPA program.

Let us consider the following: NAWAPA may require 30 (+or‐ a few) nuclear stations to be operable, some as early as 10‐15 yrs.from start and others up to 10‐15 years later. These reactors will average out at 600 to 1200 MWe. Add to this nuclear plant requirement, the following quote from the Global Energy Reporter, dated 1‐16‐11: “The industry is pursuing a variety of strategies to overcome its financial challenges, even as it projects the need for more than 46 new nuclear plants by 2030 to meet U.S. power demand and Washington’s target for reducing greenhouse gases.” Does this country have the industrial infrastructure, including the manufacturing‐fabrication capacity to do this? In my assessment, probably not, without Manhattan Project type of commitment from the U. S. government, industry and cost guarantees to utilities. Proven and tested technical expertise, quality assurance and control practices, material standards, operating and maintenance procedures, reliable equipment, operations and safety procedures and systems are paramount. These plants are massive, complex and expensive requiring demonstrated proven technology. In the case of new untested concepts, such as the PRISM, LMFBR or IFR, a demonstration plant is definitely in order.

Does the U. S. have the available U and Pu to fuel 70‐80 nuclear planned reactors in a 20 year period or does this mean we consider reprocessing some existing inventory of spent fuel? Reprocessing facilities or new fuel enrichment facilities may be required. We possibly have weapons grade material that could be blended for fuel. Also, I believe a French company has applied for approval to build an enrichment plant at INL—I do not know its current status.

In addition to the above mentioned manufacturing capability, do we have in this country the necessary and qualified engineering talent to assure these plants are built to the rigid required standard, on schedule and within the required costs? If not, we need to develop this expertise via training and education. This cost burden of nuclear plants is definitely impacted by the recent surge of building material commodity prices. Again, I quote from the Global Energy Reporter of 1‐16‐11: “In 2009, MIT doubled its forecasted construction costs of new nuclear plants, while the U. S. Information Administration increased its 2009 estimate by 37% just this past December”…”All cost estimates have a huge amount of uncertainty…there is a big unknown in how reliable the contractors are going to be in coming through with their estimated costs. And similarly, how good they’ll be at constructing them on time”. (John Parsons, director of energy and environment program, MIT Sloan School of Management)

The above views and opinions reflect some major issues and obstacles that will confront the program. These issues will be there tomorrow and into the foreseeable years to come. They probably are resolvable with adequate program planning and resources applied. It won’t be simple, but with expert leadership—doable.

Now, back to your initial request of the preferable type of reactor to power lift pumps.

BWR’s and PWR’s have demonstrated operating experience with considerable industrial capacity in the U. S. The fabrication capability of the large pressure vessels required may need to be re‐established within the U. S. The Westinghouse PWR’s, generally , are a loop type design to have 2, 3 or 4 loops of 300 MWe, each; thus allowing some sizing at a specific power site. Other PWR’s and BWR’s are of a single rating of about 1200 MWe. Economy of size is important for nuclear plant construction and operation.

In my opinion, these plants could be built on today’s well proven experience. No demonstration plant is required. The newer, safer plants are not dissimilar to the existing 100 nuclear commercial plants operating in the U. S. Reprocessing , waste disposal and some Pu proliferation issues remain. If one had to choose today for a plant(s) to be operable in a 10 to 12 year period, without further developmental costs, these concepts would suffice.

If we look to advance the U. S. nuclear program beyond the existing water reactors, liquid metal (primarily Sodium) fast reactors should be considered because they can breed, have much higher power density per unit volume of core, the fuel can be processed on‐site and significantly reduce proliferation concerns. One concept, PRISM, as proposed, would use a dissolving and electroplating concept to reprocess spent stored fuel.

Three advanced concepts exist that may be considered:

1. PRISM (Power Reactor Innovative Small Module)
2. IFR (Integral Fast Reactor)
3. LMFBR (Liquid Metal Fast Breeder Reactor)

Below, I have included a summary of these concepts, followed by a reflection on the more advanced issues.

1. PRISM (Power Reactor Innovative Small Module)

This is a new concept with significant attributes. The PRISM is a pool type sodium cooled fast reactor with 4 components: reactor core and associated pumps and heat exchanger, a hot cell to fabricate fuel, an Advanced Recycling Center (ARC) to recycle spent nuclear reactor fuel and an electrical steam generator producing 622 MWe. As proposed either 1, 2 or 3 generators, would encompass one nuclear site. The reactor core is fueled with a metallic alloy of U, Pu and Zr, easily fabricated in an attached hot cell. Proposed reactor core design and shutdown mechanisms make the reactor super safe.

Probably the most innovative and attractive attribute is the ARC. The ARC would take spent nuclear fuel, now stored at the 100 or so operating nuclear power plants, expose it to a molten LiCl pool with an adjacent electroplate anode. The uranium based fuel would be dissolved, the Uranium deposited and collected on the anode and made into new metallic fuel. The ARC, if employed, can extract and burn up 90% of the U, in lieu, 2‐3% burn‐up in light water reactors. Proliferation of U and Pu are non‐existent in this concept because it never leaves the reactor,hot cell and ARC complex.

In the U. S., 100 nuclear power plants produce 20 tons of spent fuel per plant per 60 year lifetime. Therefore, 120,000 tons of spent fuel will accumulate. ;26 PRISM‐ARC plants can consume 120,000 T of spent fuel while producing 50,000 MWe, avoiding the emission of 400,000,000 tons of CO2 every year. It would, also, consume our spent fuel inventory avoiding expensive and time sensitive storage. Only small quantities of radioactive waste would need storage at a site such as Yucca Mountain.

GE‐Hatachi has initiated preliminary licensing steps for a single reactor and 50T/yr. ARC separations facility. Their estimate of a demonstration plant available by 2020, would cost $3.2B over a 10 year period.

This is an extremely safe concept, proven reactor materials and equipment and could be a most reasonable and practical method to dispose of the tremendous spent fuel inventory now stored at the U.S. Reactor sites. The ARC concept is a patented, electrometallurgical process—but it needs to be demonstrated and verified A pound of uranium,. As found in nature, has an energy equivalent of about 7500 bbl of oil, or 1500 tons (3,000,000 lbs) of coal. Uranium in nature is 0.7% U‐235, fissionable, and 99.3% U‐238. Commercial reactor nuclear fuel is normally enriched to about 4‐6% U‐235. We now store the 40‐50 year spent fuel inventory of our commercial nuclear power plants at the power plant site because we have no fuel reprocessing plants operable. If processed, about 95% of the U‐235 fuel in this inventory is available for additional power generation. Why would we want to process this spent fuel? The spent fuel of a 1000 MWe plant over a a40 year lifetime, contains the equivalent energy of 5B gal. of oil or 37 M tons of coal. And we have the equivalent of 60 to 80 of these plants in the U. S.

2. IFR (Integral Fast Reactor)

The IFR concept is a Na cooled, pool‐type, fast reactor with a closed fuel cycle employing uranium and/or plutonium metal fuel. The EBR‐II (Experimental Breeder Reactor‐II), a similar concept, was operated successfully for about 20 years. With the proper core and blanket design, this concept can be configured to breed. With an attached fuel reprocessing cell, only chemical wastes are generated and Pu is never outside the hot‐cell reactor complex and is, therefore, unavailable for proliferation. Because of its design with metal fuel and pool sodium coolant, it is very safe in all modes of operation. Double walled heat exchanger tubing on the secondary system is generally used to avoid sodium‐water contact, adding to its capital cost. It can be designed to have a high power density, which requires smaller cores to produce a given amount of electricity, when compared to PWR’s and BWR’s. Because of its lower operating pressure, containment structures are less massive. Operating temperatures are very nominal for the fuels and materials used. Sodium melts at 208 degrees F and boils at 1621 degrees F, while the metal fueled core operates between 640‐905 degrees F.

3. LMFBR (liquid Metal Fast Breeder Reactor)

The LMFBR concept is a sodium cooled, generally a mixed oxide fueled fast flux facility. Alternately, it could be a metal fueled core, hence like the IFR. Demonstration plants exist in England, France, Japan and Russia. Many have been in operation 10 years or so. Most experience is reportedly positive. The concept was developed to breed plutonium, use it as new fuel and extend our uranium resources for hundreds of years. The materials technology and nuclear characteristics are well established. In a closed cycle like the IFR above, Pu concentration for proliferation would not be a problem. Metal fuel lends itself to the closed cycle, but the refabrication of irradiated, mixed oxides in a hot‐cell complex is relatively undemonstrated. It would be a new concept to license, hence time consuming and expensive.

The U. S. had a design and construction underway at Clinch River, TN for a demonstration plant but was cancelled in the early l980’s, for possibly political and economical reasons and the viewed lack of need for a breeder reactor at that particular time.

The FFTF (Fast Flux Test Facility), a sodium cooled fast flux test facility was built in the late 1970’s, early 1980’s at Hanford, WA. It was fueled with mixed oxides of U and Pu to test and evaluate materials, operating characteristics of the equipment and core, and reliability of equipment and other related purposes of a new concept. It operated successfully for several years. It is shut down now.

The above concepts have pluses and minuses. Many concerns would be addressed via a vigorous demonstration plant program. A demonstration plant of intermediate size, 300 MWe to 500 MWe would be in order.

Reflecting on the above Na cooled concepts, I offer the following personal views for further knowledgeable peer review and discussion.

Argonne National Laboratory (ANL) has extensive reactor design experience of a pool type, Na cooled, and metal (U and PU) fueled cores—attached to a hot cell fabricating facility, i.e. the EBR‐II facility. Metal fuel was chosen for the core for a high density fuel, high power density, high fast flux and breeding capability testing. Metal fuels originally had a problem with swelling thus limiting their core lifetime, thus effecting costs. To my best knowledge the swelling issues have been made manageable by fuel alloying and fuel element design. However, I’m not currently aware of the tested and verified maximum burn‐up achievable with metal fuels.

Industry and utilities, initially, did not embrace the Na cooled‐metal fuel concept because of the extensive experience with UO2 fuel and water coolant reactors. The design and maintenance of equipment and water cooled reactor operation was perceived as a better proven concept than the Na cooled concept.

ANL designed and operated EBR‐II. Westinghouse designed and operated the FFTF, since they were the operator of the Hanford facility at the place and time of the FFTF. General Electric was a prime contractor to the government for testing mixed‐oxide fuels in a Na cooled reactor environment. This occurred during a 20‐30 year period commencing in the 1960’s and GE has extensive experience with UO2 fueled water cooled reactor concepts for the next generation of water cooled thermal reactors.

It, therefore, is of significant note that GE is proposing as an advanced reactor concept a Na cooled metal fueled fast reactor with an attached reprocessing hot cell. This PRISM concept addresses proliferation and spent fuel reprocessing issues. Also, it implies, a major industrial vendor is endorsing a metal fueled, Na cooled fast reactor. The GE choice, with their significant experience and resources, should be a major factor in the selection of an advanced reactor design. ANL still exists to assist (if required) in the design of equipment and components of a Na pool type reactor where most fuel handling is accomplished in a non‐transparent Na pool with an Argon inert gas.

In my view, the PRISM concept, with GE’s formidable technical experience and resources may be a preferred concept for NAWAPA to advance the U.S.’s nuclear program.

The PRISM concept, as proposed, with its ARC (Advanced Recycle Center) is probably the most complex of the concepts considered. The PRISM concept without an ARC is not dissimilar to an IFR concept. Without the ARC, the ability to reprocess spent stored fuel is lost.


Dewitt Moss

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