10 Cool things about Nuclear Waste
Posted by Things Worse Than Nuclear Power × 1:45 AM
The height difference between a man and the moon is like the energy density difference between nuclear and any carbon combustion.
1) It still contains 95% of its energy. That’s like taking a couple tablespoons out of a liter of pop and throwing the rest away… if we don’t make use of it!
2) Nuclear fuel is around 2,000,000 times more energy dense than coal, oil, and biomass. This means the quantity of nuclear waste is super tiny for the super huge amount of energy it produces. 2,000,000 times is like the difference between the diameter of the moon compared to the height of an adult person.
3) It can be recycled, and actually produces energy while being recycled, instead of requiring energy to recycle! This can be accomplished with a Molten Salt Reactor configured to be an actinide burner.
The IFR- Integral Fast Reactor
4) If recycled, the waste lifetime is reduced to around 300 years. This (both point 3 and 4) is possible using fast reactors, which have been demonstrated through over 400 reactor-years of operation or with Molten Salt Reactors.
5) In fact, a cool thing about radioactive elements in general is that they go away with time– and the more radioactive they are, the quicker they go away. Some decay in a few seconds. A general rule of thumb is that after a few decades, the most dangerous elements are effectively gone.
6) You can easily detect even a single atom of radioactive material. You cannot detect arguably more dangerous pollutants such as mercury, lead, NOx or SO2 with near that kind of accuracy or ease– elements that fossil fuel plants are putting out in significant quantities every minute of every day worldwide. These dangerous pollutants do not go away with time, worse, metals like mercury actually bio-accumulate.
7) Fission products save lives. Many critical medical isotopes used for cancer treatment, diagnostics, and more, can only be produced after irradiation in reactors. Cesium 137, a fission product, can be used to protect blood in blood banks, saving lives of babies and the immuno-deficient. This is just one example of possible uses of fission products made in nuclear reactors.
8) Dry cask storage is one of the most robust structures made by man. It can survive being dropped from thousands of feet, direct impact from an airplane crash, or explosives!
Medical isotopes are produced in reactors
9) To that end, there has not been, in the history of worldwide commercial nuclear power, a known case of stolen commercially used nuclear fuel. It is too mixed up with different things and too difficult to handle to be very useful to anyone with bad intentions.
10) Even including the worst accidents in worldwide nuclear history, nuclear has the best safety record (deaths/yr) of any type of energy source, including wind, solar, natural gas, and coal. There have been no deaths in the history of U.S. commercial nuclear power due to exposure to nuclear waste.
Is it “waste” or “used” fuel with lots of value for society?
This paper outlines the investigations into molten salt reactors undertaken at the Oak Ridge National Laboratory (ORNL) in the 1960s and 1970s. The advantages of the thorium cycle are described and the reason why the work was not taken further is given.
The paper outlines the thorium cycle and assesses its potential for warship propulsion by illustrating how a medium sized surface warship might be powered by a reactor plant based on the molten-salt demonstration reactor plant designed by ORNL.
Nuclear power for submarine propulsion using moderately enriched uranium fuel in a Pressurised Water Reactor (PWR) has a long and highly successful history. But there are disadvantages: the high pressure demands high mechanical strength in large components which must be maintained through life; while the moderate operating temperature limits plant efficiency.
In comparison, thorium, used in a suitably designed reactor plant:
a. Operates at high temperature and low pressure simplifying the mechanical design and yielding increased thermodynamic efficiency with consequent reductions of component sizes.
- Offers a much improved fuel cycle with fewer and less troublesome radioactive waste products.
- Is much more abundant than all uranium isotopes.
The paper that now follows is split into two sections. The first summarises information that is available in the public domain and does not intend, or pretend, to offer new information or insights. Section 2 provides the authors’ view on how a thorium molten salt reactor might be used to provide power for a medium sized surface combatant of around 8,000 te displacement.
The post Study on the Potential of Molten Salt Reactor Use in Ships! appeared first on The Energy From Thorium Foundation.
Jiang Mianheng Why Nuclear Power in China Thorium & the Energy Outlook of China @ ThEC12
The post Video of China’s Plan to Exploit Thorium from ThEC12 appeared first on The Energy From Thorium Foundation.
Paper ID Number 10990
Presented at the 121st ASEE Annual Conference and Exposition
- Miss Yoonjo Jo Jo Lee, University of Missouri
- Mr. Matthew Paul Simones, Nuclear Science and Engineering Institute, University of Missouri
- John C. Kennedy, University of Missouri
- Hakan Us
- Mr. Philip F Makarewicz, University of Missouri
- Dr. Janese Annetta Neher, Nuclear Science and Engineering Institute-University of Missouri Columbia
- Dr. Mark A Prelas, University of Missouri, Columbia
Authors Brief Bio
Miss Yoonjo Jo Jo Lee, University of Missouri
Jo Jo Lee received her bachelor’s degree in Chemical Engineering and minor in Chemistry from the Uni-versity of Minnesota-Twin Cities in 2004. After working about a year as a process engineer, she decided to become a lawyer. She received her Juris Doctorate at Hamline University School of Law in St. Paul, Minnesota in 2008 and she was sworn in as an attorney that year. After law school, she worked as a patent analyst for advanced chemical engineering patents. In 2010, she enrolled in the NSEI doctorate program to pursue her PhD in nuclear engineering. Her research areas include graphite oxidation, candidate alloy oxidation, carbon transport in HTGRs and thorium nuclear fuel.
Mr. Matthew Paul Simones, Nuclear Science and Engineering Institute, University of Missouri John C. Kennedy, University of Missouri
Currently a Ph.D. candidate and Graduate Research Assistant in the Department of Mechanical and Aerospace Engineering at the University of Missouri. Specialize in experiments and numeric modeling of Fluid-Structure Interaction for nuclear fuel applications. Earned a M.S. in Mechanical and Aerospace Engineering in 2012.
Mr. Philip F Makarewicz, University of Missouri
Dr. Janese Annetta Neher, Nuclear Science and Engineering Institute-University of Missouri Columbia
Janese A. Neher is the mother of two sons, Max Nicklas and Stock William Neher. Janese is a Licensed Professional Engineer; a Licensed Professional Counselor in the State of Missouri, and a Missouri Bar approved Mediator/Arbitrator. She owns a small business, Counseling and Coaching to Change with Mediation Services.
She has worked in the nuclear industry for 22 years in Nuclear Oversight. She has also worked at the Missouri Department of Natural Resources and the Missouri Public Service Commission. In 2010, she received the coveted Patricia Bryant Leadership Award from the Women in Nuclear Organization, the Region IV Women in Nuclear Leadership Award, and the University of Missouri’s Chancellor Award for her support in the area of women’s diversity. She was also awarded the Ameren Diversity award in 2009 and 2010.
She has received the following degrees: Bachelor of Science in Civil Engineering, Mathematics and History; a Masters in Environmental Engineering and Education; and PhD in Nuclear Engineering from the Nuclear Science and Engineering Institute at the University of Missouri-Columbia.
Dr. Mark A Prelas, University of Missouri, Columbia
Professor Mark Prelas received his BS from Colorado State University, MS and PhD from the University of Illinois at Urbana-Champaign. He is a Professor of Nuclear Engineering and Director of Research for the Nuclear Science and Engineering institute at the University of Missouri-Columbia. His many honors include the Presidential Young Investigator Award in 1984, a Fulbright Fellow 1992, ASEE Centenial Certificate 1993, a William C. Foster Fellow (in Bureau of Arms Control US Dept. of State) 1999-2000, the Frederick Joliot-Curie Medal in 2007, the ASEE Glenn Murphy Award 2009, and the TeXTY award in 2012. He is a fellow of the American Nuclear Society. He joined the University of Missouri nuclear engineering program in 1979 as an assistant professor.
This paper was generated as part of a course on advanced nuclear fuel cycles supported through a curriculum development grant from the Nuclear Regulatory Comission. The course was graduate level and required a research component. The students in the course chose the topic of “Thorium Fuel Cycle for a Molten Salt Reactor: State of Missouri Feasibility Study.” The study consisted of developing a primer to be shared with interested parties in the nuclear community and the state. This paper was generated from this research study and the approach students took towards synthesizing the primer are annotated throughout.
The thorium fueled molten salt reactor concept is one of the most promising nuclear reactor designs currently being studied, but mostly outside of the United States. Thorium reactors are desirable due to the high availability of naturally occurring fuel. However, in order for thorium fuel to be a viable alternative to uranium, a harmonious relationship between the reactor physics and molten fluoride salt transport medium must be fully understood. Chief among these technological hurdles is the use of continuous processing of spent fuel to remove fission products while the reactor is online .
The voluminous literature on molten salt reactors mostly dates to the 1960s era. Notably, in the U.S. the Molten Salt Reactor Experiment at Oak Ridge National Laboratory was an 8 MW(th) reactor that was designed primarily to study the technical feasibility and safety of using a molten salt based fuel and coolant. In addition to demonstrating the practicality of a molten salt reactor, the Molten Salt Reactor Experiment also addressed issues of on-line refueling, fuel makeup, and salt chemistry. Towards the end of the Molten Salt Reactor Experiment, and continuing after its shutdown, research efforts focused on techniques for separation of waste products, namely protactinium . Given the prevalence of uranium based technology in the military at the time and as a matter of politics, there was little desire in the U.S. to fund nuclear research that did not provide a direct defense-related benefit.
Today, as we aspire as an industry to reduce nuclear proliferation and build safer reactors, research efforts have shifted towards reducing the amount of 235U available, particularly highly- enriched uranium. Current power reactors use low-enriched uranium (235U content of less than 4.95%). In addition to the benefits of avoiding uranium enrichment, thorium for nuclear power production is also supported by a growing demand for clean energy both in the U.S. and abroad.
Internationally, research in thorium fuels and molten salt reactors is underway in Japan, India, China, France and to a lesser extent, the United States. Presently, international research is needed in the development of molten salt reprocessing technology to allow for the active
removal of wastes from the fuel stream, while simultaneously replenishing the fuel supply. This online fuel recycling technology could lead to a significant advantage over uranium based reactors such that a fluid fuel stream allows the reactor to be continuously re-fueled, thereby significantly reducing reactor shutdown time.
Given the future economic aspirations of the nation, and in particular the state of Missouri, it is difficult to ignore the potentially huge impact of a new thorium based economy. The expansion of current mining facilities in Missouri to extract thorium could easily be achieved as thorium is present in the waste products of several mining operations already throughout the region. Moreover, the central and strategic location of the state of Missouri along the Mississippi River naturally postures the state into being a major shipping point and railway center for mineral exports. Intrastate mining, utilities, engineering, and construction companies could benefit significantly from an establishment of a thorium industry. Missouri is perfectly poised geographically, economically and academically to nurture next generation technologies for energy independence of the state and the nation as a whole.
U.S. Development of Affordable Clean Energy
that is Cheaper than Coal
“Let us not seek the Republican answer or the Democratic answer, but the right answer. Let us not seek to fix the blame for the past. Let us accept our own responsibility for the future.”
President John F. Kennedy
A problem with most traditional natural energy resources, such as coal and oil, is that they are cheap only when they are abundant, and expensive when they are not. These peaks and valleys in our energy markets are a result of accessibility and abundance. Simply put, the more abundant an energy source is, and the easier it is to access, the more affordable it is. But interrupt supply, or reduce it, and prices skyrocket.
Embracing fossil fuels, along with their constant boom and bust cycle, is a long-term energy plan that will have our economy riding the roller coaster of success and failure respectively in times of boom and bust. It is not a long-term plan to provide any type of lasting stability to the markets. Fossil fuels simply are neither abundant enough nor accessible enough.
What we need is a clean energy resource that is abundant, accessible, safe, reliable, and affordable.
Until now, energy that was cheap tended to be the dirtiest (solid and liquid fossil fuels), and energy that is clean (renewables) was the most expensive. Expensive energy puts the brakes on the economy and hurts those that can least afford it, the poor, by increasing our cost of living. Dirty energy hurts our environment and our health and is not a sustainable long-term policy either.
We are at an energy and environment crossroads: we want clean energy, but we do not want to pay any more for it. If we sacrifice our economy by the government’s mandating the use of expensive clean energy technologies to improve our health, many of us, including our poor, will not be able to afford our energy bills for gasoline and electricity. Our neighborhoods will suffer, and eventually the health effects of poverty caused by the unavailability of cheap and abundant energy will negate any benefits of a cleaner environment.
And it is not just our bills for direct energy consumption that go up. Higher energy costs amount to an increased “tax” on everything we buy and use. Everything we make, everything we grow, everything we transport, uses energy. If the cost of diesel fuel rises, truckers pay more to transport goods, and that increased cost is passed on to the eventual consumer. When a trucking company passes on its increased fuel costs to, say, McDonalds, to keep them supplied with hamburgers and buns, McDonalds covers these increased costs by increasing the price of its hamburgers. If they did not do this, they would not stay in business, and no one would have McDonalds hamburgers any more. When energy prices rise, all of us pay the price in everything we buy, not just for the gasoline and the electricity we use.
But there is a solution to this. What if we had an energy resource that was so common, so abundant, and so cheap that everyone would have the ability not only to sustain themselves, but to thrive? That resource is the element Thorium, and every nation has it! Thorium can provide us with ultra clean energy cheaper than coal!
Dr. Carlo Rubbia, who shared the 1984 Nobel Prize in Physics, described Thorium as having “absolute pre-eminence” over all other fuels, including fossil fuels and Uranium.
China, Russia, and India are all aggressively pursuing the commercialization of Thorium energy technology. A few American companies want the chance to develop this technology. We missed our first and best chance to do that ourselves. America was well on its way to commercializing Thorium technology in the 1970’s, but when there was a downturn in the economy, the technology was shelved at the time due to the cost to commercialize it.
Today, more than ever, we need affordable and accessible energy if we are to remain competitive in the world marketplace, and it must be clean energy. With abundant, low-cost energy from Thorium, our economy will stop shipping jobs overseas because our manufacturing costs will go down, wages will increase, and there will be many more jobs here.
In addition to electricity, Thorium technologies will provide inexpensive fresh water from seawater; inexpensive synthetic gasoline and diesel to get America completely off foreign oil; and two kinds of medical isotopes: one for which there is now no American supply, and is used in over 320,000diagnostic procedures in the U.S. per week; and a second isotope, now very expensive and in short supply, which is needed for cancer research and for a possible cure for many cancers. Finally, Thorium technology will produce isotopes to power NASA’s deep space probes.
To learn more goto: http://www.Th90.org
Recent molten salt reactor (MSR) developments in Europe and Russia address an advanced large power unit without graphite in the core. In new MSR concepts neutron spectrum is hard enough and the reprocessing rate strongly reduced compared to prior graphite moderated Th-U MSBR design developed at United States ORNL. The consideration done demonstrates the MSR potential as the system with flexible fuel cycle options which can operate with different loadings and make up scenarios based on transuranic elements from spent LWR fuel without and with Th-U support. This system may operate in the actinide recycler and transmuter, convertor as well as breeder modes without essential changes in the core/blanket configuration and find its place in any scenario of nuclear energy development.
We would like to know how you feel about the following relationship we have with China?
Currently, the United States is helping China to develop MSR technology. Many of our legislators have bought the “Beyond Nuclear” accusation that Thorium MSR technology is a proliferation problem.
If it were a proliferation problem why are we helping China to develop this technology?
Why isn’t “Beyond Nuclear” making more of a stink about us assisting China in the development of a ThMSR?
The post China and U.S. Global Partnership: All Right! or Outrage? appeared first on The Energy From Thorium Foundation.
D. E. Holcomb
G. F. Flanagan
B. W. Patton
J. C. Gehin
R. L. Howard
T. J. Harrison
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37831-6283
managed by: UT-BATTELLE, LLC
for the U.S. DEPARTMENT OF ENERGY
under contract DE-AC05-00OR22725
Fast Spectrum Molten Salt Reactor Options
Fast-spectrum molten-salt reactors (FS-MSRs) have been the subject of periodic investigations since the early 1960s. However, none has ever been built; and the investigations have not proceeded beyond high- level material balance, heat transfer, and chemistry exploration. While the principal concepts underlying FS-MSRs have not changed over the intervening decades, much of the underlying technology base has evolved. The principal purpose of the current report is to provide an integrated overview of FS-MSR design options reflecting the current technology status.
FS-MSRs can be employed to consume actinides from light-water reactor (LWR) fuel or, alternatively, to extend fissile resource availability through uranium-to-plutonium breeding. FS-MSR reactors are highly flexible and can be configured into modified open or full-recycle configuration. The modified open FS-MSR fuel cycle options do not include chemical processing of the fuel salt. A traditional fully open fuel cycle is not an option with a liquid fuel reactor in that the gaseous fission products inherently separate from the liquid fuel. The conversion ratio of an FS-MSR is largely determined by the fissile-to- fertile-material ratio in its fuel salt. Thus, a single reactor core design may be capable of performing both fissile resource extension and waste disposition missions.
Molten-salt reactors can operate with the salt processing and fuel addition being performed in either continuous or batch operations. For thermal-spectrum systems, it is important to remove the fission products from the salt to minimize the parasitic neutron capture that results from fission products with large capture cross sections. In a fast-spectrum system, these parasitic losses are lower since the fission product capture cross sections are lower in fast-spectrum energy range. Hence, for an actinide burner reactor to maintain criticality, only sufficient additional fuel need be added to primarily compensate for burnup. Therefore, a fast-spectrum system may result in a simpler salt-processing approach or in a batch processing approach with relatively long salt-processing intervals. Similarly, a slightly positive breeding gain converter-type reactor may be able to operate for an extended period without any material additions or removal. In comparison, a breeder reactor will require fissile material extraction once the reactor control system cannot compensate for additional reactivity.
FS-MSRs have the potential for incorporating excellent passive safety characteristics. They have a negative salt void coefficient (expanded fuel is pushed out of the core) and a negative thermal reactivity feedback that avoids a set of major design constraints in solid-fuel fast reactors. Thus, an FS-MSR can provide a high power density while maintaining passive safety. The liquid state of the core also enables a passive, thermally triggered (melt plug) core draining into geometrically subcritical tanks that are passively thermally coupled to the environment. FS-MSRs have a low operating pressure even at high temperatures, and FS-MSR salts are chemically inert, thermodynamically lacking the energetic reactions with environmental materials seen in other reactor types (e.g., hot zirconium or sodium with water).
In the context of proliferation resistance, FS-MSR fuel has a uniform isotopic concentration of actinides, including highly burnt plutonium or uranium isotopes along with other minor actinides and fission products. The local fuel processing of the breeder and burner configurations eliminates the possibility of diversion during transport. The fission-product–saturated fuel salt of the minimal fuel processing converter reactor is highly self-guarding during transportation. Further, the transport casks are massive because of the required amounts of shielding. In general, diversion of molten salt materials is difficult. The reactor operates as a sealed system with an integrated salt processing system that is technically difficult to modify once contaminated. The hot salt freezes at relatively high temperatures (450–500°C), so it requires heated removal systems. FS-MSRs operate with very low excess reactivity. Loss of a significant amount of fuel salt would change the core reactivity, which could be measured by a well- instrumented reactivity monitoring system. During operation (with the exception of deliberate fissile material removal for a breeder or addition for waste burner), the fissile materials always remain in the hot, radioactive salt. However, FS-MSRs, with integrated fuel separation, may be unsuitable for deployment in nonfuel-cycle states to minimize dispersal of separation technologies. Also, methods of inspection and materials accountability for liquid cores have not yet been fully developed.
All of the reactor-significant transuranic elements can form chemically and radiolytically stable salts with halide elements. Use of heavier halides results in a harder neutron spectrum. Harder neutron spectra improve the reactor burning and/or breeding. However, little information is available about the chemical and material properties of the actinide bromides or iodides. Hence, the FS-MSR investigation was limited to the actinide chlorides and fluorides. In addition to providing a harder neutron spectrum, chloride salts (compared with their fluoride counterparts) have higher solubility for the actinides, increasing the capability of the reactor to accommodate higher fuel loading and thus maintain criticality as fission products build up.
The two thermal-spectrum MSRs operated previously both employed a fluoride-based carrier salt. Much of the structural material information developed for the prior MSR program can be applied to fluoride salt FS-MSRs. The harder neutron spectrum of an FS-MSR, however, can cause additional atomic displacements within the nearby solid materials. Hence the lifetime of the neutron shielding material proximate to the core will be less for an FS-MSR. Additionally, nickel-based alloys embrittle when exposed to core levels of neutron flux (>1020 neutrons/cm2) at high temperatures (>500°C). Thus shielding of the primary vessel from the neutron flux is imperative.
Chloride-based salts have been employed in the fuel-reprocessing scheme developed for the integral fast reactor. However, the corrosion processes for chlorine are more complex than those for fluorine. Consequently the knowledge base for structural materials tolerant of chloride-based salts is not as mature as that for fluoride-based salts. A confident structural material selection cannot yet be performed for a chloride salt-based FS-MSR.
A light-water reactor (LWR)–transuranic burner can either make use of centralized fuel reprocessing or use much of the infrastructure of its fuel processing system to directly accept used LWR fuel, avoiding the need for a separate reprocessing plant. In addition to helium sparging to extract the gaseous fission products and mechanical filtering to remove the noble metal fission product particles, a fluoride salt– based FS-MSR would employ fluoride volatility and reductive extraction processes to separate the fission products from the fuel salt. Chloride salt–based reactors would employ electrochemical separation, zeolite ion-exchange capture, and chloride volatility processing. In either case, longer-lived fission products could be returned to the salt for fast neutron destruction, albeit with relatively low efficiency because of their primarily thermal absorption cross sections. As the separated fission products have relatively small volume, they can be left in salt form and allowed to solidify and decay in short-term storage.
A uranium–plutonium breeder would require an initial fissile material charge to initiate the breeding cycle. A liquid-fueled reactor is neutronically efficient compared with solid-fuel reactors, because it lacks in-core parasitic neutron absorbing structures. Therefore, a smaller amount of initial fissile material in the core is required to start up an FS-MSR compared with a sodium fast reactor. Note, however, that the FS- MSR will require additional fissile mass as a result of the fuel salt outside the core. The separations processes for a breeder would be nearly the same as for a burner except that the excess fissile material would be electrowon from the salt. Processing can be done in either batch or continuous form.
A minimal separation, modified open cycle, converter reactor would allow fission products to accumulate within the fuel salt. The lower melting points and much higher elemental solubilities afforded by chloride salts, as well as the resultant harder spectrum, make the chloride salts preferable for a limited-separation converter reactor. Sufficient natural (or depleted) uranium chloride or fluoride could be added to the reactor to compensate for any mismatch between fissile breeding and burnup, as well as the small amount of fission product absorption. Alternatively, the breeding ratio could be set to slightly positive to automatically compensate for the buildup of fission products. Helium sparging and mechanical filtering would be employed to separate gaseous and solid fission products. The process would continue until either the reactor vessel (or other major component) needed to be replaced, a fissile material solubility limit was reached, or the overall salt melt temperature had been so shifted by fission product dissolution that it exceeded 550°C. At this point, the fuel salt would be pumped out, poisoned, and allowed to solidify into mechanically robust rock salt. The salt containers would then be sent for disposal or centralized reprocessing. Alternatively, the salt could be allowed to solidify in the reactor vessel and the vessel as a whole transported for reprocessing and or disposal.
The post Fast Spectrum Molten Salt Reactor Options ORNL-TM-2011/105 appeared first on The Energy From Thorium Foundation.
merle [at] lpsc [dot] in2p3 [dot] fr – Professor at Grenoble INP/PHELMA and in the Reactor Physics
Group of Laboratoire de Physique Subatomique et de Cosmologie de Grenoble (CNRSIN2P3-
LPSC / Grenoble INP – PHELMA / UJF)
For the ‘MSFR Team’ of LPSC – M. ALLIBERT, M. BROVCHENKO,
V. GHETTA, D. HEUER, A. LAUREAU, E. MERLE-LUCOTTE, P. RUBIOLO
MSFR and the European project EVOL
12 European Partners: France (CNRS: Coordinateur, Grenoble INP , INOPRO,
Aubert&Duval), Pays-Bas (Université Techno. de Delft), Allemagne (ITU, KIT-G, HZDR), Italie
(Ecole polytechnique de Turin), Angleterre (Oxford), Hongrie (Univ Techno de Budapest)
+ 2 observers since 2012 : Politecnico di Milano et Paul Scherrer Institute
+ Coupled to the MARS (Minor Actinides Recycling in Molten Salt)
project of ROSATOM (2011-2013)
Partners: RIAR (Dimitrovgrad), KI (Moscow), VNIITF (Snezinsk), IHTE (Ekateriburg),
VNIKHT (Moscow) et MUCATEX (Moscow)
WP2: Design and Safety
WP3: Fuel Salt Chemistry and Reprocessing
WP4: Structural Materials
• Recommendations for the design of the core and fuel heat exchangers
• Definition of a safety approach dedicated to liquid-fuel reactors – Transposition of the
defence in depth principle – Development of dedicated tools for transient simulations of
molten salt reactors
• Determination of the salt composition – Determination of Pu solubility in LiF-ThF4 -
Control of salt potential by introducing Th metal
• Evaluation of the reprocessing efficiency (based on experimental data) – FFFER project
• Recommendations for the composition of structural materials around the core
The post PresentationThEC 2013: Introduction to the Physics of the Molten Salt Fast Reactor appeared first on The Energy From Thorium Foundation.
Doctoral Dissertation of: CARLO FIORINA
Matr. n. 754131
Background and motivation
Fast Reactors (FR) have been developed in the early stages of nuclear technology for the purpose of breeding fissile isotopes from fertile materials, thus greatly extending the fuel resources available to sustain the rapid growth of nuclear energy forecasted in the past. The U- Pu cycle has been preferred over the Th-U cycle because of the better breeding potential in a fast-spectrum and the sounder technical basis for fuel fabrication, irradiation, and reprocessing. On the other hand, Th use has been historically investigated for its capability to breed fissile material (namely, U-233) in a thermal neutron spectrum, thus possibly avoiding specific technological challenges associated to the development of FRs.
Over the course of the years, interest in fissile breeding has faded, especially in western countries, due to the slow deployment rate of new nuclear power plants and thanks to the availability of natural U resources, including the recent development of techniques for a reasonably economical extraction of U from sea water (ORNL, 2012). Conversely, waste management has emerged as one of the main problems for public acceptance of nuclear energy (Artioli et al., 2010; Salvatores and Palmiotti, 2011). Following these trends, both Th- based thermal reactors and U-based FRs started to be considered in view of their capability to operate with continuous recycle of all actinides, while potentially burning legacy TRU (TRansUranic isotope) wastes, thus drastically limiting the actinide wastes to be disposed.
Under this scenario, Th-based FRs can offer some specific advantages. The lower mass number of Th fosters a very low endogenous generation of TRUs, which could benefit public acceptance and the repository thermal performance. In addition, the low breeding capability of Th cycle may enhance the consumption of an external supply of TRUs. Past studies have also pointed out the Th potential capability to improve safety parameters (Till et al., 1980). Following these considerations, studies about Th use in FRs have started gaining momentum (Rubbia et al., 1995; IAEA, 2002; IAEA, 2005; Gruppelaar and Schapira, 2006). A new impetus to this option has been recently given by the cancellation of the Yucca Mountain nuclear waste repository project in the US, as well as by the Fukushima accident. The latter focused once again the attention of the public opinion on safety-related aspects of the nuclear energy production and on spent fuel accumulation at the reactor pools. For countries that have decided to phase out the nuclear energy option, management of the TRU legacy from Light Water Reactors has become a priority and Th may be the carrier to expedite TRU burning.
Despite the potential advantages, the implementation at an industrial scale of the Th closed cycle needs to overcome formidable challenges, including difficulties in dissolution and reprocessing of used fuel (Ramanujam, 2008), and fabrication of highly radioactive recycled fuel (Wenner et al., 2012). In particular, ThOx fuel is a particularly stable compound, which may benefit fuel disposal in once-through fuel cycle options, but necessitates dedicated dissolution processes for reprocessing. Similarly to PUREX, a nitric acid solution is used to
dissolve ThOx fuel, but addition of HF is required to reduce the dissolution time, leading to exacerbated corrosion of the equipment. As concerns fuel fabrication, problems originate from the build-up during Th-232 irradiation of U-232, mainly via (n,2n) reaction and subsequent neutron capture from Pa-231. The intensity and high energy of the gamma radiation emitted by Bi-212 (0.7-1.8 MeV) and Tl-208 (2.6 MeV), both daughters of U-232, imposes fuel handling and manufacturing behind thick shielding. In this sense, it is worth noting that the first and longest-lived isotope in the U-232’s progeny is the 1.9-years half-life Th-228. By sending Th to the waste stream during reprocessing, the fuel would be momentarily free from the U-232 progeny. Drawbacks would be the necessity of a quick fabrication after reprocessing, and the accumulation of intensely radioactive Th waste. In addition, during the TRU transmutation stage spontaneous neutron emitters (primarily Cm and Cf isotopes) would require remote fuel handling independent of the of the gamma field from U-232’s progeny.
The use of liquid fuel with online reprocessing would avoid most of the issues related to reprocessing, manufacturing and transporting highly radioactive recycled fuel in a closed cycle. The logical technology for the adoption of liquid fuel is the Molten Salt Reactor (MSR). In MSRs, a liquid fuel salt circulates through the core and transfers the heat to external heat exchangers via convection, thus playing the role of both fuel and coolant. The first MSRs developed used fluoride salts and Th fuel. They were conceived during the fifties for military purposes at the Oak Ridge National Laboratory (ORNL) in the US, and subsequently developed for two decades as graphite-moderated reactors for U-233 breeding and power production (MacPherson, 1985). In 2001, the Generation IV International Forum (GIF-IV) selected the MSR as one of the six innovative nuclear reactors with the potential to meet the compelling need for an increasingly sustainable, economical, safe and proliferation resistant nuclear energy production (GIF-IV, 2002). Few years after the selection of the MSR among the Generation-IV reactors, the concept evolved in the direction of fast-spectrum Th-based MSRs (Mathieu et al., 2006; 2009), identified as Molten Salt Fast Reactor (MSFR) (Merle- Lucotte et al., 2011) and mainly developed in the frame of the EURATOM EVOL (Evaluation and Viability Of Liquid fuel fast reactor system) Project (EVOL, 2012). The subject of the thesis work is the assessment of the novel MSFR technology as a promising route for combining the potential advantages of Th use in FRs with the unique fuel cycle flexibility fostered by a liquid fuel.
Objectives and outline of the work
Use of Th in fast-spectrum MSRs is a relatively recent proposal and only few studies are available in the open literature. Past works on fast-spectrum MSRs were mainly focused on chloride-based reactors for Pu breeding (Mourogov and Bokov, 2006), fertile-free TRU burning, or U-supported TRU burning (Ignatiev et al., 2007). The 30-year research experience developed at ORNL has a limited applicability to the MSFR case. Specifically, use of fluoride molten salts is still envisioned for the MSFR, which has allowed to rely, at least in a first development stage, on the ORNL experiences in terms of structural materials and reprocessing system. Limited studies have been instead carried out on the MSFR fuel cycle performances, waste management issues, thermal-hydraulics and safety aspects. Primary objective of the present thesis work is to offer an evaluation of the MSFR performances in
this sense, including an assessment of the MSFR potential to operate as a flexible conversion- ratio reactor (Chapter 1). A first-of-a-kind comparative analysis with traditional solid-fuelled FRs is performed (Chapters 2, 3, 4) in terms of main performance parameters, namely: breeding capabilities, decay heat and radiotoxicity of wastes, fuel management, TRU burning, safety aspects. For the comparison, two promising FR systems are selected (Appendix B), namely the sodium-cooled Toshiba-Westinghouse Advanced Recycling Reactor (ARR) (Dobson et al., 2008), and the European Lead SYstem (ELSY) (Alemberti et al., 2011). Both U- and Th-based versions of the two selected FRs are investigated in the thesis work. Although this is not a primary objective of the work, a comparison between Th and U use in traditional FRs is a necessary step for a better assessment of the MSFR performances vs traditional FRs, especially in view of the limited information available in literature in this sense (Till et al., 1980; Gruppelaar and Schapira, 2006; Touran et al., 2010). For both the MSFR and the traditional FRs, fuel cycle strategies are investigated envisioning the recycling of all the actinides in the core.
To ease the investigation while excluding major sources of biasing, a common tool is employed to evaluate the performances of the MSFR and of the traditional FRs (Appendix A). Specifically, an existing ERANOS-based (Rimpault et al., 2002) procedure, developed at the Paul Scherrer Institut (Switzerland) for the analysis of solid-fuelled FRs (Krepel et al., 2009), is employed and extended to allow the simulation of Th-containing cores, the possible use of fertile blankets and the online reprocessing system of the MSRs. In addition, dedicated sub- procedures are set up for the calculation of radiotoxicity and decay heat of wastes. The extended procedure is assessed against a tool recently developed at the Politecnico di Milano (Aufiero et al., submitted) and based on the Monte Carlo code SERPENT (Leppänen, 2007). The ORIGEN-S code (SCALE, 2006) is also used to assess decay heat and radiotoxicity calculations.
After characterizing the MSFR through a top-level comparison with the well-developed solid-fuelled counterparts, the work concentrates on two specific aspects of the MSFR that differentiate this technology from the others. The first one is the thermal-hydraulics (Chapter 5) that combines in MSRs a relatively high Prandtl number with the unique feature (in nuclear reactors) of the internal heat generation. This determines specific heat transfer characteristic and requires investigation to single out, or exclude, major impacts on the heat transfer phenomena in the reactor components. In particular, a theoretical investigation is carried out, leading to the derivation of a general form for correlations to be used to predict the Nusselt number for the forced convection of internally heated fluids flowing in turbulent regime in a straight circular channel. A generalized analytic approach previously developed at the Politecnico di Milano (Appendix C) is adopted to derive a specific correlation for the case of molten fluoride salts, and results are used to evaluate the impact of decay heat on the MSFR out-of-core components.
The second distinctive aspect of the MSFR technology is the reactor dynamics (Chapter 6). A liquid and circulating fuel impacts the reactor behavior due to 1) the direct deposition of fission heat inside the coolant, and 2) the movement of the delayed neutron precursors, causing their accumulation in low-flux regions and their partial decay out of the core. In addition, compared to graphite-moderated MSRs investigated in the past, the MSFR fuel is not restrained into graphite channels but flows freely in a wide core, with consequent flow
patterns that requires CFD (Computational Fluid Dynamics) codes for a proper characterization. These unique features of the MSFR exclude the use of tools developed in the past for solid-fuelled FRs, or for graphite-moderated MSRs. For this reason, a new dedicated model has been developed at the Politecnico di Milano in the frame of this and other 2 PhD theses. It consists of a set of non-linear and time-dependent coupled partial differential equations, which are simultaneously solved in the same simulation environment (namely, the simulation platform COMSOL Multiphysics) and describe the different “physics” (neutron transport, precursor diffusion and convection, thermo-fluid dynamics) occurring in the nuclear reactor. In the present thesis work, this model is presented and assessed against a similar model developed at the Technical University of Delft. The latter relies on a traditional coupling of dedicated neutron transport and thermo-fluid dynamic codes, in which the time- dependent solution is reached using the output from one code (e.g., the neutron kinetics code) as input to another code (e.g., the thermo-fluid dynamic code) at each time step. The results provided by the two codes are first used to investigate the steady-state core behavior. The effect of fuel movement on the precursor distribution is discussed and the core temperature field is investigated, thus extending to the reactor core the analysis of the MSFR heat transfer phenomena presented in Chapter 5. The MSFR transient behavior is then analyzed, pointing out general dynamic features and the most critical issues to be taken into account during core design and optimization.
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