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Development of closed nuclear fuel cycles in the United States
Development of closed nuclear fuel cycles in the United States
19
T.A. Todd Idaho National Laboratory, Idaho Falls, ID, USA
Acronyms An AMUSE ALSEP DAAP FCT GANEX GNEP HDEHP HEDTA HEH[EHP] Ln MPACT MT PUREX SANEX STMAS TALSPEAK TBP TRU TRUEX T2EHDGA UREX USDOE
19.1
actinide Argonne Model for Universal Solvent Extraction actinide-lanthanide separation process concept diamylamylphosphonate fuel cycle technologies program group actinide extraction concept Global Nuclear Energy Partnership bis(2-ethylhexyl)phosphoric acid, extractant for An(III) and Ln(III) in TALSPEAK process N-(2-hydroxyethyl)ethylenediamine-N,N¢ ,N¢ -triacetic acid, actinide-selective complexant used in Advanced TALSPEAK aqueous phase 2-Ethylhexylphosphonic acid mono-2-ethylhexyl ester, extractant for An(III) and Ln(III) in Advanced TALSPEAK and ALSEP processes lanthanide material protection, accountability, and control technologies metric ton plutonium uranium reduction extraction process selective actinide extraction concept Sigma Team for Minor Actinide Separations Trivalent actinide-lanthanide separations by phosphorus-reagent extraction from aqueous komplexes tri-n-butyl phosphate transuranium element (typically referring to Np, Pu, Am, Cm in used nuclear fuel) transuranic extraction process N,N,N¢ ,N¢ -tetra(2-ethylhexyl)diglycolamide, extractant for An(III) and Ln(III) in ALSEP uranium extraction process U.S. Department of Energy
Introduction
The United States has approximately 100 operating nuclear reactors, and the used fuel from each reactor is temporarily stored at each reactor site. On-site storage includes both wet and dry storage facilities, with the majority of the fuel (approximately Reprocessing and Recycling of Spent Nuclear Fuel. http://dx.doi.org/10.1016/B978-1-78242-212-9.00019-8 © 2015 Elsevier Ltd. All rights reserved.
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three-quarters of the fuel as of 2011) in wet storage; however, the percentage of fuel in dry storage is rapidly increasing. The current U.S. Department of Energy (USDOE) approach to the nuclear fuel cycle is to open a geologic repository around mid-century, and directly dispose of the used fuel inventory in the repository (http://www.energy. gov/downloads/strategy-management-and-disposal-used-nuclear-fuel-and-high-levelradioactive-waste, January 2013). This fuel amounts to nearly 70,000 metric tons of used nuclear fuel as of 2014. The fuel generated beyond approximately 2014 will be stored, either on-site or in consolidated storage facilities, until a future decision is made on a long-term approach to managing used nuclear fuel. To support this future decision, research is being performed on a number of fuel cycle options, including closed fuel cycles. It is generally recognized that partitioning and transmutation in fast spectrum reactors have significant potential benefit in terms of resource utilization and waste management. Fuel cycles that continually recycle U/Pu or U/TRU are currently the focus of research efforts in the United States. It should be noted that each nation or region will have their own set of constraints and drivers, so an approach that is ideal for one nation or region may or may not be for a different nation or region.
19.2
Future fuel cycle development requirements
The requirements for fuel cycle performance are dependent on a number of factors that vary depending on the objectives of the fuel cycle. For example, the amount of TRU losses that can be tolerated in waste streams depends on the ultimate waste form and the geology that the waste will be placed in. It is very important to look at the performance of the entire fuel cycle, not just a portion of the fuel cycle, to determine performance requirements. To focus on the TRU losses in separation processes, for example, without considering losses in fuel fabrication could lead to inaccurate analyses. The consideration of integrated processes is also important. One cannot determine requirements for the separations processes without first knowing the specifications for fuel fabrication (which drive separation purity requirements) and the limitations of waste forms and repository performance (which drive extent of recovery requirements). In the United States, the separations and waste form research activities are integrated into a single program, and while the fuels development program is separate, there are several activities to integrate development results between the two programs. A critical factor for fuel development is the amount of rare earth elements that the fuel can safely accommodate to avoid high neutron cross sections and to avoid fuel cladding chemical interaction, which could lead to early fuel failure. For metal transmutation fuels, the current estimate of the amount of rare earth elements in the fuel range from about 1% to 5%. Future testing of transmutation fuels is needed to further refine this number, as it could impact the choice of the separation processes used. A wide range of estimates exist for the needed recovery of TRU elements. Early in the U.S. Global Nuclear Energy Partnership (GNEP) program, recoveries of the TRU elements of greater than 99.9% were targeted (Wigeland et al., 2014). This value was
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largely arbitrary, and could be achieved by many of the separation processes tested at laboratory scale; however, TRU losses in the undissolved solids and in the fuel fabrication processes far exceed the 0.1% losses in the separation process. Therefore, losses in the separation process could reasonably be expected to be more, on the order of 0.5% to 1.0%, without having a significant impact on the overall fuel cycle performance. To date, a sensitivity analysis has not been performed to determine the effects of separation process losses on the overall fuel cycle performance. The U.S. Fuel Cycle Technologies (FCT) program will perform an initial sensitivity analysis in 2014, following the completion of the Nuclear Fuel Cycle Evaluation and Screening report (Wigeland et al., 2014). This study will evaluate separation recoveries in the range of 90%, 95%, 99%, and 99.9%, to determine at what recovery level fuel cycle performance is significantly altered. Another key development requirement is the need to assess and address technical maturity. Assessing technical maturity can be very useful in identifying areas that are less mature and need more development effort. This approach is helpful in identifying research needs and prioritizing research to address the less-mature aspects of a technology.
19.3
U.S. Fuel Cycle Technologies program
The U.S. Fuel Cycle Technologies program is managed under the USDOEs Office of Nuclear Energy. The various programs (or “campaigns”) are organized by function, as shown in Figure 19.1; advanced fuels development, separations and waste form development, fuel cycle options/systems analysis and integration, material protection,
NE-5 Deputy assistant secretary for fuel cycle technologies
NE-51 Office of systems engineering & integration
NE-52 Office of fuel cycle R&D
Fuel cycle options, systems analysis & integration campaign
Advanced fuels campaign
Separations and waste form campaign
MPACT campaign
NE-53 Office of used nuclear fuel disposition R&D
NE-54 Office of uranium management and policy
Used fuels disposition campaign
Figure 19.1 USDOE office of nuclear energy fuel cycle technologies program.
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accountability and control technologies, and used fuel disposition. The objectives, goals, and time lines for development of future fuel cycles is described in the USDOE Office of Nuclear Energy Research and Development Roadmap (http://energy.gov/ sites/prod/files/NuclearEnergy_Roadmap_Final.pdf, April 2010), which was issued in 2010. A revision of the roadmap is underway, but has not been published at the time this is written. The primary objectives laid out in the roadmap are (1) Develop Technologies and Other Solutions That Can Improve the Reliability, Sustain the Safety, and Extend the Life of Current Reactors; (2) Develop Improvements in the Affordability of New Reactors to Enable Nuclear Energy to Help Meet the Administration’s Energy Security and Climate Change Goals; (3) Develop Sustainable Nuclear Fuel Cycles; and (4) Understand and Minimize the Risks of Nuclear Proliferation and Terrorism. The development of reprocessing and recycling technologies falls under Objective 3: Develop Sustainable Nuclear Fuel Cycles. The roadmap lays out a strategy for developing new technologies to enable implementation in mid-century, if the decision is made to close the fuel cycle. Because of the very long time frames to develop, design, build, and startup large nuclear facilities, it is imperative that we are actively evaluating new technologies now, with testing and down selection in the near future to support this schedule. The general U.S. approach to advanced fuel cycles is to utilize aqueous processes to reprocess/recycle light water reactor oxide-based fuels. The reasons for this include the long historical experience and technical maturity, the high throughput of aqueous processing facilities (the United States currently generates about 2000 MT of used nuclear fuel per year), and the compatibility of the fuel matrix with aqueous processing. For a closed fuel cycle with fast reactors, the United States leans toward the use of metal fuel. It should be pointed out that no decision on future fuel cycle reactor types, fuels, or fuel cycles has been made by the USDOE, but these are the areas currently of most interest for research and development. The primary method for recycling fast reactor metal fuel in the United States is the electrochemical (or pyroprocessing) approach. The reasons supporting this approach include lower throughput and purity requirements needed for fast reactor recycle, the fact that the process begins with metal feed and produces metal product for recycling, and that the process has advantages in terms of radiation stability and criticality control. The primary research area for advanced aqueous processing in the U.S. program is the separation of minor actinides (neptunium, americium, curium) from the used fuel. Separation of uranium and plutonium are industrially established technologies, and recent trends to coprocess uranium with the plutonium product to reduce material attractiveness require only minor adjustments to the current industrial process. Historically, in the PUREX process, neptunium was not recovered and routed to the high-level waste fraction. This is straightforward, as neptunium primarily exists in the pentavalent state and will not extract in tri-n-butyl phosphate (TBP). There are a number of ways to manage neptunium in advanced closed fuel cycles, including extracting the neptunium with the uranium and plutonium or leaving the neptunium in a PUREX or UREX type process raffinate to be removed with the remaining TRU elements in a subsequent step. In the first approach, the key to removing the neptunium is to change its oxidation state to one that is extractable by TBP (either IV or
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VI) and ensure that the valance state can be maintained during the separation process (Gregson et al., 2012; Sarsfield et al., 2009). This is one of the preferred methods of managing neptunium in the U.S. approach, but the option to separate neptunium with the other minor actinides still exists. Due to similar chemistry, it is appealing to be able to remove the neptunium with the other tetra and hexavalent metals. The current U.S. program has not been actively investigating neptunium extraction; instead, it has chosen to model various approaches using the Argonne Model for Universal Solvent Extraction (AMUSE) code to assess the efficacy of various neptunium management approaches (Wigeland et al., 2014). Specific flow sheets to address neptunium separation with uranium and plutonium are now being developed. The separation of the trivalent minor actinides from the lanthanides has been the primary focus of the U.S. separations program now for the past five years. The Sigma Team for Minor Actinide Separations (STMAS), a multidiscipline, multilaboratory collaboration among U.S. national laboratories and universities, has been investigating multiple approaches for separating the trivalent actinides. This work is described in Chapter 11 of this book in detail, so only a brief overview will be provided in this chapter. The STMAS was conceived to formulate a program that encouraged collaboration and teamwork, while maintaining individual researcher interests and creativity. A number of leading scientists from U.S. national laboratories and one university were selected as the core team, and collaborations beyond this team with several other universities and research institutes soon followed. Until the formation of the STMAS, the U.S. approach to An(III) separations was the TRUEX process followed by the TALSPEAK process, which had been demonstrated at lab scale a number of times in the mid-2000s (Weaver and Kappelmann, 1968; Horwitz et al., 1885; Gelis et al., 2009). While demonstrated in short lab-scale tests with actual used nuclear fuel, the TALSPEAK process posed a number of chemistry and engineering challenges, as well as the cost and complexity of operating two separate solvent-extraction processes with different solvents. The goals of the STMAS were to develop simplified separation approaches, with more robust operational parameters, so that both the chemistry and engineering of the process could be well characterized and understood. The two primary focus areas investigated under the STMAS, are enhanced selectivity of An(III) over Ln(III) by selective complexation and the utilization of the higher valence states of americium (primarily V and VI) to selectively extract americium from Ln(III) and Cm(III). In the area of advanced complexation, new approaches such as the “Advanced TALSPEAK” process, which utilizes a weaker organic extractant 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester or HEH [EHP], along with a weaker aqueous soluble complexant N-(2-hydroxyethyl)ethylenediamine-N,N¢ ,N¢ -triacetic acid or HEDTA in a citrate buffer to selectively extract the lanthanides from the trivalent actinides (Braley et al., 2012). In the Advanced TALSPEAK process, the distribution coefficients for Ln(III) and An(III) are much flatter across the pH range of interest (2.8-3.5). In addition, the chemistry of the citrate buffer appears to be more predictable and less complex in comparison to the lactate buffer in the TALSPEAK process. Attempting to further simplify the separation of the trivalent actinides from lanthanides has led to the development of the actinidelanthanide separations (ALSEP) concept, which effectively combines the two step
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TRUEX-TALSPEAK functions into a single process (Gelis and Lumetta, 2014). The ALSEP process utilizes a solvent containing N,N,N¢ ,N¢ -tetra(2-ethylhexyl)diglycolamide (T2EHDGA) plus HEH[EHP] dissolved in n-dodecane. The ALSEP solvent extracts both trivalent actinides and lanthanides (except for lanthanum and possibly cerium), and then uses an aqueous soluble complexant, HEDTA, in a citrate buffer to selective strip the actinides from the lanthanides. The approach to exploit higher oxidation states of americium has been demonstrated using diamylamylphosphonate (DAAP) as the extractant in a solution where americium has been oxidized to Am(VI) using sodium bismuthate (Mincher et al., 2012). The challenge for this process is to effectively oxidize americium to the hexavalent state and be able to stabilize it in that valence state, long enough to perform the extraction process. A major focus of the program has been to search for alternative oxidation/stabilization agents for americium. Another area of great importance to the understanding of these solvent extraction approaches is the radiation chemistry of the solvents and the systems, particularly the stability of the solvent due to radiolysis and hydrolysis, as well as the effect of radiation and free radical formation on the valence states of multivalent cations (such as neptunium and possibly americium, if oxidized). Developing a fundamental scientific understanding of the chemistry of separations processes, including the radiation chemistry, thermodynamics, and kinetics of the systems, is an integral part of the U.S. Separations and Waste Form Campaign research effort. In the area of electrochemical processing of fuel, the primary effort in the U.S. Fuel Cycle Technologies program is part of a collaboration between the United States and the Republic of Korea to demonstrate at the kilogram scale the electroreduction of oxide fuel and the electrorefining of the resultant metal to recover uranium and U/TRU fractions. This research program is planning to process a large number of sequential kg-scale batches of fuel in an integrated process. This demonstration is currently scheduled for the 2016-2017 time frame. Additional research in the U.S. domestic electrochemistry program focuses on the recovery of U/TRU on nonalloying metal cathodes. As described in an earlier chapter, recovery of uranium from spent Experimental Breeder Reactor-II fuel has been ongoing at the Idaho National Laboratory for about two decades, processing over 4 tons of driver and blanket fuels.
19.4
Future trends
Only a decade ago, many approaches to the partitioning of trivalent actinides from lanthanides were still based on two-step processes with TALSPEAK-type chemistry, utilizing bis(2-ethylhexyl)phosphonic acid (HDEHP) as an extractant. There is a very common trend between U.S. and European research programs to simplify and combine processes to a single-process approach (after a process to remove uranium or uranium, plutonium (and possibly neptunium)). There are a number of similarities to the SANEX and GANEX approaches studied in Europe with the ALSEP process chemistry. Most notable is the trend that many approaches are converging toward
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the use of diglycolamide extractants and aqueous soluble complexants to selectively strip An(III) or americium. While these approaches were arrived at somewhat independently (with a healthy awareness of published literature), the collaboration between international research programs is increasing. The U.S. Separations and Waste Form program is planning to perform laboratory-scale flow sheet tests in the near future with simulant feeds and eventually used nuclear fuel, and these tests will most likely involve international support or participation. While the United States appears to be the only group actively investigating higher oxidation states of americium, there is a common interest in processes that separate americium without curium, to potentially simplify the transmutation fuel fabrication process. French and European research programs have made significant progress in americium extraction processes based on selective complexation. In the area of electrochemical processing, there is much common interest in the general approach, but less convergence on the specific technologies. The European Union programs are investigating solid alloying cathodes for the recovery of U/TRU. There are many similarities in approaches by the Republic of Korea and Japan with the U.S. approach, but each has unique aspects aimed at addressing issues of importance for each country. The U.S. Fuel Cycle Technologies program objectives are to develop technical options to support future decisions by the U.S. government on nuclear fuel cycles. Currently, the focus of the U.S. program is the near-term storage and disposal of existing light-water reactor fuel. The long-term research and development focus is the development of separation technologies, transmutation fuels, and advanced waste forms, for a closed fuel cycle. The overall time line, according to the current Nuclear Energy R&D Roadmap is to deploy advanced fuel cycles, if selected, by about 2050. In support of this long-term goal, the separations program is researching advanced separation technologies leading to laboratory-scale testing of these technologies with simulated feeds in the 2016-2017 time frame and with actual used fuel in the 2017-2020 time frame. Longer-term activities include scale-up testing at the engineering scale with simulated and eventually actual fuel, approximately in the mid-2020s time frame. This testing will provide data that will be used to evaluate the technical and economic viability of advanced closed fuel cycles in support of future nuclear energy policy decisions by the U.S. government. For this reason, a primary driver for the separations and waste form program is to develop technologies that have robust operational conditions and are as simple as possible to make them industrially useable and cost effective.
References Braley, J.C., Grimes, T.S., Nash, K.L., 2012. Ind. Eng. Chem. Res. 51, 629–638. Gelis, A.V., Lumetta, G.J., 2014. Ind. Eng. Chem. Res. 53 (4), 1624–1631. Gelis, A.V., Vandegrift, G.F., Bakel, A., Bowers, D.L., Hebden, A.S., Pereira, C., Regalbuto, M., 2009. Radiochem. Acta 97, 231–232. Gregson, C., Boxall, C., Carrott, M., Edwards, S., Sarsfield, M., Taylor, R., Woodhead, D., 2012. Procedia Chem. 7, 398–403. http://dx.doi.org/10.1016/j.proche.2012.10.062.
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Horwitz, E.P., Kalina, D.G., Diamond, H., Vandegrift, G.F., Schulz, W.W., 1885. Solvent Extr. Ion Exch. 3, 75–109. Mincher, B.J., Martin, L.R., Schmitt, N.C., 2012. Solvent Extr. Ion Exch. 30, 445–456. Sarsfield, M., Sims, H., Taylor, R., 2009. Solvent Extr. Ion Exch. 27 (5-6), 638–662. Weaver, B., Kappelmann, F.A., 1968. J. Inorg. Nucl. Chem. 30, 263–272. Wigeland, et al., 2014. Nuclear Fuel Cycle Evaluation and Screening Final Report, INL/EXT14-31465. Idaho National Laboratory.
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T.A. Todd Idaho National Laboratory, Idaho Falls, ID, USA
Acronyms An AMUSE ALSEP DAAP FCT GANEX GNEP HDEHP HEDTA HEH[EHP] Ln MPACT MT PUREX SANEX STMAS TALSPEAK TBP TRU TRUEX T2EHDGA UREX USDOE
19.1
actinide Argonne Model for Universal Solvent Extraction actinide-lanthanide separation process concept diamylamylphosphonate fuel cycle technologies program group actinide extraction concept Global Nuclear Energy Partnership bis(2-ethylhexyl)phosphoric acid, extractant for An(III) and Ln(III) in TALSPEAK process N-(2-hydroxyethyl)ethylenediamine-N,N¢ ,N¢ -triacetic acid, actinide-selective complexant used in Advanced TALSPEAK aqueous phase 2-Ethylhexylphosphonic acid mono-2-ethylhexyl ester, extractant for An(III) and Ln(III) in Advanced TALSPEAK and ALSEP processes lanthanide material protection, accountability, and control technologies metric ton plutonium uranium reduction extraction process selective actinide extraction concept Sigma Team for Minor Actinide Separations Trivalent actinide-lanthanide separations by phosphorus-reagent extraction from aqueous komplexes tri-n-butyl phosphate transuranium element (typically referring to Np, Pu, Am, Cm in used nuclear fuel) transuranic extraction process N,N,N¢ ,N¢ -tetra(2-ethylhexyl)diglycolamide, extractant for An(III) and Ln(III) in ALSEP uranium extraction process U.S. Department of Energy
Introduction
The United States has approximately 100 operating nuclear reactors, and the used fuel from each reactor is temporarily stored at each reactor site. On-site storage includes both wet and dry storage facilities, with the majority of the fuel (approximately Reprocessing and Recycling of Spent Nuclear Fuel. http://dx.doi.org/10.1016/B978-1-78242-212-9.00019-8 © 2015 Elsevier Ltd. All rights reserved.
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three-quarters of the fuel as of 2011) in wet storage; however, the percentage of fuel in dry storage is rapidly increasing. The current U.S. Department of Energy (USDOE) approach to the nuclear fuel cycle is to open a geologic repository around mid-century, and directly dispose of the used fuel inventory in the repository (http://www.energy. gov/downloads/strategy-management-and-disposal-used-nuclear-fuel-and-high-levelradioactive-waste, January 2013). This fuel amounts to nearly 70,000 metric tons of used nuclear fuel as of 2014. The fuel generated beyond approximately 2014 will be stored, either on-site or in consolidated storage facilities, until a future decision is made on a long-term approach to managing used nuclear fuel. To support this future decision, research is being performed on a number of fuel cycle options, including closed fuel cycles. It is generally recognized that partitioning and transmutation in fast spectrum reactors have significant potential benefit in terms of resource utilization and waste management. Fuel cycles that continually recycle U/Pu or U/TRU are currently the focus of research efforts in the United States. It should be noted that each nation or region will have their own set of constraints and drivers, so an approach that is ideal for one nation or region may or may not be for a different nation or region.
19.2
Future fuel cycle development requirements
The requirements for fuel cycle performance are dependent on a number of factors that vary depending on the objectives of the fuel cycle. For example, the amount of TRU losses that can be tolerated in waste streams depends on the ultimate waste form and the geology that the waste will be placed in. It is very important to look at the performance of the entire fuel cycle, not just a portion of the fuel cycle, to determine performance requirements. To focus on the TRU losses in separation processes, for example, without considering losses in fuel fabrication could lead to inaccurate analyses. The consideration of integrated processes is also important. One cannot determine requirements for the separations processes without first knowing the specifications for fuel fabrication (which drive separation purity requirements) and the limitations of waste forms and repository performance (which drive extent of recovery requirements). In the United States, the separations and waste form research activities are integrated into a single program, and while the fuels development program is separate, there are several activities to integrate development results between the two programs. A critical factor for fuel development is the amount of rare earth elements that the fuel can safely accommodate to avoid high neutron cross sections and to avoid fuel cladding chemical interaction, which could lead to early fuel failure. For metal transmutation fuels, the current estimate of the amount of rare earth elements in the fuel range from about 1% to 5%. Future testing of transmutation fuels is needed to further refine this number, as it could impact the choice of the separation processes used. A wide range of estimates exist for the needed recovery of TRU elements. Early in the U.S. Global Nuclear Energy Partnership (GNEP) program, recoveries of the TRU elements of greater than 99.9% were targeted (Wigeland et al., 2014). This value was
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largely arbitrary, and could be achieved by many of the separation processes tested at laboratory scale; however, TRU losses in the undissolved solids and in the fuel fabrication processes far exceed the 0.1% losses in the separation process. Therefore, losses in the separation process could reasonably be expected to be more, on the order of 0.5% to 1.0%, without having a significant impact on the overall fuel cycle performance. To date, a sensitivity analysis has not been performed to determine the effects of separation process losses on the overall fuel cycle performance. The U.S. Fuel Cycle Technologies (FCT) program will perform an initial sensitivity analysis in 2014, following the completion of the Nuclear Fuel Cycle Evaluation and Screening report (Wigeland et al., 2014). This study will evaluate separation recoveries in the range of 90%, 95%, 99%, and 99.9%, to determine at what recovery level fuel cycle performance is significantly altered. Another key development requirement is the need to assess and address technical maturity. Assessing technical maturity can be very useful in identifying areas that are less mature and need more development effort. This approach is helpful in identifying research needs and prioritizing research to address the less-mature aspects of a technology.
19.3
U.S. Fuel Cycle Technologies program
The U.S. Fuel Cycle Technologies program is managed under the USDOEs Office of Nuclear Energy. The various programs (or “campaigns”) are organized by function, as shown in Figure 19.1; advanced fuels development, separations and waste form development, fuel cycle options/systems analysis and integration, material protection,
NE-5 Deputy assistant secretary for fuel cycle technologies
NE-51 Office of systems engineering & integration
NE-52 Office of fuel cycle R&D
Fuel cycle options, systems analysis & integration campaign
Advanced fuels campaign
Separations and waste form campaign
MPACT campaign
NE-53 Office of used nuclear fuel disposition R&D
NE-54 Office of uranium management and policy
Used fuels disposition campaign
Figure 19.1 USDOE office of nuclear energy fuel cycle technologies program.
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accountability and control technologies, and used fuel disposition. The objectives, goals, and time lines for development of future fuel cycles is described in the USDOE Office of Nuclear Energy Research and Development Roadmap (http://energy.gov/ sites/prod/files/NuclearEnergy_Roadmap_Final.pdf, April 2010), which was issued in 2010. A revision of the roadmap is underway, but has not been published at the time this is written. The primary objectives laid out in the roadmap are (1) Develop Technologies and Other Solutions That Can Improve the Reliability, Sustain the Safety, and Extend the Life of Current Reactors; (2) Develop Improvements in the Affordability of New Reactors to Enable Nuclear Energy to Help Meet the Administration’s Energy Security and Climate Change Goals; (3) Develop Sustainable Nuclear Fuel Cycles; and (4) Understand and Minimize the Risks of Nuclear Proliferation and Terrorism. The development of reprocessing and recycling technologies falls under Objective 3: Develop Sustainable Nuclear Fuel Cycles. The roadmap lays out a strategy for developing new technologies to enable implementation in mid-century, if the decision is made to close the fuel cycle. Because of the very long time frames to develop, design, build, and startup large nuclear facilities, it is imperative that we are actively evaluating new technologies now, with testing and down selection in the near future to support this schedule. The general U.S. approach to advanced fuel cycles is to utilize aqueous processes to reprocess/recycle light water reactor oxide-based fuels. The reasons for this include the long historical experience and technical maturity, the high throughput of aqueous processing facilities (the United States currently generates about 2000 MT of used nuclear fuel per year), and the compatibility of the fuel matrix with aqueous processing. For a closed fuel cycle with fast reactors, the United States leans toward the use of metal fuel. It should be pointed out that no decision on future fuel cycle reactor types, fuels, or fuel cycles has been made by the USDOE, but these are the areas currently of most interest for research and development. The primary method for recycling fast reactor metal fuel in the United States is the electrochemical (or pyroprocessing) approach. The reasons supporting this approach include lower throughput and purity requirements needed for fast reactor recycle, the fact that the process begins with metal feed and produces metal product for recycling, and that the process has advantages in terms of radiation stability and criticality control. The primary research area for advanced aqueous processing in the U.S. program is the separation of minor actinides (neptunium, americium, curium) from the used fuel. Separation of uranium and plutonium are industrially established technologies, and recent trends to coprocess uranium with the plutonium product to reduce material attractiveness require only minor adjustments to the current industrial process. Historically, in the PUREX process, neptunium was not recovered and routed to the high-level waste fraction. This is straightforward, as neptunium primarily exists in the pentavalent state and will not extract in tri-n-butyl phosphate (TBP). There are a number of ways to manage neptunium in advanced closed fuel cycles, including extracting the neptunium with the uranium and plutonium or leaving the neptunium in a PUREX or UREX type process raffinate to be removed with the remaining TRU elements in a subsequent step. In the first approach, the key to removing the neptunium is to change its oxidation state to one that is extractable by TBP (either IV or
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VI) and ensure that the valance state can be maintained during the separation process (Gregson et al., 2012; Sarsfield et al., 2009). This is one of the preferred methods of managing neptunium in the U.S. approach, but the option to separate neptunium with the other minor actinides still exists. Due to similar chemistry, it is appealing to be able to remove the neptunium with the other tetra and hexavalent metals. The current U.S. program has not been actively investigating neptunium extraction; instead, it has chosen to model various approaches using the Argonne Model for Universal Solvent Extraction (AMUSE) code to assess the efficacy of various neptunium management approaches (Wigeland et al., 2014). Specific flow sheets to address neptunium separation with uranium and plutonium are now being developed. The separation of the trivalent minor actinides from the lanthanides has been the primary focus of the U.S. separations program now for the past five years. The Sigma Team for Minor Actinide Separations (STMAS), a multidiscipline, multilaboratory collaboration among U.S. national laboratories and universities, has been investigating multiple approaches for separating the trivalent actinides. This work is described in Chapter 11 of this book in detail, so only a brief overview will be provided in this chapter. The STMAS was conceived to formulate a program that encouraged collaboration and teamwork, while maintaining individual researcher interests and creativity. A number of leading scientists from U.S. national laboratories and one university were selected as the core team, and collaborations beyond this team with several other universities and research institutes soon followed. Until the formation of the STMAS, the U.S. approach to An(III) separations was the TRUEX process followed by the TALSPEAK process, which had been demonstrated at lab scale a number of times in the mid-2000s (Weaver and Kappelmann, 1968; Horwitz et al., 1885; Gelis et al., 2009). While demonstrated in short lab-scale tests with actual used nuclear fuel, the TALSPEAK process posed a number of chemistry and engineering challenges, as well as the cost and complexity of operating two separate solvent-extraction processes with different solvents. The goals of the STMAS were to develop simplified separation approaches, with more robust operational parameters, so that both the chemistry and engineering of the process could be well characterized and understood. The two primary focus areas investigated under the STMAS, are enhanced selectivity of An(III) over Ln(III) by selective complexation and the utilization of the higher valence states of americium (primarily V and VI) to selectively extract americium from Ln(III) and Cm(III). In the area of advanced complexation, new approaches such as the “Advanced TALSPEAK” process, which utilizes a weaker organic extractant 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester or HEH [EHP], along with a weaker aqueous soluble complexant N-(2-hydroxyethyl)ethylenediamine-N,N¢ ,N¢ -triacetic acid or HEDTA in a citrate buffer to selectively extract the lanthanides from the trivalent actinides (Braley et al., 2012). In the Advanced TALSPEAK process, the distribution coefficients for Ln(III) and An(III) are much flatter across the pH range of interest (2.8-3.5). In addition, the chemistry of the citrate buffer appears to be more predictable and less complex in comparison to the lactate buffer in the TALSPEAK process. Attempting to further simplify the separation of the trivalent actinides from lanthanides has led to the development of the actinidelanthanide separations (ALSEP) concept, which effectively combines the two step
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TRUEX-TALSPEAK functions into a single process (Gelis and Lumetta, 2014). The ALSEP process utilizes a solvent containing N,N,N¢ ,N¢ -tetra(2-ethylhexyl)diglycolamide (T2EHDGA) plus HEH[EHP] dissolved in n-dodecane. The ALSEP solvent extracts both trivalent actinides and lanthanides (except for lanthanum and possibly cerium), and then uses an aqueous soluble complexant, HEDTA, in a citrate buffer to selective strip the actinides from the lanthanides. The approach to exploit higher oxidation states of americium has been demonstrated using diamylamylphosphonate (DAAP) as the extractant in a solution where americium has been oxidized to Am(VI) using sodium bismuthate (Mincher et al., 2012). The challenge for this process is to effectively oxidize americium to the hexavalent state and be able to stabilize it in that valence state, long enough to perform the extraction process. A major focus of the program has been to search for alternative oxidation/stabilization agents for americium. Another area of great importance to the understanding of these solvent extraction approaches is the radiation chemistry of the solvents and the systems, particularly the stability of the solvent due to radiolysis and hydrolysis, as well as the effect of radiation and free radical formation on the valence states of multivalent cations (such as neptunium and possibly americium, if oxidized). Developing a fundamental scientific understanding of the chemistry of separations processes, including the radiation chemistry, thermodynamics, and kinetics of the systems, is an integral part of the U.S. Separations and Waste Form Campaign research effort. In the area of electrochemical processing of fuel, the primary effort in the U.S. Fuel Cycle Technologies program is part of a collaboration between the United States and the Republic of Korea to demonstrate at the kilogram scale the electroreduction of oxide fuel and the electrorefining of the resultant metal to recover uranium and U/TRU fractions. This research program is planning to process a large number of sequential kg-scale batches of fuel in an integrated process. This demonstration is currently scheduled for the 2016-2017 time frame. Additional research in the U.S. domestic electrochemistry program focuses on the recovery of U/TRU on nonalloying metal cathodes. As described in an earlier chapter, recovery of uranium from spent Experimental Breeder Reactor-II fuel has been ongoing at the Idaho National Laboratory for about two decades, processing over 4 tons of driver and blanket fuels.
19.4
Future trends
Only a decade ago, many approaches to the partitioning of trivalent actinides from lanthanides were still based on two-step processes with TALSPEAK-type chemistry, utilizing bis(2-ethylhexyl)phosphonic acid (HDEHP) as an extractant. There is a very common trend between U.S. and European research programs to simplify and combine processes to a single-process approach (after a process to remove uranium or uranium, plutonium (and possibly neptunium)). There are a number of similarities to the SANEX and GANEX approaches studied in Europe with the ALSEP process chemistry. Most notable is the trend that many approaches are converging toward
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the use of diglycolamide extractants and aqueous soluble complexants to selectively strip An(III) or americium. While these approaches were arrived at somewhat independently (with a healthy awareness of published literature), the collaboration between international research programs is increasing. The U.S. Separations and Waste Form program is planning to perform laboratory-scale flow sheet tests in the near future with simulant feeds and eventually used nuclear fuel, and these tests will most likely involve international support or participation. While the United States appears to be the only group actively investigating higher oxidation states of americium, there is a common interest in processes that separate americium without curium, to potentially simplify the transmutation fuel fabrication process. French and European research programs have made significant progress in americium extraction processes based on selective complexation. In the area of electrochemical processing, there is much common interest in the general approach, but less convergence on the specific technologies. The European Union programs are investigating solid alloying cathodes for the recovery of U/TRU. There are many similarities in approaches by the Republic of Korea and Japan with the U.S. approach, but each has unique aspects aimed at addressing issues of importance for each country. The U.S. Fuel Cycle Technologies program objectives are to develop technical options to support future decisions by the U.S. government on nuclear fuel cycles. Currently, the focus of the U.S. program is the near-term storage and disposal of existing light-water reactor fuel. The long-term research and development focus is the development of separation technologies, transmutation fuels, and advanced waste forms, for a closed fuel cycle. The overall time line, according to the current Nuclear Energy R&D Roadmap is to deploy advanced fuel cycles, if selected, by about 2050. In support of this long-term goal, the separations program is researching advanced separation technologies leading to laboratory-scale testing of these technologies with simulated feeds in the 2016-2017 time frame and with actual used fuel in the 2017-2020 time frame. Longer-term activities include scale-up testing at the engineering scale with simulated and eventually actual fuel, approximately in the mid-2020s time frame. This testing will provide data that will be used to evaluate the technical and economic viability of advanced closed fuel cycles in support of future nuclear energy policy decisions by the U.S. government. For this reason, a primary driver for the separations and waste form program is to develop technologies that have robust operational conditions and are as simple as possible to make them industrially useable and cost effective.
References Braley, J.C., Grimes, T.S., Nash, K.L., 2012. Ind. Eng. Chem. Res. 51, 629–638. Gelis, A.V., Lumetta, G.J., 2014. Ind. Eng. Chem. Res. 53 (4), 1624–1631. Gelis, A.V., Vandegrift, G.F., Bakel, A., Bowers, D.L., Hebden, A.S., Pereira, C., Regalbuto, M., 2009. Radiochem. Acta 97, 231–232. Gregson, C., Boxall, C., Carrott, M., Edwards, S., Sarsfield, M., Taylor, R., Woodhead, D., 2012. Procedia Chem. 7, 398–403. http://dx.doi.org/10.1016/j.proche.2012.10.062.
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Horwitz, E.P., Kalina, D.G., Diamond, H., Vandegrift, G.F., Schulz, W.W., 1885. Solvent Extr. Ion Exch. 3, 75–109. Mincher, B.J., Martin, L.R., Schmitt, N.C., 2012. Solvent Extr. Ion Exch. 30, 445–456. Sarsfield, M., Sims, H., Taylor, R., 2009. Solvent Extr. Ion Exch. 27 (5-6), 638–662. Weaver, B., Kappelmann, F.A., 1968. J. Inorg. Nucl. Chem. 30, 263–272. Wigeland, et al., 2014. Nuclear Fuel Cycle Evaluation and Screening Final Report, INL/EXT14-31465. Idaho National Laboratory.