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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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4

Fuel Cycle Development for Advanced Nuclear Reactors

This chapter responds to the first charge of the statement of task calling for an evaluation of the merits and an assessment of the viability of different nuclear fuel cycles, including fuel cycles that may use reprocessing, for both existing and advanced reactor technology options. It builds from the discussions in Chapters 2 and 3 on fuel cycle options for commercial light water reactors (LWRs) and research and development (R&D) programs and needs for advanced reactors, respectively. This chapter addresses the requirements in both the front and back ends of the fuel cycle to support the development and deployment of advanced nuclear reactors. To determine which factors have the greatest impact on the viability of advanced fuel cycles, the chapter examines the proposed merits of advanced fuel cycles and the challenges of achieving those potential benefits, including technical challenges as well as cost and safety aspects.

The chapter begins with a summary that includes the committee’s findings and recommendations (Section 4.1). It is then divided into two main sections for the front end (Section 4.2) and back end (Section 4.3) of the fuel cycle. The front-end section examines aspects of mining and milling (Section 4.2.1), conversion (Section 4.2.2), enrichment (Section 4.2.3), and fuel fabrication (Section 4.2.4) relevant to advanced reactor development. In particular, it focuses on the required infrastructure and capabilities for producing and fabricating high-assay low-enriched uranium (HALEU) fuel, which is needed for almost all of the advanced reactor designs examined (Section 4.2.3). Beginning with an introduction to advanced fuel cycles (Section 4.3.1), the back-end section discusses and analyzes reprocessing and recycling options for advanced reactors, under the assumption that plutonium and uranium are considered strategic assets and essential to sustain nuclear power production into the future.1 It introduces the concept of reprocessing (Section 4.3.2) and then focuses on the recovery and reuse of fissile material, in particular plutonium, from spent nuclear fuel to allow extraction of the maximum energy content possible, analyzing monorecycling of uranium and plutonium in LWRs (Section 4.3.3), multirecycling of plutonium in LWRs (Section 4.3.4), management of minor actinides (Section 4.3.5), and reprocessing methods for advanced reactor fuel (Section 4.3.6). Various motivations, potential benefits, and challenges of implementing advanced fuel cycles (Sections 4.3.8–4.3.11) are considered. Section 4.4 presents the committee’s assessment of what is needed for cost estimation for different nuclear fuel cycle options. Finally, Section 4.5 provides information relevant for safety considerations of fuel cycles.

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1 In contrast, Chapter 5 focuses on the direct disposal of spent nuclear fuel as waste.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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4.1 CHAPTER 4 SUMMARY, FINDINGS, AND RECOMMENDATIONS

A primary concern for the front end of the fuel cycle for advanced reactors is the production of HALEU fuel. The committee highlights here that the United States will likely not have any significant reliable domestic supply of HALEU for at least one decade—maybe even longer. A sufficient U.S. domestic commercial supply would alleviate the risk of supply reliability, given the great uncertainty over whether the current sole commercial HALEU supplier, Russia, will be allowed to provide HALEU to U.S.-based advanced reactor developers. In response to the Russian war against Ukraine, the U.S. government may impose sanctions against Russian nuclear companies; however, as of March 2022 (as the committee was completing its report), these sanctions had not occurred (see Box 4.1). Additionally, the United States has no commercial-scale facilities for producing other fuel types proposed for advanced reactors (e.g., TRistructural ISOtropic [TRISO] particle, metallic, nitride, mixed oxide, carbide, molten salt liquid, and thorium fuels), although plans for a TRISO fuel fabrication facility are under development. This analysis led the committee to the following finding and recommendation for the front end of the fuel cycle:

Finding 7: There is no current domestic capacity to supply high-assay low-enriched uranium (HALEU) to meet the projected needs of U.S.-based advanced reactor developers over the next decade. Therefore, if reactor projects requiring HALEU continue to advance, identifying a reliable supply of the material will be crucial. Otherwise, many developers will likely initially acquire HALEU from foreign sources, such as Russia, raising concern about ensuring reliable supply. Reliance on foreign sources of HALEU or HALEU feedstock (as many advanced reactor developers had planned to do prior to the invasion of Ukraine by Russia) without a reliable domestic supply could have serious energy and national security implications if advanced reactors using HALEU are adopted widely.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Recommendation C: Given the uncertainty of foreign supply arrangements of high-assay low-enriched uranium (HALEU) for advanced reactors, the U.S. Department of Energy should prepare contingency plans that may include (1) scheduled delays in the development, demonstration, and deployment of these systems; (2) a schedule for industry as to when and what level of federal support will be available; and (3) the release of stockpiles of highly enriched uranium for downblending until domestic and secure supplies are available.

On the back end, the advanced fuel cycles described herein represent scenarios where, in the limit, all plutonium and the minor actinides (neptunium, americium, and curium) are recycled and only fission products are left to be disposed of in a geologic repository. Although these advanced fuel cycles are theoretically possible, from a practical point of view, they are costly and challenging to implement as a whole in commercially licensed facilities. These fuel cycles require the construction of complex reprocessing and fuel fabrication facilities, whose reliable, safe, and efficient operation in most cases is well beyond the existing commercial-scale experience base in the United States. While reprocessing options are currently not economically viable in the United States, future events, including decreased availability and increased cost of uranium or major cost escalations of other sources of energy, might nullify these current cost penalties. From that perspective, maintaining modest R&D support for reprocessing technologies would be valuable.

Implementing advanced fast reactors and their associated fuel cycles in order to effectively reduce, but not eliminate (because of inevitable process loss), the quantity of long-lived actinides destined for geologic disposal would have to be operated for many decades to achieve the permanent benefits to the repository and other parts of the nuclear fuel cycle. This necessitates a commitment to nuclear power for several centuries (National Research Council, 1996) and significant financial investment. Implementing a fully closed fuel cycle that includes reprocessing does not eliminate the need for a geologic repository because fission products will still require disposal, material losses will inevitably occur in reprocessing and in other parts of the fuel cycle, and contaminated fuel and process hardware will be generated; therefore, no fuel cycle can be considered as perfectly closed.

Furthermore, for a nuclear fuel cycle supporting any reactor technology to be viable, it has to be industrially sustainable. Although different options are available, most would be dramatically different from the current U.S. situation. An industrially sustainable nuclear fuel cycle would require all its elements to be demonstrated individually and together. For that reason, an evolutionary pathway is more practical than a revolutionary approach that attempts to solve all potential issues at the same time.

The committee makes the following findings and recommendations regarding fuel cycles:

Finding 8: For a nuclear fuel cycle supporting any reactor technology to be viable, it has to be industrially sustainable. Although many fuel cycle options are possible, most differ dramatically from the current situation in the United States—the once-through fuel cycle. All elements of a sustainable nuclear fuel cycle would have to be fully demonstrated both individually and together, because what works with computer-aided designs would not necessarily translate to industrial-scale deployment. For that reason, an evolutionary, progressive approach is likely more practical than a revolutionary approach that attempts to solve all potential issues at the same time with advanced technologies. The evolutionary approach is more important for commercial deployability and will require the majority of investment efforts; nonetheless, some investments in high-risk, high-reward approaches may be worth pursuing. The committee agrees with the 1996 National Research Council report Nuclear Wastes: Technologies for Separations and Transmutation, which states that advanced fuel cycles will require substantial investment and take many decades to more than a century of continuous recycling using a separations and transmutation system of appropriate scale, in order to potentially achieve the full benefit of plutonium recycling and partitioning and transmutation of minor actinides.

Recommendation D: The current U.S. policy of using a once-through fuel cycle with the direct disposal of commercial spent nuclear fuel into a repository should continue for the foreseeable future. The once-through fuel cycle is the baseline, and any new fuel cycles should have advantages over that baseline for them to be deployed. However, so as not to preclude these options in the future, the U.S.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Department of Energy (DOE) should continue fundamental studies to evaluate the feasibility of using recycling and transmutation for closing fuel cycles. Specifically, DOE should develop and implement a phased, long-range research and development program that focuses on advanced separations and transmutations technologies.

Finding 9: As proposed for some advanced reactor closed fuel cycles, reprocessing and recycling of spent nuclear fuel introduces additional safety and environmental considerations over the management of open-cycle light water reactor oxide fuels. In assessing the safety and environmental performance of advanced reactors, the risks and environmental impacts will require optimization over the entire fuel cycle, including front-end processes (mining, enrichment, and fabrication), back-end processes (reprocessing and recycling together), and disposal (interim and final). Currently, advanced reactor developers focus primarily on the safety aspects of the reactor and its operation, and put less priority on the safety aspects of other parts of the fuel cycles.

Recommendation E: Congress and the U.S. Department of Energy should incentivize safety improvements across the supporting fuel cycle.

Finding 10: Because of the absence of current commercial operational experience with advanced reactor technologies in the United States, reliable cost data and estimates for these technologies and their associated fuel cycle components are lacking. The costs of advanced reactors and their associated fuel cycles could range from at least several billion dollars—for pilot-scale non–light water advanced reactors and their fuel cycle facilities—to hundreds of billions of dollars—for full deployment of an alternative fuel cycle that would replace the existing once-through cycle and existing light water reactors. Congress and the U.S. Department of Energy will need better understanding of the cost estimates for various scenarios of reactor deployment and supporting fuel cycle requirements to aid their decision making as to what technologies to support in the coming years.

Recommendation F: Congress and the U.S. Department of Energy should obtain an independent assessment of cost estimates of various scenarios for potential deployment of advanced reactor technologies and related fuel cycle components. The independent assessor should have expertise in evaluating large-scale construction projects; examining project management challenges; and understanding technological and financial risks, as well as their uncertainties.

4.2 FRONT END OF THE FUEL CYCLE TO SUPPORT FUEL PRODUCTION FOR ADVANCED NUCLEAR REACTORS

The objective of the front end of the fuel cycle is to make fuel for reactors. Chapter 2 covers the steps of the front end relevant for the once-through and monorecycling fuel cycles for LWRs. This section focuses on aspects of the front end relevant for non-LWR advanced reactors, as defined in Chapter 3. One major aspect is the use of HALEU, which is a requirement for fueling most of the advanced reactor designs under development (see Table 3.1).

In this report, HALEU is defined as enrichment of uranium-235 greater than or equal to 10 percent but less than 20 percent. The commercial LWR fleet currently uses low-enriched uranium (LEU) fuel with enrichments between 3 and 5 percent uranium-235. Enrichments from 5 percent to just under 10 percent are considered LEU+. While the interdependent, international market for the front-end services has worked effectively to supply U.S. LWR fuel needs, developers have expressed concern to the U.S. Department of Energy (DOE) about whether the same will be true for advanced nuclear reactor fueling requirements—particularly whether HALEU will be available (Caponiti, 2020a).

Advanced reactors use HALEU fuels for different reasons. Some small and microreactor designers plan to use HALEU because the higher enrichment is needed to counteract neutron leakage from the smaller reactor cores. Also, fast reactors cannot achieve criticality with lower-enriched fuel. In contrast, thermal reactors such as high-temperature gas-cooled reactors (HTGRs) do not require HALEU, but some designers choose to use HALEU to allow the fuel to reach higher burnups, which can improve fuel utilization, or to reduce the frequency of refueling.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

However, use of HALEU fuel generally reduces natural uranium utilization efficiency because of the larger amount of mined uranium needed to generate the HALEU, unless higher burnups are achieved.

The following subsections provide information on U.S. and international capacities for the specific steps of the front end: mining and milling, conversion, enrichment, and fuel fabrication, with particular emphasis on addressing the challenges of supplying HALEU.

4.2.1 Mining and Milling

As introduced in Chapter 2, the mining and milling of uranium provide the raw uranium material required for nuclear fuel. Concerns about increasing U.S. dependency on uranium imports from foreign state–owned enterprises (e.g., those in Russia, China, and Kazakhstan), led two U.S. domestic uranium mining companies to petition the U.S. Department of Commerce (DOC) in January 2018 to investigate the effects of state-subsidized foreign uranium supplies on the domestic mining industry and the potential impacts of these imports on national security (Energy Fuels, 2018). In May 2019, DOC reported to President Donald J. Trump that these imports posed a threat to national security and recommended action under Section 232 of the Trade Expansion Act of 1962 (19 U.S.C. § 1862), which gives the president the authority to restrict certain imports if DOC determines that these “impair national security” (Larson, 2020).

While President Trump did not concur with DOC’s findings, the administration expressed significant concerns and responded by creating the Nuclear Fuel Working Group (NFWG), which was assigned to examine options for reviving domestic nuclear fuel production, as well as the entire nuclear supply chain. In early 2020, the Trump administration proposed $150 million in its fiscal year (FY) 2021 budget to Congress to build a strategic uranium reserve that would (1) provide a supply of domestic uranium to commercial nuclear power production in the event of an international market disruption and (2) ensure sufficient domestic uranium capacity for U.S. defense needs. However, according to the Congressional Research Service, DOE did “not identify what would constitute a market disruption, criteria for providing uranium material to commercial utilities, or the price at which the uranium from the stockpile would be sold” (Larson, 2020). Using the then-current uranium spot price of $24.90 per pound milled uranium oxide (U3O8), the requested $150 million could purchase about 2.7 million kilograms of U3O8 from domestic suppliers (Larson, 2020), which would only meet about 13 percent of the annual demand from U.S. nuclear power plants. At the time, $75 million was provided (WNN, 2020a).

Protections sought by U.S. uranium mining companies and Urenco, the only active commercial uranium enrichment company in the United States, were realized with the October 2020 agreement between DOC and Rosatom, the Russian state nuclear energy corporation, to extend the suspension of the U.S. antidumping investigation until 2040 (NEI, 2020a). This agreement allows Russia to continue exporting enriched uranium to the United States, but it reduces the proportions from approximately 20 percent of U.S. annual demand to no higher than 15 percent from 2028 to 2040. The amendment also limits the natural uranium and uranium conversion services from Russia to an amount equivalent to no more than 5 percent of U.S. annual enrichment demand from 2026 to 2040. This agreement is relevant for meeting the fuel requirements for advanced reactors because it provides “predictable rules” and would not significantly impact Russia’s ability to supply fuel services of the market for existing U.S. reactors and the potential market for U.S. advanced reactors, according to Rosatom after the agreement was concluded (WNN, 2020b). However, because of the Russian war against Ukraine, the U.S. government may impose sanctions on imports of Russian uranium and nuclear fuel services. For additional information on what was known as of March 2022 as the committee was completing its report, see Section 4.2.3.3 and Box 4.1.

4.2.2 Conversion

As described in Chapter 2, milled uranium oxide (U3O8) is converted to uranium hexafluoride (UF6) prior to enrichment. The October 2020 U.S.–Russian amended agreement on antidumping of Russian uranium also affected the business decision of Honeywell-ConverDyn, the sole U.S. uranium conversion company, about restarting its conversion plant. In 2018, the company had stopped production at its Honeywell Metropolis Works (MTW) conversion plant in Illinois (capacity of 7,000 MT [metric tons] uranium per year) because of low demand and high UF6 inventories

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

globally. In early 2021, Honeywell announced plans for the MTW to restart in 2023 after completing plant upgrades. Honeywell chief executive officer (CEO) Malcolm Critchley stated in a press interview that before spending the $150 million required for the plant’s upgrades, he and his team needed to assess the business case, which involved several pieces. One was assurance that the Russian conversion capacity would be kept at a relatively low level for supplying the U.S. reactor fleet. Second was the continuation of the NFWG under the Biden administration, indicating bipartisan support for reviving domestic nuclear fuel production. Third, in 2020, the company’s customers expressed strong interest in long-term contracts, thereby helping to lower the financial risk of restart. Finally, in 2020, the U.S. Nuclear Regulatory Commission (U.S. NRC) approved the license extension of MTW for 40 years—in the past, the license extensions had been valid for only 10 years. This license approval further reduced financial uncertainty around reopening the MTW (WNN, 2021a). Having a U.S. uranium conversion company in long-term operation is an important step toward a domestic nuclear fuel supply chain. However, its benefit to U.S. advanced reactor developers would be limited unless domestic HALEU enrichment capacity can meet demand, a challenge discussed in the next section.

4.2.3 Enrichment, Production, and Supply of HALEU

Establishing commercial-scale production and supply of HALEU will be an essential enabling factor in the deployment of advanced reactors, since, as noted above, most advanced reactors under development require HALEU for their fuel. The U.S. government has ongoing efforts to develop HALEU production capability through enrichment of natural uranium, downblending of highly enriched uranium excess to nuclear weapons, and reprocessing and downblending of highly enriched uranium spent nuclear fuel from research and test reactors. Production of HALEU via enrichment of natural uranium is also being explored by private industry.

Congress recognized the potential need for HALEU in the Energy Act of 2020, which authorized DOE to establish a HALEU Availability Program to “support the availability of HALEU for civilian domestic research, development, demonstration, and commercial use” (DOE-NE, 2021e). DOE released a request for information in the form of an extensive questionnaire regarding this program in December 2021, with an extended due date of February 14, 2022, for receipt of information (DOE-NE, 2021e). Despite these efforts, significant timing, cost, and regulatory challenges remain for building a robust domestic HALEU supply chain. This section examines the status and challenges of HALEU production and supply, provides discussion about ways to address production and supply, and concludes with information on safety and security requirements for HALEU.

4.2.3.1 Domestic HALEU Production Capabilities

The sole operating commercial enrichment facility in the United States, the Urenco plant in New Mexico, is currently not ready to produce HALEU. Urenco, a foreign-owned company, claims that its New Mexico plant has scope for expansion to accommodate a facility meeting Category II security requirements (see Sidebar 4.1) to produce HALEU and that “if detailed design, site permits, and contractor selection were undertaken in parallel with the regulatory licensing process,” it could have a HALEU production unit ready within 24 months of regulatory licensing approval (Fletcher, 2020).

DOE is working to stimulate the development of a domestic U.S.-based HALEU production capability, most notably through a 3-year, $170 million cost-shared demonstration project with Centrus Energy Corporation to show the viability of advanced centrifuges for making HALEU. As of 2021, Centrus had built 16 of these centrifuges with the aim of producing up to 600 kg of HALEU by June 2022. In June 2021, the U.S. NRC approved Centrus’s request to amend its license to produce HALEU up to 20 percent enrichment (DOE-NE, 2021f). In November 2021, Centrus announced that supply chain difficulties due to the COVID-19 pandemic have affected the production schedule and stated that production would not begin until mid-2022 (Patel, 2021). Beyond the June 1, 2022, end date for Centrus’ current contract with DOE, the company could receive additional funding through DOE’s proposed HALEU Availability Program, to be started in FY2022, although this would require them obtaining a new, competitively awarded contract (Patel, 2021). However, several members of the House of Representatives have expressed concern that the competitive bidding process will not be truly competitive (Kramer, 2022). In particular, the presolicitation announcement on February 7, 2022, by DOE’s Office of Nuclear Energy (DOE-NE) limits the

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

bids to those who can produce HALEU at the 16-centrifuge facility and specifically the “place of performance will be at the DOE-owned American Centrifuge Plant in Piketon, Ohio” (DOE, 2022d).

An alternative HALEU production method is downblending, the process by which higher enrichment material is mixed with lower-enrichment material to obtain the desired specified level of enrichment. The United States has declared 374 MT of highly enriched uranium as excess to nuclear weapons; 152 MT have been set aside to fuel U.S. naval reactors on submarines and aircraft carriers; and 28 MT were made available to have U.S.-origin material to fuel the Watts Bar Nuclear Power Plant, which has the mission of producing tritium, a hydrogen radioisotope used in nuclear weapons. In addition, the United States has been using the remaining part of the declared excess highly enriched uranium to meet the needs of fueling research reactors and isotope-production reactors (DOE, 2015b). Because the enrichment levels of the highly enriched uranium stocks are classified, the committee was not able to calculate how much HALEU could be produced from downblending available highly enriched uranium stocks.

Currently, two facilities have conducted downblending activities: BWXT’s Nuclear Fuel Services (NFS) facility in Erwin, Tennessee, a designated Category I facility licensed by the U.S. NRC to handle highly enriched uranium, and the DOE’s Y-12 facility at Oak Ridge, Tennessee. The NFS is the only U.S. commercial company licensed and capable of downblending MT quantities of highly enriched uranium, but as of April 2020, no funding was available for this downblending, although the company has been able to show production of demonstration quantities of HALEU (e.g., for TRISO fuel) (Nagley, 2020). BWXT claims that, with adequate funding, NFS would be prepared to add equipment to its existing capacity and could downblend 1–2 MT of highly enriched uranium annually to produce up to 10 MT of HALEU. In addition, BWXT has demonstrated deconversion capabilities to turn the downblended uranium compounds into oxides, metals, and various other compounds including nitrides and silicides. Deconversion on-site would avoid shipping HALEU as UF6 (Nagley, 2020). In a January 2021 briefing to the committee, BWXT mentioned that the company has developed a near-term strategy to use U.S.sourced natural uranium feedstock for producing about 3 MT/year of unobligated HALEU that would be free of peaceful-use obligations (which is not a requirement for most power-related uses of HALEU). BWXT estimates it will take 5–6 years to construct, commission, and begin production at the proposed new facility (Nygaard, 2021).

Another method for having a near-term U.S. capacity for HALEU is the use of reprocessed uranium, though there are some significant technical challenges with this option. Reprocessed uranium can come from research and test reactors, as well as isotope-production and naval reactors fueled with highly enriched uranium. For example, the Savannah River Site has capability for reprocessing at its H-Canyon facility and could meet some needs for HALEU (Bates, 2020). From 2003 to 2011, the Savannah River Site produced 301 MT of 4.95 percent–enriched uranium for commercial reactors at the Tennessee Valley Authority. The recycling operations have been performed on irradiated fuels from research, test, and isotope production reactors. The Savannah River Site is storing highly enriched uranium solution that could be converted into about 2 MT of 19.75 percent HALEU by late FY2022, and is awaiting a decision about what to do with that material. In addition, the site could produce even greater amounts of HALEU: in particular, up to 19 MT from reprocessing DOE’s aluminum-based spent nuclear fuel, which is stored at the site, and up to 13 MT from reprocessing Idaho National Laboratory’s spent nuclear fuel. Production of 1 to 1.5 MT per year could start in FY2023 (Bates, 2020).

However, the Defense Nuclear Facilities Safety Board has documented safety issues at H-Canyon that would have to be addressed to allow for extending operations at this facility (DNFSB, 2021). In April 2022, DOE decided to not pursue recovery of highly enriched uranium from 29.2 tonnes of heavy metal of research-reactor spent fuel currently in storage at the Savannah River Site for downblending to HALEU, but it instead will dissolve the spent fuel (containing highly enriched uranium) and immobilize it into borosilicate glass for storage until a repository is available (DOE, 2022e).

Seeking additional options for HALEU fuel production, Idaho National Laboratory (INL) undertook a study in 2019 to evaluate the technical and cost requirements for establishing a metallic or ceramic/intermetallic-type (e.g., pellet) HALEU fuel fabrication line in each of three existing buildings on-site, using as feedstock sodium-bonded metallic highly-enriched uranium fuel that was irradiated in Experimental Breeder Reactor-II (EBR-II) and electrometallurgically treated (Crawford et al., 2019). INL’s planning assumed a 2.5-MT throughput,2 and because the

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2 The actual production rate for a typical commercial fuel fabrication facility is 1–2 orders of magnitude higher than INL’s planning assumption.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

starting material was known to contain residual contamination from both transuranic elements and fission products, use of inert gas–filled glove box enclosures was required. The rough order-of-magnitude cost estimate for building modifications started at $10 million and increased commensurately with other activities depending on the specific building, such as relocation of existing programs, decontamination, and decommissioning. The cost estimate for fabrication equipment and glove boxes totaled $22.5 million for the metallic fuel line and $28 million for the ceramic/intermetallic fuel fabrication line, with uncertainties of −20 to +50 percent. The annual operating cost for a fuel fabrication line was estimated to be around $7 million in 2019 dollars. Also important to note, industry feedback to DOE-NE indicated that HALEU derived from EBR-II spent fuel could not be used for TRISO fabrication because “(1) The residual impurities and radioactive contamination impose the requirement for excessive transportation, handling, shielding, and protective equipment that will significantly add to the fuel fabricator’s manufacturing costs. (2) The EBR-II HALEU, produced at the INL, cannot leave the site, so fuel fabrication would have to be performed there. There are no current plans for TRISO based fuel manufacturing at INL” (Caponiti, 2020a).

To satisfy defense-related needs for enriched uranium, in 2017, DOE’s National Nuclear Security Administration (NNSA) initiated efforts to establish a domestic uranium-enrichment capability to address the need for a reliable supply of unobligated LEU at 19.75 percent enrichment. Three specific needs were called out: (1) HALEU for research, test, and demonstration reactors by 2030; (2) low-enrichment uranium reactor fuel for tritium production by 2038; and (3) highly enriched uranium for naval reactor fuel by 2060. A 2018 assessment of NNSA’s plan by the U.S. Government Accountability Office (GAO) recommended

  1. “The NNSA Administrator should revise the scope of the mission need statement to clarify which mission need it seeks to achieve and, as appropriate, adjust the range of options considered in the analysis of alternatives process.
  2. The NNSA Administrator should—following clarification of the scope of the mission need statement—ensure that the agency’s cost estimates for whichever options it considers going forward are aligned with the scope of the mission need that the enrichment capability is intended to fulfill and that they are developed consistent with best practices” (GAO, 2018).

However, NNSA disagreed with these recommendations.

4.2.3.2 Challenges and Opportunities for Establishing Domestic HALEU Supply

Major steps that would be needed to produce HALEU fuel for advanced reactors were outlined at INL’s Gateway for Accelerated Innovation in Nuclear workshop on HALEU in April 2020:

  • Develop, on concurrent schedules, the enrichment, deconversion, and fabrication facilities.
  • Make sure the regulatory resources are available to support the licensing framework for the HALEU fuel cycle.
  • Match the timing of the front-end fuel cycle development for next-generation fuels with the forecasted aggregate demand from the advanced reactor vendors.
  • Provide sufficient assurance of a reasonable and necessary economic return for companies investing in HALEU facilities (Fletcher, 2020).

Industry feedback to DOE-NE, shared at the same INL workshop, reflected these challenges and proposed some potential solutions. In particular, multiple industry representatives urged DOE to support the entire infrastructure for fuel production for advanced reactors, including enrichment, fabrication (of various fuel types), transportation, and licensing (Caponiti, 2020a). Industry representatives also stressed the importance of HALEU being commercially available at an economically competitive price in a timely fashion (within 3–4 years). The industry’s recommended actions for DOE included guaranteeing an interim supply of HALEU for demonstration reactors, assessing future procurement of HALEU for a fuel bank, establishing fuel fabrication facilities to handle HALEU and accommodate multiple fuel types, and licensing a HALEU shipping container (Caponiti, 2020b).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

During a briefing to the committee in June 2021, Daniel Poneman, CEO of Centrus, indicated that the primary dilemma for HALEU supply is somewhat of a “chicken and egg” problem: who will buy U.S. advanced reactors if the United States lacks a guaranteed fuel supply—but who will invest in the fuel supply infrastructure without a guaranteed customer base? He put forward the model of leveraging U.S. government demand for HALEU in order to build domestic capacity to meet national security needs, as well as needs for advanced nuclear reactor development, noting that “DOE should lead the way” (Poneman and Cutlip, 2021). Another challenge, however, comes from the different economies of scale for low-enriched uranium versus HALEU: there is a known and reliable demand of about 15 million SWU/year (separative work units per year) for low-enriched uranium for the LWR fleet, but there is an uncertain future demand for HALEU, and the government’s near-term need is less than 100,000 SWU/year. In addition, as a Category II nuclear material (see Sidebar 4.1), HALEU has additional security costs. To help reduce this cost and risk, Centrus plans to collocate enrichment and deconversion on one site (Poneman and Cutlip, 2021). Similarly, Urenco plans to collocate enrichment and deconversion for security reasons (Fletcher, 2020).

Recognizing the challenges and concerns regarding HALEU availability, DOE-NE plans to continue its coordination with other agencies and partnerships with industry to make progress on HALEU production in FY2022 (Griffith, 2021). In particular, DOE-NE will work with NNSA on the recovery and downblending of limited quantities of highly enriched uranium to HALEU. It will also “develop criticality benchmark data to support the design and licensing of transportation packages” and “acquire transportation packages for Department of Energy-owned HALEU” (Griffith, 2021). Furthermore, DOE will continue support for the HALEU-enrichment demonstration facility, sharing costs with industry, and will provide additional support to this demonstration contractor “for impacts related to COVID, such as supplier issues and contractual impacts” (Griffith, 2021). DOE-NE will also work “with industry to understand and help enable commercialization of long-term private-sector HALEU production” (Griffith, 2021); however, it does not intend to be the sole funder of HALEU production and thus requires private-industry investment. Finally, DOE-NE plans to “initiate National Environmental Policy Act activities supporting HALEU availability” (Griffith, 2021). As this report was being finalized, DOE-NE was receiving responses to its request for information in the Federal Register (DOE-NE, 2021g).

4.2.3.3 Foreign HALEU Supply

The one currently available foreign supplier of HALEU is Russia’s Rosatom and its company TENEX, which supplies uranium products from the Russian Nuclear Fuel Complex (NFC). This complex includes the Novosibirsk Chemical Concentrates Plant (fuel and targets fabrication), the Siberian Group of Chemical Enterprises (conversion services), and the Production Association Electrochemical Plant (enrichment and deconversion services). The Russian NFC has many years of experience producing relatively small quantities of HALEU for research reactors and has certified shipping containers for this material. It would need to have higher-capacity certified containers for commercial-reactor quantities of HALEU, but TENEX officials believe that the experience with the existing TUK-159 containers (35 kg uranium oxide or up to 50 kg uranium metal) could form the basis for developing higher-capacity containers. At INL’s April 2020 HALEU Workshop, these officials stated that the Russian NFC could produce HALEU in metal or oxide forms within 6–9 months after receipt of an order (Newton and Kolosovskaya, 2020). See Box 4.1 for the political and legal situation as of March 2022 regarding potential U.S. sanctions on importation of Russian nuclear fuel services.

Potentially, Orano could also produce and supply HALEU, though its enrichment facilities in France are not currently configured for this purpose. During the committee’s September 2021 information-gathering meeting, Amir Vexler, CEO of Orano USA, acknowledged that Russia is presently the only commercial supplier of HALEU and said that “Orano is committed to fueling the future of nuclear energy through the development of commercial platforms that secure HALEU production capacity and the associated logistics infrastructure.” But the diversity of reactor operations’ enrichment and deconversion requirements will need “a flexible platform for HALEU production,” leveraging “integrated advanced chemistry applications backed by Orano’s expertise and delivered through industrial partnership.” Orano is also closely monitoring demand signals from the United States and needs “greater clarity” about commercial commitments and “the availability of policy tools that can accelerate investment in this critical infrastructure” (Vexler, 2021).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

If advanced non-LWR developers deploy demonstration reactors in the next decade, they will need amounts of HALEU that current U.S. sources might not be able to supply. As stated above, currently the only source of commercial HALEU is Russia, where legal and political barriers are in flux (see Box 4.1). In his remarks to the committee, Daniel Poneman of Centrus Energy expressed that U.S. utilities would want at least one assured domestic source before they made a commitment on a HALEU-fueled reactor for 30–60 years (Poneman and Cutlip, 2021). In other words, foreign imports can provide price competition, but not fuel assurance. Furthermore, although downblended or reprocessed highly enriched uranium can provide some limited amounts of HALEU, this source is insufficient for fuel assurances spanning the operational life of utilities’ advanced reactors (Poneman and Cutlip, 2021).

4.2.3.4 Safety and Security Requirements for HALEU

In developing a HALEU supply chain to support advanced reactors, safety and security considerations must be addressed. First and foremost, enrichment and fuel fabrication facilities will have to conform to radiological and safety standards. HALEU will require Category II special nuclear material facilities (see Sidebar 4.1 and Chapter 6). Other issues include having available U.S.-certified shipping containers for enrichments between 10 and 19.75 percent (for UF6, uranium oxide, and uranium metal, and as manufactured fuel forms); criticality benchmarks for licensing facilities and transport packages; and updated regulatory requirements and guidance for nuclear material control and accounting, as well as physical security, that address issues specific to HALEU. These transportation and regulatory issues are discussed in greater detail in Chapter 6.

Use of higher enrichments in HALEU means that margins to inadvertent criticality are reduced. Criticality accidents could occur if there is inadequate regulatory oversight, lack of an appropriate safety culture, or deficient worker training and qualification ( U.S.NRC, 2000). Such problems have occurred; for example, in 1999 at the JCO Fuel Fabrication Plant in Japan, an inadvertent criticality event occurred when operators attempted to process uranium fuel at 18.8 percent enrichment using a technique they had successfully used previously to process uranium with 6 percent enrichment.

Fuel fabrication facilities that process HALEU will have other significant requirements beyond criticality control associated with whether the final form of the fuel is metallic, ceramic (e.g., UO2, UN, UC), intermetallic (e.g., U3Si2), or alloy (e.g., UMo, UZr). These might include fire protection for pyrophric materials and additional chemical safety and radiological controls depending on the specific process being used. Chemical, radiological, and criticality safety standards are likewise important for on-site storage and transportation of HALEU-containing materials at fuel fabrication facilities. All of these changes will also be accompanied by another level of nuclear material control and accounting and physical security requirements because of HALEU. These requirements may prohibit current Category III facilities from reasonably transitioning to a Category II facility without massive modifications and retrofits or substantial redesign of the existing facility. (See Sidebar 4.1 for category definitions.) It might be most cost effective to design, construct, and license a new Category II facility for fuel fabrication rather than modifying an existing Category III facility to a Category II facility and amending the license. (See Chapter 6 for more discussion about security risks.)

4.2.4 Fuel Fabrication

Domestic fuel fabrication is the only piece of the uranium supply chain that is currently sufficient to meet the needs of the U.S. commercial power industry for uranium oxide fuel with less than 5 percent enrichment. However, the existing domestic fuel fabrication facilities are not equipped to handle HALEU or to produce non–uranium dioxide fuels. Thus, fuel fabrication capabilities will have to be built to meet the needs of advanced reactors requiring HALEU and/or using fuel types other than uranium dioxide. A primary challenge for starting an industry for advanced fuel production, as identified by the Nuclear Energy Institute, is that the reactor designers and fuel producers cannot proceed early in the process unless each side is certain that the other will reach commercial deployment (Nuclear Energy Institute, 2018). Other challenges include the lengthy time required to

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

garner commercial support for funding, to address regulatory issues involving both the reactor designs and the fuel enrichments, and to build up the overall fuel cycle infrastructure (Nuclear Energy Institute, 2018).

As mentioned in Chapter 2, three Category III fuel fabrication plants are currently licensed by U.S. NRC and fabricate low-enriched uranium fuel that is sold worldwide to the LWR community: Global Nuclear Fuel-Americas in Wilmington, North Carolina; Westinghouse Columbia Fuel Fabrication Facility in Columbia, South Carolina; and Framatome, Inc., in Richland, Washington. In addition, as noted in Section 4.2.3, two U.S. facilities are licensed to fabricate highly enriched uranium fuel primarily for defense applications: Nuclear Fuel Services in Erwin, Tennessee, and BWXT Nuclear Operations Group in Lynchburg, Virginia. Both are designated as Category I facilities. They have produced both high- and low-enriched uranium fuel for the naval reactors program as well as greater than 5 percent–enriched fuel for nonpower reactors (test, medical isotopes, and research [and training] reactors) by downblending highly enriched uranium. These facilities, in principle, could produce HALEU fuels with modest amendments to their licenses; however, having commercial and defense-related enrichment activities at the same location is not desirable, and the United States has long avoided mixing defense and commercial nuclear activities so as not to encourage this behavior in other countries.

The fabrication of new fuels for advanced reactors will likely require different processes than those described in Chapter 2 for fabricating uranium dioxide fuel assemblies for the existing LWR fleet. New fuels, most of which are based on HALEU as the starting material,3 are being developed to complement advanced reactor designs that are intended to perform more efficiently at high temperatures and to be more accident tolerant under off-normal conditions. The remaining parts of this section focus on fabrication of non-LWR fuels. Because several advanced reactor developers want to use TRISO-type fuels, including X-energy, which received a major award under DOE-NE’s Advanced Reactor Demonstration Program (ARDP), Section 4.2.4.1 has an extensive discussion of their fabrication. Similarly, Section 4.2.4.2 includes substantial coverage of metallic fuel fabrication because of the interest in this fuel type proposed by TerraPower’s Natrium, also a major-award recipient under the ARDP. Section 4.2.4.2 also provides information relevant for understanding fuel fabrication challenges of the other advanced reactor fuel types, including nitride, mixed oxide, carbide, and molten salt liquid fuels.

4.2.4.1 TRISO Fuel Development and Production

TRISO-coated particle fuel is at the heart of many high-temperature reactor designs. These designs include high-temperature gas-cooled reactors (HTGRs), such as X-energy’s Xe-100 helium-cooled pebble-bed high-

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3 The enrichment limitation of 5 percent for LWRs is also under study at the request of Member States of the International Atomic Energy Agency, who want to realize potential economic benefits from higher fuel burnup, longer fuel cycle operation, and reduced used fuel inventory (IAEA, 2020d).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

temperature reactor and Framatome’s Steam Cycle High Temperature Gas-Cooled Reactor, as well as fluoride salt–cooled high-temperature reactors (FHRs), such as Kairos’ XP-X fluoride-cooled [FLiBe]4 pebble-bed high-temperature reactor.

TRISO fuel was first developed in the late 1950s to support the Dragon Reactor in the United Kingdom. The initial process involved applying a pyrocarbon5 layer to protect the uranium carbide fuel kernels during fabrication, but this technique rapidly evolved to the application of many layers to prevent the loss of fission products from the kernel (Demkowicz et al., 2019). Current TRISO fuel particle designs use a spherical kernel with a diameter of 350–500 microns, which contains the nuclear fuel (e.g., uranium dioxide or UCO, a mixture of uranium oxide and uranium carbide), which is then surrounded by three layers of carbon and one layer of silicon carbide (SiC). The first layer is a porous carbon buffer that provides space for fission gases to migrate and accommodates kernel swelling. The next is a pyrolytic carbon layer that acts as a barrier for diffusion of fission products and serves as the mechanical substrate for the SiC layer. The SiC layer functions both as the primary barrier for nongaseous fission products and as the structural, load-bearing component of the fuel particle. The fourth and final layer is made of pyrolytic carbon that further aids in the retention of fission gases, protects the underlying SiC layer during handling, and provides a surface for bonding to the fuel-form graphite matrix. After the graphite is added, the mixture is pressed into its final shape (pebbles or prismatic compacts). For pebbles, pressing is done in two stages: the first pressing is of the fuel center, where the coated particle fuel kernels are evenly distributed throughout the volume. The second pressing then produces an outer fuel-free zone of ~5 mm (Mulder, 2021; Nygaard, 2021; Pappano, 2021).

The fuel types within the kernels of TRISO particles can vary widely depending on the reactor design. The fuel types can be either fissile (e.g., UC2, PuO2, (Th, U)C2, (Th, U)O2, UO2, UCO) or fertile (e.g., ThC2, ThO2, UO2, UCO) materials. The fission reaction using UCO as the fuel produces less carbon monoxide (a product of irradiation-induced chemical reactions), thereby generating less internal gas pressure in the pyrolytic layers of TRISO fuel particles under reactor operating conditions. As a result, UCO has shown superior fuel performance at high burnup. For this reason, many high-temperature reactor developers are choosing UCO as the reference fuel type for their TRISO particles (EPRI, 2019), typically in one of two fuel configurations (see Figure 4.1). In the first configuration, TRISO particles are fabricated into cylindrical pellets or compacts and embedded in channels in prismatic graphite blocks. In the second, TRISO fuel particles are dispersed in a graphite matrix and formed into spheres or pebbles about the size of a billiard ball.

Their structural resistance to neutron irradiation, corrosion, and oxidation makes TRISO fuels very robust. Furthermore, because they do not melt under extreme temperatures, such as those found in loss-of-coolant scenarios, TRISO fuels are considered accident tolerant. TRISO developers also claim the added benefit that, because of the layering process, fission products are retained in the TRISO particles; thus, the fuel form acts as its own containment vessel. Although the layered approach used in making TRISO fuel offers a high degree of containment for fission products, the containment is not perfect, and there is limited operating experience at commercial scale using this fuel. Additional information on TRISO fuel, particularly with respect to its storage and disposal as waste, can be found in Section 5.5.2 in Chapter 5 and Appendix G.

Two companies, BWXT and X-energy, are taking steps to manufacture TRISO fuels in the United States. BWXT announced in November 2020 that its TRISO nuclear fuel line project is actively producing fuel at its Lynchburg, Virginia, facility with the intent of meeting application needs for the Department of Defense and demonstration needs for the National Aeronautics and Space Administration (ANS, 2020). Plans for this facility include scaling up TRISO production with full scrap recovery to ~1 MT per year, if needed, by downblending highly enriched uranium to HALEU in the short-term. BWXT is actively considering adding additional TRISO manufacturing capacity in 1 MT/year modules at another of its existing U.S. NRC–licensed facilities.

X-energy is currently making TRISO-coated particles at its TRISO-X Pilot Facility located at Oak Ridge National Laboratory (ORNL); this facility was developed using private and government funding through DOE’s

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4 FLiBe = 27LiF-BeF2.

5 Pyrocarbon, or pyrolytic carbon, is a robust, graphite-like material deposited from gaseous hydrocarbon compounds on suitable underlying substrates (carbon materials, metals, ceramics) at high temperatures in the absence of oxygen.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.1 Reactor configuration using TRISO fuel particles as pebbles in a pebble-bed reactor or a cylindrical fuel compact embedded in graphite blocks in a prismatic reactor.
SOURCE: Adapted from Demkowicz (2019), slide 10.

Advanced Reactor Concepts Cooperative Agreement, awarded to X-energy in 2016. The Xe-100 fuel form is a pebble of 60 mm diameter, weighing ~200 g, and containing ~1 g uranium-235 with dilution of uranium throughout for proliferation resistance. The fuel core of the TRISO pebble contains roughly 19,000 TRISO particles uniformly throughout, with an exterior 5-mm-thick fuel-free zone.

Operational since 2018, X-energy’s TRISO-X Pilot Facility was developed through a public–private partnership. ORNL supplies 5,000 sq. ft. of processing floor space, but the processing equipment is owned by X-energy. Most of the work to date has been processing with depleted and natural uranium, but the processing area is authorized to handle research quantities of HALEU to support fuel design, manufacturing, and licensing for the Xe-100 reactor. The facility is referred to as a “pilot facility” because only one fuel fabrication production line is being developed and tested under the American Society of Mechanical Engineers’ Nuclear Qualification Assurance Standard. However, all of the equipment in the pilot facility has been developed to be used at commercial scale, which has the potential to significantly reduce risks associated with fuel qualification and simplify the process of scaling production capacity in a commercial setting by simply replicating the existing equipment and increasing the number of production lines in a HALEU-licensed facility.

In November 2021, X-energy announced that its preliminary design for its fuel fabrication facility was completed (WNN, 2021c). X-energy believes this facility will be the first Category II licensed facility capable of handling HALEU of up to 19.75 percent enrichment. X-energy plans to obtain the needed HALEU feedstock from commercial sources. The company anticipates the complete U.S. NRC licensing review to take 36 months, and the facility is scheduled to be operational in early 2025 to meet the Xe-100 schedule for deployment in 2026–2027. In April 2022, X-energy selected the Horizon Center Industrial Park in Oak Ridge, Tennessee, as the location of its first commercial TRISO-X fuel fabrication facility (X-energy, 2022). X-energy states that the ARDP funds all elements of the Xe-100 plant deployment, including the commercial deployment of the TRISO-X Fuel Fabrication Facility (TF3): license application submittal, U.S. NRC license review, fuel processing and material-handling equipment procurement, setup, operational readiness reviews, and manufacturing a sufficient number of fuel pebbles to meet the first core load requirements of the Xe-100 reactor plant (Pappano, 2021).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

4.2.4.2 Metallic Fuel Development and Production

Although metallic fuels are not currently in use in power reactors, they are being researched actively in Russia, the United States, China, and Japan. The United States has previous experience with these fuels—for example, in the operation of EBR-I, EBR-II, the Fast Flux Test Facility, and Fermi-1. TerraPower plans to use sodium-bonded metallic fuel in its Natrium demonstration reactor, slated for initial operation by 2027, and to eventually move to a metallic fuel form with no sodium bond (Hejzlar, 2021; Neider, 2021). TerraPower is conducting irradiation testing of its advanced fuel and, to date, has demonstrated good behavior to over 180 GWd/MT (TerraPower, 2022). ARC Clean Energy and Oklo also plan to use metallic fuel in their ARC-100 and Aurora reactors, respectively. Lightbridge Corporation is developing metallic fuel for use in LWRs (Totemeier, 2021).

Compared with oxide fuels, metallic fuels (U-Pu-Zr) have very high thermal conductivity, but they can experience high swelling and melt at a relatively low temperature (1,160°C). A typical fuel design is shown in Figure 4.2, where the metallic fuel slug is contained within cladding, with the two components being bonded by liquid sodium to improve heat transfer while accommodating the differences in thermal expansion between the metallic fuel and cladding (FRWG, 2018). However, given the risk of cladding failure, metallic fuels are incompatible with lead coolant because of their solubility in lead. Fabrication of metallic fuels, as outlined in Figure 4.3, involves (1) preparing the fuel feedstock from ore or spent nuclear fuel; (2) reducing the feedstock to metallic form; (3) alloying and casting by arc melting, vacuum induction melting, or microwave melting; (4) thermomechanical processing via rolling and coextrusion to remove defects; and (5) heat treatment to obtain the mechanical and material properties necessary to achieve the desired fuel performance (LaHaye and Burkes, 2019; Wood et al., 2020). Metallic fuel fabrication processes can generate large quantities of unrecoverable scrap (e.g., from the reaction of molten fuel with crucibles and fuel molds during the casting process) (see Carmack et al., 2017), which decreases resource utilization and provides challenges for safeguarding the material (Moore and Severynse, 2020). The Environmental Impact Statement for the Versatile Test Reactor assumes that up to 27 percent of the metallic fuel feedstock could be lost as waste (DOE-NE, 2022b). All steps of the metallic fuel fabrication process introduce pyrophoricity, chemical, radiation, and criticality hazards, as detailed in LaHaye and Burkes (2019).

A 2019 workshop identified near- and long-term research needs for the development and use of uranium-zirconium-based metallic fuels (Aitkaliyeva et al., 2020). The near-term research gaps related to swelling and fission gas release, fuel-cladding chemical interactions, phase diagram development and characterization, and thermal conductivity during reactor operation. The long-term needs identified to optimize fuel performance were better understanding of fuel creep/plasticity for individual fuel phases and lanthanide transport in the fuel during reactor

Image
FIGURE 4.2 Schematic of a typical metallic fuel element sodium bonded to the cladding for improved heat conduction.
SOURCE: Adapted from FRWG (2018). Courtesy of Oklo, Inc.
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.3 Process flow for fabrication of metallic fuel.
SOURCE: LaHaye and Burkes (2019). Courtesy of Pacific Northwest National Laboratory.

operation. For each gap, the workshop participants recommended experimental and modeling studies to obtain the requisite data for addressing the research needs. To facilitate further study of metallic fuels, the participants also recommended combining all existing data on such fuels—from experience at EBR-II, the Fast Flux Test Facility, and the Transient Reactor Test Facility, especially—into a single database.

The Advanced Fuels Campaign (AFC), part of DOE’s Nuclear Fuel Cycle and Supply Chain Program, is focusing its efforts related to advanced reactor fuel development on metallic fuels, given their proposed use in several advanced reactor designs (INL, 2021b). Its primary goals in this area include qualifying existing metallic fuel designs and developing and qualifying sodium-free metallic fuel. Another key focus is establishing a licensing basis for metallic fuel, which will require compilation of existing data, as well as additional testing and data collection. The AFC aims “to develop and establish the qualification basis for a sodium-free metallic fuel design with extended temperature performance by 2027,” noting that this initiative will benefit from collaborations with industry stakeholders and other DOE programs (INL, 2021b).

4.2.4.3 Other Advanced Reactor Fuel Types

Other fuel types being considered for advanced reactor designs include nitride fuels, mixed oxide (MOX) fuels, carbide fuels, and molten salt liquid fuels.6 For a discussion of Th fuels and Th-based fuel cycles, see Section 3.2.5 in Chapter 3. Considerations for the fabrication and use of each of these fuel types are discussed below. In general, the production of both nitride and carbide fuels is more complex than that of MOX or metallic fuels. Additionally, fabrication of fast reactor fuels incorporating minor actinides (which could be present in reprocessed fuel materials; see Section 4.3.5 below) may require a much higher degree of complexity and introduce additional occupational safety considerations. These complications are illustrated in Table 4.1 later in this chapter.

Nitride fuels (UN-PuN) are being researched in the United States, Russia, Japan, and Sweden. Their high melting temperature (2,762℃), high thermal conductivity (10 times that of oxide fuels), and high density of fissile atoms enable the potential advantages of larger power uprates, longer fuel cycles, and higher burnup (Wood et al., 2020). Drawbacks include low oxidation resistance and poor hydrothermal corrosion resistance, but additives can help to increase corrosion resistance. The presence of 14N, the most abundant natural isotope of nitrogen (99.6 percent), can negatively affect fuel performance and result in production of 14C, whose half-life of 5,700 years can pose a radiation hazard. Using nearly pure 15N can improve fuel performance and avoid 14C production but requires expensive isotopic enrichment processes that are currently infeasible at scale (Wood et al., 2020). Russia’s High Technology Research Institute of Inorganic Materials reportedly “has patented a technique for enrichment in 15N, annual demand for which is expected to be several tonnes” (WNA, 2021d). Also, as mentioned in Section 3.2, LeadCold has stated that it has developed plans for acquiring enriched 15N and manufacturing nitride fuel (Wallenius, 2021). Without prospects for large demand, these specialty enriched materials will likely remain expensive.

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6 See Table 3.1 for the fuel types being proposed by specific advanced reactor developers.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

A number of methods for synthesizing nitride fuel have been studied, with the two most common being carbothermic reduction of uranium oxide followed by nitridation (CTR-N) and the direct thermal hydride-dehydride-nitride (H-D-N) synthesis (Wood et al., 2020). The CTR-N method is the only nitride fuel fabrication technique demonstrated at large scale; however, its requirements for large volumes of nitrogen and high temperatures, as well as long reaction times, lead to high processing costs and create challenges with working in inert atmosphere glove boxes. The H-D-N synthesis requires less nitrogen and shorter reaction times than CTR-N, but the resulting UN powder has a propensity for oxidation and therefore must be handled in moisture- and air-free environments, making process scale-up challenging. Following any synthesis method, the UN powder is consolidated to produce a dense fuel form. This process is more challenging with nitride fuels than with oxides or carbides, but several successful methods have been reported (Wood et al., 2020).

Mixed oxide fuel (UO2-PuO2) consists of a mixture of PuO2 and depleted or reprocessed UO2 powders that have been sintered into pellets, inserted into Zr-alloy fuel rods, and bundled into assemblies. It has low thermal conductivity and a low density of fissile atoms, but it does not react with Na or Pb. It is the reference fuel form for several reactors. For future advanced reactors, this fuel form could potentially be useful for and used in fast reactors, which would require a higher content of plutonium in their MOX fuel as compared with MOX fuel for LWRs. Chapter 2 provides more details about MOX for LWRs, and this subsection gives some brief relevant information about MOX fuel fabrication. Experiments have demonstrated that the oxide fuel is able to reach very high burnups: in the reactor Rapsodie, a burnup of above ∼240 Gigawatt-days per metric ton of heavy metal was achieved in the 1970s. Commercial production of MOX fuel occurs at the Melox plant in France using PuO2 from the La Hague reprocessing facility. This MOX fuel is distributed for use in 25–30 reactors worldwide, which operate on 30–50 percent MOX in combination with standard UO2 fuel (Orano, 2021a; WNA, 2017a). Japan is currently constructing a MOX Fuel Fabrication Plant, scheduled to be completed in 2024, that will also use PuO2 obtained by reprocessing spent UO2 fuel (JNFL, 2020). As discussed in Chapter 2, U.S. efforts to construct a MOX fuel fabrication facility at Savannah River National Laboratory were unsuccessful, as the project faced significant cost, schedule, and budget challenges; it was terminated in FY2018 (Holt and Nikitin, 2017).

Carbide fuels (UC-PuC) have “high thermal conductivity and a high density of fissile atoms, but high swelling and poor compatibility with air and water” (Wood et al., 2020). The several types of carbide fuels include UC, UC2, U2C3, and (U, Pu)C. These fuels are most commonly synthesized via carbothermic reduction (CTR) using a UO2 precursor, but arc melting synthesis with uranium metal and graphite is also being explored (Wood et al., 2020). Research on UC fuels is being pursued in India, and India’s fast breeder test reactor has run on mixed carbide fuel (70 percent PuC, 30 percent UC) since 1985 (Kumar et al., 2011).

Molten salt liquid fuels, in which the fissile or fertile material is dissolved in a molten chloride or fluoride salt, have been studied since the Molten Salt Reactor Experiment at Oak Ridge National Laboratory in the 1960s. The fuel fabrication syntheses will vary for the different reactor core designs proposed. As an example, Terrestrial Energy plans the following fuel fabrication process for its Integral Molten Salt Reactor: reduction of enriched UF6 (4.95 percent 235U) to UF4 with hydrogen gas, followed by mixing with commercially available fluoride salts and further processing to remove impurities, oxides, and moisture (Terrestrial Energy, 2021). Many molten salt fuel syntheses will involve electrochemical or pyrochemical methods, as these are considered the most viable for industrial-scale processing (Wood et al., 2020). In all cases, molten salt fuel fabrication must be performed under an inert atmosphere, as the salts form corrosive acids in the presence of water.

4.3 BACK END OF THE FUEL CYCLE

The remainder of this chapter discusses fuel cycles that support, or conceptually could support, the existing LWR fleet and potential future advanced reactors starting with the reference case, the once-through fuel cycle. The introduction of fuel reprocessing in a few countries has resulted in the monorecycling of both uranium and plutonium in existing LWRs. However, fast reactors that would be able to use uranium and plutonium extracted from spent LWR fuel have not become available in the expected time frame, resulting in a need to mitigate the buildup of the extracted plutonium. The deployment of fast reactors in combination with advanced reprocessing

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

technologies would close the fuel cycle with the promise of efficient consumption of fertile uranium-238, multirecycle of plutonium, and concurrent management of the minor actinides. The goals of the multirecycle strategy are to (1) allow full utilization of the fertile uranium-238 by transmutation into plutonium-239 fuel; (2) multirecycle plutonium-239 to regenerate plutonium-239 at the expense of uranium-238; (3) remove or partition elements of interest—most notably, the minor actinides from the waste stream during spent fuel reprocessing; and then (4) destroy or transmute these elements into shorter-lived or stable species by recycling them in transmutation devices, such as advanced fast reactors.

While the goal of multirecycling plutonium is to maximize the energy extracted from nuclear fuel and subsequently natural uranium, the goal of partitioning and transmutation (P&T) endeavors is to minimize the alpha-bearing, radiotoxic, and heat-generating transuranic elements in the wastes requiring geologic disposal. In a multirecycle, fully closed fuel cycle scenario, if advanced reactors and fuels cycles supporting them could separate (partition) all of the actinides from spent fuel (uranium and transuranic elements) via reprocessing and transmute them (burn or fission them in advanced reactors) such that they are completely removed from the waste stream going to a geologic repository (no process losses), then all that is left is radioactive fission products (both short-lived, such as 137Cs and 90Sr/90Y, and long-lived [see the table in Sidebar 1.2 in Chapter 1]). These radioactive fission products would require management and isolation from the biosphere (geosphere) for long periods of time. In the limit of P&T with 100 percent efficiency (a highly unlikely situation because processes are not perfect), the remaining high-level waste would be just radioactive fission products. As the efficiency of partitioning and transmutation decreases, due to process losses during partitioning and incomplete reactions during transmutation, more actinides become part of the high-level waste, which at late times adds to the heat load and the radiotoxic inventory of the high-level waste.

4.3.1 Introduction to Advanced Fuel Cycles

Advanced fuel cycles seek to maximize the utilization of natural uranium resources and reduce the high-level waste burden on permanent geologic repositories while maintaining the economic viability of power generated by nuclear systems and conforming to the highest levels of safety and nonproliferation. This is a tall order, and in practice, these goals will be extremely challenging to achieve simultaneously.

As noted in Chapter 1, Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (NEA-OECD) describes advanced nuclear fuel cycle options in terms of a simple basis set of three fuel cycles: (1) once-through, (2) monorecycle, and (3) multirecycle options. Fuel cycle options can be further delineated as open, partially closed, and fully closed depending on the types of material sent for permanent disposal in lieu of recycling. In its 2010 report Advanced Nuclear Fuel Cycles—Main Challenges and Strategic Choices, the Electric Power Research Institute (EPRI) defined open, partially closed, and fully closed fuel cycles with respect to the management of Pu, noting that these terms are “often associated with different understandings by different authors” (EPRI, 2010b). EPRI’s definitions follow:

  • An “open fuel cycle is a fuel cycle in which Pu [spent fuel] is sent to a geologic repository for permanent disposal”; they include the once-through and the mono-recycle fuel cycles, when the spent LWR-MOX fuel is eventually disposed of in a geologic repository.
  • A partially closed cycle is a fuel cycle in which Pu is continuously recycled while the minor actinides are sent to a geologic repository for permanent disposal;7 and
  • “A fully closed cycle is a fuel cycle in which no Pu or minor actinides are sent to a geologic repository for permanent disposal,” but are being continuously recycled (EPRI, 2010b).

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7 This applies for the situation where spent MOX is stored pending reprocessing for future use in fast-spectrum reactors. But if spent MOX is eventually sent to a repository, monorecycle of Pu is an open cycle. Partially closed means multirecycle of Pu, or closed with respect to Pu management.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

Other entities define fuel cycles in other ways, such as the French program, which emphasizes the fate of uranium and plutonium.8

Because heat generation due to plutonium isotopes and their decay products can impact a geologic repository, the EPRI definition focusing on the disposition of plutonium has been adopted for the purpose of this report. Importantly, material losses will inevitably occur with reprocessing and in other parts of the fuel cycle; therefore, no cycle can be perfectly closed.

The simplest and most straightforward fuel cycle is the once-through cycle that uses low-enriched uranium in LWRs and directly disposes of the spent nuclear fuel in a deep geologic repository. The once-through cycle, shown in Figure 4.4, is an open fuel cycle and is the reference to which all other fuel cycles are compared. As noted previously, the United States and most other nuclear power–producing countries—including Canada, Sweden, Finland, Switzerland, Spain, and Germany—have opted for the once-through cycle for their existing LWR fleets because of its relative simplicity, lower implementation cost in the short term, and anticipated higher proliferation resistance. No advanced facilities are required to support the once-through fuel cycle, as fuel reprocessing is not required. However, as indicated by Table 3.1 in Chapter 3, almost all advanced reactor developers intend to use a once-through cycle for the first decade or more of reactor operations.

4.3.2 Background on Reprocessing and the PUREX Process

Reprocessing refers to a set of operations or steps, both mechanical and chemical, that are required to remove the irradiated fuel from the fuel assemblies and cladding and to separate the spent fuel into process streams of reusable fuel material and waste. A reprocessing system or plant consists of four main components: (1) head-end processes (e.g., processes to separate fuel matrix from assembly structural materials and make the fuel form compatible with subsequent chemical separation steps); (2) chemical separation and conversion steps (e.g., processes designed to remove elements to be recycled and then convert them into a form compatible with the subsequent fuel fabrication process, such as oxide powder for aqueous-based processing); (3) waste management (e.g., off-gas treatment, fission product vitrification, metallic waste compaction); and (4) additional plant support systems (e.g., reagent and solvent recycle systems, methods for process control and accountability, robotics, in-cell maintenance capabilities).

Separations processes are the core of a reprocessing plant, as they are the chemical processes designed to specifically remove or isolate elements of interests based on their chemical properties from the mixture of elements in spent nuclear fuel. The separations are often difficult and complex because the mixture of elements in spent nuclear fuel represents roughly one-third of all of the elements in the periodic table (Nash and Lumetta, 2011; Taylor, 2015). The separation processes can be characterized broadly as aqueous- (hydro-) or nonaqueous-based (pyrochemical and pyroelectrochemical) processes.

Image
FIGURE 4.4 Schematic of an open or once-through fuel cycle for a light water reactor (LWR).
NOTE: Nat. U = natural uranium; Udep = depleted uranium; UF6 = uranium hexafluoride; UOX = uranium oxide.
SOURCE: Adapted from MIT (2011).

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8 In France, a partially closed fuel cycle is where spent nuclear fuels are reprocessed for recycling fissile and fertile materials (uranium and plutonium) to increase the energy production from the same initial material. Monorecycling where irradiated mixed oxide fuels are irradiated once and disposed of falls in this category. The French definition of a closed fuel cycle is where uranium and plutonium are continuously recycled, while the minor actinides can either be recycled or sent to a geologic repository for permanent disposal (Poinssot and Boullis, 2012).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

To date, the only commercial-scale, industrially deployed process for reprocessing spent nuclear fuel is the well-known PUREX (plutonium uranium reduction extraction) process, a mature aqueous-based separation process9 with an operational experience base of more than 60 years. PUREX has been the industry standard for separating uranium and plutonium from both oxide and metallic fuels. It involves the extraction, reduction, and back extraction of uranium and plutonium between a nitric acid solution and an organic phase containing tributyl phosphate. PUREX results in two separate pure products streams, uranium and plutonium, and a nitric acid solution containing the minor actinides (Np, Am, and Cm) and fission products. The nitric acid solution is designated as high-level waste and is further treated to immobilize the minor actinides and nonvolatile fission products in a glass matrix using vitrification. In the vitrification process, the radioactive materials are dispersed within the glass and chemically bonded in the glass network. While still hot, the vitrified high-level waste is poured into a standardized stainless-steel container and allowed to cool and solidify. Reprocessing also generates a solid waste stream mostly consisting of metallic fuel assembly hardware (e.g., fuel cladding and fuel assembly structural materials). At La Hague, these solid wastes are compacted to up to 65 percent of the metal density and placed in another standardized stainless-steel container. These uniform stainless-steel containers serve as the waste packages for both the high-level waste and metallic residues. The waste packages are designed to provide a safe, stable, and highly reliable compact form for long-term storage (>300 years) or until a geologic repository is available.10

Reprocessing spent fuel generates other process and operational wastes, including solid (residues, structural materials and equipment, resins, filters, and personal protective equipment), liquid (organic and aqueous), and gaseous (process off-gases) waste (CEA, 2008, 2009; NEA-OECD, 2006b). Each type of waste needs to be treated and conditioned in accordance with industry standards and applicable regulations for transportation, storage, or disposal for that particular waste category. These wastes will be low-level waste, as well as GTCC (Greater-than-Class-C waste). (See Appendix D for definitions of the various waste categories.) By volume, reprocessing produces more low-level waste and GTCC than high-level waste (ANDRA, 2006). (See Chapter 2 for more details and industry data about the amounts of these wastes from PUREX reprocessing.)

Since its inception, the PUREX process used for LWR fuel has been continually optimized for maximum recovery (>99.9 percent) and purification (decontamination factors versus fission products >105) of the uranium and plutonium product streams. These improvements have also led to a significant reduction in the volume of waste requiring geologic disposal, reduction of in-process water consumption, and the recovery and reuse of many of the reagents and solvents used in the reprocessing. Other advances have been incorporated into plant operations to increase process efficiency (e.g., redesign of liquid–liquid extraction equipment, such as mixer-settlers, pulsed columns, and centrifugal extractors for reliability and ease of maintenance), as well as to decrease secondary waste (e.g., optimization of process flow sheets and procedures). These improvements have been carried out concurrently while increasing plant safety and implementing many additional safeguards outlined by the International Atomic Energy Agency (IAEA) (CEA, 2008; Poinssot, 2021).

Future reprocessing plants are expected to face additional, more stringent, and costly requirements to improve safety and further reinforce safeguards, security, and proliferation resistance while reducing environmental impact and waste production. The cost-effectiveness of reprocessing and recycling is a controversial issue with studies presenting various conclusions, likely because of the difficulty of estimating and accounting for the cost of reprocessing and geologic disposal in the very long-term, as well as comparing short- (reprocessing) with long-term (disposal) costs. For more detailed information on the state of the art in reprocessing at La Hague, see CEA (2008, 2009).

As described in the next sections, implementing the mono- and multirecycle fuel cycles requires reprocessing.

4.3.3 Monorecycling of U and Pu in LWRs

Monorecycling of Pu in LWRs using PUREX reprocessing is the only fuel cycle besides the once-through fuel cycle that has been commercially deployed to support LWR operations (CEA, 2015). The U and Pu product

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9 PUREX was developed at Oak Ridge National Laboratory. See Long (1967).

10 The waste packages for both high-level waste and metallic residues described are those used by the French Reprocessing program at La Hague, France, in CEA (2008).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

streams from PUREX can be recycled as reprocessed uranium oxide (UrepOX) fuel and MOX fuel, respectively.11 As mentioned in Chapter 2, Urep must first be reenriched before it can be recycled. Reenrichment of Urep is performed in a dedicated cascade because of potential cross contamination by 232U. An overenrichment of ~0.5 percent is required to counteract the presence of 236U after enrichment.12 Another option for reenrichment involves downblending by using some of the existing stockpiles of excess-defense-related highly enriched uranium. This approach is technically preferable because it limits the buildup of 236U. Figure 4.5 illustrates the fuel cycle steps for monorecycling Urep in LWRs, which provides ~8 percent savings in natural U resources compared with the once-through cycle13 (EPRI, 2009b; IAEA, 2007a, 2009).

On the other hand, Pu can be mixed with Udep and directly used to make MOX (U/Pu) fuel for recycle in the LWRs (LWR-MOX). Implementing a fuel cycle for the monorecycle of Pu requires the addition of a MOX fuel fabrication facility to the supporting infrastructure. The current industrial practice is to recycle LWR-MOX fuel only once in LWRs because of the buildup of nonfissile, even Pu isotopes. Figure 4.6 shows the steps for monorecycling Pu in LWRs.

Because LWR-MOX fuel typically contains 7–10 percent Pu, the fissile inventory and minor actinide content of spent LWR-MOX fuel is higher than that of spent UOX fuel.14 As a result, spent LWR-MOX fuel exhibits characteristics additional to those of spent UOX that must be managed appropriately. These include a higher decay heat,15 an increased potential for criticality, and (because of the increased minor actinide content) a higher

Image
FIGURE 4.5 Steps in monorecycling uranium (U) in light water reactors (LWRs). Spent UOX (1) refers to the spent fuel resulting from the initial load of UOX fuel. Spent UrepOX (2) refers to the spent fuel resulting from the initial load of UrepOX fuel.
NOTE: FP = fission product; HLW = high-level (radioactive) waste; MA = minor actinide; Nat. U = natural uranium; Udep = depleted uranium; UF6 = uranium hexafluoride; UOX = uranium oxide; Urep = reprocessed uranium; UrepOX = reprocessed uranium oxide.
SOURCE: Adapted from MIT (2011).

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11 As implemented in France, reprocessing and recycling can be adjusted to mitigate any accumulation of separated materials. According to the French nuclear industry, the current accumulation of separated materials in the French program is temporary and related to the ongoing adjustment of the MOX manufacturing process to accommodate new U powder. See Table 2.1 in Chapter 2 and references therein about the accumulated stockpiles of separated Pu in France and other countries.

12 This creates a constraint for plants that operate with fresh fuel enriched at >4.5 percent, because the higher reenrichment would bring the required enrichment value to >5 percent, which is over the enrichment limit of present LWR fuel fabrication plant licenses.

13 Savings estimates can vary by a few percent depending on the assumed reference scenario (e.g., initial fuel enrichment, burnup, residual enrichment at discharge, assigned 236U enrichment penalty).

14 See Chapter 2, table in Box 2.1, where the masses of Pu isotopes and the main MAs in a total of seven spent LWR fuel rods are compared to one spent LWR-MOX rod 3 years after discharge.

15 See Chapter 2, Figure 2.2, which compares decay heat of spent UOX to that of spent MOX fuel irradiated to the same burnup.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.6 Steps for monorecycling Pu fuel in LWRs. Spent UOX (1) refers to the spent fuel resulting for the initial load of UOX fuel. Spent LWR MOX (2) refers to the spent fuel resulting from the initial load of LWR-MOX fuel.
NOTE: FP = fission product; HLW = high-level (radioactive) waste; MA = minor actinide; Nat. U = natural uranium; Udep = depleted uranium; UF6 = uranium hexafluoride; UOX = uranium oxide.
SOURCE: Adapted from MIT (2011).

radiation source term (mainly alpha, but gamma and neutrons as well). Because recycle of MOX into a fast reactor (FR-MOX) is not currently an industrially available option, spent LWR-MOX fuel is being stored, awaiting a final decision to either be recycled, should fast reactors become available, or to be disposed of in a geologic repository. Monorecycling of Pu in LWRs should be classified as “open,” just like the once-through UOX fuel cycle, when the spent MOX fuel is not reprocessed and is instead disposed of as high-level waste in a geologic repository. When both U and Pu are recycled, this fuel cycle provides ~20 percent savings in natural resources compared with the once-through cycle (roughly 10 percent U and Pu each) (EPRI, 2010b; Poinssot and Boullis, 2012).

4.3.4 Fuel Cycle for the Multirecycling of Pu in LWRs

In the absence of fast reactors, a potential next step in the progression of fuel cycles could be multirecycling of Pu in LWRs. In this case, spent fuel (UOX and MOX) is always reprocessed, and the recovered Pu is, at least theoretically, indefinitely recycled in an LWR fleet, providing a path to stabilization16 of spent fuel and Pu inventories.17 However, multirecycling of only Pu in MOX-fueled LWRs has limitations, because the fissile Pu content decreases with each recycle. Compensating for the decrease in fissile Pu quality by increasing the overall pluto-

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16 The number of fuel assemblies decreases by a factor of 6–7 when spent UOX fuel is reprocessed. When MOX fuel assemblies are also reprocessed and the recovered Pu is recycled by adding fissile 235U makeup, MOX is no longer stored cumulatively. At this point the number of spent fuel assemblies no longer increases, so all of the spent fuel is stabilized. How “fast” this situation will be realized depends on the implementation details (i.e., the rate at which reprocessing of the spent MOX and MOX-EU assemblies can be accomplished). Note that the description is relevant for space savings in the number of spent fuel assemblies that need to be stored either at a reprocessing facility or at a temporary spent fuel facility. The geologic repository footprint would have to take into account the additional heat load from the spent MOX fuel assemblies.

17 La Hague has already reprocessed spent MOX fuels with a residual fissile content greater than that in low-enriched uranium fuel by using dilution, which consists of processing MOX (high residual fissile content) concurrently with LEU-UOX (low residual fissile content). With such an approach, the capacity for processing spent MOX fuel is only part of the total capacity of the plant. Reprocessing MOX fuels at industrial scale would therefore require adding workshops to the current reprocessing plant to avoid this dilution step.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

nium content is limited to a maximum of ~12 percent Pu, at which value the void coefficient becomes positive (Martin et al., 2018). To overcome this limitation, multirecycling of Pu requires the addition of enriched rather than depleted uranium in the fabrication of the MOX fuel (LWR-MOX-EU). The realized savings in natural U do not differ significantly from what monorecycling can achieve, but multirecycling Pu in an LWR can provide a way to manage the inventory of Pu; it allows the extraction of additional energy while helping to minimize the buildup of Pu requiring storage.

While, in theory, no Pu would be sent for permanent disposal, there are always some losses during reprocessing. Compared with Pu monorecycling, there is an expected increase of ~70 percent in the accumulation of minor actinides going into the repository (Martin et al., 2018). Multirecycling of Pu constitutes a partially closed fuel cycle as classified by EPRI (2010b), in that the fuel cycle would be closed with respect to Pu (except for process losses) but open with respect to the minor actinides. As discussed in Section 2.3.2.2, Urep is not currently being recycled in LWRs but rather stored, pending decisions on whether to recycle or dispose of it. Such a fuel cycle is shown in Figure 4.7.

France has recently announced its intention to assess the feasibility of such a fuel cycle (only for one or two additional recycling loops) as a means of stabilizing its spent fuel and Pu inventories as a transient stage before a potential deployment of fast neutron reactors (Landais, 2021). Similarly, Russia is working toward multirecycling Pu in LWRs via its REMIX program (described in Chapter 2).

4.3.5 Management of Minor Actinides

As early as the 1970s, studies, particularly in Europe, focused on the feasibility of a fully closed fuel cycle with multirecycling of plutonium and P&T of the minor actinides for the management of nuclear waste (Serco, 2011). Most fully closed fuel cycles using P&T have focused on a goal of a 100-fold reduction in the inventory of transuranic radionuclides in the resulting high-level waste, which implies recovery efficiencies greater than 99.9 percent (less than 0.1 percent process loss).

As first envisioned, P&T involved partitioning the minor actinides (and possibly select long-lived fission products) and placing them in some sort of transmutation device in which large quantities of neutrons react with these isotopes via fission or capture reactions to form shorter-lived or even stable isotopes. Four basic types of transmutation devices have been considered and evaluated: fast reactors, thermal reactors, accelerator-driven systems (ADSs), and fusion/fission hybrids (FFHs) (National Research Council, 1996). ADSs are hybrid accelerator/reactor devices in which a beam of particles from an accelerator strikes a subcritical target producing a burst of high-energy neutrons in a process called spallation. The neutrons interact with the isotopes to be transmuted in a subcritical blanket assembly causing them to fission. An FFH is also a subcritical reactor, but it differs from an ADS in that a fusion core provides the high-energy neutrons to fission the isotopes that are to be transmuted in the surrounding blanket. Consideration of the ADS and FFH technologies are not within the scope of this study, and they are not discussed further in this report. For additional information on ADS and FFH, see NEA-OECD (2002), IAEA (2015a), and DOE (2009b).

Neutron-induced fission is a transmutation reaction that, for minor actinides, results in much lighter, shorter-lived isotopes that would likely decay by beta emission or be stable. As discussed in Chapter 2, the neutron capture reaction in LWRs leads to the buildup of significant inventories of even-mass-number higher actinides (e.g., curium isotopes and even californium-252), because of their propensity for neutron capture (Carbonnier, 2006). Many of these isotopes, especially californium-252, are intense neutron emitters that would make reprocessing or fuel fabrication much more difficult—“the neutron source term for multi-pass recycling of MAs in LWRs is more than 2,000 times higher than that for fast reactors” (EPRI, 2010b). Consequently, transmutation of the minor actinides in thermal reactors would not be a very efficient process. This is not the case for fast neutrons, however, as shown in Figure 4.8.

With fast neutrons, all plutonium isotopes can fission and contribute to neutron production, although plutonium-239 and -241 provide the dominant contribution. The fission of even plutonium isotopes in a fast reactor has the added benefit of decreasing the production of minor actinides during reactor operation, whereas in thermal reactors, isotopes of minor actinides with even numbers of protons18 continue to accumulate. Consequently,

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18 The sentence was corrected following release of a prepublication version of the report to clarify which isotopes of minor actinides accumulate in thermal reactors.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.7 Multirecycling of Pu in light water reactors (LWRs) using mixed oxide fuel plus enriched U (MOX-EU).
NOTE: EU = enriched uranium; FP = fission product; HLW = high-level (radioactive) waste; MA = minor actinide; MOX = mixed oxide; Nat. U = natural uranium; Udep = depleted uranium; UF6 = uranium hexafluoride; UOX = uranium oxide; Urep = reprocessed uranium.
SOURCE: Adapted from MIT (2011).

management of minor actinides has been evaluated primarily in the framework of closed fuel cycles with fast reactors (EPRI, 2010b). The P&T strategy for closing the fuel cycle by managing the MAs is a two-step process: (1) partitioning or isolating these elements in high yield and (2) efficiently transmuting them in fast reactors.

Closing the fuel cycle to recycle plutonium and destroy the minor actinides requires the development of advanced separation processes and fuel fabrication technologies. Fully closing the fuel cycle would require processing of advanced fuels that are much more radioactive than conventional uranium oxide fuel. For example, advanced fuels at discharge can have on the order of 100 times more decay heat and 1,000 times stronger radiation fields. These conditions are well beyond the current industrial experience base of reprocessing and fuel fabrication technologies, making it that much more difficult to perform these operations with low actinide losses, especially using aqueous reprocessing methods.

However, countries such as Russia and France have already gained limited experience on recycling fast reactor spent fuels. For example, the French program has demonstrated the capacity of recycling irradiated sodium-cooled fast reactor mixed oxide fuels by recycling more than 27 MT through their APM (Marcoule) and UP2-400 (La Hague) plants in the 1980s and 1990s (Poinssot and Boullis, 2012). In addition, France has acquired some experience with reprocessing, fabrication, and burning of spent fuel from the Phénix fast reactor

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.8 Probability of fission per neutron absorbed in actinide isotopes for thermal and fast spectra (e.g., for 239Pu, probability 0.65 in a thermal spectrum and 0.87 in a fast spectrum).
SOURCE: Adapted from Wade and Hill (1997).

(Guidez, 2013, 2017). To achieve a 100-fold reduction in long-term hazards associated with spent nuclear fuel, fuel burnup in advanced reactors would need to be high, and reprocessing and fuel fabrication processes would have to be demonstrated on an industrial scale to have high recovery efficiencies with minimal (<0.1 percent) process loss. Dedicated R&D programs have been developed in Europe and Japan particularly to demonstrate the possibility of reaching such high recovery and recycling performances based on innovative and more efficient extractive molecules, such as monoamides (Mahanty et al., 2019; Prabhu et al., 1997; Taylor, 2015; Wang and Zhuang, 2019).

Also considered in early studies of minor actinide management was the possibility of P&T of some long-lived fission products (LLFPs), which are high-yield fission products with particularly troublesome behavior in the environment. LLFPs (including iodine-129, technetium-99, cesium-135, chlorine-36, and selenium-79) constitute only about 0.4 percent of spent fuel mass (compared with about 1 percent for plutonium and the minor actinides), but they are of concern for geologic disposal because of their predicted high mobility in the geosphere and biosphere due to their high water solubility, high diffusivity, and low sorption propensity. However, management of LLFPs is far less amenable to P&T techniques compared with that of minor actinides.

Because fission products do not undergo fission by thermal or fast neutrons, neutron capture reactions offer the only path for transmutation in reactors. Fission products are a direct consequence of the fission process and are relatively insensitive to whether thermal or fast neutrons initiate the fission reaction. Recycling is not a reasonable option for fission products, as they do not produce energy or add to the neutron inventory in the reactor. In fact, the result is just the opposite—transmutation of fission products is a net consumer of neutrons and thus a drain on the neutron economy in any type of reactor, fast or thermal. In theory, a transmutation strategy applied to LLFPs could be cost effective if neutron capture reactions (followed by beta decay) would lead to shorter-lived or stable transmutation products without having to first perform the costly step of isotope separation (Kailas et al., 2015).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

In practice, efforts to transmute the LLFPs with either thermal or fast neutrons have been difficult due to a combination of their relatively small neutron capture cross sections and the associated isotopic compositions for these LLFPs.19 However, research carried out by the French Atomic Energy Commission (CEA) shows that LLFP extraction is technically feasible by modifying the PUREX process; this is subject to important limitations and considerations (Zaetta et al., 2005).

With neutrons, only the transmutations of iodine-129 and technetium-99 have been considered seriously because of their large isotopic abundances and relatively large thermal capture cross sections (Chiba et al., 2017; EPRI, 2010b). CEA studies indicate that if iodine could be transmuted, its chemical form would be unstable under irradiation in a reactor. Thus, P&T of iodine-129 appears to be an unattractive option for the current technology. France, which supported a large national research program on the transmutation of LLFP from 1991 to 2006, concluded the transmutation of LLFPs was not viable and closed the door to this alternative (Poinssot, 2021). Although partitioning of LLFPs might become feasible, it will most likely be restricted to the most challenging fission products (EPRI, 2010b).

4.3.5.1 Plutonium Burning and P&T of Minor Actinides in Fast Reactors

As was previously mentioned, all plutonium isotopes (odd and even) in a fast neutron spectrum can fission and contribute to neutron production, although the plutonium-239 and -241 isotopes provide the dominant contribution. Furthermore, in fast reactors, the production of higher actinides, many of which are intense neutron emitters, is relatively minor compared with the buildup and accumulation of higher actinides during multirecycling in thermal LWRs. Consequently, any attempt to manage the minor actinides—that is, to reduce their mass for disposal in a geologic repository—only makes sense in the framework of closed fuel cycles with fast reactors (Szieberth et al., 2013; Tuček et al., 2008).

With the introduction of fast reactors, one possible scenario would involve the transition from an existing fleet of LWRs to a fleet of all fast reactors in a two-tier approach, coupling an LWR with a fast reactor for reprocessing the spent LWR-MOX fuel. The resulting plutonium would be recycled as fuel, and the minor actinides would be partitioned and transmuted in the fast reactor. As mentioned in Chapter 3, fast reactors can operate in a number of different modes based on the conversion ratio (CR), or ratio of fissile material produced versus that consumed. For this fuel cycle, the reactor is assumed to be operating as a burner reactor with a CR <1, thus it is consuming more fissile material than it produces. A schematic diagram of such a two-tier coupled LWR/fast reactor fuel cycle is shown in Figure 4.9, with the components of the fuel cycle before LWR-MOX fuel fabrication omitted for clarity.20

In 2009, a detailed time-dependent analysis of a two-tier coupled LWR/fast reactor fuel cycle (as shown in Figure 4.9) was conducted jointly by EPRI in the United States and Électricité de France (EDF), the French electric utility company. The objective of this study was to quantify the efficiency of FRs operating as burner reactors for recycling Pu and destroying the MAs for a coupled LWR/FR system (Machiels et al., 2009). The results indicated that, in addition to the decades needed to deploy advanced fast reactors and their associated fuel cycle facilities, many decades of continuous operation of the multirecycle fuel cycle are required to reach equilibrium for significant waste management benefits to be achieved. Furthermore, the results suggested that a more efficient approach than the two-tiered system would be to deploy fast breeder reactors that can produce new fissile fuel, thereby saving natural uranium resources as well as destroying the minor actinides.

Several countries, most notably France, Russia, China, and Japan, are developing transitional plans toward fully closed fuel cycles that first couple thermal LWRs with fast breeder reactors and advanced fuel cycles as a

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19 An LLFP is a specific isotope of a given element. Spent nuclear fuel most likely contains one or more isotopes of that element in addition to the specific LLFP. The isotopic composition is important because efforts to reduce the half-life of an LLFP should not lead to the creation of the same isotope or some others having longer half-lives through nuclear reactions induced on the other stable or short-lived isotopes of the element being transmuted.

20 For simplicity, the components of the fuel cycle before LWR-MOX fuel fabrication in this figure have been omitted. The omitted parts are exactly the same as that shown in Figure 4.6 for monorecycling plutonium in LWRs with the exception that the spent LWR-MOX is no longer sent to a geologic repository but rather reprocessed to recycle plutonium and separate the minor actinides (neptunium, americium, and curium) for transmutation in a fast reactor.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.9 Schematic diagram of the fuel cycle that combines LWRs and fast reactors for multirecycling Pu and partitioning and transmuting minor actinides. If the fast reactor in the diagram is operating as a burner (CR <1), there would be no need for blanket fuel. Blanket fuel is required for breeding (where more fissile material is produced than consumed).
NOTE: FBR = fast breeder reactor; FP = fission product; HLW = high-level (radioactive) waste; MA = minor actinide; MOX = mixed oxide; UrepOX = reprocessed uranium oxide.
SOURCE: Adapted from MIT (2011).

means of recycling reprocessed nuclear materials. These transitional plans cover a time period of roughly 50–100 years over which thermal LWRs continue to generate electricity but are slowly phased out and replaced by advanced, new fast spectrum reactors (see Chapter 2, Section 2.5.1).

4.3.5.2 Plutonium Breeding and P&T of Minor Actinides in Fast Reactors

Breeder reactors operate with a CR ≥1 such that they both breed (via blanket fuel21) and burn plutonium and transmute (via burning) the minor actinides. Because breeder reactors both produce new fuel and can transmute minor actinides, they are the reactors of choice for P&T. In theory, a fleet of 100 percent fast breeder reactors would not need any additional uranium enrichment and would not consume any new natural uranium resources because they use existing depleted uranium stockpiles (depleted uranium tailings from previous enrichment operations) and would generate all of the plutonium needed for power production. The use of depleted uranium tailings would allow extending the current electricity production for several millennia (Poinssot and Boullis, 2012). The French program intends to pursue a strategy of operating fast reactors at a CR = 1 to develop a self-sustaining fast reactor fleet without any increase in the plutonium stockpile, while consuming the minor actinides.

For the case of plutonium breeding and P&T of the minor actinides in fast reactors, the same fuel cycle strategy as shown in Figure 4.9 is employed, except that the fast reactor is operating as a breeder/converter with a CR ≥1 and uses MOX fuel. The MOX fuel for the fast breeder/converter reactor (FBR-MOX) would have considerably higher plutonium content compared with LWR-MOX fuel (20–22 percent plutonium compared with 8–10 percent). The difficulty of reprocessing and recycling spent FBR-MOX increases with increasing plutonium

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21 Blanket fuel surrounds the reactor core in fast reactor designs and is composed of fertile material (e.g., uranium-238 or thorium-232) that forms fissile material (e.g., plutonium-239 or uranium-233) upon neutron capture. For the fissile material bred in the blankets to be used as new fuel, the blankets must be reprocessed to extract the fissile material from the remaining fertile material.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

content, especially for aqueous processes, because of radiolysis and criticality concerns. With proper attention and procedures, these concerns may be manageable. Implementation of this fuel cycle requires advanced technologies (reactors, processing plants, and fuel fabrication facilities) that do not currently exist at industrial scale, although, as previously mentioned, reprocessing FBR-MOX has been demonstrated as feasible at significant scale of 27 MT in France (Poinssot and Boullis, 2012).

Depending on the fast reactor technology deployed, the fuel cycle for multirecycling plutonium and the P&T of minor actinides using an all–fast reactor fleet fuel cycle could look like that shown in Figure 4.10.

4.3.5.3 Homogeneous and Heterogeneous Recycling and Transmutation of the Minor Actinides

Two recycling approaches, homogeneous and heterogeneous, are being evaluated for the transmutation of minor actinides in fast reactors. These approaches differ in the location of the materials in the core and the manner in which they are irradiated (NEA-OECD, 2012).

Homogeneous recycling: In homogeneous recycling, the minor actinides, extracted by advanced reprocessing using either element-specific or grouped separations technology, are homogeneously diluted with uranium/plutonium as an integral component of the commercial nuclear driver fuel22 (metallic, nitride, MOX, etc.). As an integral component of the driver fuel, the minor actinides can impact the reactivity and kinetic behavior of the core. Thus, loading minor actinides using this method will likely be constrained to no more than a few percent. This constraint also helps to limit the dose and decay heat that must be managed during fuel fabrication, transport, and handling. The experience base for homogeneous recycling is similar to that of MOX recycling, so reactor operations are not expected to be significantly altered by the homogeneous addition of a few weight percent of minor actinides to the driver fuel. Multirecycling of minor actinides in fast reactors generates higher actinides, although the accumulation is significantly less than that from LWRs. One major shortcoming is that this method contaminates the entire fast reactor fuel cycle with minor actinides. This disadvantage is mitigated by dilution, which avoids concentrating minor actinides in any one process or material form.

Image
FIGURE 4.10 Notional fuel cycle for multirecycling plutonium and partitioning and transmuting minor actinides using an all–fast reactor fleet.
NOTE: FBR-MOX = mixed oxide fuels for fast breeder reactors; FP = fission product; HM = heavy metal.
SOURCE: Adapted from MIT (2011).

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22 Driver fuel, located within the reactor core, consists of the fissile material (e.g., used to sustain the nuclear chain reaction) and serves as the primary source of heat for electricity generation.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

Transmutation of minor actinides in a homogeneous mode cannot be accomplished in a single pass but requires multirecycling. During the recycling, some higher actinides—many of which are intense neutron emitters—are generated, but this buildup is relatively minor compared with that from multirecycling in LWRs. The actual elemental and isotopic composition will be determined by which particular actinides are recycled. Notably, transmutation of minor actinides (Np, Np+Am, or Np+Am+Cm) does not eradicate them instantaneously. Their management and radiation exposure burden shifts to other parts of the fuel cycle, thereby making fuel fabrication, spent fuel reprocessing, and reactor operations more difficult. The extent of this challenge can be evaluated by determining the total inventory of transuranic elements by location at some arbitrary time in the fuel cycle. The inventory can be represented in terms of decay heat, γ dose, or neutron emission source term, as shown in Table 4.1 (Zaetta et al., 2005).

The values in Table 4.1 show factors by which a given problem (e.g., decay heat, γ dose, and neutron sources term) is magnified compared with a reference FR-MOX fuel during fuel fabrication and reprocessing. Content of neptunium, americium, and curium is increased by 2.5 percent compared with its content in the reference FR-MOX fuel. Based on the data in Table 4.1, neptunium does not present serious and unmanageable concerns for either fuel fabrication or reprocessing. Operations involving americium present greater difficulty during fuel fabrication because of the strong γ emission from neptunium-239 (daughter of americium-243), necessitating the use of shielded hot cells with remote handling systems to fabricate the fuel. Fuel fabrication is also difficult with curium recycling because of the high energy γ emissions from curium-243 and -244, and the neutron emissions from curium-244. In general, as seen in Table 4.1, the impacts of these minor actinides are much less significant for reprocessing than for fuel fabrication, because fission products dominate radiation exposure during reprocessing (EPRI, 2010b).

Heterogeneous recycling: Heterogeneous recycling involves loading minor actinides in special targets that are separate and distinct from the main driver fuel. This method requires selective element-specific separations (extractions) of minor actinides to place the preferred minor actinides in specific targets. To accomplish transmutation, these special targets containing relatively high concentrations of minor actinides (up to ~20 percent) are placed in the reactor to exploit excess neutrons produced by the main driver fuel. Due to the higher content of minor actinides, helium generation from alpha decay could be an issue affecting the behavior of the special targets during irradiation. The heterogeneous recycling method provides flexibility in that the location and residence times of the minor actinide targets in the core can be optimized for maximum transmutation.

Fabricating, reprocessing, and recycling of special minor actinide–containing targets are complex technical activities and would need additional development and demonstration to reach industrial-scale deployment. One strategy being considered for heterogeneous recycling that avoids the buildup of higher actinides is the complete burning of a target in a single cycle, rather than multirecycling, on a timescale similar to the fast reactor assemblies through the core. Such a deep burning strategy would require the use of advanced fuels and materials not yet available that are capable of withstanding high neutron doses (displacements per atom of >200) and harsh irradiation conditions for much longer periods of time than for typical fast reactor fuel assemblies. The downside

TABLE 4.1 Impact of Partioning and Transmutation of Minor Actinides on Fuel Fabrication and Reprocessing

Actinide Content of FR Fuel → (comparison with reference FR-MOX fuel) 2.5% Np 2.5% Am 2.5% Cm
Decay heat × 1 × 4 × 12
Fuel Fabrication γ dose × 4 × 80 × 500
Neutron source × 1 × 2 × 1,700
Decay heat × 2 × 3 × 6
Reprocessing γ dose × 1 × 1 × 1
Neutron source × 1 × 4 × 8

NOTE: FR = fast reactor; MOX = mixed oxide.

SOURCE: Adapted from Zaetta et al. (2005).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

of this strategy is that if the deep burn is not successful (100 percent consumption of the target material), then the spent target fuel will become a waste problem (NEA-OECD, 2012).

Both homogeneous and heterogeneous recycling have advantages and disadvantages related to the reprocessing flow sheet, fuel fabrication and handling, and reactor operation and costs. Like all nuclear fuels that are to be recycled, fuels containing minor actinides must be robust with respect to the containment of fission products during and after irradiation and compatible with the capabilities of the reprocessing system used. Heterogeneous recycling tends to be more expensive than homogeneous recycling because it requires remote handling in more heavily shielded facilities for the fabrication of special target minor actinide fuel (NEA-OECD, 2018b).

4.3.5.4 Management of Minor Actinides: Conclusions

Implementation of advanced fuel cycles with plutonium recycling and the P&T of minor actinides requires the development of new actinide separation schemes that, in theory, may offer a wide variety of options for both material reuse and waste reduction. Plutonium and the minor actinides are produced in nuclear fuel under irradiation in either thermal or fast neutron spectra, and the buildup of each isotope is a strong function of burnup. As a result, spent nuclear fuel subjected to different burnup conditions can result in large isotopic variations (i.e., large differences in the ratios of isotopes of plutonium, neptunium, americium, and curium) that can have very different nuclear properties (Forsberg and Greenspan, 2003). Managing the minor actinides, in particular, is technically complex and depends on the specific objectives of the P&T strategy employed. The P&T objectives define which actinides are to be transmuted and lead to a large array of options, including the following:

  • Partitioning:
    • Choice of reprocessing technology (e.g., aqueous- or nonaqueous-based, selective-element specific or a group separation of elements)
  • Transmutation:
    • Method of recycling (homogeneous or heterogeneous, and if heterogeneous, single or multicycle)
    • Fuel fabrication technology (e.g., need for shielded hot cells equipped with automation and remote handling systems)
    • Fast reactor technology (e.g., sodium-cooled, gas-cooled, lead-cooled, or molten salt reactors) (e.g., the combination of reprocessing and the fast reactor technology chosen for transmutation)

As was concluded in the 1996 National Research Council report Nuclear Wastes: Technologies for Separations and Transmutation (National Research Council, 1996), the bottom line is that implementing a fully closed, multirecycle strategy of plutonium, and P&T of the minor actinides, requires a long-term and sustained commitment involving construction and operation of large, complex reprocessing facilities, equally complicated fuel fabrication facilities (because of higher levels of radioactivity), and transmutation devices such as advanced fast reactors. To realize the expected benefits of maximum use of uranium resources and a significant reduction in the waste source term due to the minor actinides requires the successful deployment and operation of not one, but all of these technologies (reactors, reprocessing, and fuel fabrication facilities). Aggressive pursuit of such a program is unrealistic to consider at this time, as it is inconsistent with current U.S. national security policies and does not have favorable economics. It would be of interest only in the context of a sustained national or international program with a time horizon of a century or longer.

4.3.6 Processing Advanced Reactor Fuels

Over the past several decades, research around the world has focused on the development of a vast array of novel separation processes to support various options for a closed fuel cycle. For aqueous processes (which are by far the most mature), ongoing R&D has focused on more efficient, simple (potentially single-cycle), and environmentally friendly processes that use salt-free aqueous and organic phases and employ highly selective ligands that contain only carbon, hydrogen, oxygen, and nitrogen (Baron et al., 2019). This R&D has further focused on

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

leveraging the current aqueous reprocessing experience by (1) allowing processing streams to be easily passed from one process step to the next with minimal process adjustments and (2) utilizing the same operationally tested liquid–liquid extraction contactor technology (mixer-settlers, pulsed columns, centrifugal extractors) in current use.

In addition to aqueous processes, nonaqueous (pyroprocessing) processes are also being investigated to handle fuels with higher plutonium content and shorter cooling times. The main countries participating in these efforts are the United States, Russia, the European Union, China, Japan, and the Republic of Korea (IAEA, 2008; NEA-OECD, 2018a). Several reviews published on the technical readiness levels (TRLs) of the various separation processes note that their technological maturity varies considerably (Baron et al., 2019; Collins et al., 2014; Joly and Boo, 2015; NEA-OECD, 2018a). Most separation processes are in the proof-of-principle range (TRLs 4–6), and still others are only at the proof-of-concept stage (TRLs 1–3). In addition, other processes are being developed that show promise for reducing waste by, for example, improving solvent and reagent recycling or recovering and recycling cladding material, such as zirconium. It is too early in technological development to determine which specific separations and recycle methods are achievable and could improve the economics and waste generation associated with fuel reprocessing.

As described previously, plutonium recovered from spent LWR oxide fuel can be recycled as mixed oxide fuel in LWRs (LWR-MOX). Reprocessing the spent LWR-MOX uses similar head-end and aqueous processes as spent uranium oxide fuel, with a couple of exceptions. During reprocessing of spent LWR-MOX, care must be taken to adjust the extraction process to account for a higher content of plutonium because of criticality concerns. Additionally, due to higher fuel burnup, the dissolution process may result in more intractable residues ending up as process losses that go to waste, although specific processes have already been developed in France to overcome this risk (Miguirditchian and Taylor, 2021; Miguirditchian et al., 2017). To date, ~70 MT of LWR-MOX fuel has been successfully reprocessed at La Hague (Todd, 2020).

Head-end processes would need to be developed for non–oxide-based fuels (e.g., nitride, carbide, metal, and metal-alloys) and TRISO fuel (see Appendix G). These advanced fuel types would have to be converted into standardized and convenient chemical forms compatible with subsequent aqueous or nonaqueous processing schemes, while minimizing process losses and efficiently capturing volatile and semivolatile off-gases containing fission products.

The concept of fuel reprocessing to support liquid-fueled and liquid-cooled molten salt reactors is somewhat different from that for solid-fueled reactors. Some molten salt reactors require an extensive chemical processing plant intimately coupled to the reactor. Head-end processes for liquid-fueled or liquid-cooled molten salt reactors differ in that they are used to prepare the spent nuclear fuel for subsequent use in molten salt reactors. Depending on the specific reactor type, fluorination or chlorination is used to convert spent oxide fuels to either fluoride or chloride salts. The head-end processes ensure the fuel salts are clean and free from impurities such as oxides, water, and potentially other oxidants (NEA-OECD, 2018b). The functions of chemical processing plants for reprocessing spent fuel salt for molten salt reactors are described in Section 4.3.6.5.

4.3.6.1 Advanced Reprocessing Strategies for Aqueous-Based Partitioning of the Actinides

Modifications to the aqueous reprocessing flow sheets for recycling of plutonium and partitioning the minor actinides are under development with the intent of preventing or significantly reducing the risks of nuclear proliferation. As a first step, these efforts have focused on separation schemes that do not produce a pure stream of plutonium that could be diverted for weapons use (DOE-NE, 2007). For example, the COEX process, developed by the French program, exploits evolutionary changes in the PUREX reprocessing flow sheet to address proliferation concerns. In this process, uranium and plutonium are processed together to produce a mixed solid solution of (U, Pu)O2 and avoid having a pure plutonium product stream at any time during reprocessing (Paviet-Hartmann et al., 2011). Such coprocessing of uranium and plutonium also provides some advantages for the fabrication of mixed oxide fuel for recycle (Castelli et al., 2009). Variations of COEX are also possible where the coconversion can be done with either neptunium or americium ending up in the mixed oxide fuel—(U, Pu, Np)MOX or (U, Pu, Am)MOX—to be recycled (Drain et al., 2008). Solvent extraction processes, such as GANEX (Grouped ActiNide EXtraction), based on altogether different extractants are also being developed by the French program

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

with the goal of recovering all of the transuranic elements from a solution of dissolved spent fuel. Heterogeneous recycling, on the other hand, requires the development of even more advanced aqueous separation processes that add complexity to the reprocessing plant, as described below.

Assuming that neptunium can be managed as part of uranium/plutonium COEX or a similar process (e.g., UREX developed in the United States), the remaining waste stream contains americium, curium, and fission products. Americium, curium, and lanthanide fission products all exist as trivalent ions in solutions, making the chemical separation of americium and curium from the fission product waste stream challenging. Two-step processes have been developed to achieve such separations, such as DIAMEX-SANEX in France and TRUEX-TALSPEAK in the United States. In the first step, americium, curium, and the lanthanide fission products as a group are isolated from the other fission products; a second, more difficult, separation step recovers americium and curium as a group from the lanthanides. Individually separating americium from curium is yet an even more difficult task but potentially desirable, as removal of curium reduces the heat load and neutron source term during the fabrication of special americium transmutation targets. The French program has demonstrated the feasibility of the EXAm process for the isolation and heterogeneous recycling of americium (Poinssot et al., 2017a). The curium fraction is allowed to decay in storage, and after about 100 years, the decay products (mostly plutonium) can be either disposed of or recycled as mixed oxide fuel.

One of the most important technologies used in fuel reprocessing is the conditioning or processing of the waste prior to disposal. Currently, vitrification is the technology of choice for aqueous-based reprocessing of uranium oxide fuel, as it defines the resulting waste form in terms of its chemical composition (degradation behavior under near-field repository conditions) and its volume (limited by the heat-loading fraction in the specific glass composition). Conditioning technologies appropriate for advanced fuels will have to be developed and qualified if advanced reactors are deployed in the future, considering that borosilicate glass has already been demonstrated to be compatible with the high-level waste stream coming from the advanced reactors fuels reprocessing (CEA, 2009). Of particular note is the progress made over the past three decades by the French (CEA) R&D nuclear waste management program, which is studying potential benefits of aqueous fuel reprocessing to reduce both long-term volume and toxicity of wastes. The French program has established the “feasibility of homogeneously or heterogeneously recycling of the minor actinides in a Gen IV fast neutron reactor” using hot cell demonstration tests with actual solutions of spent fuel; however, to make these processes operational at commercial scale, more R&D is required (Poinssot et al., 2017b). Table 4.2 lists some of the more mature aqueous processes organized by element separated and the motivation for their development. Included in the table is the separation of cesium-137 (half-life ~30 years) and strontium-90 (half-life ~29 years), which along with their decay products, account for more than 90 percent of the decay heat from all of the fission products. If cesium and strontium were to be separated, they would likely be stored and managed separately from the repository.

4.3.6.2 Advanced Reprocessing Strategies for Nonaqueous, Pyroprocessing-Based Partitioning of the Actinides

Nonaqueous, or pyroelectrochemical, processes used for recycling spent nuclear fuel rely on refining techniques23 conducted in molten chloride (or fluoride) salts at elevated temperatures (500–900°C). Countries currently engaged in pyroprocessing R&D programs include France, India, Japan, the Republic of Korea, Russia, and the United States (IAEA, 2021d). There are two primary pyroelectrochemical processing methods: (1) electrorefining, in which an electrical current of sufficient voltage separates elements in a molten salt by anodic dissolution concurrently with metal deposition at the cathode, and (2) liquid–liquid reductive extraction, or separations based on the selective reduction and extraction of a metal from a molten salt phase into an immiscible liquid metal phase containing the reductant (NEA-OECD, 2018b; Rodrigues et al., 2015). Both electrorefining and liquid-liquid reductive extraction can be used for either metallic or oxide fuels if an appropriate head-end treatment is added to the flow sheet.24 Electrorefining of oxide fuel requires that the oxide fuels are first reduced to the metallic state, which

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23 Refining techniques are methods developed for reducing impurities in a substance, in this case a metal.

24 This paragraph was modified after release of a prepublication version of the report to correct details concerning electrorefining and liquid-liquid reductive processes.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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TABLE 4.2 Summary of Some of the More Mature Aqueous Separations Under Development to Support Various Fuel Cycle Options

Category of Separation Element(s) Separated/Recovered Example Process(es) Motivation
U-only U GANEX (1st cycle), NEXT UREX Recover remaining fissile U for possible reenrichment
Major actinide corecovery U and Pu separately PUREX—Industry standard Recover fissile U for reenrichment and Pu for recycle as MOX fuel
U/Pu or Pu/Np as a group COEX, UREX, UREX+ Aid MOX fuel fabrication and recycle
U/Pu/Np as a group TBP extraction in NEXT process U/Pu/Np mixed oxide fuel recycle, send Am/Cm to waste with lanthanides and other fission products
Transuranic actinide separation Pu/Np/Am/Cm GANEX (2nd Cycle) Group extraction of transuranic elements for homogeneous fast reactor transmutation
Lanthanides/actinides as a group DIAMEX, TRUEX Remove trivalent lanthanides and actinides from PUREX raffinate
Lanthanides/actinides separation leaving Am/Cm as a group DIAMEX-SANEX, TRUEX-TALSPEAK Avoid high neutron absorbers in fuel (La-Tb and Y) allowing heterogeneous recycling of Am/Cm as a group; lanthanides sent to waste
Am from Cm EXAm Heterogeneous recycle of Am without the fuel fabrication problems related to decay heat and SF neutron issues with Cm
Cs/Sra separation Cs/Sr as a group Cs Treat, Sr Treat Separately manage short-term heat generators outside of a deep geologic repository; Cs, Sr Treat are decontamination processes in use for Cs/Sr-contaminated equipment

a Cs-137 and Sr-90 with their daughter decay products Ba-137 and Y-90 alone provide over 90 percent of the decay heat from all fission products.

SOURCES: Baron et al. (2019); NEA-OECD (2018b); Okada (1985); Rodríguez-Penalonga and Moratilla Soria (2017); Vandegrift et al. (2004).

can be done by chemical or electrochemical methods, before being subjected to the normal pyroelectrochemical processes used with metallic fuel.25 To date, no commercial-scale pyroprocessing facilities have been built, but two chloride salt–based, pyroelectrochemical metallurgical processes have been implemented at pilot “engineering” scales: (1) electrowinning26 of oxide fuels at RIAR in Russia and (2) electrorefining of metals, pioneered by Argonne National Laboratory and ongoing since 1996 for the electrometallurgical treatment of metallic spent EBR-II reactor fuel27 (National Research Council, 2000; NEA-OECD, 2018b).

Other nonaqueous methods, such as those based on fluoride (and chloride) volatility, have also been investigated and are being considered for online fuel reprocessing in molten salt reactors (NEA-OECD, 2018b; Pereira, 2020). Also see Section 4.3.6.4 for a discussion of fluoride-based volatility methods for molten salt reactors.

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25 Prior to reduction of uranium oxide fuel, an additional process step called “electrowinning” (see next footnote) can be carried out to significantly reduce the volume of material to be treated, as is being developed in Japan (NEA-OECD, 2018b). In this step, uranium dioxide is electrolytically dissolved from an anode basket and reelectrolyzed as “pure” uranium dioxide on the cathode. The pure oxide is converted to metal, typically by an electrochemical reduction step, and then further processed using electrorefining.

26 Electrowinning is a generic term for electrodeposition of a species from an electrolyte solution onto a cathode and is accomplished by applying a potential across two electrodes to separate the species based on its reduction potential.

27 The initial kg-scale electrometallurgical treatment operations on spent EBR-II fuel were carried at Argonne-West, which later became part of the Idaho National Laboratory.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

4.3.6.3 Electrometallurgical Treatment of Sodium-Bonded Spent Nuclear Fuel

Argonne’s electrometallurgical treatment (EMT) used to treat EBR-II fuel28 is described briefly to illustrate pyroprocessing of metallic uranium spent fuel.29 Key to the EMT process is the electrorefining step. Broadly, electrorefining involves the anodic dissolution of metallic fuel into a molten salt electrolyte (typically, an LiCl-KCl eutectic mixture) followed by, depending on the electrode material used and the cell potential, the electrodeposition of uranium only onto the cathode. The driving force for the separations is the free energy of formation (ΔGfº) of the various metal chlorides in the spent nuclear fuel. When a constant controlled current is allowed to flow between the anode basket containing chopped fuel elements (including the stainless-steel cladding) and the cathode, the uranium, plutonium, transuranic elements, alkalis, alkaline earth, and rare earth (lanthanide) metals oxidize and dissolve into the molten salt as cations. Left behind in the anode basket are elements such as the noble metals (small ΔGfº, least stable chlorides) that do not dissolve.30 The alkalis, alkaline earth, and rare earth (lanthanides) elements form very stable chlorides (large ΔGfº) that are easily oxidized but not easily reduced, so they remain in the salt. The actinides (intermediate ΔGfº) are efficiently transported in the molten salt to the cathode. By controlling the electrode material and redox potential, metals can be reduced and deposited on the cathode. In the case of the EMT process, the anode and cathode potential are controlled to allow only the reduction of U3+ to uranium metal at the cathode. Because salts tend to be hygroscopic and metals tend to be pyrophoric, pyroelectrochemical (electrorefining) processes must be carried out under an inert atmosphere in which the levels of oxygen and water are tightly controlled, adding another level of complexity to an industrial-scale pyroelectrochemical processing plant.

A schematic of the EMT processes used to recover relatively “pure” uranium is shown in Figure 4.11. This process uses a steel cathode to separate “pure” uranium. Further purification of the uranium occurs in a cathode processer and involves the removal of residual salts by distillation. The final step in the process is to cast the uranium metal product into an ingot for storage. A variant of the process can be used to recover the transuranic elements. After most of the uranium has been deposited, the steel cathode is removed and replaced with a liquid cadmium cathode (LCC) on which the transuranic elements and remaining uranium (and some residual lanthanide fission products) can be codeposited at the desired cathodic potential. The adhering salt and cadmium can be distilled, leaving the actinides to be further processed and remotely fabricated into fuel to be recycled.

Two main waste streams are generated from the EMT pyroprocess. The first consists of the stainless-steel cladding hulls, which contain any unoxidized material, including fission products, and the noble metals (i.e., undissolved solids) remaining in the anode basket. After removing the adhering salt from these materials, some Zr is added to form a lower-melting alloy. The resulting material is cast into a Fe/Zr metal waste form. Periodically, some electrorefiner salt-containing transuranics, some residual uranium, alkali and alkaline earth fission products, and rare earth (lanthanide) fission products are removed and mixed with zeolite. (Methods are available to recover the uranium and transuranic elements from the salt prior to its disposal).31 Upon heating, the salt reacts with the zeolite to form sodalite (e.g., zeolite + sodium chloride forms sodalite). Finally, glass is added, and the mixture is heated and undergoes pressureless consolidation to form a glass-bonded sodalite ceramic waste form. Like aqueous reprocessing systems, pyroprocessing-based systems must manage fission product gases such as xenon and krypton, as well as fuel assembly hardware that becomes activated during irradiation. Typically, the atmosphere control system for a pyroprocessing facility would be used to collect the fission product gases released during decladding and other head-end processes; activated fuel assembly hardware would be separated and compacted or melted to reduce volume for disposal most likely as high-level waste (Frank et al., 2015; IAEA, 2008; National Research Council, 1996, 2000; Williamson, 2020). For fuels with zircaloy cladding, because of the large volumes involved, other pyroprocesses are being developed to potentially recover and reuse zirconium.

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28 EBR-II driver fuel was approximately 65 percent–enriched in uranium-235, metallic fuel alloyed with ~10 percent zirconium, depleted uranium blanket fuel, and stainless-steel cladding. Sodium was used to bond the fuel to the cladding for better heat conduction.

29 This sentence was altered after release of a prepublication version of the report to remove incorrect information about pyroelectrochemical processes.

30 Noble metals demonstrate outstanding resistance to oxidation even at high temperatures. Although the group is not defined strictly, it usually includes the metals of groups VIIb, VIII, and Ib of the second and third transition series of the periodic table: rhenium, ruthenium, rhodium, palladium, silver, osmium, iridium, platinum, and gold. On occasion, mercury and copper are considered noble metals.

31 Electrowinning methods can be used to recover the uranium and transuranic elements remaining in the salt after electrorefining (Choi and Jeong, 2015).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.11 Schematic of pyroelectrometallurgical treatment used for metallic fuel for the Experimental Breeder Reactor II. Block diagram with (left) and without (right) the cadmium electrode. “Pure” uranium in the diagram denotes uranium that is relatively pure but requires additional purification steps to reach the desired purity.
NOTE: FP = fission product; TRU = transuranic element.
SOURCE: National Research Council (2000).

Efforts to pyroprocess sodium-bonded highly enriched uranium metallic fuel from both EBR-II and Fast Flux Test Facility (FFTF) using the EMT process have continued at INL in concert with Argonne National Laboratory since the end of the demonstration program in 2000 (Patterson, 2021; Williamson, 2020). Valuable lessons have been learned guiding further innovations throughout the EMT processes. Examples include the following:

  • Process—elimination of alloy-forming liquid metal cadmium cathode in favor of a nonalloying solid metal cathode for (co)-recovery of uranium and transuranic elements; staggered-batch continuous-electrorefiner operations.32
  • Processing hardware—scraped cathode to collect and consolidate recovered uranium more efficiently; new multifunctional furnace to reduce bottlenecks, improve efficiency, and provide redundancy in distillation and casting steps.
  • Facility design—integration of facility, process, and safeguards, including increased automation.
  • Safeguards—in situ process monitoring using voltammetry, input accountancy, and inventory modeling.
  • Waste forms—promising iron-phosphate glasses with higher waste density.

Idaho National Laboratory also sees continued processing of EBR-II driver fuel as a near-term solution to the supply of HALEU for some of the advanced reactor concepts; see Section 4.2.3.1 for some of the challenges and limitations of this approach.

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32 This bullet point was corrected following prepublication release of the report to refer to (co)-recovery rather than recovery.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

4.3.6.4 Other Nonaqueous Pyroprocesses

In addition to pyroelectrochemical methods (e.g., electrorefining and reductive extraction), nonaqueous processes based on fluoride or chloride volatility have been investigated (Uhlíř and Marecek, 2009). Halogenation with F2 gas generates volatile hexavalent fluoride compounds of U, Pu, or Np that are then captured, condensed, and reduced to the tetrafluoride species for recycle.33 Volatility methods are being considered in combination with electrochemical methods to process34 salts in molten salt reactors (MSRs) (Pereira, 2020). Other processes under development include ion exchange, melt crystallization, vacuum distillation, oxidative precipitation, phosphorylation, and dehalogenation, to name a few (IAEA, 2021d). Many of these processes are also being evaluated for separating and recycling MSR-type salts (Riley et al., 2018, 2019). Most of these processes are in early stages of development and require considerably more R&D (Baron et al., 2019).

Another dry process for recycling spent fuel is the voloxidation process (or Atomics International Reduction Oxidation [AIROX]), which uses only gaseous and solid materials. In this process, fuel rods are punctured to expose the spent fuel and allow O2 to react with the UO2 in an argon atmosphere at high temperature to form U3O8. The U3O8 expands, increasing the rupture of the cladding and pulverizing the fuel. H2 gas is then used to reduce U3O8 back to UO2. This oxidation-reduction cycle is repeated two times or more to achieve the desired particle size distribution. During the process, volatile fission products (Kr, Xe, and I) along with tritium are released and trapped, while the more refractory fission products (rare earths, lanthanides, and actinides) remain in the fuel. The resulting UO2 fuel can be blended with enriched U or other recycled fuel, sintered into new fuel pins, and reclad into new fuel. During the sintering process, some semivolatile fission products, such as Cs and Ru, are also released (Majumdar et al., 1992). General Atomics is considering voloxidation (AIROX), or a variation of the process just described, as a part of an advanced fuel cycle in the future for recycling spent fuel from their Energy Multiplier Module (EM2) gas-cooled fast reactor design. (Back and Schleicher, 2021).

Using pyroprocess separations technologies to partition spent nuclear fuel has several advantages (IAEA, 2021d). The nonaqueous medium used for pyroprocessing consists of inorganic salts, typically alkali (Li, Na, K) or alkaline earth (Mg, Ca, Sr, Ba) metal chloride or fluoride salts, which become liquid (molten) at relatively high temperatures (between 500 and 900°C). Where radiolysis-induced degradation of organics and water is a problem for aqueous processing in a high-radiation field, molten ionic salts have no structure in solution and thus do not decompose or otherwise degrade under high-radiation fields. This allows shorter cooling times between reactor discharge and reprocessing, as well as the potential to process higher-burnup spent nuclear fuel. Because these molten salts do not contain elements such as H or C that act as effective neutron moderators, the potential for critical accidents during processing is much reduced compared with aqueous processing. Furthermore, because molten salt systems, especially chloride-based systems, have a sufficiently high cross section for neutron absorption, they can tolerate feeds with higher fissile content compared with aqueous processes. This allows for more compact processing equipment to be used, which, in turn, leads to smaller footprints for facilities that use pyroprocessing compared with those that use aqueous separation technologies. An additional advantage of this smaller size is that the facility can more easily be collocated with reactors. Many MSR developers are likely to choose pyroprocessing for on-site recycling for this reason. Collocation also avoids some of the transportation issues for spent nuclear fuel, and existing physical protection systems and engineered safeguards can be leveraged across the entire plant site. From a nonproliferation viewpoint, collocation of spent nuclear fuel storage, reprocessing, and fabrication of recycled fuel—all within one protected facility—is thought to be a more effective safeguards strategy compared with performing partial separation of actinides and fission products (using, e.g., UREX and COEXTM) at a centralized reprocessing facility.

Pyroprocessing also has disadvantages compared with aqueous reprocessing. A significant drawback for pyroelectrochemical processing is that it is inherently a batch process, which makes scale-up difficult. Lack of continuous processing capability limits the throughput of a process and, unless redundant or duplicative process equipment is in use, the overall process throughput will be hampered by single-point failures of the individual

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33 Actinide elements heavier than Pu (e.g., Am and Cm) do not form hexafluoride compounds.

34 In processing, fluoride volatility is used in combination with other chemical/redox methods that recover the nonvolatile or less-volatile fluorides, such as adsorption/condensation or reductive extraction. Electrolysis, where the desired separations are achieved electrochemically, is an alternative method to halide volatility.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

unit processes. Additionally, pyroelectrochemical processing lends itself to grouped recovery of the uranium and transuranic elements because of their similar redox potentials. As a result, the selectivity and product purity are relatively low (for fission products, separation factors are ~1,000 times smaller) compared with aqueous processes (IAEA, 2021d). Additionally, compared with the decades of experience with material accountancy and process monitoring for aqueous reprocessing, nuclear safeguards and security for pyroprocessing systems remain in their infancy (Coble et al., 2020). However, R&D based on electrochemical and spectroscopic techniques show promise for process monitoring and control, and for supplementing nuclear material accounting methods (Williamson, 2020). Lastly, to achieve the overall benefit of P&T using pyroprocessing technology, actinide losses to waste must be small (<0.1 percent). The committee was unable to find information on material recoveries and decontamination factors now being achieved by the EMT process, but past experience shows relatively poor and less efficient separations from pyroelectrochemical processing. In that case, a more extensive actinide drawdown operation by electrolysis or chemical reduction before passing the process salt on to waste form production steps would be needed to minimize actinide losses.35

4.3.6.5 Molten Salt Reactor: An Obvious Application of Pyroelectrometallurgical Technology

Liquid-fueled and liquid-cooled MSRs are unique in that the molten salt acts as both the fuel and the heat transfer fluid, and the reactors are adaptable to a wide range of fuel cycles (Hombourger et al., 2019). As mentioned in Chapter 3, MSRs can be designed to use an array of actinide fuels (thorium, uranium, plutonium) and operate over a wide range of neutron energies from thermal (with graphite used as a moderator) to fast (Holcomb et al., 2011). They can be operated as breeders (to produce startup fissile materials for other reactors), converters (no net production of fissile material), or burners (requiring periodic addition of fissile material); they can use a one-fluid system with the fissile and fertile materials in the same fluid or a two-fluid system—a more complex design with separate fluids for fissile and fertile materials. During MSR operation, the salt(s) flow(s) between the reactor core and heat exchanger(s) located external to the core where the heat is extracted and converted to steam for power production.36 Some MSRs are designed to operate with fuel processing to remove fission products, while others run within a sealed environment where no fissile material is removed, and only denatured low-enriched uranium (by the addition of uranium-238) is added (LeBlanc, 2010). MSRs tend to be self-regulating because of the thermal expansion that causes the fission reaction rate to decrease with increasing temperature (e.g., large negative temperature coefficient of reactivity). However, MSRs pose a number of challenges, such as a higher-radiation-field environment leading to potential first-wall issues, more difficult maintenance and inspection operations, the use of unconventional methods of fissile material accountancy, and fewer barriers to radionuclide release. On the other hand, MSRs require no fuel fabrication in the traditional sense, as required by solid fueled reactors; avoid in-core materials limitations, such as that exhibited by cladding at high burnup; and have minimal excess reactivity, the ability to breed, and the potential to burn transuranic elements (Holcomb, 2015).

The choice of the fuel salt for an MSR is constrained by several factors, some of which include (1) relatively low melting temperature (typically around 500℃ or below); (2) for thermal MSRs, a low capture cross section; (3) for fast MSRs, a low scattering cross section; and (4) a relatively high solubility for actinides. Since actinide elements are known to form chemically and radiolytically stable salts with the halides, and halide salts have an acceptably low parasitic neutron absorption, they are good candidates for use as both fuel salts and coolants in MSRs. As a result, much of the early work on MSRs (e.g., the Molten Salt Reactor Experiment at Oak Ridge National Laboratory) employed halide salts and in particular, fluoride salts, which has led to a more extensive literature available for fluoride than for chlorides salts. However, different properties can be exploited depending on the specific MSR design. Whereas the use of fluoride salts results in softer neutron spectra, the heavier halide (chloride) provides a harder neutron spectrum, which tends to favor breeding and burning of minor actinides. MSRs for actinide burning will favor the use of molten chloride salts because of their higher actinide solubilities.

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35 This sentence was altered after release of a prepublication version of the report to clarify treatments needed before processing.

36 Several options are being explored in addition to steam turbine generation of electricity via the conventional Rankine cycle, including advanced heat exchangers using other heat transfer fluids such as secondary molten salts or a gas such as helium (Brayton cycle gas turbine).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

There are other constraints on the choice and composition of fuel salts. For molten salt chloride fast reactors, enrichment of chlorine-37 is needed not only to reduce neutron absorption and enhance breeding but also to limit the production of the long-lived (3 × 105 years) chlorine-36 isotope, which is problematic given that chlorine is readily soluble in water. Currently, chlorine-37 enrichment is not performed at commercial scale anywhere in the world (Napier, 2020). On the other hand, tritium production and containment are important issues for thermal MSRs that use fluoride salts containing lithium, such as 2LiF-BeF2 (FLiBe), since at elevated temperatures, tritium can permeate structural alloys (Holcomb, 2017). Neutron reactions on lithium-6 (which is 7.6 percent abundant) within the fuel salt generate copious amounts of tritium (1 Ci tritium/MWth [megawatts thermal]/day), some of which would be released to the cover gas during normal operations (McFarlane et al., 2020; Sorensen, 2021). By using enriched 7LiF (≥99.99 percent), the neutron absorption problem can be mitigated to a large extent. (See Box 4.2 for details on current industrial uses of lithium-7.)

Fission products in MSRs pose a challenging problem in that, depending on their half-lives (some of which are very short), as well as their chemistry and volatility, they can be found almost anywhere within the reactor volume and require different treatment methods. Volatile noble gas fission products (krypton and xenon) will escape the salt and end up within the reactor cover gas. Since many fission products have krypton or xenon precursors (decay daughters cesium, barium, rubidium, and strontium), a significant fraction of cesium, strontium, and even iodine can end up in the off-gas, so managing the volatile effluents from an MSR is an important design consideration. Off-gas treatments, such as those involving in situ helium sparging, can be used to separate, capture, and store (for decay) the volatile fission products (e.g., continuous removal of krypton and xenon fission products) (Riley et al., 2019). The more noble metal fission products, which plate out as metals on reactor components and end up as particulates in the salt, can be filtered out physically, as they tend to come to the surface during sparging. Most of the fission products, as well as the actinides (including the lanthanide fission products), remain dissolved in the salt. For MSRs that require processing the fuel salt to remove fission products (e.g., MSRs breeders using thorium-232), a fraction of the salt can be removed and processed online or in batches to remove fission products and, if breeding thorium to uranium-233, to isolate protactinium.

The fuel processing operations are expected to be carried out in what amounts to a chemical processing facility (CPF) directly attached to the reactor (Fredrickson et al., 2018). The CPF is appropriately sized to support reactor operations and serves both as a reprocessing facility for MSR fuel and a means for managing the chemistry of the reactor’s molten salt. Performance of the reactor and the CPF are intimately and complexly coupled for MSRs, so a key function of the CPF is to ensure that the properties of the salt are appropriate for optimal reactor performance. Salt processing is important for controlling such impurities as oxygen and water in the salt, as well as the redox potential of the salt, to limit corrosion (Frederickson et al., 2018). Furthermore, fuel salt chemistry in MSRs is complex, and its composition (e.g., quantities of fissile and fertile isotopes; fission, transmutation, radiolysis, and corrosion products) is continually changing with time. Actinide management in MSRs can take a variety of forms depending on the reactor design (Forsberg, 2007; Forsberg and Greenspan, 2003). Theoretically, the actinides could remain in the fuel and be burned (as planned for Terrestrial Energy’s IMSR-400 with no salt processing or the net breed-and-burn molten chloride fast rector being developed by TerraPower), or they could be isolated in the CPF and reintroduced later as fuel (as planned by developers of ThorCon’s thermal thorium-converter reactor, Flibe’s thermal liquid-fluoride breeder reactor, and Moltex’s stable salt-waste burner reactor) (Delpech et al., 2009; Holcomb et al., 2011; Hombourger et al., 2019). As mentioned in Section 3.2.3.5, a 2021 EPRI report noted that salt lifetimes in fast chloride MSRs may be limited by the accumulation of fission products in liquid fuel molten salts over time unless steps are taken to clean up the fuel salts of fission products (EPRI, 2021a). For example, fuel salt processing would be useful for removing parasitic neutron-absorbing fission products, maintaining the desired salt chemistry, managing corrosion potentials, and ensuring the actinides remain soluble.

The wide range of MSR design concepts each has numerous variants with their own potential advantages and challenges. Basic elements of fuel cycles that might support MSRs have been identified, and some technologies have been demonstrated to some level of success. However, it is too soon in the development of MSRs to discuss and analyze their associated fuel cycles. If and when MSR designs become more mature, their associated fuel cycles will become more obvious, as they will depend on the objectives of reactor operation (e.g., power production, breeder with online salt processing, actinide burning).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

4.3.7 Analyzing the Main Motivations for LWR-Inspired Advanced Fuel Cycles

Historically, the two main drivers for developing advanced fuel cycles that build on LWR technology have been better utilization of natural resources and, at a later time, reduction in the long-term radiotoxicity of the resulting waste, which comes primarily from contained transuranic elements (TRU) (see Box 4.3 on radiotoxicity).

Table 4.3, adapted from EPRI (2010b), shows several potential fuel cycles and their expected impact on natural uranium consumption and on the mass of TRU waste being eventually disposed of in a geologic repository. Five fuel cycles were selected for illustration in Table 4.3:

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
  • LWR—Once-through fuel cycle (OTC): uranium dioxide reference cycle
  • LWR—Monorecycling of plutonium in the form of MOX, followed by disposal of the spent MOX fuel in a geologic repository
  • LWR—Multirecycling of plutonium in the form of MOX fuel with enriched uranium (MOX-EU)
  • LWR + FR (fast reactor)—Multirecycling of plutonium in an FR in the form of MOX-FR, with partitioning of americium and curium, once-through transmutation of americium, and 100-year storage of curium
  • FR—Multirecycling of plutonium in the form of MOX-FR with partitioning and homogeneous transmutation of transuranic elements

These fuel cycles were selected first on the basis of their improved plutonium management in LWRs and second on their management of minor actinides (neptunium, americium, and curium) by introducing fast reactors and integrating partitioning and transmutation technologies in the fuel cycle.

The analyses were performed using a steady-state approach, that is assuming that all reactors operate at constant power and all mass flows have reached an equilibrium. Importantly, the equilibrium phase has to be preceded by a deployment phase. The deployment phase ends when the heavy metal inventory of the fuel cycle has built up to the point where the fuel composition has reached equilibrium. The deployment phase lasts many decades for a transmutation strategy using multirecycling (National Research Council, 1996; NEA-OECD, 2006b).

Natural uranium consumption: A fleet of all PWRs (100 percent) operating in a once-through fuel cycle is used as a reference and assigned the relative value of 1 for natural U consumption. Assuming the same amount of electrical energy generation, the natural U consumption rate of the other fuel cycles is a fraction of the reference PWR fleet value and varies from slightly lower than 1 for various recycles of Pu in PWRs to <0.01 for a fleet of fast reactors operating in the breeder mode in which all TRUs are continuously recycled. Calculated values for such a cycle yield a natural U consumption of 0.036 with no U recycle and 0.004 with U recycle (NEA-OECD, 2006b). Table 4.3 illustrates that fast reactors are required for extending the use of natural U resources by a significant factor.

TRU content going to the repository: This exercise assumes that the TRU masses going to the repository are the result of a 1-GWe plant operating at 100 percent capacity for 1 year. For the once-through cycle, all TRUs are from spent UOX fuel. For the monorecycle option, TRUs are from reprocessing (~15 percent) and spent MOX fuel (~85 percent). For the multirecycle option of Pu in PWRs, TRUs are from reprocessing. For the scheme involving fast reactors and partitioning and transmutation, most of the TRUs are coming from reprocessing losses.

Table 4.3 shows that the waste management benefits of Pu monorecycling in LWRs, when assuming disposal of the spent MOX fuel in a geologic repository, are limited. Most of the waste management benefits accrue over time when fast reactors are part of the fleet of nuclear power plants.

TRU cycle inventory: Also shown in Table 4.3 are the in-process and in-reactor TRU inventory in the equilibrium fuel cycle at any given time, assuming a single 1-GWe plant operating at 100 percent capacity. Recycling of fuels leads to high in-pile and out-of-pile TRU inventories.

There is an inverse correlation between the TRU mass flow going into the geologic repository and the TRU mass in the fuel cycle. In other words, positive benefits for waste management become negative attributes when reactor operations (licensing challenges), fuel fabrication (remote operation versus glove-box operation), and reprocessing (TRU partitioning, radiological protection due to increase in TRUs, criticality safety) are considered. Closing the fuel cycle with partitioning and transmutation of the TRUs in fast reactors eliminates, with the exception of process losses, their disposal in a geologic repository. However, it is done at the expense of handling, irradiating, and storing large TRU inventories in close proximity of places where people live (public acceptance issues). As further discussed in Chapter 5, if all TRU is directly disposed of in a geologic repository, the majority is not expected to make it to the biosphere because of multiple barriers—all of which allow time for decay and retard or prevent dispersion.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

TABLE 4.3 Comparison of (1) Natural Uranium Consumption Compared with the Once-Through Cycle; (2) TRU Mass Going into a Geologic Repository; (3) TRU Mass in the Equilibrium Fuel Cycle; and (4) Requirements for Advanced Reactors and Facilities for Five Different Fuel Cycle Optionsa

Nuclide LWR Once-Through Cycle (OTC) LWR Monorecycling of Pu LWR Multirecycling of Pu LWR + FR Multi- (Pu/Np) & Mono- (Am/Cm) Recycling FR Multi (Pu/MA) Recycling
Reactor Fleet PWR (100%) PWR (100%) PWR (100%) PWR (44%) and FR (56%) FR (100%)
Natural U consumption (compared with OTC)
1 0.89 0.87 0.44 <0.01
TRU Content Going to Repository Assuming 0.1% Loss in Separation Processes [kg/year]
Pu 230 153 0.37 2.10 1.25
Np 16.2 16.6 14.4 0.02 0.0066
Am 6.35 16.2 39.4 0.35 0.055
Cm 3.3 8.11 19.7 2.06 0.013
Total TRU 256 194 74 4.53 1.32
TRU Cycle Inventory (reactor + fabrication + reprocessing) [kg]
Pu 767 3,285 4,818 10,293 17,520
Np 53 131 116 241 88
Am 22 88 307 438 701
Cm 11 44 158 263 175
Total TRU 853 3,548 5,399 11,235 18,484
Requirements for Advanced Reactors and Facilities
None required None required ALWRsb and reprocessingc Fast reactors and advanced FBR fuel reprocessing Fast reactors and advanced FBR fuel reprocessing

a Values in this table were derived from the information in (NEA-OECD, 2006b); values were calculated by the French Alternative Energies and Atomic Energy Commission (CEA) and documented in “Synthèse des Résultats des Recherches sur l’Axe” 1, 2005. Small differences between the NEA and CEA reports are due to the different assumed values for some of the input parameters, such as reactor efficiency, burnup, and storage times. For all fuel cycles involving LWRs, the same burnup of 60 GWd/MTHM and efficiency of 34 percent were assumed. In all cases, reprocessing losses sent to waste were assumed to be 0.1 percent.

b Advanced LWRs, such as the third-generation European Pressurized Water Reactors (EPRs), licensed for handling fuel with high Pu content, and/or LWRs with a higher moderator-to-fuel ratio compared with existing LWRs.

c Compared with the current situation in France, where only low-enriched U fuels are recycled, new facilities may be required for processing Pu-rich spent fuel at the required industrial throughput.

NOTE: ALWR = advanced light water reactor; FBR = fast breeder reactor; FR = fast reactor; LWR = light water reactor; MA = minor actinides; PWR = pressurized water reactor; TRU = transuranic element.

SOURCE: Adapted from EPRI (2010b).

Requirements for advanced reactors and facilities: The need for advanced, not-yet-available facilities required for fuel cycle operation is an indication of technological challenges and a potential detriment to economic competitiveness. As stated in Chapter 2, the U.S. Energy Information Administration notes that renewable energy incentives and falling technology costs support robust competition in the electricity mix (EIA, 2021b). Advanced nuclear technologies will have to prove that they can be cost competitive for providing energy in a carbon-constrained world (MIT, 2018).

4.3.8 Long Timescales to Implement LWR-Inspired Advanced Fuel Cycles

The establishment of an equilibrium fuel cycle for advanced schemes involving multirecycling can take a long time. For example, the 1996 National Research Council report on separation and transmutation showed that some advanced fuel cycles require several centuries of sustained operation before achieving the very low transuranic

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

mass rates in a repository (National Research Council, 1996).37 In addition, an MIT analysis of alternative fuel cycles for nuclear power growth scenarios through 2100 concluded that fuel cycle transitions take 50 to 100 years [emphasis added] (MIT, 2011). Also importantly, the simplest form of the existing LWR fuel cycle (i.e., the once-through fuel cycle) is still incomplete in the United States, as there is no geologic repository. Its completion is expected in Finland, Sweden, and France during this decade or the next; thus, it will have taken about 60 to 75 years, in the best cases, to complete the once-through LWR fuel cycle. In the United States and several other countries, completion of the once-through LWR fuel cycle will likely require more than a century. Based on the French experience, implementing the LWR-inspired fuel cycle involving fast reactors fueled with plutonium, if it is indeed pursued to completion, cannot be expected prior to the end of this century. Any fuel cycle involving processes that require administrative controls and public/policy support for time periods longer than several decades will be challenging to implement.

4.3.9 Implications of Using Multirecycling with the P&T Strategy and Fast Reactors to Reduce Transuranic Wastes and Repository Heat Load

Does multirecycling using P&T offer advantages for waste repositories? Notably, only transmutation strategies with fully closed fuel cycles can reduce transuranic (TRU) waste by 100-fold. Partially closed fuel cycles, such as the multirecycling of Pu in LWRs, are easier to implement but cannot achieve high TRU reductions. The other potential advantages for multirecycle P&T considered here are (1) decreasing the high-level waste burden in terms of the inventory of long-lived TRU radionuclides as well as the time required to safely sequester the waste and (2) reducing the heat load of the wastes. Whereas closing the fuel cycle makes recycling fuel (i.e., reprocessing and fuel fabrication) much more difficult, removing the actinides, which includes uranium, from the waste stream makes handling and managing of the waste in a repository somewhat easier and less complicated, because only about 3 percent of the mass of the spent nuclear fuel, consisting of the fission products, goes to the repository.

4.3.9.1 Radiotoxicity Reduction

Reducing both the volumes and radiotoxicity and shortening the time required for waste storage are the primary current motivations for developing P&T scenarios. Discussions of the impact of the radiotoxicity of actinides in radioactive waste for various P&T scenarios are commonly found in the literature. Often, the radiotoxicity of waste from spent fuel is compared with that of uranium ore as a reference point and called “relative radiotoxicity” to avoid the use of the dose unit. The comparison to uranium ore, as pointed out by Piet (2013), is not a regulatory concept but is used for reasons such as (1) the concept is easier to explain to a nontechnical audience; (2) it seeks to compare the hazard as greater or lesser than natural uranium existing in the environment38; and (3) useful comparisons can be made if no site-specific dose assessments are available. However, some of the radiotoxicity calculations are incorrect because they do not include the decay products of all isotopes in the waste that either captured neutrons or did not fission while in the reactor. See Box 4.3 for details on radiotoxicity’s use and applications.

Magill et al. (2003) reported on the impact of P&T scenarios on the radiotoxicity of actinides in radioactive waste. In this work, the efficiency of P&T was calculated for an initial inventory of actinides (for fuel with 4.2 percent enrichment, 50 GWd/MT burnup, and 6 years cooling time) in grams per ton of spent PWR fuel through

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37 See for example, Figure 4-1 in National Research Council (1996), which plots the “transuranic (TRU) ratio,” y(t), as a function of time of transmutation operation. The TRU ratio, y(t), is defined as the ratio of the total inventory of transuranics sent to waste disposal as a function of time “for the reference once-through LWR fuel cycle (no fuel reprocessing, no recycle, and no transmutation) to the total inventory of TRUs at time t in the transmuter itself (i.e., fast reactor), in its fuel cycle, and in process wastes” (National Research Council, 1996). Assuming an advanced fuel cycle based on fast reactors (advanced light water reactors) operating “at constant power for 100 years and then terminated,” the TRU ratio would reach “a value of 6.9. Reducing the TRU inventory by only a factor of 6.9 below that of the reference once-through fuel cycle is far from the goals proposed for transmutation” (National Research Council, 1996). Based on a linear extrapolation of the plot in Figure 4-1, achieving y(t) = 20 or 70 would require ~300 years or 1,000 years, respectively, based on a fast reactor fuel cycle operating at constant power.

38 As pointed out by Piet (2013), “there is a sense (sometimes implicit) that if one can be adequately protected against natural ore hazards, then it is possible to be adequately protected against waste with comparable (or lower) radiotoxicity.”

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

all successive burnups, reprocessing and recycling, and sums all of the reprocessing losses at each step. These data were used along with effective dose coefficients to calculate total actinide ingestion radiotoxicity as a function of time by three independent research groups: ITU and FZK in Germany, and CEA in France.39 The agreement among the groups on the actinide mass inventories was better than ±10 percent for U and Pu, better than ±20 percent for Am isotopes, ±25 percent for the lighter Cm isotopes (242 and 244), and ±70 percent for the heavier Cm isotopes (245 and 246). The groups agreed on radiotoxicity with time, with small differences unnoticeable on a logarithmic plot.

Figure 4.12 shows these results for the evolution of ingestion radiotoxicity in a geologic repository as a function of time for several cases.

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39 ITU, Institute for Transuranium Elements, Karlsruhe, Germany, FZK. Forschungszentrum, Karlsruhe, Germany, and CEA-Cadarache, DER/SPRC/LEPh, France.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

As seen from Figure 4.12, the radiotoxicity of spent nuclear fuel in a geologic repository will remain above the reference level for (1) 130,000 years for the once-through fuel cycle; (2) between 500 and 1,500 years for high-level waste after P&T, depending on assumptions regarding P&T efficiencies; and (3) 270 years when all actinides are removed in a fully implemented P&T scenario in which only fission products are sent for geologic disposal. For case (3), 270 years represents a lower limit, as process losses will inevitably occur throughout the fuel cycle; therefore, no fuel cycle can be considered perfectly closed. This exercise is meant to demonstrate that if P&T could be implemented at the required efficiencies, it could have an impact on the time required to safely sequester the high-level waste in a geologic repository. Magill et al. (2003) caution that “the reprocessing of spent nuclear fuel inevitably results in the production of secondary waste” (e.g., contaminated resins, fuel hulls and end pieces), and the management of such waste and its impact on geologic disposal require further detailed study.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Image
FIGURE 4.12 Evolution of ingestion radiotoxicity as a function of time for an initial inventory of actinides for fuel with 4.2 percent enrichment, 50 GWd/t burnup, and 6 years cooling time. The radiotoxicity reference level (horizontal dashed line) is set at 7.83 MT of natural U, or the amount of natural U required to produce 1 MT of fresh fuel enriched to 4.2 percent 235U.
NOTE: P&T = partitioning and transmutation.
SOURCE: Magill et al. (2003). Courtesy of the Nuclear Institute.

As pointed out in Box 4.3, radiotoxicity is only one of several parameters that can be used to evaluate the impact of different fuel cycle scenarios on repository performance. Any discussion of radiotoxicity regarding waste only considers the consequences of release of radioactive material in the absence of a repository. Chapter 5 discusses how the design of a repository, including the choice of the geologic setting and engineered barriers, can significantly mitigate the potential release and migration of certain elements away from the repository.

The actual hazard posed is a stronger function of the mobility of the radionuclides in and around the geologic repository than the radiotoxic inventory of the radionuclides in the repository. For example, the radiotoxicity due to the actinides when in a repository with reducing conditions is minimal since their mobility in the biosphere is much less compared with that of some long-lived fission products that exist as anions (Grambow, 2008). Reducing the minor actinide inventory in a chemically reducing geologic repository is therefore only important in the case of the scenario of inadvertent or accidental intrusion.

4.3.9.2 Heat Load Reduction

Fully closing the fuel cycle using fast reactors—that is, by multirecycling of Pu and P&T of the minor actinides—can reduce the heavy metal mass by more than three orders of magnitude compared with the once-through fuel cycle. But what is the impact of the heat load on the repository if the actinides are removed? The residual heat power associated with spent nuclear fuel follows essentially the same trend as radioactive decay, being driven by short-lived fission products for the first 50–70 years and then by the alpha-emitting actinides (Pu, then minor actinides) at later times. Because of their decay characteristics, Pu and the minor actinides represent the largest contributor to the heat load in a geologic repository, particularly at longer times (>100 years). Although the total residual heat from the actinides in spent nuclear fuel is higher than that from the decay of fission products,

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

there is little if any temperature rise expected in a repository at later times because of the long-half lives of the actinides; their rate of heat release is much lower and released over a much longer time period (>100,000 years).

In contrast, the relatively short half-lives of fission products provide a significant thermal pulse and temperature rise in the repository early on (from emplacement to roughly 300–400 years, peaking around 40–50 years) as their heat is released over a much shorter period (a few hundred years). It is the thermal load along with the thermal conductivity of the geologic setting that determines the temperature rise in a repository. Temperature increases may be able to be managed by appropriate spacing of spent nuclear fuel or high-level waste canisters, but the amount of decay heat per disposal canister will be limited depending on the backfill and the local geologic environment of the repository.

Because the repository temperature (e.g., heat load), not the waste volume, is a key factor in repository design, a properly designed repository will have a disposal density (canister spacing) that manages the thermal pulse from the fission products from spent nuclear fuel, and doing so should be well within a conservative envelope to handle the late thermal load from the actinides. As mentioned previously, recycled fuel, such as spent mixed oxide fuel, has a higher heat content than spent uranium oxide fuel. Direct disposal of spent mixed oxide could largely negate any benefit from reprocessing with regard to decreasing decay heat per unit electrical energy generated and incur costs for long-term storage of spent mixed oxide outside of a repository (CBO, 2007). Factors that could influence the total costs of this option are the costs of disposing the low-level wastes, including GTCC, generated from reprocessing and any engineering or management strategies deployed to handle the higher heat load.

4.3.10 Observations on the Back End of the Fuel Cycle

When considering advanced fuel cycle options, the question to be asked is, What has changed in the United States that requires a reevaluation of the decision not to reprocess spent fuel?

In 1996, at the request of DOE, the National Academies convened a committee of experts to evaluate state-of-the-art separations technology and transmutation (S&T) systems (National Research Council, 1996). As introduced in Chapter 1, the principal recommendations from that committee were as follows:

  • “None of the S&T system concepts reviewed eliminates the need for a geologic repository. DOE should continue efforts to develop a geologic repository for spent LWR fuel.
  • The current policy of using the once-through fuel cycle for commercial reactors, with disposal of the spent fuel as HLW, should be continued.
  • Fuel retrievability should be extended to a reasonable time (on the order of 100 years) to avoid foreclosing alternative fuel strategies that may be in the national interest.
  • Research and development should be conducted on selected topics to support the cost-effective future application of S&T of commercial spent fuel and separations for defense waste applications.” (National Research Council, 1996)

The report went on to say:

A sustained, but modest, and carefully focused program of research and development over the next decade could prepare the technical basis for advanced separation technology for the radionuclides in spent LWR fuel and for decisions on the possible applications of S&T as part of the more efficient future use of fissionable resources. The research and development effort should focus on the factors that strongly influence fuel-cycle economics, especially the costs of reprocessing spent LWR fuel, minimalization of long-lived radionuclides to secondary wastes in the reprocessing cycle, and on the need to minimize the possible increase in proliferation risks that could result from the commercial use of plutonium in recycle fuels. (National Research Council, 1996)

The report also found that “The construction and operation of an S&T system would require, in addition to several new types of facilities, the resolution of major institutional, public policy, and public acceptance issues” (National Research Council, 1996). It further concluded that, should these issues be overcome, “An S&T system

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

of appropriate scale must be operated for many decades to achieve the permanent benefits to the repository and other parts of the nuclear fuel cycle” (National Research Council, 1996).

The report also concluded that implementing P&T to reduce substantially the amount of long-lived minor actinides that would be placed in a geologic repository necessitates a commitment to this approach for more than a century. Such an effort would require “a cohesive national intent and commitment” on the part of the U.S. government, else S&T operations could be shut down early and “facilities related to recycling (i.e., reprocessing and fuel fabrication plants) are likely to be a loss” (National Research Council, 1996).

It has now been 26 years since that report was issued, but its conclusions remain even more valid now than before, given the diminishing contributions of LWR technology to electricity generation in the United States. The availability and price of uranium have remained stable and relatively inexpensive, so there is no compelling reason at this time to deploy reprocessing technologies and advanced reactors on the sole basis of economic considerations (NEA and IAEA, 2020).40 The once-through fuel cycle remains the safest and most cost-effective option in the short term. However, this option hinges on the siting, construction, and operation of a geologic repository.

4.4 COST ESTIMATION OF DIFFERENT FUEL CYCLE OPTIONS

The committee was tasked with examining the potential costs of the different nuclear fuel cycles required for advanced nuclear reactors that could be commercially deployed by 2050. A number of published reports and papers have addressed the economics of nuclear power, including fuel cycles (e.g., Black and Peterson, 2018; Black et al., 2019; Bunn et al., 2003; EPRI, 2007, 2009b, 2010a,c,d; Huff, 2019; MIT, 2011; NEA-OECD, 1994, 2006b, 2013; Recktenwald and Deinert, 2012; Rothwell et al., 2014; Schneider et al., 2009; Shropshire et al., 2021).

In the 2011 MIT study “The Future of the Nuclear Fuel Cycle,” three fuel cycles were modeled in detail: (1) once-through LWR fuel cycle; (2) monorecycling of Pu in LWRs with direct disposal of recycled MOX spent nuclear fuel; and (3) a closed fuel cycle using reprocessing where spent uranium and TRU are recycled back to the fast reactors operating with different conversion ratios (CRs). These fast reactors and their CRs are an actinide burner (CR <1), a self-sustaining converter with CR = 1, and a breeder reactor with a CR >1 with excess TRU used to start additional fast reactors. For that study, the levelized cost of electricity (LCOE) was the measure of cost, since LCOE is the “constant price that would have to be charged in order to recover all of the costs expended to produce the electricity, including a return on capital” (MIT, 2011). LCOE was divided among the following main components: cost of the front-end fuel cycle (cost of raw U and conversion, enrichment, and fuel fabrication); reactor capital costs and nonfuel operating and maintenance (O&M) costs; and the cost of the back end of the fuel cycle.41 The MIT (2011) study authors assessed:

  • For the once-through cycle, the reactor capital costs (81 percent) and the reactor (non-fuel) O&M costs (9 percent) dominate the LCOE with front-end and the back-end costs coming in at 8 percent and 1.6 percent, respectively. For closed fuel cycles with fast reactors, the reactor capital and O&M are expected to be an even higher portion of the LCOE.
  • The “most important conclusion from a comparison of the LCOE across the three fuel cycles is the differences between them are small relative to the total cost of electricity” with the highest being the fast reactor cycle and the lowest the once-through cycle (with less than 3 percent range from the highest to the lowest).

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40 Information on fissile content in the discharged fuels from advanced reactors using HALEU was not provided to the committee, so consideration of recovery and recycling of uranium-235 would be only speculative at this time. Most advanced reactor designers are opting for a once-through cycle as their initial approach.

41 For the once-through fuel cycle, the back-end fuel costs included above-ground interim spent nuclear fuel storage, transportation, and the cost of disposal in a geologic repository. For the monorecycling fuel cycle, reprocessing, transportation, mixed oxide fuel fabrication, and high-level waste disposal costs were added, along with positive and negative credits for such things as fuel utilization and increased costs of spent mixed oxide disposal. For the closed fuel cycle, higher costs for reprocessing, fuel fabrication, and high-level waste disposal were added, along with similar positive and negative credits. In the case of the closed fuel cycle, a higher capital and O&M cost for fast reactors relative to LWRs was included.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
  • “The benefits to resource extension and waste management of mono-recycling in LWRs using mixed oxide fuel, as is being done in some countries, are minimal.”
  • “A conversion ratio near unity is acceptable and opens up alternative fuel cycle pathways” with different reactor choices and use of lower enriched uranium (<20 percent) rather than highly enriched U or Pu, eliminating the need to reprocess LWR spent nuclear fuel for closed fuel cycle startup.
  • “The most important fact to keep in mind in considering any estimate of the cost of alternative fuel cycles is the high degree of uncertainty about key components of each cycle. First, there is uncertainty about the cost of disposing of the high level wastes from each cycle. Second, there is great uncertainty about the cost of reprocessing spent fuel and the cost of fabricating the recycled fuel. Third, there is enormous uncertainty about the construction and operating costs for fast reactors, which are at the core of many alternative fuel cycles.”

Furthermore, the MIT study cautions: “A second elusive factor that can play a large role in the economic calculations is the cost of capital (discount rate)” (MIT, 2011).

A similar study by the NEA-OECD was undertaken to “assess the available knowledge from different countries on the costs of the various options for the long-term management of spent nuclear fuel and, to the extent possible, compare the cost estimates of different countries on a common basis” (NEA-OECD, 2013). Because of major differences between countries (e.g., discount rates and government subsidies), “a direct cross-country comparison of SNF/HLW management costs was not deemed feasible,” but instead simulations of a generic, idealized system operating on the same three fuel cycles were carried out with specific cost input data provided by member countries (NEA-OECD, 2013). Many of the same conclusions as from the MIT study resulted from the NEA-OECD (2013) modeling:

  • Total fuel cycle costs were lower for the once-through fuel cycle, increasing successively through the monorecycle and closed fuel cycles with the difference being small compared with the LCOE.
  • “Cost estimates for future facilities, including repositories, entail many uncertainties, which will only be reduced as experience is gained in implementing the necessary infrastructure.” Factors that dominate the relative cost estimates are uranium price, reprocessing cost, and fast reactor capital cost premium. “Overall, the uncertainties related to the full recycling option remain the largest since only sparse data are available for these systems and no commercial system is in current operation.”

Notably, relatively recent studies looked at the cost of current generation reprocessing using open data for the Thermal Oxide Reprocessing Plant (THORP) facility (Recktenwald and Deinert, 2012; Schneider et al., 2009). Recktenwald and Deinert (2012) concluded, “The analysis suggests a total life-cycle cost of 2.11 ± 0.26 mills/kWh, with a 90% and 99% confidence that the overall cost would remain below 2.45 and 2.75 mills/kWh respectively. The most significant effects on cost come from the efficiency of the reactor fleet and the growth rate of nuclear power. The analysis shows that discounting results in life-cycle costs decreasing as recycling is delayed. However, the costs to store spent fuel closely counter the effect of discounting when an intergenerational discount rate is used.” This result is important because it shows that the 1 mill/kWh nuclear waste fee that had been levied in the United States would be insufficient to cover even the costs of conventional reprocessing. Advanced aqueous separations that partition the higher actinides typically operate on the PUREX raffinate. These would be additions or adaptations to the conventional PUREX process and would likely only increase reprocessing costs.

The committee notes that the 2014 DOE report on the Nuclear Fuel Cycle Evaluation and Screening Study provides a cost range based on that study’s simulations:

Alternatives to the current U.S. fuel cycle in the promising Evaluation Groups require R&D to bring the enabling technologies up to the level of successful engineering demonstration including pilot-scale facilities, which the Study results indicate as requiring several billion dollars over 10-25 years. Similarly, further development up to the first-of-a kind commercial facilities would require an additional several billion dollars. Any transition to a new fuel cycle would take decades to achieve, although some fuel cycle performance benefits such as wastes destined for deep

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

geologic disposal would accrue more quickly. Fully deploying an alternative fuel cycle would likely require several hundred billion dollars or more, comparable to the cost of continuing with the current U.S. fuel cycle as new reactors replace existing reactors. (Wigeland et al., 2014)

Importantly, the committee notes that “the promising Evaluation Groups” are in fact closed fuel cycles with multirecycling: “EG23—continuous recycling of U/Pu in fast reactors,” “EG24—continuous recycling of U/TRU with fast reactors,” and “EG-30—continuous recycle of U/TRU in both fast and thermal reactors”42 (all cases are with new natural uranium fuel).

Like all such studies, the specific methodologies vary, and there are many important assumptions made regarding the input data used to calculate costs, making cost comparisons challenging. A common conclusion reached across many cost and modeling studies is that, while spent fuel management represents a relatively small fraction of the total LCOE, differences in that area could result in large absolute costs depending on the size of the nuclear program and the duration of electricity generation.43

Studies such as those described can go only so far using economic simulations without real-world data to validate them. In the United States, more than 40 years have elapsed since the most recent commercial reprocessing activity. Despite the lack of recent actual commercial operating experience, other recent experiences in the United States shed light on the challenges of fuel cycles’ implementation. For example, Chapter 5 discusses the U.S. government’s less-than-stellar experience with large construction projects—specifically, the actual versus projected cost overruns (more than tens of billions of dollars) at the Hanford Waste Treatment Plant (vitrification facilities) and the canceled mixed oxide fuel fabrication facility at the Savannah River Site. Chapter 5 also provides information on the additional costs to the government when there are delays in opening geologic waste repositories; for example, about $600 million annually is being paid out of the Judgment Fund to utilities for costs of continued storage of spent nuclear fuel at nuclear power plant sites, since they cannot be stored at a disposal facility. Unlike simulations, delays and interruptions to actual fuel cycles can produce adverse effects such as backlog of spent fuel waiting for disposal and buildup of separated plutonium (as experienced by France, Japan, and the United Kingdom) when there are not enough (or none in the case of the United Kingdom) reactors using the material (see Table 2.1 in Chapter 2 that lists the holdings by country of civil separated plutonium).

In the early stages of its study, the committee became aware that cost data were considered proprietary by the reactor developers, especially with regard to reactor designs. Because the focus of this committee’s study is the supporting fuel cycles and not the reactors themselves, the problem was further exacerbated by the lack of information provided by reactor developers on fuel cycle components required to support their designs, with the exception of the need for HALEU for almost all proposed designs and facilities for fabricating HALEU-based fuels. In addition, because the primary criteria and functional requirements for specific fuel cycles to support advanced reactors are either not defined or defined at such a high level, the committee was unable to find sufficient, publicly available, third-party cost information to make reliable cost comparisons. As a result, the committee was unable to conduct a true economic analysis of specific fuel cycles to support various advanced reactor designs.

As discussed in Chapter 3, advanced reactor developers claim that small modular reactors—generating notionally less than 300 MWe with water- or non-water-cooled designs—offer many advantages due to smaller and simpler designs with shorter deployment schedules, compared with other advanced reactor designs. Among the advantages claimed are scalability; load-following flexibility; ease of remote siting; and, in particular, factory-built, modular construction with lower capital construction costs compared with those of large power reactors. If true, this potential cost advantage might make small modular reactors cost competitive relative to nonnuclear energy sources. While this type of analysis is the purview of the parallel National Academies study, the committee calls attention to this economic issue because it impacts the viability of small modular reactors for commercial deployment. Given that none of the proposed small modular reactors have yet to reach the operational demonstration

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42 The sentence was revised following a prepublication version of the report to correct quoted text from the cited source.

43 The potentially significant capital expenditures associated with management of spent nuclear fuel, such as construction of reprocessing and storage facilities, are amortized over a long period of time based on the period of electricity generation and waste storage. This limits the impact of large absolute costs on the annualized LCOE.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

phase, let alone commercial deployment, real cost data is obviously lacking. Commercial viability will depend on understanding whether there is an optimal size for a small modular reactor from an economic point of view and when the break-even point will be reached for the construction of an nth-of-a-kind reactor for a particular type of small modular reactor to become economically competitive. In other words, the learning curve for both small modular reactor construction costs and deployment needs to be understood. Of particular concern to this committee is the impact of small modular reactors on the fuel cycle with emphasis on waste management and disposal, whether once-through or closed fuel cycles are used.

To gain some insights regarding costs, the committee took a graded approach that reflected the reactor choices and fuel cycle options provided by the advanced reactor developers interviewed by the committee. Most of the nonwater-cooled advanced reactor developers expressed plans to use a once-through fuel cycle combined with direct geologic disposal for the foreseeable future with the option of transitioning to a closed fuel cycle at a later time; only a few advanced reactors expressed plans to close their fuel cycles upon initial deployment. See Table 3.1 in Chapter 3 for details for each reactor type and fuel cycle choice.

Because of its base technology similarity to commercial LWRs, the development of small modular integral pressurized water reactors (iPWRs) represents the least costly and most near-term path for developing and licensing a specific reactor technology to the point where assessing its economic viability would be meaningful. With the exception of a geologic repository, all of the fuel cycle facilities required to fuel and manage the waste via interim-to-indefinite storage for iPWRs already exist.

Demonstrating one or two promising nonwater-cooled reactors using a once-through cycle would require the addition of two new fuel cycle components—specifically, enrichment facilities capable of producing HALEU (up to 20 percent uranium-235) and fuel fabrication facilities capable of handling the higher-enrichment fuels. As with current enrichment facilities for low-enriched uranium, the new enrichment facility would likely be based on existing centrifuge technology. The associated cost of a separative work unit is expected to be greater than the costs for existing low-enrichment facilities that support the LWR fleet, in order to accommodate the higher level of enrichment; such a facility would need to conform to radiological and safety standards, including criticality required for Category II special nuclear facilities.

Currently, two Category I facilities (see Section 4.2.3.1 for more details) are licensed in the United States for greater than 20 percent–enriched material; these facilities could perform both the downblending of highly enriched uranium to produce HALEU and fabrication of HALEU-based fuels, although depending on the specific fuel type, some facility modifications may be necessary for fuel fabrication. Notably, for HALEU-based TRISO fuel, two companies, BWXT and X-energy, have already taken steps to manufacture this fuel type in the United States (see Section 4.2.4.1 for details).

Additional costs will likely result from the imposition of chemical, radiological, and criticality safety standards important for storage and transportation of HALEU-containing materials while onsite at fuel fabrication facilities. All of these changes will also be accompanied by increased costs related to a higher level of nuclear material control and accounting and physical security requirements. These requirements may make it difficult for current Category III facilities to transition to Category II without significant modifications and retrofits or substantial redesign of the existing facility. It might be most cost effective to design, construct, and license a new Category II facility for fuel fabrication, rather than to upgrade an existing Category III facility to Category II and amend its license. Additionally, critical infrastructure is needed to support R&D and fuel qualification for advanced reactors, such as material testing capabilities, as discussed in Chapter 3.

Higher levels of spending will be required to support fuel cycles for nonwater-cooled reactors using a once-through fuel cycle compared with iPWRs that also use once-through fuel cycles. This prompts the question, Who should pay for the infrastructure and fuel cycle facilities to support the development of advanced reactors to the point where their economic viability could be firmly established? Notably, when President Ronald Reagan lifted the ban on commercial reprocessing in 1981, he also stated, “It is important that the private sector take the lead in developing commercial reprocessing services” (Reagan, 1981). To date, as discussed in Chapter 2, commercial reprocessing has not been economically viable in the United States.

Advanced reactor concepts that aim to close the fuel cycle by recovering and reusing the fissile material in the fuel and/or breeding new fuel require construction and operation of fuel reprocessing and fuel fabrication

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

facilities, with the exception of some molten salt reactor designs. Reprocessing and recycling are highly dependent on the choice of the type of reactor technology. Reprocessing will likely be a costly addition to the fuel cycle, and notably, a single reprocessing technology will not support the wide array of advanced reactor designs. For example, developers of solid-fueled, gas-cooled fast breeder reactors propose to use different reprocessing technologies to close their fuel cycles than those of solid-fueled, liquid metal–cooled fast reactors. The former are planning on a cyclic, dry pyrochemical oxidation/reduction process for actinide recycling from uranium carbide or nitride fuels, while the latter are proposing to use a nonaqueous electrometallurgical process for actinide recycling from metallic uranium or uranium alloys.44 See Section 4.3.6.4 for more details of these processes.

Similarly, fuel fabrication facilities for the type of reactors just described will be different and use different technologies because the fuel types are different and will depend on whether (1) a homogeneous or heterogeneous approach is chosen for recycling the minor actinides and whether (2) blanket fuel is used for breeding. In any case, these facilities will require additional funding for more shielding and remote handling to ensure worker safety.

Molten salt reactors (MSRs) have very different fuel cycle requirements from those of liquid metal– and gas-cooled fast reactors operating on a closed fuel cycle. Section 4.3.6.5 describes fuel cycle processes relevant for MSRs, and the essential points are highlighted here to illustrate what is required for understanding these advanced fuel cycles. Liquid-fueled and -cooled MSRs are unique in that the molten salt acts both as the fuel and as the heat transfer fluid; no fuel fabrication is required. They are adaptable to a wide range of fuel cycles, given that breeding can occur using both thermal and fast neutrons. Because there is no fuel cladding to contain fission products, MSRs need to have off-gas treatment capabilities to continuously manage volatile effluents. For MSRs that operate with salt or fuel processing, what amounts to a chemical process facility (CPF) is attached directly to the reactor. Performance of the reactor and the CPF are intimately and complexly coupled for MSRs. The CPF serves as a reprocessing plant for MSR fuel and manages the chemistry of the reactor’s molten salt for optimal reactor performance. Some MSRs operating as breeders use a one-fluid system in which the fissile and fertile materials are in the same fluid, or a two-fluid system, which is a more complex design with separate fluids for fissile and fertile materials. Some MSRs designs have no salt processing while other designs are for breed-and-burn reactors. Because of the large number of MSR designs and supporting fuel cycle options, making cost estimates is impractical until one or two specific designs emerge and their fuel cycle requirements can be defined. The cost of fuel cycles to support advanced fast reactors or molten salt reactors will be specific to the reactor technology and will potentially exceed several tens of billions of dollars as the number of deployed reactor designs increases.

During this study, the committee recognized the important concept of trade-offs when assessing potential merits and viabilities of different advanced reactors and their fuels and fuel cycles. All of the advanced reactors first need to be built and operated for a sufficient time before their technical viability can be meaningfully assessed. As these advanced reactors move from conceptual design to engineering demonstration, the supporting fuel cycle requirements need to be defined concurrently to allow for a credible economic analysis. The U.S. government and industry going forward will have to decide which features or attributes of advanced reactors best align with U.S. energy needs without increasing proliferation risks, having an adverse impact on the environment, or imposing an unacceptable economic burden on current and future generations.

4.5 FUEL CYCLE SAFETY CONSIDERATIONS

The radiological safety of any nuclear fuel cycle operation or facility is ensured by (1) maintaining subcriticality; (2) appropriately containing the radionuclides involved; (3) removing decay heat and preventing autoignition (fires); and (4) radiation shielding. In addition, safety requirements for fuel cycle facilities are consistent with those of other industrial chemical facilities. This section provides a brief discussion of safety in the current once-

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44 Sodium-cooled fast reactors in the French program, such as Phénix and Superphénix, used uranium oxide fuels that were reprocessed using the well-known aqueous process, PUREX. The recovered plutonium was used to fabricate fast reactor mixed oxide fuel.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

through LWR fuel cycle and then describes in greater detail safety considerations associated with the deployment of advanced fuel cycles. Particular emphasis is placed on safety considerations for the operation of fuel processing facilities. Safety aspects of advanced reactor designs will be discussed in the parallel National Academies report Laying the Foundation for New and Advanced Nuclear Reactors in the United States.

4.5.1 Safety of the Current U.S. Nuclear Fuel Cycle

The mining, conversion, enrichment, fuel fabrication, and transportation of uranium oxide fuels and, to a lesser extent, mixed oxide fuels have occurred safely over the past 50 years. The industry and technology have developed to improve both operational efficiency and process safety in these facilities. Effluent releases of radioactivity from front-end processes are significantly below those of operating nuclear power plants (National Research Council, 2012). The highest safety risk is the release of gaseous uranium hexafluoride due to leaks or handling incidents. Normal discharges are relatively low and represent a low risk to the public.

The current open LWR fuel cycle in the United States is characterized by the use of oxide fuels and the absence of spent nuclear fuel processing. As discussed in Chapters 2 and 5, the lack of a geologic repository requires that spent nuclear fuel be stored at reactor sites throughout the country. The spent fuels currently being stored are chemically stable and maintain the fission products and actinides in a nondispersible form. Fissile material remains in a fixed geometry in which heat removal and subcriticality can be ensured throughout a variety of conditions and postulated events. Moreover, once spent nuclear fuel is moved to dry cask storage, the safety functions (prevention of criticality, protection from external hazards, decay heat removal, and confinement) are maintained through passive mechanisms. There is a long history of safe storage of these fuels. As such, there are essentially no environmental releases, and the risk of inadvertent criticality or release of radioactive material is low. Chapter 5 discusses the safety aspects of associated with storage and transportation in greater detail.

4.5.2 Safety Considerations for Advanced Fuel Cycles

The choice of fuels and fuel cycles for advanced reactors introduces additional safety considerations. In considering the safety advantages inherent to advanced reactors and their passive safety features, it is important to recognize that other processes introduced by the associated choice of fuel cycle may pose additional risks not present in the once-through uranium fuel cycle. As discussed further below, introducing processing into the nuclear fuel cycle increases the risk of a release of radioactive material into the environment due to the physical and chemical processes used in conjunction with dispersible radioactive material resulting from extended reactor operations. Recycled uranium has significantly higher gamma activity than natural uranium because of trace fission products and the presence of the decay daughters of uranium-232 (half-life = 1.9 years), which continue to build up to equilibrium in approximately 10 years. Recycled uranium also has higher alpha activity than natural uranium from the presence of uranium-234 and traces of plutonium and neptunium. In addition to increased radiation exposure from these decays, the decay products also generate additional heat in excess of that from natural uranium. If the feedstock also contains plutonium, additional gamma activity will result from the buildup of americium-241. The source term for reprocessed fuels will, thus, contain other radioactive isotopes and decay products that must be considered in the evaluation of on-site and off-site consequences. Fuel cycle facilities must implement safety features to protect workers from these hazards, as well as designs to minimize the risk to the off-site public and environment. Several presenters to the committee discussed the safety features of their advanced reactor designs, as well as the potential waste reduction and fuel utilization benefits of closing the nuclear fuel cycle through fuel reprocessing. However, no presenters discussed advances in the safety of processing facilities, such as passive cooling or inherently safe processing techniques, suggesting that the safety considerations have been limited to operational reactors with little attention paid to improving safety margins in the rest of the fuel cycle. In fact, by introducing reprocessing into a nuclear fuel cycle, it is not clear to what extent the safety benefits of new reactor technology are offset by the increased safety risk inherent to fuel reprocessing.

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

4.5.2.1 Safety Considerations for the Front End of the Fuel Cycle

On the front end of the fuel cycle, additional criticality controls are required for processing HALEU to manage risks in fuel fabrication. The potential impact of advanced fuel cycles on the safety of front-end processes results primarily from differences in the source material radioactivity and heat loading and the increased enrichments of HALEU fuel, which reduce margins to inadvertent criticality. Incidents that result in solidification must be prevented or shown to remain subcritical by geometry. An example of such an inadvertent criticality event occurred in 1999 at the JCO Fuel Fabrication Plant in Japan, when operators attempted to process uranium fuel at 18.8 percent enrichment using a technique they had successfully used previously to process uranium with 6 percent enrichment. It resulted in three workers suffering from acute radiation sickness, which was fatal in two of the cases. In addition, several other facility workers and members of the public received radiation doses, more than 100 members of the public were evacuated, and hundreds of thousands of residents of the prefecture were instructed to shelter in place for 18 hours because of the off-site release of radiation (IAEA, 1999a).

Moreover, the neutron emissions of even isotopes of plutonium, particularly plutonium-238, must be accounted for to reduce potential for criticality and limit radiation doses to workers. In addition to controlling the material’s amount and geometry, careful design of facilities should preclude the introduction of moderation and reflection, particularly from water ingress. As such, design and controls should minimize the risk from both internal (fire piping and other liquid sources) and external flooding.

4.5.2.2 Safety Considerations for Fuel Processing

On the back end of the fuel cycle, spent nuclear fuel processing introduces risk associated with complex industrial chemical or electrochemical processes that involve treating significant quantities of fissile material with large quantities of other hazardous material. Complex industrial chemical processes also introduce additional potential failure modes; for instance, reprocessing facilities can experience failures involving inadvertent criticality, gaseous and liquid leaks, fires, and adverse chemical reactions, as well as susceptibility to external hazards and prolonged loss-of-power events. Additionally, although processing operations are generally conducted at lower temperatures and pressures than reactor operations, they involve active handling of highly radioactive fissile material, fission products, and other material in dispersible form that is subject to physical and chemical (and often vigorous) processes, as shown in Table 4.1; the associated risks to site workers, off-site populations, and the environment must be properly managed. Improvements in technologies and processes have reduced these risks over time; for example, between 1985 and 2005, the average annual occupational radiation exposure at reprocessing facilities decreased from 10 mSv (millisievert) to 1.5 mSv per person (IAEA, 2005c).

Potential safety hazards associated with fuel reprocessing facilities include release of radioactive materials and accumulation of pyrophoric or highly reactive species. For example, mechanical processes at the head-end release nonretained gases, such as radioactive krypton-85 and iodine-129, and can accumulate pyrophoric metals (such as zirconium) from fuel shearing and cladding removal. Additional processes can release radioactive tritium, technetium-99, several iodine isotopes, carbon-14, and nitrogen-15. The separation of fissile material in many processes involves dissolution in nitric acid and chemical treatment of pyrophoric material, reactive chemicals, and flammable solvents. In addition to increased radiation exposure from the decay of isotopes (such as uranium-232, uranium-234, americium-241) inherent in reprocessed fuel, the decay products also generate heat in excess of that from natural uranium.

Throughout all of the processes, decay heat and other sources of self-heating must be effectively removed to ensure the chemical and physical form of various process streams are maintained within controllable and predictable bounds. The presence of fissile material requires deliberate management of material inventories and geometries in order to ensure that adequate margin to subcriticality is maintained. The accumulation of undesirable species in recycled reagents and high-level waste must also be carefully managed to avoid severe overheating. For example, in processes such as PUREX using concentrated nitric acid or heavy metal nitrates (uranyl or plutonium nitrate), an organic (such as tributyl phosphate) and kerosene can combine in a radiation flux to produce nitrated organics often referred to as red oil. If process and chemical conditions are not properly maintained, violent nitration oxidation reactions involving red oil can occur. Such red oil events have occurred (Savannah River Site in 1953

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

and 1975 and Tomsk, Siberia, in 1993) when organic solutions and kerosene used in PUREX processing mix with nitric acid at high temperature (>120°C). Modern fuel cycle facilities have effectively minimized the explosive risks from red oil, as well as hydrogen (from radiolysis) and other explosive gases resulting from operations, by controlling temperature and other process parameters within safe bounds and limiting buildup of susceptible material through control of the reaction rate and use of engineered controls.

Often, process and facility design require balancing competing safety considerations. For example, liquid mixing of fissile material reduces hazards associated with radiotoxic dust, but it requires additional attention in the design to prevent inadvertent criticality. Thorium fuel reprocessing is complicated by the presence of crystalline ThO2 and unirradiated PuO2, both of which are difficult to dissolve in nitric acid. Fluorine can be added to improve dissolution, but doing so can increase the possibility of leaks and vessel failures because of its corrosive nature and incompatibility with common materials in reprocessing equipment (IAEA, 2019e).

Because of the risks associated with processing spent nuclear fuel, a graded approach to defense in depth45 is applied to processing facilities, similar to that applied to the nuclear power plants they support (IAEA, 2017b). Processing facilities are properly sited to minimize risks of external hazards (fires, floods, and seismic events) and limit the off-site population and areas affected by potential radioactive releases. On-site risk is minimized through careful process and facility design and the use of codes and standards to ensure high quality of construction. Such design includes heavy radiation shielding and passive geometries to prevent inadvertent criticality. Facility design must include passive features, automated controls, and operators that maintain parameters within specific bounds during normal operations and in response to anticipated operational occurrences. For example, inert, leak-tight enclosures are used to minimize explosive risks and protect personnel from radiotoxicity associated with plutonium dust. Facility designs must include liquid and gaseous radioactive waste processing for controlling radioactive effluent releases incidental to operations within regulatory limits. In the event of an accident, additional engineered safety systems mitigate the effects, including multilevel confinement systems designed to prevent large off-site releases of radioactive material. Some of these safety systems, such as dynamic confinements and tank cooling systems, are active systems and rely on emergency power and other support systems to function. These systems are needed to ensure that chemical processes are safely shut down, nuclear material remains subcritical, hazardous concentrations of explosive or flammable gases do not accumulate, and radiological material is cooled and confined. Should an off-site release occur, the final layer of defense in depth ensures that off-site emergency response organizations are trained, equipped, and prepared to mitigate these consequences and protect public health and the environment.

Several large processing facilities have been safely operated throughout the world, but the risk associated with these facilities is real, and accidents have occurred46 (Bixler et al., 2017; IAEA, 1996).

  • Mayak Production Association, Russia (1957)—high-level waste tank cooling system failure led to a nitration oxidation reaction and an explosion (equivalent to 74 tons of TNT) and subsequent release of over 740 PBq (petabecquerel) of radioactivity.47 This event was classified as a Level 6, Serious Accident, the second most severe level on the International Nuclear Event Scale (INES).
  • Windscale Reprocessing Plant, United Kingdom (1973)—explosion and off-site release due to an exothermic reaction in a reprocessing tank. Classified as INES Level 4, Accident without Significant Offsite Risk.

Safe operation of reprocessing facilities in the United States (within the nuclear weapons complex for material production and West Valley Demonstration Project) and internationally has demonstrated that these risks can be managed safely (IAEA, 2005c). In order to fully realize the safety benefits associated with the deployment of

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45 As defined by the U.S. NRC, defense in depth is “An approach to designing and operating nuclear facilities that prevents and mitigates accidents that release radiation or hazardous materials. The key is creating multiple independent and redundant layers of defense to compensate for potential human and mechanical failures so that no single layer, no matter how robust, is exclusively relied upon. Defense in depth includes the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures” (U.S. NRC, 2021c).

46 For summaries and data of previous safety incidents at fuel cycle facilities, see IAEA (1996) and Bixler et al. (2017).

47 By comparison, the accident at Chernobyl was estimated to have released approximately 1,000 to 2,000 PBq (IAEA, 1986).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×

existing or advanced reactors, the design of the fuel cycle processes should seek to optimize safety over the entire fuel cycle, leverage international experience, and improve safety margins through implementation of passive and inherently safe facility and process designs.

4.5.2.3 Safety Considerations for Transportation and Storage of Radioactive Materials

While the type and design of facilities vary, the processing of spent nuclear fuel includes the risks associated with transport and storage of spent nuclear fuel awaiting processing and introduces additional safety considerations beyond those of an open fuel cycle. For example, while reprocessing reduces the inventory of radioactive and fissile material ultimately disposed in a geologic repository, it requires managing large inventories of these materials at processing facilities, as illustrated in Table 4.4. In addition to the risks associated with the processing of spent nuclear fuel, processing facilities will also serve as interim storage locations for spent nuclear fuel awaiting processing and for the storage of low- and high-level radioactive wastes resulting from reprocessing. The risks to spent nuclear fuel would be the same as when stored at individual reactor sites, but would occur in larger quantities, potentially putting more material at risk to external events.

Several vendors noted that they would be using sodium-bonded metallic fuel for initial demonstration or until they develop more advanced metallic fuel designs. Sodium is highly reactive and pyrophoric in both air and water even at relatively low temperatures. These reactions are exothermic and generate caustic sodium hydroxide and explosive H2, which presents challenges to qualification for storage both interim and in a deep geologic repository. Therefore, in the absence of innovative storage or disposal options not currently available, processing is required, even for a once-through fuel cycle, to avoid adverse chemical reactions or explosions, which—in combination with dispersible radioactive fission products, actinides, and other material—would result in significant environmental contamination. Special procedures and safety protocols must be implemented in processing sodium-bonded fuels and other sodium-wetted components. Sodium must be removed before working on materials, as even residual sodium will react with air and cause fires that generate caustic aerosols. Nitride and carbide fuels are also pyrophoric and require similar safety protocols (IAEA, 2007b).

4.5.2.4 Operational Experience: Effluents

The separation of fissile material during spent nuclear fuel processing also generates other waste streams. Some of these waste streams are captured in waste forms for disposal in low-level waste disposal facilities or a geologic repository, while others are released to the environment either in liquid or gaseous forms. Of particular concern for processing facilities is the release of 3H, 85Kr, 129I, 14C, 15N, and traces of alpha emitters including Pu.

85Kr, a noble gas with a 10.76-year half-life, represents the most significant release of gaseous radioactivity incidental to normal operations. Kr is released during head-end operations and throughout the separation stages. While bench-level methods for removing 85Kr have been explored, none have been demonstrated at an industrial level (Croff et al., 2008). Krypton is not, therefore, removed by off-gas treatment in industrial facilities. Because the quantities released exceed the practical capacity of current removal and retention technologies, retention is not practical, and 85Kr is released directly to the atmosphere (NEA-OECD, 2005).

Iodine, with 129I as the predominant isotope of concern, is scrubbed in off-gas treatment systems that involve passing it through a sodium hydroxide solution or capturing it in charcoal adsorber beds and filter trains. Internationally, iodine captured in sodium hydroxide has usually been discharged as liquid radwaste to the ocean. Design of reprocessing facilities in the United States has captured iodine in other media and retained for disposal.

Tritium (3H) is released to the environment, either directly via the liquid waste pathway or evaporated and released via the ventilation pathway.

In the United States, 40 CFR 190 provides federal limits for the total quantity of “radioactive material entering the general environment from the entire uranium fuel cycle.” Limits are provided per gigawatt-year of electrical generation for 85Kr, 129I, and 239Pu, and other alpha emitters. Table 4.4 below compares discharges from the La Hague facility in 2008 to calculated 40 CFR 190 limits and typical releases from U.S. nuclear power plants. The

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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most significant release from the La Hague facility is 85Kr, and such levels would challenge current U.S. regulatory limits should a similar facility be built and operated in the United States. Tritium releases from processing are comparable to those from a single nuclear power plant.

TABLE 4.4 Comparison of Radioactive Material Release Limits to Typical Discharges at the La Hague Reprocessing Facility and a Typical U.S. Nuclear Power Plant

40 CFR 190.10 Limit (per GWe-yr) 40 CFR 190.10 Limit at 45 GWe-yra,b La Hague effluent releases (2008)a Typical Nuclear Power Plant Effluent Release per Yearc
85Kr 1.85 × 106 GBq (50,000 Ci) 8.30 × 107 GBq 1.55 × 108 GBq Total Noble Gas 1 × 101 GBq to 1 × 105 GBq
129I 0.185 GBq (5 mCi) 8.32 GBq 6.76 GBq
239Pu and other alpha emitters 0.018 GBq (0.5 mCi) 0.81 Gbq 1.83 × 10–3 GBq
Tritium 4.64 × 104 GBq BWR: 1.1 × 103 GBq PWR: 1.85 × 104 GBq

a Data from Van der Stricht and Janssens (2010).

b Based on a total of 63.2 GWe of nuclear capacity in France in 2008 running at a 72 percent capacity factor.

c Data from National Research Council (2012).

Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Page 132
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Page 133
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Page 134
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Page 135
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
×
Page 136
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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Page 137
Suggested Citation:"4 Fuel Cycle Development for Advanced Nuclear Reactors." National Academies of Sciences, Engineering, and Medicine. 2023. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. doi: 10.17226/26500.
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The United States has deployed commercial nuclear power since the 1950s, and as of 2021, nuclear power accounts for approximately 20 percent of U.S. electricity generation. The current commercial nuclear fleet consists entirely of thermal-spectrum, light water reactors operating with low-enriched uranium dioxide fuel in a once-through fuel cycle. In recent years, the U.S. Congress, U.S. Department of Energy, and private sector have expressed considerable interest in developing and deploying advanced nuclear reactors to augment, and possibly replace, the U.S. operating fleet of reactors, nearly all of which will reach the end of their currently licensed operating lives by 2050. Much of this interest stems from the potential ability of advanced reactors and their associated fuel cycles - as claimed by their designers and developers - to provide a number of advantages, such as improvements in economic competitiveness, reductions in environmental impact via better natural resource utilization and/or lower waste generation, and enhancements in nuclear safety and proliferation resistance.

At the request of Congress, this report explores merits and viability of different nuclear fuel cycles, including fuel cycles that may use reprocessing, for both existing and advanced reactor technologies; and waste management (including transportation, storage, and disposal options) for advanced reactors, and in particular, the potential impact of advanced reactors and their fuel cycles on waste generation and disposal.

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