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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges CHAPTER 3 FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE Primary Issues: B1. Compare the uranium recovery by extraction plus (UREX+), the plutonium and uranium recovery by extraction (PUREX) process, and other processes being considered by the Russian Federal Agency for Atomic Energy for separation of fissile and other materials from spent or irradiated nuclear fuel. Consider the resulting waste streams and what can and should be done with these waste streams. B2. Compare the burn up and the number of cycles needed to reach an acceptable level of destruction of actinides in the conceptual advanced burner reactor proposed in the U.S. Global Nuclear Energy Partnership (GNEP) and in the Russian BN-600 and BN-800 reactors. COMPARING NUCLEAR OPTIONS: THE NEED FOR A SYSTEMS APPROACH The joint committees believe that a comparison to make choices among different fuel cycle options (reactors, fuel types and sources, spent fuel management, and processing) must use a systems approach. Such analyses would consider the entire life cycle of proposed nuclear energy systems, integrating assessments of fuel processing, fabrication, reactor design, and more. Only in this way can key trade-offs be made among different parts of the system. It is likely that the best technologies for processing spent fuel will be different depending on the specific reactors in which the processed materials will be irradiated, and the fuel fabrication approaches for them. In the U.S. case, for example, the Experimental Breeder Reactor II (EBR-II) Program was successful because fuel fabrication, reactor design, and spent fuel processing were done in an integrated way, making it possible to optimize choices for the system as a whole. Good decisions among different proposed processing-fabrication-reactor systems require clear, consistent, and well-thought-out criteria, based on justifiable system objectives. Picking a particular numerical target for some system characteristic (such as 99.99 percent purity for uranium separated from spent fuel) without careful analysis of the overall system benefits and costs of meeting that goal leads to poorly optimized systems. Building in assumptions or early decisions, such as a requirement for either a once-through or a closed fuel cycle or a particular reprocessing technology, allows a systems analysis to consider only variants of the already-chosen approach. A good goal would be an integrated reactor fuel cycle system that offers the best combination of economics, safety, security, proliferation resistance, environmental impact,
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges process operability, and sustainability, given the situation that exists for a nation at a particular time. In many cases some systems may offer more promise on some of these criteria, while others look better with respect to other criteria, making trade-offs inevitable. Whether more emphasis should be given, for example, to saving money or to reducing environmental impact is not a technical decision but one based on values, which must ultimately be made by society, through a political process. The role of designers and technical experts is to make clear the choices and trade-offs that need to be made, outline the benefits and downsides of each of the leading approaches, and do their best to ensure that the decisions ultimately made are well informed and carefully considered. Criteria for Comparison Each of the key criteria mentioned above can be specified in more detail, so as to provide more detailed guidance to those designing and assessing these systems. Economics. Each system can be compared based on its life-cycle electricity cost. Additional criteria may include the degree of uncertainty of those cost estimates; the system’s contribution to the costs of spent fuel and nuclear waste management; initial capital costs and the resulting level of financial risk in implementing and operating a system; the variability and reliability of the electrical output; and the system’s attractiveness or unattractiveness to the private sector (along with the scope of required government subsidies or regulations needed to make the system competitive). Safety. Each system can be compared based on the overall risk of a significant accident it poses (including both the probability and the consequences of the various types of plausible accidents in the system); accident reports by regulatory agencies and others can provide insight into risks. Radiation doses to the public and industrial safety during normal operations are also considerations, though these risks are low for most proposed systems. Because the risks of significant accidents may be difficult to estimate rigorously and compare among systems that have never been built, decision makers may choose to focus on the degree to which known risk factors are present and how they are addressed (such as positive coefficients of reactivity, which can result in power excursions), or the degree to which known safety factors are present (such as “passive safety systems”). Security. Thorough security comparisons would examine how difficult it would be for adversaries to cause a major radioactive release through sabotage, or through the theft of material that could be used to make a nuclear device. Systems that continuously maintain the nuclear materials in their cycle in forms that could not be used in weapons without either isotopic enrichment or extensive chemical processing using heavy shielding rank better on this criterion. Reactors with greater degrees of inherent safety and widely separated redundant safety systems so that they would be more difficult to sabotage simultaneously are also more inherently secure, according to this measure. Proliferation resistance. The proliferation resistance of alternative nuclear systems depends on how difficult it would be for a nation or a subnational group to use a facility or material to make a nuclear explosive device. No chemical processing facility can be constructed to make it impossible to change its product streams, but it can be designed to make changes costly, lengthy, and detectable. Proliferation resistance can be judged by criteria related to the material streams and the processes, including the extent to which (a) access to the material,
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges facilities, and technologies used in the proposed cycle would reduce the time, cost, and observability of producing weapons-usable material;1 (b) the personnel and experience involved in operating the proposed system would reduce the time and cost to produce weapons-usable material (not only at the facilities in the proposed system but at other, possibly covert, facilities); (c) the difficulty of ensuring against sensitive leakage of technology might increase or decrease if the proposed fuel cycle were implemented; (d) the number of safeguards inspection days per gigawatt-day (GW-d) generated would increase or decrease in the proposed fuel cycle, compared to other systems; and (e) the uncertainty in meeting safeguards goals would increase or decrease for the proposed system compared to other systems. In addition, one needs to consider how the adoption of the proposed system by some countries might affect other countries’ decisions to pursue sensitive technologies such as enrichment or reprocessing.2 With fuel cycle facilities and processes in particular, useful objectives include ensuring that conversion of material from reactor fuel material to directly usable weapons material would be difficult, time consuming, and have a high probability of being detected (see Box 3.1). A facility that achieves these objectives would have no separation or processing facilities that (a) have directly usable material in storage, (b) have directly usable material at any other point in the fuel cycle, (c) offer a way to produce directly usable material by simple process changes, (d) offer a way to produce directly usable material without substantial equipment replacement or major modifications, (e) offer a way to carry out such equipment or plant modifications with facilities and components normally onsite, or (f) offer a way to carry out equipment or plant modifications without plant decontamination or entry into extremely high radiation fields. In addition, such a facility would (g) have uranium-handling equipment for all stages of the fuel cycle that are designed for criticality safety when handling low-enriched uranium (LEU), but not when handling highly enriched uranium (HEU), so as to deter using it for higher enrichments than those for which it was designed; and (h) provide a high likelihood of timely warning—that is, the length of time required after likely detection of a diversion effort and before sufficient material was available for a small nuclear arsenal would be such that there is time for national and international bodies to respond. Environmental impact. All proposed systems would be expected to meet all applicable environmental, safety, and health requirements. The environmental impacts of a fuel cycle depend sensitively on the details of the fuel cycle and how it is implemented and operated, and it is difficult to argue for holding today’s proliferation and other problems at risk for tomorrow’s unknown problems. A system can therefore be evaluated based on whether it would significantly increase existing environmental, safety, or health risks beyond those that would exist if it were not implemented. Thorough comparisons among different nuclear systems would be based on expected harms to the public, workers, and the environment throughout the life cycle of the system from both radiation and other industrial or chemical impacts. This would include both normal operations and plausible accident scenarios. Variations among doses of radiation that are all very low may not be particularly important discriminators between one system and another, however. 1 A related metric is how amenable the process is to safeguards, particularly the relative ability to meet International Atomic Energy Agency (IAEA) goals for timely detection of diversion of a significant quantity of weapons-usable material. 2 For a discussion of similar criteria, see Bunn 2007, and Nuclear Energy Agency for the Generation IV International Forum, 2006.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges Resource utilization. Proposed systems can be compared on the basis of how long they could continue to generate electricity economically given likely future system constraints, including the cost of uranium and repository capacity. It will not be urgent to shift toward closed fuel cycle systems that utilize uranium more efficiently until the cost of fuel from fresh uranium persistently exceeds the full cost of fuel from recycled fissile material or other factors, such as constraints on repository capacity, become overriding factors. Technical feasibility and maturity. Admiral Hyman Rickover pointed out the perils of comparing “academic reactors” and “practical reactors.”3 Comparisons of proposed future systems must take into account their respective levels of technological development, as it is often the case that as work focuses on a specific design, problems arise that were not anticipated at earlier stages of development. Proposed systems can be compared based on the presence or absence of required steps whose technical feasibility is not yet established, on the level at which individual steps and the total system have been designed and demonstrated, and on the estimated years and resources that would be required to prepare the system for commercial deployment. Advanced safeguards and security technologies could play a critical role in pursuing the nonproliferation goals mentioned above. In particular, in providing increased capabilities to detect covert nuclear facilities; highly accurate near-real-time monitoring of material flows in bulk processing plants with reduced intrusiveness, increasing confidence that any diversion would be detected; low-cost real-time monitoring that would set off an immediate alarm if stored nuclear material were tampered with or removed; effective protection against sophisticated outsider and insider theft and sabotage threats at reduced cost; and design of facilities to simplify and increase the effectiveness of safeguards. A study group of the American Physical Society concluded that a reinvestment in research and development on safeguards and security technologies is needed (APS, 2005), and the joint committees agree. 3 In 1953, in the face of criticism of the U.S. Atomic Energy Commission plan to develop pressurized water reactors rather than exploring the multitude of other reactor options, Admiral Rickover wrote (Rockwell, 2002; Kuliasha and Zucker, 1992, p. 271; and Rickover, 1970, p. 1702): An academic reactor or reactor plant almost always has the following basic characteristics: (1) It is simple. (2) It is small. (3) It is cheap. (4) It is light. (5) It can be built very quickly. (6) It is very flexible in purpose. (7) Very little development will be required. It will use off-the-shelf components. (8) The reactor is in the study phase. It is not being built now. On the other hand, a practical reactor can be distinguished by the following characteristics: (1) It is being built now. (2) It is behind schedule. (3) It requires an immense amount of development on apparently trivial items. (4) It is very expensive. (5) It takes a long time to build because of its engineering development problems. (6) It is large. (7) It is heavy. (8) It is complicated.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges BOX 3.1 DIRECTLY USABLE MATERIAL The joint committees use the term “directly usable” to mean that the material could be used to fabricate a nuclear explosive without extensive chemical processing using heavy shielding or isotopic enrichment. As examples, fresh LEU fuel and spent fuel from a typical power reactor would not be directly usable weapons materials by this definition, as LEU would require isotopic enrichment before it could support an explosive nuclear chain reaction, and spent fuel from typical power reactors could only be processed if some form of heavy shielding were used. By this definition, unirradiated mixed-oxide (MOX) or transuranic (TRU) fuel or uranium-aluminum HEU research reactor fuel would be considered directly usable, because, while each would require chemical processing before it could be used in a nuclear explosive, that chemical processing would not have to be done remotely and would pose fewer challenges.* The joint committees’ use of directly usable weapons material is very similar to the International Atomic Energy Agency’s (IAEA) term “unirradiated direct-use material,” which refers to direct-use material (including chemical mixtures such as MOX) “which does not contain substantial amounts of fission products; it would require less time and effort to be converted to components of nuclear explosive devices” than would, for example, plutonium in spent nuclear fuel. *For a discussion of the relative availability of different types of adversaries to recover material usable in a weapon from different types of materials, see NRC, 2000. NOTE: For the IAEA definition, see IAEA Safeguards Glossary, accessed at www.pub.iaea.org/MTCD/publications/PDF/nvs-3-cd/Start.pdf on July 19, 2005. The Generation-IV International Forum (GIF) has outlined an approach that is similar in some respects to the system-level, criteria-based approach advocated here. GIF’s “technology roadmap” emphasizes the need to focus on entire nuclear energy systems, including “the nuclear reactor and its energy conversion systems, as well as the necessary facilities for the entire fuel cycle from ore extraction to final waste disposal” (DOE, 2002, pp. 5-6). GIF has specified several ambitious goals for such systems (though it remains unclear whether any single system can meet all of these objectives simultaneously). Sustainability. The goals are to develop systems that will “provide sustainable energy generation that…promotes long-term availability of systems and effective fuel utilization for worldwide energy production,” and “minimize and manage their nuclear waste and notably reduce the long-term stewardship burden, thereby improving protection for the public health and the environment.” Economics. The goal is a system that “will have a clear life-cycle cost advantage over other energy sources,” and “a level of financial risk comparable to other energy projects.” To complete the economic analysis, a discount rate must be selected and its basis carefully explained. Safety and reliability. Goals for Generation IV systems are to “excel” in safety and reliability, and in particular to have “a very low likelihood and degree of reactor core damage” and to “eliminate the need for offsite emergency response.” (The goal of eliminating all reliance on emergency responses outside the site is an example of setting very specific goals within an overall category, possibly without adequate consideration of the costs and benefits of that particular objective.)
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges Proliferation resistance and physical protection. GIF set the goal of “increasing the assurance” that these systems would be “a very unattractive and the least desirable route for diversion or theft of weapons-usable materials,” and that they would “provide increased physical protection against acts of terrorism.” As stated, these are notably less specific than the goals for economics or safety and reliability. EVALUATING CURRENTLY PROPOSED SYSTEMS Nations that have led technological development of nuclear fuel cycles, including France, Japan, Russia, the United Kingdom, and the United States, have developed a variety of technological options for processing spent nuclear fuel. Some processes, including the only ones deployed on a large scale, initially were developed and optimized for the military purpose of extracting plutonium for nuclear weapons. Some of those processes have been adapted for nonmilitary applications, specifically for processing different types of commercial nuclear fuels. Each of the processes is actually a family of processes (variants on the overall process approach; no two PUREX lines are identical). The most important of these families are PUREX, COEX, UREX(+), pyroprocessing, fluoride volatility, REPA, TRUEX, and supercritical carbon dioxide (CO2). Among these, PUREX, COEX, UREX+, and pyroprocessing garner the most attention today in nations with grand nuclear energy ambitions. Box 3.2 gives descriptions of these processing options. The descriptions are necessarily at a high level because many variations within the same family are possible, and two variants can have important differences (see Box 3.3 for an illustration of a processing family, UREX+). One of the reasons why variants exist within a family is that it is necessary to tailor a given process specifically to deal with each different fuel type, or even to deal with very different burn-ups of the same fuel type. For this reason, even with this set narrowed, it is not really possible to carry out a detailed comparison among the options, as described in greater detail below.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges BOX 3.2 MAJOR TECHNOLOGICAL OPTIONS FOR PROCESSING SPENT NUCLEAR FUEL PUREX The PUREX process coextracts and then individually separates to desired purity uranium and plutonium from fission products and other transuranics. Those transuranics and fission products become part of the waste stream. The plutonium can be used in fabrication of mixed-oxide or metallic fuel. The uranium can be reused, too, but uranium from commercial reactors typically is not reused, because the isotopic mix of irradiated uranium is not optimal and fresh uranium is relatively inexpensive. However, uranium recovered from research and propulsion reactors is sometimes recycled. COEX The COEX process is a modified version of PUREX that coextracts roughly equal amounts of uranium and plutonium for fabrication into MOX fuel. Minor actinides go to the high-level waste product along with the remaining fission products. UREX and UREX+ The UREX process removes uranium in an initial extraction step. That uranium is purified for disposal as low-level waste or for reuse. The remaining stream of transuranic constituents, including plutonium, is maintained as a group and destined for fabrication into fast-reactor fuel. Fission products are a separate stream, but some of them may be separated further (UREX+). For example, in some schemes the plan is to separate cesium and strontium from the other fission products and store them for decay, to reduce repository heat load, which for some repositories may increase effective repository capacity. Lanthanide fission products may be retained with the transuranics if they are deemed to provide some self-protection radiation barrier, or they may be left with the other fission products. Pyroprocessing There are different processes that were initially developed in Russia and the United States. Each country is continuing to develop its own approach, and France and Japan are also conducting research on their own approaches. U.S. process: Spent fuel, if oxide, is reduced to a metallic form and immersed in a bath of molten salt floating on a liquid cadmium cathode, which attracts plutonium and the minor actinides. Uranium can be deposited on a solid cathode. This process, never deployed at any significant scale, would be most readily applied to metallic fuel. The United States also developed a melt-refining process for pyroprocessing a special sodium-bonded fuel from the EBR-II, and ran an extensive processing campaign for several years, but the direct applicability of this process to other types of fuels is probably limited. Russian process: Spent fuel is dissolved in molten salts and crystal plutonium dioxide or electrolytic plutonium, and uranium dioxides are recovered from the melt on a solid cathode. Uranium and plutonium remain together. This process is most readily applied to oxide fuel.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges BOX 3.3 THE UREX+ FAMILY OF PROCESSING OPTIONS SPENT NUCLEAR FUEL Table Stages and Products From UREX+ Variants* Process 1st Product 2nd Product 3rd Product 4th Product 5th Product 6th Product 7th Product UREX+1 U (highly purified) Tc, I (LLFPs, dose issue) Cs, Sr (short-term heat mgmt.) Other FPs TRU+Ln (temporary storage) UREX+1a U (highly purified) Tc, I (LLFPs, dose issue) Cs,Sr (short-term heat mgmt.) FPs (including lanthanides) TRU (group extraction) UREX+2 U (highly purified) Tc, I (LLFPs, dose issue) Cs,Sr (short-term heat mgmt.) Other FPs Pu+Np (for FR recycle fuel) Am+Cm+ Ln (temp. storage) UREX+3 U (highly purified) Tc, I (LLFPs, dose issue) Cs,Sr (short-term heat mgmt.) FPs (including lanthanides) Pu+Np (for FR recycle fuel) Am+Cm (heterogeneous targets) UREX+4 U (highly purified) Tc, I (LLFPs, dose issue) Cs,Sr (short-term heat mgmt.) FPs (including lanthanides) Pu+Np (for FR recycle fuel) Am (heterogeneous targets) Cm (storage) UREX+1 and UREX+1a are designed for homogeneous recycling of all transuranics to fast-spectrum reactors. UREX+2, +3, and +4 are designed for heterogeneous recycling, possibly as an evolutionary step, to preclude the need for remote fabrication of fuel. * SOURCE: Laidler, 2007. Table UREX+ Variants and Their Associated Technologies and DOE-assessed Technological Maturity† Process Fuel Type Fabrication Technology Technological Maturity UREX+1 (Interim storage only) - - UREX+1a FR mixed oxide Remote, hot cell Low UREX+1a FR metal Remote, hot cell Low UREX+2 (Interim storage only) - - UREX+3 LWR mixed oxide Glovebox High UREX+3 FR mixed oxide or metal Glovebox High UREX+3 Am/Cm transmutation target Remote, hot cell Low UREX+4 LWR mixed oxide Glovebox High UREX+4 FR mixed oxide or metal Glovebox High UREX+4 Am transmutation target Remote, possibly glovebox Low UREX+4 Interim storage of Cm † SOURCE: Finck, 2006.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges The joint committees’ statement of task calls for a comparison of the PUREX reprocessing process, the UREX family of processes, and other spent fuel treatment processes being considered or developed in the United States and Russia. The joint committees found that insufficient information was available for realistic comparisons. First, as noted above, meaningful comparisons consider entire nuclear energy systems, rather than being based on a single component of those systems, such as fuel processing. Second, while PUREX is an established industrial process that has been used at a large scale for decades in several countries, the UREX family of processes is still at an early stage of development, and the features depend very much on the details of the process and the fuel to be processed.4 PUREX itself is not a single process but a series of solvent extraction steps with several variants, somewhat different in each incarnation. The development and selection of the technology options requires clear goals. PUREX was initially developed to separate high-purity plutonium for nuclear weapons. Variations on PUREX may try to improve the process with respect to other objectives, but the process inescapably bears some features of that original design goal that both cause proliferation concerns if the technology spreads and result in waste streams that have proven problematic. Alternative methods for processing are being designed to other goals, but those goals are not always clear or compatible. Key decisions about the specific process under consideration, whether PUREX, UREX, or some other process, strongly affect issues such as the radiation levels from the materials to be recovered for recycling and the characteristics of expected waste streams. In general, the UREX family of processes involves additional separation steps not included in PUREX, and is therefore likely to be somewhat more complex and expensive, and may increase the difficulty of material accountancy, though there may be potential for further optimization of whatever processes are eventually developed. Studies to date suggest that the material recovered for recycling in these processes would be more radioactive than the plutonium recovered in the PUREX process, but not radioactive enough to be a substantial barrier to theft and subsequent processing for use in a nuclear explosive. Pyrochemical processes have been pursued in Russia and the United States, and elsewhere, with a particular emphasis on processing fast-reactor fuels. Russia’s process is well along in development, and Russia has decided to use this process for processing spent fuel from the BN-800 fast reactor now under construction. Russia has decided to use pyroprocessing combined with vibropacking the fuel to produce assemblies for the BN-800 fast reactor. These technologies complement each other well and produce fission materials with a sufficiently high level of radioactivity at each processing stage, in a mixture with minor actinides and certain fission products. Its high radioactivity drives the application of remotely controlled and fully automated fuel manufacturing processes in a closed fuel cycle, so that the fuel is very difficult and very costly to remove for other purposes. There is far less experience with the Russian pyrochemical process than there is with PUREX, however, and estimates of costs for widespread deployment are still difficult to make. It appears that the wastes from the process can be made suitable for geologic disposal. The material recovered from the Russian process, sometimes called dirty fuel in Russia, includes 4 “The characteristics, treatment, and final disposition requirements of several waste streams from spent fuel reprocessing is not completely known at this time. This is because (a) different separations and fuel fabrication options are still being evaluated, (b) waste stream generation from the proposed separations options is uncertain and unprecedented, and masses, volumes, and compositions remain uncertain… The UREX suite of separation technologies can result in many different waste streams.” (DOE, 2008a, p. 24).
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges several actinides and some fission products, and Russian sources report that even a kilogram of the recovered material emits several Sv/hr, which is above the international standard for self-protection. Most of this radiation, however, comes from fission products with short half-lives. The radiation barrier that remained if the material were stored until the short-lived fission products decayed would still be too high for hands-on operations in normal commercial environments, but not too high for determined terrorists to attempt to use the material for weapons. (The same is true of commonly discussed variants of the UREX process.)5 The joint committees believe that additional fuel treatment processes, not currently being actively pursued in the United States or Russia, deserve additional exploration, including, for example, processes making use of supercritical CO2 and fluoride volatilization. Decision makers still need to know whether these processes can overcome any of the most important cost, proliferation, safety, and security issues associated with the traditional PUREX process. The joint committees’ statement of task also calls for a comparison of the Russian BN-600 and BN-800 fast reactors to the types of fast reactors under consideration in the U.S. Global Nuclear Energy Partnership (GNEP) Program. This is, in a sense, an apples-to-oranges comparison, as these reactors are at very different stages of development and being pursued with very different purposes in mind. The Russian reactors are breeders, designed to produce more plutonium than they consume to address long-term concerns over limited uranium resources, while the proposed GNEP concepts are burners, designed to burn up stockpiles of plutonium and other actinides in the minimum number of cycles.6 The BN-600 reactor has been operational for decades, and the BN-800 is under construction, while the proposed GNEP reactors are still paper concepts. While the BN-600 and BN-800 reactors have breeding ratios just over 1.0, some Russian designers envision future reactors with breeding ratios in the range of 1.6, which would be a major technical challenge; GNEP, by contrast, envisions burners with conversion ratios in the range of 0.25-0.5, also a major technical challenge. While the number of cycles required to achieve any given level of actinide destruction can be calculated for burners of any given conversion ratio, it makes no sense to compare the proposed GNEP reactors to the Russian designs in this respect, since the Russian designs are not intended for this purpose. As currently planned, the BN-800 will operate with oxide fuel, and the spent fuel will be pyroprocessed and new fuel produced with a vibropack process, demonstrating these approaches on an industrial scale. (More detail on these Russian fuel cycle plans is provided in Appendix C.) With advanced computer modeling, it may be possible to design a fast reactor for which it can be demonstrated that the reactor would shut itself down automatically in response to any of the plausible transients in the system; if so, this would be a substantial safety advantage, and might make it possible to eliminate some of the redundant safety systems now used in light-water reactors, potentially reducing costs. This possibility and the cost impacts that might result from it, however, both remain to be demonstrated. Here, too, the joint committees believe that continued funding for research and development on fast-reactor concepts and other reactor types not currently being actively 5 PUREX was designed to separate out plutonium, which is a nuclear weapons material. 6 The Russian fast reactors can be configured to burn (that is, have a conversion ratio of less than 1), but they have not been operated in that way or with minor actinide-bearing fuels and are not optimized for this configuration and mode of operation. Some fuel tested at the Russian BOR-60 test reactor with minor actinides (neptunium and americium) were “semi-industrial” rather than laboratory studies. Many difficulties arise, however, in building commercial-scale facilities even with semi-industrial-scale experience.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges pursued in the United States and Russia would be desirable, including such concepts as lead-cooled systems, nonfertile fuels, thorium fuel cycles, and molten salt reactors. Finding 8a Both Russia and the United States are working on new technologies for processing spent fuel, intended to reduce the economic costs and proliferation risks of traditional reprocessing approaches and improve waste management. The technologies being proposed would still pose significant proliferation concerns if deployed in countries that did not previously have reprocessing capabilities. The new technologies under development will take significant time before being ready for demonstration at commercial scale. Finding 8b In most cases, reprocessing is not economic under current conditions. When the world’s economically recoverable uranium resources diminish compared to demand or there is widespread deployment of fast reactors, then reprocessing may become economically attractive. Recommendation 8 Developers of nuclear fuel cycle technologies should assess the technologies’ proliferation risks and projected economic costs and benefits as critical elements of design. As new technologies are developed, it will be important for developers to consider the proliferation hazards and work with the IAEA to develop appropriate safeguards. Finding 9 Excess stocks of plutonium separated from spent fuel, beyond plutonium that would be needed for making MOX fuel for use in the near term, pose security risks. Recommendation 9 States should end the accumulation of stockpiles of plutonium separated from spent fuel as soon as practicable, and begin to reduce existing stocks. Spent fuel should only be reprocessed when its constituents are needed for fuel, or when reprocessing is necessary for safety reasons. WHY “ACCEPTABLE LEVEL OF DESTRUCTION OF ACTINIDES” IS NOT WELL DEFINED TECHNICALLY Actinide destruction, more properly actinide fissioning or more commonly actinide burning, has been stated as one of the main objectives of the advanced technologies for nuclear energy in the United States, and has been considered as a central objective of programs in Japan and Europe. As articulated in GNEP, actinide burning is meant to support three main goals: extracting more energy from the earth’s uranium resources, reducing the quantity and hazard of radioactive waste in a deep geologic repository, and reducing the potential for fuel cycle material to be used to make nuclear weapons. In the joint committees’ view, each of these is a worthy
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges TABLE 3-2 Reactor Types, Fuel Types, and Purposes of Fuel Reprocessing in the Thorium Fuel Cycle Stages of Thorium Fuel Cycle Reactor type, fuel accommodation Fuel Purpose of reprocessing blanket and core fuel Stage 1 Accumulation of U-233 BN-800 Blanket Metallic Th Extraction of U-233 ThO2 Core MOX Recovery of NPhP VVER-1000 Core PuO2-ThO2 Extraction of U-233, Recovery of NPhP 233UO2-235UO2 Recovery of NPhP UO2-ThO2 Recovery of NPhP Stage 2 Closed Thorium Cycle BN-800 Blanket Metallic Th Extraction of U-233 Adjustment of U-233 content Core UO2-ThO2 + MA + LLFP Recovery of NPhP, Introduction of MA and LLFP VVER-1000 Core UO2-ThO2 Recovery of NPhP LLFP − long-lived fission products NPhP − nuclear-physical properties such as the physical integrity and composition of the fuel matrix. SOURCE: Provided by joint committees. DRY METHODS FOR FUEL SEPARATIONS The Russian nuclear effort in dry methods for separation of nuclear fuel constituents is divided into two main categories: (1) pyroelectrochemical, which are the most compact, but provide only partial separation and purification; and (2) halogenide distillation, which can achieve high levels of purification of uranium (mainly) and plutonium from fission products. An integrated technology—a combined reactor and fuel-processing unit, such as a molten salt reactor—has the advantage of easy fuel preparation and recycling, because the fluid nature of the fuel provides extra flexibility and a simpler back-end fuel cycle. The molten salt reactor concept appears to have substantial promise not only as a transmuter of transuranics, but also as an advanced TRU-free system operating with the uranium-thorium cycle. Pyroelectrochemical Processes Basic research on molten salt systems has enabled Russian facilities to develop processes for production of granulated uranium and plutonium oxides and mixed uranium and plutonium
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges oxides. Pyrochemical technology is able to carry out all of the deposit production operations in one apparatus—a chlorinator-electrolyzer—which simplifies the process. Russian pyrochemical reprocessing consists of three main stages: dissolution of initial products or spent nuclear fuel in molten salts precipitation of plutonium dioxide or deposition of electrolytic uranium and plutonium dioxides from the melt processing of the material deposited on the cathode or precipitated at the bottom of the melt for granulated fuel production The process can recover the cathode deposits without changing their chemical composition or redistributing the plutonium. Three alternatives were considered and are now under development for reprocessing irradiated nuclear fuel at the Research Institute of Atomic Reactors (RIAR): reprocessing uranium fuel with the production of uranium dioxide for recycling reprocessing MOX fuel for only plutonium recycling as the most valuable component reprocessing MOX fuel with production of MOX fuel All products are reprocessed with the goal of having a complete recycle of plutonium, neptunium, americium, and curium. Vibropacking technology is applied to the manufacture of fuel pins. VIBROPACKING PROCEDURE RIAR has used vibropacking technology for about 20 years to fabricate granulated fuel in glove boxes or hot cells. The main advantages of the vibropacking technology and fuel rods with vibropacked fuel are as follows: The production process is simple and reliable because it has a relatively small number of subprocesses and control operations, which facilitates automation and remote control. The granular form of the fuel feedstock enables vibropacking technology to use both homogeneous compositions and mechanical mixtures for heterogeneous compositions. The thermal-mechanical stress on the cladding is lower for vibropacked fuel than for pellet-stacked fuel. Vibropacked fuel tolerates relaxed requirements for the inner diameter of fuel rod cladding. Vibropacked fuel is made by agitating a mechanical mixture of (U, Pu)O2 granulate and uranium powder, which binds up excess oxygen and some other gases (that is, operates as a getter) and is added to the fuel mixture in proportion during agitation. The getter resolves problems arising from fuel-cladding chemical interactions. The process allows fabricators to
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges control the distribution of plutonium and density along the fuel column length, with the getter distributed uniformly throughout. CLOSING THOUGHTS ON NEW TECHNOLOGIES If we are to achieve anything with technology, what is needed is a set of specific objectives that can be used to guide the research and development programs. Finding 10 Many of the technologies for improved nuclear fuel cycles are not areas that will advance without directed research specifically focused on the nuclear fuel cycle; advances in other areas of science and engineering will help, but are not sufficiently linked to nuclear fuel cycles to solve the technical challenges described here by themselves. Research is needed in the areas of processing of irradiated nuclear fuel and nuclear fuel design (beyond the incremental improvements in uranium oxide fuel for light water reactors), as well as in improved approaches to disposal of wastes or spent fuel, and reduced-cost recovery of uranium from low-grade sources. Additional research and development is also needed to develop advanced safeguards and security technologies that can provide increased capabilities to detect covert nuclear facilities; highly accurate near-real-time monitoring of material flows in bulk processing plants with reduced intrusiveness, increasing confidence that any diversion would be detected; low-cost real-time monitoring that would set off an immediate alarm if stored nuclear material were tampered with or removed; effective protection against sophisticated outsider and insider theft and sabotage threats at reduced cost; and design of facilities to simplify and increase the effectiveness of safeguards. Recommendation 10 The U.S., Russian, and other governments should take the lead in a cooperative international effort to make additional research and development investment in advanced safeguards and security technologies. A focused effort should be made to make the results of this research and development available to the international community to ensure that new facilities are more secure and readily safeguarded. The international community also should adopt the philosophy of designing high levels of security and safeguards into new nuclear systems and facilities from the outset, including both the inherent technical characteristics of the process and the institutional measures to be taken. Finding 11 It is not possible today to construct an entire, operational international fuel cycle program.14 Such a program will have to be built incrementally. However, elements of that program currently exist and the groundwork for other elements has been laid. Recommendation 11 For new technologies, the U.S., Russian, and other governments should 14 This would be run internationally and include all elements of the fuel cycle.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges continue to invest in research and development on advanced approaches to once-through and closed fuel cycles that offer the potential to improve proliferation resistance, safety, security, economics, resource utilization, and/or waste management. utilize a systems approach to developing and assessing these technologies, with clear objectives and technically justifiable criteria for decision making. Use systems analysis to identify potentially promising approaches before proceeding to build pilot or larger facilities. take all relevant proliferation risks into account when assessing proliferation resistance, including how the availability of the materials, facilities, and expertise associated with a particular fuel cycle approach would affect the time, cost, uncertainty, and detectability of a nuclear weapons program. The implementation of those elements that are feasible, for example, assurance of fuel supply, should not be delayed while other options are being refined or explored both institutionally and technically. Secondary Issues: B4. Compare the fuel to be produced from the processes examined in (B1) for use in appropriate reactors (light-water reactors, high-temperature gas-cooled reactors, and fast reactors). What are the advantages and disadvantages of each type of fuel? B5. Compare the repository requirements for the waste produced by the processes proposed in the GNEP concept with that from a system based on PUREX and one based on Russian plans. Handling the fuel after use in a reactor is difficult. Only Finland has an approved process to build a repository for spent fuel and Sweden may be close to having a site acceptable to a local community.15 The three largest users of nuclear power, France, Japan, and the United States, do not have operating sites and only the United States has selected a site for a repository. Several billions of dollars have been spent in the United States, and on June 3, 2008, the U.S. Department of Energy submitted to the U.S. Nuclear Regulatory Commission an application for a license to construct a high-level radioactive waste repository at Yucca Mountain. The final standard for evaluating the license application has not yet been issued, and the regulator’s review is still pending. Although it is often argued that a closed fuel cycle reduces the volume of waste from nuclear energy, the amount of radioactive material requiring long-term storage depends upon the processes, the country’s regulatory requirements, and even the definitions of waste.16 Pool storage for 5 years followed by dry cask storage has been approved by the U.S. Nuclear Regulatory Commission as being safe storage for many decades. Nevertheless, some countries, such as France and Japan, are pursuing the option of reprocessing, which they believe offers 15 Both Russia and, on a smaller scale, the United States have injected liquid radioactive waste underground as a means of disposal, but both countries now regard this practice as undesirable for future disposal. 16 For an explanation and argument that the closed cycle produces more waste, see Schneider and Marignac, 2008.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges waste management and resource extension advantages. Separating direct-use material by reprocessing significantly raises the proliferation risk from a nuclear program, but various forms of separation and recycling are nonetheless an important feature of some proposed fuel assurance programs. Countries embarking on nuclear energy programs should examine the approaches to management and disposal of radioactive wastes that they will pursue. None of the fuel assurance programs discuss the possibility of taking the spent fuel to encourage a new country not to build a reprocessing plant. Unfortunately, except for the Russian program, there is little likelihood that any of the other programs will be able to offer to take the spent fuel. This gradually may become a difficulty in maintaining credibility of the programs. COMPARISON OF PROCESSES FOR SEPARATION OF FISSILE AND OTHER MATERIALS FROM SPENT OR IRRADIATED NUCLEAR FUEL Currently operating reprocessing plants all use variations on the PUREX process. In this process, spent nuclear fuel is chopped and cladding hulls are separated. The chopped fuel assemblies are dissolved in nitric acid, and the solution is prepared with organic flocculating agents and filtration for the extraction process. Extraction of uranium, plutonium, and neptunium is accomplished by tributyl phosphate (TBP) solutions in hydrocarbon dissolvent. Uranium and plutonium products of the process are almost entirely free of fission product. Uranium and plutonium are separated from each other to better than 1 part in 7×105, with waste losses of uranium, plutonium, and neptunium less than or equal to 0.01 percent, 0.025 percent, and 0.5 percent, respectively (Myasoedov, 2007). In addition to plants built for separating weapons plutonium, large plants of this kind are separating plutonium from civilian fuel in France, Japan, Russia, and the United Kingdom; two small plants are operating in India; and China has recently built a pilot plant. The Russian plant, RT-1, located at the Mayak Production Association in the town of Ozersk, was launched in 1976, and processes fuel from both propulsion and power reactors. The United States and Russia have accumulated large stocks of spent nuclear fuel. The United States every year adds 2,000 MTHM of spent nuclear fuel to its stored inventory, which reached 58,000 MTHM in 2007. By 2016, the inventory will be about 77,000 MT, which is over the 63,000 MTHM legal limit for commercial power-reactor waste to be disposed in the first high-level waste repository in the United States.17 The Russian Federation adds 700 MTHM of spent nuclear fuel each year to its stores, which now are at about 16,000 MT. By 2016, Russia anticipates it will have more than 25,000 MTHM of spent fuel in storage. To develop options for these stocks of spent fuel and for future fuel cycles, several research programs have examined partitioning of key radionuclides to improve the overall performance of the repository. Development of improved processes for extracting key radionuclides from spent fuel and of improved reactor and fuel technologies would be needed to achieve the ambitious goals for reducing the repository burden that GNEP and some other national programs have set. In both cases the partitioning of radionuclides has the potential to make changes in waste streams that could improve repository performance. Most important among the ones relevant to the fuel cycle options considered here are improved waste forms and reduced total actinide 17 The technical or geologic limit at the proposed site, Yucca Mountain, is expected to be larger than the legal limit. The Electric Power Research Institute (EPRI) has estimated the technical capacity to be four to nine times greater (EPRI, 2007).
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges content, which lowers the heat loading and the long-term radiotoxicity in the repository. The mobile radiotoxicity (which is more relevant than the radiotoxicity itself and is repository dependent) could be lowered in the context of Yucca Mountain if actinides are burned.18 The Yucca Mountain Program has, however, stated that the repository will meet its licensing requirements without reductions of radiotoxicity within the legal capacity of the repository. If heat load is a limiting factor in repository capacity,19 then reducing the heat load in the waste streams would enable a country to dispose of the waste from more nuclear electricity generation within a repository of fixed capacity (though most countries are planning on repository sites that would be readily expandable). Given the difficulties already encountered in siting and opening a repository, there may be a significant benefit in extending a repository’s capacity. How much of a difference recycling can make depends very much on the details of the burn-up, the waste streams, the waste forms, and the specific repository design and environment, so only a scenario-based approach to analysis works, and right now there is not enough information to know which scenarios are most likely. This approach to increasing repository capacity or reducing repository hazards, however, entails a trade-off with the siting and hazards associated with additional facilities for handling and processing the materials aboveground in the closed fuel cycle. B6. Are new laws and/or regulations required for either the U.S. or the Russian approach to the internationalization of the fuel cycle? Will either approach require any existing laws or regulations to be repealed or changed? As noted in Section A8, there are many laws, regulations, and legal instruments that would need to be revised to reduce “road blocks” to proliferation threat reduction. Key among those is bringing into force a civilian nuclear cooperation agreement (known as a 123 agreement for the relevant section of the U.S. Atomic Energy Act (AEA); see Box 3.5) with Russia and any other nation that is critical to the successful implementation of international fuel cycles involving transfer of spent nuclear fuel. Because a substantial fraction of the world’s stock is U.S.-obligated fuel, which cannot be transferred to another party without both a 123 agreement and U.S. approval, any international scheme for spent fuel management is necessarily limited by the lack of a civilian nuclear cooperation agreement with the United States. Such an agreement would be necessary for a future international center for spent fuel management to be able to 18 For many years, analyses of the proposed repository at Yucca Mountain have identified neptunium-237 as the dominant contributor to potential dose from groundwater consumption in long time frames (beyond several tens of thousands of years), with technetium-99, carbon-14, and iodine-129 dominating in earlier time frames. However, estimates of actinide contributions to potential dose in the long term have been reduced very recently (DOE, 2008b, pg. 5-6) because the U.S. Department of Energy applied revised International Commission on Radiological Protection (ICRP) weighting factors for calculation of individual doses (ICRP, 2001). Now “[t]he estimated mean annual individual dose [beyond 10,000 years] at the [reasonably maximally exposed individual] location would consist of approximately 30 percent from plutonium-242, about 20 percent from each of iodine-129 and neptunium-237, about 15 percent from radium-226, and about 8 percent from technetium-99.” (DOE, 2008, p. 5-30; ICRP, 2001) 19 Some argue that long-term heat load need not be a limiting factor for repositories, because repositories in the saturated zone (those located below the underground water table) have abundant water to absorb and carry away heat, and repositories in the unsaturated zone (those located above the water table) can be left open with air circulating to remove heat. However, all repository designs have some heat considerations. For example, some repositories in saturated zones use bentonite clay to inhibit water flow past waste packages and retard contaminant transport from the waste; but the clay properties worsen as the clay temperature rises (see, e.g., Neall, 2008).
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges operate effectively in Russia. Politically the United States is unlikely to be able to take back spent fuel itself for many years to come.20 20 Under U.S. law, such take-backs would require congressional approval, though they are not prohibited in principle; such approval is unlikely to be forthcoming, except in special cases, such as the ongoing return of irradiated research reactor fuel, which is part of a program to reduce proliferation risks by eliminating highly enriched uranium (HEU) from as many research reactors as possible.
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges BOX 3.5 NUCLEAR COOPERATION WITH THE UNITED STATES: AGREEMENTS ON PEACEFUL USES OF NUCLEAR WEAPONS BETWEEN THE UNITED STATES AND RUSSIA U.S. Atomic Energy Act of 1954, Section 123 Significant nuclear exports from the United States are only legally permitted under Section 123 of the U.S. Atomic Energy Act (AEA) of 1954 as amended, 42 U.S.C., Section 2153, in accordance with an agreement for peaceful nuclear cooperation with the recipient.* Such agreements are frequently referred to as 123 agreements.† Exports deemed significant include power reactors, research reactors, nuclear source material (including reactor fuel), and four major components of reactors (pressure vessels, fuel charging and discharging machines, complete control rod drive units, and primary coolant pumps). A 123 agreement between the United States and another country establishes a framework for exports and cooperation, but does not obligate the United States to provide nuclear exports to the recipient country, or to engage in specific cooperative activities. Section 123 of the AEA requires that the following key conditions and requirements be included in a U.S. agreement for peaceful nuclear cooperation:a a guarantee by the cooperating party that safeguards will be maintained with respect to all nuclear materials and equipment transferred, and with respect to all special nuclear material used in or produced through the use of such nuclear materials and equipment a guarantee that no nuclear materials and equipment or sensitive nuclear technology will be used for any nuclear explosive device, or for research on or development of any nuclear explosive device, or for any other military purpose except in agreements with nuclear weapon states, a stipulation that the United States shall have the right to require the return of any nuclear materials and equipment transferred to the recipient country and any special nuclear material produced through the use thereof if the cooperating party detonates a nuclear explosive device or terminates or abrogates an agreement providing for International Atomic Energy Agency safeguards a guarantee that any material or any restricted data transferred pursuant to the agreement and, except in specific cases, any production or utilization facility transferred pursuant to the agreement or any special nuclear material produced through the use of any such facility or through the use of any material transferred pursuant to the agreement, will not be transferred to unauthorized persons or beyond the jurisdiction or control of the cooperating party without the consent of the United States a guarantee that adequate physical security will be maintained with respect to any nuclear material transferred and with respect to any special nuclear material used in or produced through the use of any material, production facility, or utilization facility transferred a guarantee that no material transferred and no material used in or produced through the use of any material, production facility, or utilization facility transferred will be reprocessed, enriched, or otherwise altered in form or content without the prior approval of the United States
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges a guarantee that no plutonium, no uranium-233, and no uranium enriched to greater than 20 percent in the isotope 235, transferred pursuant to the agreement or recovered from any source or special nuclear material so transferred or from any source or special nuclear material used in any production facility or utilization facility transferred pursuant to the agreement, will be stored in any facility that has not been approved in advance by the United States a guarantee that any special nuclear material, production facility, or utilization facility produced or constructed under the jurisdiction of the cooperating party by or through the use of any sensitive nuclear technology transferred will be subject to all the requirements specified above In addition to the full list of specified requirements, it is not uncommon for 123 agreements to also apply reciprocal nonproliferation conditions, assurances, and controls. Although not required by U.S. law, the United States may accept the obligations contained in the agreement on a reciprocal basis should it import materials or equipment from the cooperating party. Proposed 123 agreements are to be negotiated by the secretary of state, “with the technical assistance and concurrence of the secretary of energy and after consultation with the (Nuclear Regulatory) Commission.” Following negotiations, the proposed agreement is to be submitted to the President for review. The President must submit an agreement for cooperation to Congress for a statutory review period of 90 days continuous session; however, the actual review period may extend over several more months, depending on the congressional schedule. The Russian Federation and the United States signed an agreement on nuclear energy cooperation, which the United States considers a 123 agreement, on May 6, 2008. Approval and enactment of a 123 agreement does not require the approval of Congress, but Congress may enact legislation to disapprove the agreement. If there is no prohibitory legislation, an agreement may be brought into force following the close of the congressional review period. Once an agreement for cooperation has been brought into force, exports made under the agreement require a license from the U.S. Nuclear Regulatory Commission and must be consistent with other sections of the AEA (Sections 127 and 128) pertaining to the U.S. nuclear export criteria. * Atomic Energy Act, 1954. † Currently the United States has 123 agreements with 19 individual countries plus Taiwan and 2 international organizations, the International Atomic Energy Agency and Euratom (which includes 27 individual countries). a For a comprehensive list of requirements, see Atomic Energy Act, 1954. The United States already has such agreements with21 Argentina, Australia, Bangladesh, Brazil, Canada, China, Colombia, Egypt, the European Atomic Energy Community (Euratom),22 Indonesia, the International Atomic Energy Agency (IAEA), Japan, Kazakhstan, the Republic of Korea, Morocco, Norway, South Africa, Switzerland, Taiwan,23 and Thailand. As noted above, 21 Information about current agreements is taken from “123 Agreements for Peaceful Cooperation” an information sheet available (as of August 31, 2008) at http://nnsa.energy.gov/nuclear_nonproliferation/123_agreements_peaceful_cooperation.htm. 22 Euratom comprises the following member states: Austria, Belgium, Bulgaria, Cyprus, the Czech Republic, Denmark, Estonia, Finland, France, Germany, Greece, Hungary, Ireland, Italy, Latvia, Lithuania, Luxembourg, Malta, the Netherlands, Poland, Portugal, Romania, Slovakia, Slovenia, Spain, Sweden, and the United Kingdom. 23 Pursuant to Section 6 of the Taiwan Relations Act, P.L. 96-8, 93 Stat. 14, and Executive Order 12143, 44 F.R. 37191, all agreements concluded with the Taiwan authorities prior to January 1, 1979, are administered on a
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Internationalization of the Nuclear Fuel Cycle: Goals, Strategies, and Challenges the United States and Russia negotiated such an agreement, but the U.S. Congress did not vote on it and on September 8, 2008, President George W. Bush withdrew it from consideration (Rice, 2008). The United States and Russia are leaders in nuclear technology. The vast majority of nuclear energy technology currently developed worldwide was developed in Russia and the United States. These two nations also have the most developed technologies and technical capabilities to support nuclear nonproliferation. Both have invested a great deal of time and energy in developing concepts to advance the concept of a safer, more secure international nuclear fuel cycle program. Russia and the United States are able to conduct civilian nuclear energy cooperation with the other leaders in nuclear energy, but not with each other, and the lack of a U.S.-Russian agreement restricts those partners’ cooperation on nuclear energy with Russia and the United States. It is difficult to see how such an international program could move forward without the active participation and (probably) cooperation of these two countries. But the appropriate mechanism must be in place to allow this kind of cooperation. Other considerations beyond the scope of this study will factor in to the decisions by the U.S. President and Congress whether or not to bring the signed agreement into force (Einhorn et al., 2008; Alvarez, 2008). The joint committees recognize that it is unlikely that the U.S. government will bring the agreement into force in an environment of worsening relations between the United States and Russia. It is the joint committees’ hope that current disagreements that have recently emerged will not interfere with the United States and Russia working together toward their common goal of inhibiting nuclear weapons proliferation as nuclear energy use grows across the world. Finding 12 The United States and the Russian Federation have signed an agreement on peaceful nuclear cooperation, but it must still be allowed to come into force. The lack of a U.S.-Russian agreement in force is interfering with joint efforts to reduce proliferation. U.S.-Russian cooperation on nuclear energy technology that involved the transfer of nuclear materials, major elements of reactor designs and technology, or major elements of fuel cycle designs and technology from the United States to Russia is only possible under a bilateral agreement on nuclear cooperation (called a 123 agreement in the United States). The expanded cooperation in nuclear energy research and development and commercial implementation that such a bilateral cooperation could make possible could serve both countries’ interests in expanding the use of nuclear energy while meeting safety, security, and nonproliferation objectives. Approval of such an agreement could help establish an atmosphere of cooperation that will strengthen prospects from cooperative international approaches to the fuel cycle and other nonproliferation problems. In particular, under U.S. law, international fuel cycle approaches that involved take-back of fuel to Russia (the only country that yet has a legal structure in place for such take-back) would have to exclude all U.S.-obligated material until a civil cooperation agreement had been put in place. nongovernmental basis by the American Institute in Taiwan, a nonprofit District of Columbia corporation, and continuation of any official relationship with Taiwan.
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