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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, 57
58 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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 systemâs 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 systemâs 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 âpassive safety systemsâ). 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,
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 59 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 countriesâ 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 warningâthat 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 todayâs proliferation and other problems at risk for tomorrowâs 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.
60 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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 âacademic reactorsâ and âpractical reactors.â3 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.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 61 BOX 3.1 DIRECTLY USABLE MATERIAL The joint committees use the term âdirectly usableâ 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 committeesâ use of directly usable weapons material is very similar to the International Atomic Energy Agencyâs (IAEA) term âunirradiated direct-use material,â which refers to direct-use material (including chemical mixtures such as MOX) âwhich does not contain substantial amounts of fission products; it would require less time and effort to be converted to components of nuclear explosive devicesâ 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. GIFâs âtechnology roadmapâ emphasizes the need to focus on entire nuclear energy systems, including âthe nuclear reactor and its energy conversion systems, as well as the necessary facilities for the entire fuel cycle from ore extraction to final waste disposalâ (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 âprovide sustainable energy generation thatâ¦promotes long-term availability of systems and effective fuel utilization for worldwide energy production,â and âminimize and manage their nuclear waste and notably reduce the long-term stewardship burden, thereby improving protection for the public health and the environment.â Economics. The goal is a system that âwill have a clear life-cycle cost advantage over other energy sources,â and âa level of financial risk comparable to other energy projects.â 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 âexcelâ in safety and reliability, and in particular to have âa very low likelihood and degree of reactor core damageâ and to âeliminate the need for offsite emergency response.â (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.)
62 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE Proliferation resistance and physical protection. GIF set the goal of âincreasing the assuranceâ that these systems would be âa very unattractive and the least desirable route for diversion or theft of weapons-usable materials,â and that they would âprovide increased physical protection against acts of terrorism.â 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.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 63 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.
64 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE BOX 3.3 THE UREX+ FAMILY OF PROCESSING OPTIONS SPENT NUCLEAR FUEL Table Stages and Products From UREX+ Variants* Process 1st 2nd 3rd 4th Product 5th Product 6th 7th Product Product Product Product Product UREX+1 U (highly Tc, I Cs, Sr Other FPs TRU+Ln purified) (LLFPs, (short-term (temporary dose issue) heat storage) mgmt.) UREX+1a U (highly Tc, I Cs,Sr FPs (including TRU (group purified) (LLFPs, (short-term lanthanides) extraction) dose issue) heat mgmt.) UREX+2 U (highly Tc, I Cs,Sr Other FPs Pu+Np (for Am+Cm+ purified) (LLFPs, (short-term FR recycle Ln (temp. dose issue) heat fuel) storage) mgmt.) UREX+3 U (highly Tc, I (LLFPs, Cs,Sr (short- FPs Pu+Np (for Am+Cm purified) dose issue) term heat (including FR recycle (heterogen- mgmt.) lanthanides) fuel) eous targets) UREX+4 U (highly Tc, I Cs,Sr FPs (including Pu+Np (for Am Cm (storage) purified) (LLFPs, (short-term lanthanides) FR recycle (heterogen- dose issue) heat fuel) eous mgmt.) targets) â¢ UREX+1 and UREX+1a are designed for homogeneous recycling of all transuranics to fast-spectrum reactors. â¢ 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 Maturityâ Process Fuel Type Fabrication Technological Technology 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 Low glovebox UREX+4 Interim storage of Cm - - â SOURCE: Finck, 2006.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 65 The joint committeesâ 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. Russiaâs 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 âThe 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 uncertainâ¦ The UREX suite of separation technologies can result in many different waste streams.â (DOE, 2008a, p. 24).
66 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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 committeesâ 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 âsemi-industrialâ rather than laboratory studies. Many difficulties arise, however, in building commercial-scale facilities even with semi-industrial-scale experience.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 67 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 worldâs 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 technologiesâ 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 âACCEPTABLE LEVEL OF DESTRUCTION OF ACTINIDESâ 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 earthâs 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 committeesâ view, each of these is a worthy
68 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE goal. They cannot, however, simply be addressed by pursuing actinide burning. The actinides are not a single species. The specific goals for the various species differ depending on the larger fuel cycle system into which actinide burning is being deployed. While all of the transuranium actinide nuclides can undergo fission, some are more useful for reactor systems than others, and some reactors are better matched with particular nuclides over others (more on this below). Similarly, in a geologic repository, some nuclides are greater contributors to risk than others, and which ones are the main contributors to risk depends on the repository system design and environment. The quantity of waste that can be loaded into a repository depends in part on the heat output of the waste to be emplaced, and also on the characteristics of the repository system. And finally, the technical objectives to serve nonproliferation and safeguards depend on the kinds of scenarios that cause concern. Radiation barriers and the presence of other actinides with plutonium in a material stream could present significant obstacles to terrorist groups, but are unlikely themselves to be major obstacles for a nation seeking nuclear weapons. Without a clearly articulated technical objective, there is no credible technical basis for answering the question, What is an acceptable level of burn-up? Burn-up is expressed as either the fraction of the initial heavy metal that has fissioned or as the energy released per ton of initial heavy metal in the fuel (for example, 5 percent or about 55,000 MWd/MTHM). This single number, however, is unlikely to provide the information needed about how the actinides have been burned for a given goal. Actinides, and even isotopes of the same actinide element, do not burn uniformly. The nuclides have different reaction rates7 for a given neutron spectrum, so adjacent nuclei of plutonium-239 and plutonium-240 will be destroyed at different rates in the same neutron flux. The reaction rates are functions of â¢ the cross sections (essentially the reaction probabilities), which are fixed for a given neutron flux spectrum â¢ the concentration of the nuclides â¢ the neutron flux spectrum and magnitude (energy distribution and strength) But as the reactions change the concentrations of different nuclides, this affects the neutron spectrum, which changes the relevant cross sections, and all of the reaction rates change. The neutron flux and energy spectrum, then, cannot be seen as fixed for a given reactor because it evolves with the changing composition of the fuel, and varies spatially across the reactor core. Finally, the neutron energy spectrum is different for different reactors depending on the coolant (for example, sodium or lead), the fuel type (for example, metal or oxide, and precisely which metals), the cladding, the operating mode, and the configuration (surrounded by a reflector, a fissile blanket, or a fertile blanket). While all reactors have a range of possible neutron spectra, that range tends to be larger for fast reactors. A reactor using fuel initially loaded with 15 percent plutonium that reaches high fuel burn-up, for example, 135,000 MWd/MTHM (around 15 percent burn-up), may still have substantial amounts of plutonium and other actinides in the spent fuel. This is because not all of the fissions occur in the plutonium. In uranium-plutonium dioxide (U-PuO2) fuel, most of the 7 The term burn is a nontechnical reference to fissioning a nucleus, but the nuclei also undergo other reactions that do not burn that nucleus. These reactions compete with each other for each impinging neutron, and while some actinide nuclei will split, others will be transformed into other heavy nuclides.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 69 fissions occur in the plutonium, but some of the fissions occur in the uranium, and neutrons absorbed in the uranium create more plutonium and other heavy nuclides that may fission or remain in the spent fuel.8 Such a reactor can be designed with a blanket of natural uranium or depleted uranium that produces plutonium, and overall the system may have more plutonium (more higher actinides) after a cycle than it had at the beginning of the cycle. One scheme for effecting improved burnout of particular species, such as neptunium and americium, is to separate them when processing spent fuel and load them into burning targets distinct from the reactor fuel. These targets could have compositions tailored to the chemistry and neutronic characteristics of the actinide atoms they hold. The parts of the core may then have modified spectra that are better for burning the targets, and the targets could reside in the core for longer or shorter times than a fuel assembly does. The range of required irradiation times or fluxes might dictate (economically) different target hardware, such as cladding. For systems that are designed to multicycle fuel, almost certainly different isotopes will require a different number of cycles to achieve desired reductions in actinides inventory. For such systems, analysts must also examine the efficiency and effectiveness of the processing facility. The product stream and the waste stream both matter for the resource utilization, waste hazard and footprint, and nonproliferation goals. At a more basic level, many of the critical reaction cross sections, and their variation with spectrum, are currently based on theory and analysis rather than measurements. Experimental measurements would seem to be essential before proceeding to actually plan the program. This issue could in fact have a major impact on the preferred type of reactor to build. The location in the reactor where the targets are placed will have a major influence on the results. In developing reactors for recycling of actinides, the United States and Russia have focused on sodium-cooled reactors. The burner reactor currently proposed in GNEP has not yet reached conceptual design (the decision whether to use metal or oxide fuel, which is essential to the design of the reactor core, has not been made). The Russian BN-600 has operated for 25 years on HEU fuel with some tests using MOX fuel. The BN-800 is under construction and is anticipated to be commissioned in 2012, and the government recently decided that the reactor will operate with vibropacked MOX fuel. It will start with recycled weapons plutonium. There is not now a plan for it to burn any of the higher actinides. B3. What impact could new technologies have on these proposals? The present commercial nuclear fuel cycle includes the mining and extraction of uranium, the purification of uranium ore, the conversion to uranium hexafluoride, uranium enrichment, fuel fabrication (including the conversion of uranium hexafluoride to uranium dioxide). Irradiation in a nuclear reactor is then followed by storage and either reprocessing or disposal of the irradiated fuel. France now accomplishes what GNEP envisions for the United States, to recycle some of the by-product plutonium in light-water reactors (LWRs) as mixed- oxide fuel (uranium dioxideâplutonium dioxide [UO2-PuO2]). It is postulated that in the future 8 An alternative is to use a different fuel material, such as thorium-zirconium-plutonium. Irradiation of thorium produces fissile uranium-233, but does not produce appreciable amounts of the higher actinides while it burns out the plutonium and its activation products. This could reduce the weapons-usability of the material dramatically and could reduce the quantities of radionuclides that dominate risk from some repositories (for example, neptunium- 237). Its benefits with respect to other goals, resource utilization, and repository heat load, are less clear. Please see below for further discussion of this point.
70 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE fast-neutron reactors will be used both to produce fuel (plutonium from uranium-238 or uranium- 233 from thorium) and to burn up (destroy by irradiation) minor actinides and some fission products. The fuel cycles of these fast reactors will be quite different from those of LWRs. There would be no need for uranium enrichment and much less need for uranium ore, but there would be a need to process breeder reactor blankets, and possibly to process burn-up targets, and irradiated fuel (possibly three or more quite different processes). All aspects of both the current fuel cycle and the future postulated fuel cycles are subject to significant changes due to new technologies. New, improved technologies could increase the attractiveness of nuclear power and support its widespread expansion. Although no technologies themselves solve the fundamental problems of internationalized fuel cycles, new technologies could improve nuclear fuel cycles in utilization of fuel resources, reducing quantities and hazards of radioactive waste, broadening the options for a high-level waste repository, making nuclear reactors more economically competitive in more situations, and reducing the proliferation hazards associated with fuel cycles. The critical questions about new technologies are how much of an improvement do they make, what new risks might they pose, over what time frames could they be realized, and do these improvements make a difference in the overall desirability of future fuel cycles? The joint committees can only begin to offer answers to these questions and make some general observations. First among the observations is that these topics are not areas of technology 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. Research is needed in the areas of processing of irradiated nuclear fuel and nuclear fuel design, as well as in improved approaches to disposal of wastes or spent fuel, and reduced-cost recovery of uranium from low-grade sources. Nuclear science and technology has reaped great benefits from improvements in computing for simulations and control systems and improvements in techniques for fabrication and processing materials (for example, improving the corrosion resistance or hardness of metals, and high-accuracy manufacturing or machining to very small tolerance). But many of the science and technology needs within the nuclear energy sector, especially those affecting the improved design and fabrication of nuclear fuel, reactor design and operation, and irradiated fuel processing, are unique to this sector. The relevant nuclear reaction cross-section data are unlikely to be gathered for other industrial sectors. Satellite manufacturers and space agencies are interested in âradiation resistantâ materials, but making materials that are durable in a reactor environment is a challenge that is unique to nuclear reactor engineering. Irradiated nuclear fuel comprises a diverse set of chemical constituents. Carrying out separations on these constituents, some of which are processed in no other context, in the high-radiation environment they generate poses a unique challenge. These challenges belong squarely to the nuclear energy sector, and a committed effort to research and development in nuclear science and technology is needed to make progress on them. Some nationsânotably France, Japan, Russia, and South Koreaâhave made progress in development of nuclear reactors, nuclear fuels, nuclear fuel processing, and their nuclear energy enterprise. The UK experience with reprocessing has not been favorable.9 The governments of 9 Martin Forwood reports that âAfter a projected slow ramp-up period, THORP was to achieve 900 tons/yr in the sixth year of operation (1999) with fuel burn-up ranging from 3.1 to 24.0 GWd/MTU for AGR fuel, 7.4 to 28.8 for
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 71 these nations have been committed to making progress, in some cases at great cost. Russia has made important progress on several fronts, creating institutional structures and arrangements in parallel with commitments to expand its use of current-generation and next-generation nuclear power plants, and focused research and development to address several technological challenges. Many other nations have shown less consistent support and little commitment to nuclear energy. The United States increased its commitment in the first half of this decade. The domestic component of the U.S. Global Nuclear Energy Partnership would require substantial funds to achieve its ambitious goals. A recent National Research Council report (NRC, 2007) criticized this program largely because to achieve almost any combination of its stated goals, the program would have to rely on new technologies, but the program was framed to move rapidly toward construction of facilities using near-term technologies. This brings us back to the critical questions about new technologies: How much of an improvement do they make? Over what time frames could they be realized? Do these improvements make a difference in the overall desirability of future fuel cycles? The following discussion describes several areas of technology in which improvements could have a substantial impact on the options available for international fuel cycles: Better fast reactors; small, self-contained reactors; new technologies for enrichment; high burn-up fuels (one-pass reactors and multirecycled transuranic fuels); thorium fuel cycles; dry methods for fuel separations; and economic new sources of uranium. IMPROVED FAST REACTORS It may be possible to show that some fast-reactor designs would not require some of the safety systems required for LWRs. Some reactor experts have long suspected that fast reactors, through choices of fuel, configuration, and coolant, can be designed to be passively safe10 and store a relatively small amount of energy in their primary coolant systems. This could not be satisfactorily demonstrated in the past,11 but with a combination of new experiments and the computing power now coming available, more definitive answers to these questions may become available. If indeed reactors can meet these hopes, then in addition to the inherent benefits, the reduction of required safety systems might make these fast reactors cost-competitive with LWRs. BWR fuel, and 16.9 to 40.0 for PWR fuelâwith higher burn-up fuels reserved for later years of the base-loadâ¦.. [T]hroughput has neither been reliable nor to specificationâwith just over 5,000 tons completed during the first ten years of operation. Contributing to schedule slippage have been a range of equipment failures and accidents including acid spills, pipe leaks and blockages and problems with the plantâs sole high-level waste evaporator. By the end of the official base-load period, with plant closure scheduled for 2010/11 âwith all contracts completed,â THORP was running some two years late. No new orders were secured and none are currently in the pipeline.â (Forwood, 2008, p. 20). âThe UK reprocessing program has produced an accumulated separated plutonium stock of over 100 tons as of the end of 2006. It is considered to be an asset of âzero valueâ and it is as yet undecided whether the UK will treat it as a waste product or a future energy asset. The stockpile will increase to 133 tons if current reprocessing contracts are completed. Some 100 tons will be from UK-origin spent fuel and 33 tons from foreign fuel. The plutonium from foreign spent fuel is to be returned to overseas customers as MOX fuel,â (Forwood, 2008, p. 35). 10 Here passively safe means that the reactor has negative reactivity coefficients in cases of voids and temperature rises in the fuel and in the primary coolant, and that the primary coolant can remove sufficient heat from the core in believable accident scenarios. 11 Experiments on EBR-II demonstrated performance only for metal fuels.
72 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE SMALL, SELF-CONTAINED, DEPLOYABLE REACTORS Reactor designers in several countries have been developing designs for small- and medium-sized reactors with improved nonproliferation features for deployment in developing economies and specialized applications (IAEA, 2005a). An interesting subset of these is factory built and fueled reactors. All commercial, light-water power reactors worldwide have historically been built at the site of operation and are refueled by opening the reactor, removing fuel that is spent and placing it in temporary storage, and loading the reactor core with fresh fuel. For a typical light-water reactor core, the reactor must be shut down every 18 or 24 months to change out one-third of the fuel. Some other power reactors (RBMKs, the CANDUs, and some others) refuel online without opening the reactor. But different designs for small reactors enable the reactor vendor to construct the reactor core and primary coolant system at a factory and then deliver this self-contained unit to the site of reactor operation, which may be on land or on an offshore floating power station. The size of the reactor and its weight would allow vendors to deliver modules by river barge, rail, or airplane. With a long-life core (ranging from 7 to 25 years, based on design), the unit can be provided fully fueled and need never be refueled at the site of operation. There are safeguard advantages to a reactor that is sealed from the day it is made at the factory until the day it returns to the factory, at the end of its useful life. As noted elsewhere in this report, under a fuel-leasing agreement, the irradiated fuel would not remain in the lesseeâs possession, and so fissile material could not be separated from the fuel without breaching the agreement. Under a reactor-leasing agreement, the fuel would never leave the reactor core, which makes breaching the agreement even more readily detected. In part for these reasons, the GNEP goals included a call for small reactors, although the U.S. government has not sponsored much work in this area. The Russian Federation is much further along in developing compact, factory built reactors than any other country, having nearly completed construction of a floating demonstration reactor in Severpodvinsk in the Arkhangelsk Region based on one of its several detailed designs (KLT-40C, which has a 3-year refueling interval; see Table 3-1). The economic viability and attractiveness of these reactors will become clearer after one or more customers have operated them for more than one fueling cycle. The cost of electricity from such reactors right now is expected to be substantially higher than electricity from fossil fuel plants and more conventional nuclear power plants, so Russia is marketing them for specialized situations and applications. For example, mining and refining operations in remote locations are potential niches, as delivery of fossil fuels in Siberia is costly, and use of natural gas for recovery of oil from tar sands in Canada is inefficient. Remote populations could also be markets for district heat, desalination, or electricity. Countries that want nuclear power on a small scale might find attractive these simplified reactors with passive safety systems that could function in environments where there is not a highly developed nuclear infrastructure. Such reactors might become more attractive if production economies of scale bring the prices down. Japanâs Toshiba is also marketing a small sodium-cooled reactor with a long-life core. Toshiba 4S (Super Safe, Small, and Simple) nuclear power system is a 10-MWe reactor designed to operate for 30 years without refueling. The fuel is 24 percent uranium and 10 percent plutonium in a zirconium matrix. Like the Russian design, the 4S (also called the nuclear
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 73 battery) could be transported in modules by barge. A remote community in Alaska in the United States is considering purchasing one of these power systems.12 TABLE 3-1 Small Reactors Power Capacity Small NPP Electricity, Cogeneration Refueling Fuel MW interval, years enrichment, % Electricity, Heat, Gcal/h MW ABV-6 2x8.5 2x6 2x12 12 19.5 SVBR-10 2x12 2x6 2x25 12 18.7 Uniterm 2x6.6 2x2.5 2x17.2 25 19.5 KLT-40C 2x38.5 2x19.5 2x73 3 17.4 Ruta - - 60.2 3 3 VVER-300 300 220 450 2 3.3 VBER-300 2x340 2x215 2x460 1.5 19.5 VK-300 - 250 400 2 4 SVBR-100 4x101.5 4x95 4x130 8 16.5 4S 1x10 30 Pu-10% SOURCE: Adapted from IAEA 2005a. HIGH BURN-UP FUELS Advanced fuel technologies could have an impact on the options available for nuclear fuel cycles in that they are essential to the technical feasibility of several of the options. As is described elsewhere in this report, nuclear fuel cycles need to be considered as systems and evaluated against criteria or goals. Different nuclear fuels and fuel cycles can accomplish different goals in different ways. A fuel cycle may achieve high energy utilization by achieving high burn-up in the fuel in a once-through fuel cycle (such as plutonium-thorium-zirconium fuel for sodium-cooled reactors and silicon-carbideâcoated [Si-Câcoated] fuel for gas reactors) or by multirecycling fuel (transuranic fuel from breeding and burning fuel cycles). The fuel matrix may be metal, oxide, nitride, or carbide, or even some dispersion combination with other materials. These examples are given as illustrations; among experts there are different opinions about each fuelâs state of development and which options best fulfill specified goals. Reactor designers must overcome significant materials challenges with the assistance of materials 12 See Overview of Galenaâs Proposed Approach to Licensing a 4S Nuclear Reactor Based Power Generation Facility, 2008.
74 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE scientists before any of these fuel technologies can be demonstrated to fulfill their promise and be deployed. By high burn-up fuel, the joint committees mean fuel that achieves more than 200 MWd/kg, or at least four times the level of current light-water reactor fuel. High burn-up fuels must have two characteristics: They must be able to endure the physicochemical and radiation conditions over their long irradiation lifetimes, and they must contain or make enough fissionable material to sustain energy production over the fuel lifetime. Each of the fuels mentioned above has different advantages: A long-life plutonium- thorium-zirconium fuel would burn out nearly all of its fissile plutonium isotopes and sustain itself on uranium-233 bred from the thorium-232 in the fresh fuel. The proliferation aspects of uranium-233 fuel are described later in this report. This fuel is highly refractory, and it is therefore difficult to separate its constituents. Silicon-carbideâcoated fuel spheres are likewise difficult to process for separations, and they have attractive features for reactor safety (especially the high coefficient of thermal expansion, which affects reactivity, and very high melting temperature) and are durable even under extensive irradiation. Both fuels can accommodate a range of actinide compositions and can be designed to accommodate fission products that accumulate during irradiation, and their robustness enables them to survive the radiation damage. Multirecycling that burns the actinides requires transuranic fuel; that is, a fuel with high transuranic content that maintains stability but can also be processed for separations. Setting aside economic viability of these systems, design and fabrication of such fuel has been identified as the greatest technical challenge for fuel cycles considered under the advanced nuclear energy development program proposed in the United States in recent years. A system that retains the higher actinides within the fuel materials to reduce the direct usability of the materials streams in weapons and to reduce the actinide content of the waste streams faces the challenge of creating fuels that have never been fabricated and run before. For metal fuels, a major challenge is retaining americium and curium in the fuel during fabrication because those elements are volatile in the temperature range in which the tested fuel fabrication techniques operate. There is some experience with the EBR-II from the 1960s and 1970s, but this provides only a starting point and some lessons. For oxide fuels, the fabrication problems are somewhat easier but still have not been demonstrated. A heterogeneous reactor core containing both fuel rods and target rods (containing the americium and curium) would simplify meeting the actinide burning goal, but compromises on the goal of maintaining the actinides together in the materials streams.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 75 BOX 3.4 SAMPLE LESSONS FROM THE EXPERIMENTAL BREEDER REACTOR II By Milton Levenson Perhaps the most important lesson from the Experimental Breeder Reactor II (EBR-II) was that to be successful the entire system had to be considered in design and operation. Originally EBR-II was three independent projectsâthe reactor, the fuel, and the fuel cycleâeach being pursued in a different division at Argonne National Laboratory. After the completion of the Title I design, Dr. Walter Zinn, director of Argonne, instituted a single design review of all three projects at the end of which he combined the three into a single project under a single project manager. During the combined review many changes were made in all aspects so as to optimize the total project. Some examples include the following. Before the reactor was built, an analysis was done to determine the fuel composition after infinite recycling, considering the residues that would be left in the fuel by pyroprocessing. This composition, called fissium, was then the original fuel composition using stable isotopes in the original fuel. By this means there was very little change in the fuelâs chemical composition with recycling, and the process always saw the same material as feed. The fuel was considered as a three-component system with three services: (1) to power the reactor, (2) to be remotely disassembleable, and (3) to be remotely fabricable. The components were the fuel matrix, the clad material, and the geometry. The final fuel was different from any original design, but one that satisfied all needs. The very high burn-ups (200,000 MWd/MTHM) were achieved by adjusting all three. Fuel to clad gaps were adjusted so that the fuel could swell only enough to allow fission gas bubbles to connect and so vent to the cladding. This stopped the swelling pressure that otherwise would rupture the cladding. Research continued on both cladding and fuel, but not independently. To address the question as to whether a âventedâ fuel element might be used in the futureâor whether a cracked cladding might be considered dangerousâone subassembly was fabricated without any cladding and run in the reactor at full power with no significant adverse affects. Some iodine migrated into the sodium as sodium iodide and was removed in the cold trap. Some xenon and krypton migrated into the cover gas closed system. However, this success is relevant only to a metal fuel alloy that is thermodynamically stable when in contact with sodium, which illustrates the point about a systems- design approach. In addition to the fuel matrix itself, other fuel materials must be able to perform reliably throughout the fuelâs residence within the reactor core. Most fuels13 have a metal casing, called cladding, that separates the fuel matrix from the primary coolant. Cladding performance has been the limiting factor for burn-up of fuels in the past. Minor defects in manufacturing of cladding can lead to cladding breaches (failed fuel), which releases radioactive material into the coolant and allows the coolant to interact directly with the fuel, as happened commonly with boiling water reactors until the 1990s. Even as manufacturing quality improves, however, reactor designers and operators must grapple with more fundamental limitations of the cladding materials, as the accumulated radiation damage for high burn-up fuel exceeds the claddingâs ability to self-anneal and maintain its integrity. Mechanical damage, too, can be a factor, as pressure and flow-induced vibrations strain cladding and other fuel materials. 13 The silicon-carbide spheres mentioned above are an exception, although some of these fuels are encased in larger graphite shells. Liquid fuels such as molten salts are another exception.
76 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE For the postulated fast-reactor cycle, new technology is required to bring it into being. Such cycles are much more interactive than those of the LWRs: The fuel, the cladding, the blanket, the coolant, the fuel processing, and the blanket processing must be treated as a single system if the result is to be an effective solution (see Box 3.4 for some lessons learned from EBR-II). Significant pieces of a fast-reactor fuel cycle do exist, and the proof-of-principle as a power source has been established, but there does not exist a fuel-reactor-process integrated system that can be demonstrated to be either economic or operationally successful as long as there is not an agreed set of metrics and criteria that defines success. THORIUM FUEL CYCLES The use of thorium in reactors has been studied for several decades. The AVR 15 MWe experimental pebblebed reactor operated at Julich, Germany, from 1967 to 1988. Based on the AVR, the 300 MWe thorium high-temperature reactor (THTR) in Germany operated from 1983 until 1989. In the United States, the Fort St. Vrain 330 MWe high-temperature gas-cooled reactor (HTGR) operated from 1978 to 1989 with thorium/HEU fuel. It never operated well, but the problems were not associated with the use of thorium in the fuel. Also in the United States, the Shippingport LWR operated as a breeder reactor from 1977 to 1982 using the Radkowsky seed-and-blanket design. In India, Kakrapar 1 and 2 use some thorium fuel (WNA, 2008). World resources of thorium are four to five times greater than those of uranium. Introduction of thorium fuel cycles would tap those resources for power generation and could reduce the waste disposal and proliferation hazards of nuclear power engineering, depending on how such cycles were implemented. Neither uranium-233 nor plutonium is found in significant quantities in nature, and so they must be produced in a reactor to acquire enough material to fuel a reactor. One approach that has been proposed in recent years is to use uranium-thorium seed-and- blanket fuels in existing LWRs. This would result in lower total quantities of spent fuel per unit of electricity generated and lower inventories of plutonium isotopes, making the spent fuel a less attractive source of plutonium for weapons (see, e.g., Galperin and Todosow, 2001). Another approach that has been discussed extensively is to use thorium-fueled molten salt reactors, with continuous partial removal of fission products, so that no weapons-usable uranium-233 is separated from the liquid fuel (see, e.g., Gat et al., 1993a). Moreover, due to the absence of uranium-238, accumulation of minor actinides in the thorium fuel cycle is considerably slower than that in the uranium fuel cycle. Thorium-uranium-233 fuel cycles generally build up small concentrations of uranium- 232, whose decay products, bismuth-212 and thallium-208, emit hard gamma rays. This would mean that those who worked for many hours with typical uranium-233 would receive doses beyond worker health and safety limits, so a commercial operation, even at the early stages of introduction, would require the use of heavy biological radiation shielding and automated remote-controlled equipment. But the uranium-232 concentrations are typically insufficient to prevent advanced states from making weapons from these materials using frequent worker rotation insufficient to prevent terrorists from making crude bombs from uranium-233. Uranium-233 has a much lower critical mass than uranium-235, and unlike plutonium, it has low-enough neutron generation that it can be used in simple gun-type bombs. It is therefore a
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 77 dangerous potential nuclear weapons material requiring stringent security and accounting measures. Advances in aqueous and nonaqueous technologies could be introduced in future thorium fuel cycles. As noted earlier, there are no significant uranium-233 resources in nature, so for a thorium fuel cycle to be introduced, a plant must breed uranium-233, initially relying on some other fissile fuel to power the reactors. The accumulated uranium-233 can then be used to fuel industrial fast and thermal reactors. In the initial stage of thorium fuel cycle development, Russia envisions reprocessing the following fuel types: thorium metal or thorium dioxide (ThO2) of the BN-800 reactor blanket; PuO2-ThO2 fuel of the VVER-1000 reactor; 233UO2-235UO2 fuel (hereinafter UO2 fuel) of the VVER-1000 fuel; UO2-ThO2 fuel of the VVER-1000 reactor. The purpose of reprocessing a fast-breeder-reactor-irradiated thorium blanket is to extract uranium-233 and return the remaining thorium back to the reactor for further irradiation. Other fertile fuels rely on fissile plutonium or uranium as the initial source of fission reactions in the fuel. PuO2-ThO2 fuel for thermal reactors can extend the use of the fuel with good nuclear- physical properties, burning out the plutonium without uranium-233 extraction. Fuel composed of 233UO2-235UO2 can be reprocessed to purify uranium from minor actinides and fission products and recover uranium reactivity by adjusting its isotopics. Finally, UO2-ThO2 fuel can be reprocessed to extract uranium from minor actinides and fission products and adjust fuel reactivity according to the thermal reactor requirements. For the second stage of Russiaâs thorium fuel cycle development, fast and thermal reactors with UO2-ThO2 fuel are being explored. The BN core is expected to operate using fuel containing minor actinides and long-lived fission products. In this case, the purpose of fuel reprocessing is its purification and adjustment of the composition to conform to the requirements of a uranium-233 fueled reactor. The uranium-233 accumulates in the fast breeder reactor alongside the metallic thorium blanket. The purpose of the BN metal blanket reprocessing remains the same, that is, extraction of uranium-233 and return of thorium into the reactor for further irradiation in the blanket. However, in a closed thorium fuel cycle, the purpose of thorium-blanket reprocessing is only to increase the uranium-233 fraction to the requirements of the reactor for which this fuel is produced. âDryâ technologies of a thorium fuel cycle can be based on the following processes: â¢ hydrogenation of metallic fuel â¢ chlorination of metallic and oxide fuel â¢ sublimation and vacuum distillation of thorium and uranium tetrachlorides â¢ electrolysis of molten salts â¢ concentration of minor actinides and fission products â¢ production of fuel compositions, fuel elements, and fuel assemblies These options are summarized in Table 3-2, which presents the types of reactors to be implemented during both stages of thorium fuel cycle development in Russia, and the purpose of reprocessing the blanket and core fuel.
78 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE TABLE 3-2 Reactor Types, Fuel Types, and Purposes of Fuel Reprocessing in the Thorium Fuel Cycle Stages of Reactor type, fuel Fuel Purpose of reprocessing Thorium Fuel accommodation blanket and core fuel Cycle Metallic Th Blanket Extraction of U-233 BN-800 ThO2 Stage 1 Core MOX Recovery of NPhP Accumulation Extraction of U-233, PuO2-ThO2 of U-233 Recovery of NPhP VVER-1000 Core 233 UO2-235UO2 Recovery of NPhP UO2-ThO2 Recovery of NPhP Extraction of U-233 Blanket Metallic Th Stage 2 Adjustment of U-233 content Closed BN-800 Recovery of NPhP, Thorium UO2-ThO2 Core Introduction of ÐÐ and Cycle + ÐÐ + LLFP LLFP VVER-1000 Core UO2-ThO2 Recovery of NPhP LLFP â long-lived fission products NPhP â 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 technologyâa combined reactor and fuel-processing unit, such as a molten salt reactorâhas 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
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 79 oxides. Pyrochemical technology is able to carry out all of the deposit production operations in one apparatusâa chlorinator-electrolyzerâwhich simplifies the process. Russian pyrochemical reprocessing consists of three main stages: 1. dissolution of initial products or spent nuclear fuel in molten salts 2. precipitation of plutonium dioxide or deposition of electrolytic uranium and plutonium dioxides from the melt 3. 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): 1. reprocessing uranium fuel with the production of uranium dioxide for recycling 2. reprocessing MOX fuel for only plutonium recycling as the most valuable component 3. 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: â¢ 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. â¢ The granular form of the fuel feedstock enables vibropacking technology to use both homogeneous compositions and mechanical mixtures for heterogeneous compositions. â¢ The thermal-mechanical stress on the cladding is lower for vibropacked fuel than for pellet-stacked fuel. â¢ 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
80 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 81 â¢ 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. â¢ 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. â¢ 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 regulatorâs 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 countryâs 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.
82 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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 7Ã105, 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).
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 83 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 repositoryâs 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 âroad blocksâ 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 worldâs 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 â[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.â (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).
84 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE 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.
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 85 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.â 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:Âª â¢ 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 â¢ 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 â¢ 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 â¢ 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 â¢ 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 â¢ 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
86 INTERNATIONALIZATION OF THE NUCLEAR FUEL CYCLE â¢ 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 â¢ 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, âwith the technical assistance and concurrence of the secretary of energy and after consultation with the (Nuclear Regulatory) Commission.â 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. â 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). Âª 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 â123 Agreements for Peaceful Cooperationâ 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
FUEL REGENERATION OPTIONS TO SUPPORT AN INTERNATIONAL NUCLEAR FUEL CYCLE 87 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 partnersâ 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 committeesâ 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 countriesâ 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.