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Nuclear Wastes: Technologies for Separations and Transmutation APPENDIX F TRANSMUTATION CONCEPTS ADVANCED LIQUID-METAL REACTOR Introduction Liquid-metal reactor technology has been under development for over 40 years in the United States and several other nations. Its principal potential application would be as an advanced nuclear power reactor, capable of breeding new fuel from natural or depleted uranium. It has generally been assumed that the advanced liquid metal reactor (ALMR) will be needed in some future era when low-cost resources of uranium are depleted, so a breeder could avoid the rising fuel costs of light-water reactors. Although natural uranium is abundant and readily available on the world market, some nations, such as France, the United Kingdom, and Japan, have stated their intent to develop breeders to reduce their dependence on imports of natural uranium. Although some of the early fast-spectrum reactors ran into serious problems, the more recent experience in the United States presents a very positive picture of fast-reactor technology. The performance of the EBR-II, in operation regularly since the early 1960s, and the 400 MWt (megawatt thermal) Fast Flux Test Facility, commissioned in 1979, have demonstrated that a medium-size liquid-metal-cooled fast reactor can be highly reliable. The fast spectrum that makes an ALMR an efficient producer of fissile plutonium can also be used to transmute plutonium and minor actinides as an aid to waste disposal. Studies of fast reactors for transmutation were carried out in the late 1970s and early 1980s. In the late 1980s, the Department of Energy (DOE) project to develop the ALMR as an advanced power reactor focused also on applying the ALMR as a transuranic burner. France and Japan have shown similar interest. In the United States, this work has been conducted by the Argonne National Laboratory (ANL) and General Electric (GE), under DOE funding. For this application, core configurations that would operate at breeding ratios varying from 1.25 to 0.22 have been studied.1 The reactor technology for transuranic transmutation is similar to, but more demanding than, that for the more familiar breeder application. Consequently, there is a large body of experience to aid the development of the ALMR as a burner of transuranic waste. 1 Breeding ratio is generally defined as the ratio of new fissile atoms produced to fissile atoms consumed. The ALMR project arbitrarily defines breeding ratio as the ratio of the number of atoms of new 239Pu and 241Pu formed to the number of atoms of 239Pu and 241Pu consumed by neutron absorption. The definition is arbitrary, because all atoms of plutonium, as well as the minor actinides, fission in a fast spectrum. With this definition, an ALMR core containing no 238U could still have a finite breeding ratio because of neutron absorption in 238Pu and 240Pu. For an ALMR with no 238U ANL estimates a breeding ratio of 0.22.
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Nuclear Wastes: Technologies for Separations and Transmutation In this section, the ALMR work of ANL and GE is summarized, including the ANL proposal for an integral fast reactor (IFR), an ALMR with integrated fuel reprocessing, for power generation and waste transmutation. Reference ALMR Designs GE has the responsibility for reactor design for the ALMR project. GE's power reactor, innovative, small module (PRISM) concept is a modular ALMR generating 1,395 MWe (megawatts electric), comprising nine reactor modules arranged in three power blocks of 465 MWe each. As shown in Figure F-1, each power block in turn comprises three compact, sodium-cooled, pool-type fast reactor modules of 471 MWt with its own steam generator that is heated by secondary sodium piped from the intermediate heat exchangers in the reactor module. The design goals emphasize passive shutdown heat removal for loss-of-cooling accidents and passive reactivity reduction to a safe, stable state for undercooling or overpower events with failure-to-scram. The current PRISM-type ALMR uses a metal-alloy fuel that is being developed concurrently by ANL. An ALMR can be designed as a breeder, such that the rate of production of new fissile material (plutonium and other transuranics) by neutron absorption in uranium equals or exceeds the rate of destruction of fissile material by fission and neutron capture. Such a breeder could be started in one of four ways: (1) with transuranic elements (TRUs) recovered from reprocessing spent light-water reactor (LWR) fuel, (2) with 25-35% enriched uranium, (3) with surplus military plutonium made available from the disarmament program, or (4) with excess plutonium produced by breeders of earlier generations. By reducing the amount of neutron absorption in uranium in the core and blankets, an ALMR can be designed to operate at a breeding ratio less than unity. This would require a continuous make-up of TRUs (e.g., from spent LWR fuel) as well as the start-up inventory, thereby increasing the mass rate of consumption of TRUs recovered from spent LWR fuel. Two core configurations for the PRISM-breeder application and proposed TRU burner are shown in Figure F-2. The ''actinide burner" core contains no radial and internal blanket assemblies to reduce the generation of new 239Pu by neutron absorption in 238U, resulting in a lower breeding ratio. The amount of TRU start-up inventory required depends somewhat on the breeding ratio at which an ALMR is operating. For a 1.4 GWe PRISM-type ALMR, the TRUs for a start-up core plus two reloads ranges from 27.2 Mg for a breeding ratio of 1.11 to about 14.6 Mg for a breeding ratio of 0.65. ALMR designs with a wide variation of breeding ratios, over the range of 1.25 to 0.22, have been considered as part of the overall ALMR design study. Variations (Thompson et al., 1991) of the reference ALMR breeder design involve elimination of external blankets containing fertile material to reduce the conversion of uranium to TRUs. To increase neutron leakage, and thereby to reduce fissile breeding in the core, these burner designs require shortening the core height, with minor adjustment in fuel composition but without changes in fuel-rod diameter or in the number of fuel rods. This approach to enhance TRU depletion is based on the key constraint that the diameter of the reactor vessel remain constant, and it is obviously limited by
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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE F-1 ALMR power plant and ALMR reactor facility. SOURCE: Quinn and Thompson (1992).
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Nuclear Wastes: Technologies for Separations and Transmutation
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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE F-2 Core layout comparison.
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Nuclear Wastes: Technologies for Separations and Transmutation the maximum allowable power density or by the linear heat generation rate of the fuel rods, so that safety characteristics are not compromised. With reduced fissile breeding and reduced heavy metal inventory, the burner designs also result in increased reactivity swing over a fuel cycle. This then requires larger control rod worths and hence entails potentially larger positive reactivity insertions and degraded performance in transient overpower events. As the breeding ratio decreases, there is less reactivity insertion resulting from sodium voiding in a power excursion. However, with decreasing breeding ratio, less negative reactivity is available from Doppler broadening of the neutron absorption resonances that occurs when the fuel is heated in a power excursion. Based on these considerations, GE concludes that a TRU burner with a breeding ratio of 0.60 and a core height of 0.76 m (30 in.) is the lowest possible breeding ratio configuration that would have acceptable safety features. Designs with breeding ratios varying from 0.22 to 1.25 appear in the ALMR/IFR project literature (Chang, 1992b; M.L. Thompson, 1991 and private communication, 1991; K. Wu, private communication, 1991). Recycle of minor actinides (MA) in an ALMR results in greater decay heat and radioactivity in discharge fuel assemblies. The ALMR project anticipates that the fuel handling equipment, including transport casks, can accommodate the higher levels of decay heat and radioactivity without major design modifications (Thompson et al., 1991). The ALMR project (Johnson et al., 1990; M.L. Thompson, private communication, 1991) reports preliminary designs for ALMRs fueled entirely with TRUs, with no 238U for breeding or internal conversion. One design is fueled with a mixture of plutonium and separated MAs, with a breeding ratio of 0.22. Another is fueled with MAs, with a breeding ratio of 0.85.2 Such an MA burner could be useful if separated plutonium were to be recycled to LWRs as mixed-oxide (MOX) fuel—as is being done in Europe and planned in Japan, and if it were found desirable to construct a separate fast-spectrum reactor to transmute the minor TRUs separated in chemical reprocessing of discharge MOX fuel. These designs obviously indicate higher mass depletion rates of TRUs supplied to the liquid-metal reactor (LMR) fuel cycle for start-up and make-up, per unit of thermal power generated, compared with those of the designs of transmutation devices without internal breeding from 238U or 232Th fertile material (e.g., the designs for the accelerator transmutation of waste [ATW] without thorium) considered in the Los Alamos Concepts of accelerator-driven transmutation systems. However, eliminating internal breeding from fertile material results in shorter fuel residence time in the reactor and lower fuel burn-up in an irradiation cycle. Consequently, the TRU inventories in the external fuel cycle increase, resulting in larger inventories required for start-up before steady-state recycle is achieved. Effects of the start-up inventories and TRU make-up rates on overall transmutation performance are discussed under Transmutation Performance. 2 The higher breeding ratio results from neutron absorption in 237Np, producing 238Pu that then absorbs a neutron to form 239Pu.
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Nuclear Wastes: Technologies for Separations and Transmutation Transmutation Performance Requirements Performance requirements for TRU-burning ALMRs have changed markedly in recent years and have not been clearly defined. Varying lines of objectives and arguments have been presented during recent years by DOE and its contractors, ANL and GE. To illustrate the varying lines of argument, in January 1990, DOE (Griffith, 1990) outlined the objectives of the ALMR project with regard to improving waste disposal of spent LWR fuel, stating that the ALMR project expects to develop a new pyrochemical process that will remove 99.9 to 99.999% of the actinides from waste streams, corresponding to process decontamination factors of 103 to 105.3 This expectation was said to result in a thousandfold reduction in the quantity of actinide waste going to a geologic repository. The principal benefit claimed for this reduction in the quantity of TRUs is that the resulting wastes would be expected to be "benign after 200 to 300 years," when the 90Sr and 137Cs fission products will have decayed, reducing the risks from geologic disposal (ANL, 1991; Berglund et al., 1990; Hannum, 1992). ANL (Till, 1990) has stated that the waste from actinide-burning ALMRs "… will not have any actinides in it such as uranium, plutonium or any of the minor actinides such that, after 200 years, the cancer risk to man will be no greater than the original form, i.e., natural uranium." Indeed, if the risks from a given quantity of TRUs in a geologic repository over a given period of time are considered to be serious, one must also consider the risks from significant quantities of TRUs that would remain as yet untransmuted in the ALMR's above-ground fuel cycle—in storage, in reactors, and in reprocessing and fuel refabrication facilities—during the long time period that the transmutation would have to be accomplished.4 Apparently, the objective of such a large reduction in TRU inventory in repository waste is based on the ALMR project's desire to reduce the "radiological risk" of the TRUs (Till and Chang, 1989) in a repository to the "cancer risk level" (Berglund et al., 1990) of the original uranium ore mined to fuel the LWRs that generated the spent fuel. The ALMR project apparently assumes that these "risks" are measured by toxicities. The errors in using toxicities and uranium ore comparisons are discussed in Chapter 4. DOE has clarified some important consequences of the ANL claims. In testimony to Congress in April 1990, DOE stated that the TRU-burning fast reactor "… would reduce the life of the remaining waste to about 200 years, instead 10,000 years, and the waste could then 3 Although the transmutation by ALMRs is said to deal with "actinides," the DOE ALMR project actually limits the calculation to transmutation of TRUs. Transmutation of the considerable quantities of uranium recovered from LWR spent fuel is not considered. Because TRU transmutation is spoken of popularly as "actinide burning" within reactor programs in this country and abroad, that terminology is occasionally used herein. 4 Estimates presented in Chapter 5 show that the time to achieve only a hundredfold reduction in total TRU inventory by the proposed ALMRs would be about 2,800 years for a series of ALMRs of a breeding ratio of 0.65, operating at constant power, assuming sufficient production of TRUs from LWRs to furnish TRU make-up during this period, or about 200 years in a declining-power scenario in which TRU inventory at the end of each reactor life is used to start and fuel a smaller number of subsequent ALMRs. The times for inventory reduction are even longer for other breeding ratios reported by the ALMR project.
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Nuclear Wastes: Technologies for Separations and Transmutation be stored above ground until it dies out" (Watkins, 1990). More recently, ALMR proponents have stated that a geologic repository will still be needed for some of the waste streams from the reprocessing cycles, because the process decontamination factor is not sufficiently high to qualify the waste as Class C "non-TRU" waste (i.e., with an alpha activity of less than 100 nanocuries per gram [Till and Chang, 1989]). Another way of approaching performance requirements is in terms of the DOE schedule (Griffith, 1990) that showed the introduction of 1.4 GWe of new TRU-burning ALMRs in the year 2005, followed by the introduction of an additional 1.4 GWe ALMR per year for the next 40 years. All of the 62,000 Mg of LWR spent fuel accumulated by 2011—the time of opening the first repository—could be reprocessed and used to fuel these ALMRs, together with the yearly discharge of additional spent fuel from LWRs assumed to operate during this 40-year period. By the year 2045, there would be no inventory of LWR spent fuel. However, a geologic repository would still be needed for defense wastes and for the radioactive wastes produced in ALMR operation. The ALMR project does not propose to develop separation processes with decontamination factors sufficiently high that TRU-containing wastes could be disposed of as low-level waste, nor does it propose to transmute fission products present in LWR spent fuel and generated by ALMR operation.5 The Department of Energy's 1991 National Energy Strategy (DOE, 1991) assumed a similar rate of introduction of TRU-burning ALMRs, beginning in 2012 (Goldner et al., 1991), achieving a level of 19 ALMR power plants by 2030. Each ALMR was assumed to generate 1,395 MWe and to operate at a breeding ratio of 0.6 or 1.01. The National Energy Strategy also assumed a growth of total U.S.-installed nuclear capacity from the present installed capacity of about 110 to 195 GWe by 2030, primarily with advanced LWRs, introduced at a rate much greater than that assumed by Griffith (1990). The cases analyzed show that with such ALMR growth, all spent fuel discharged from LWRs from 2012 to 2030 could be processed to fuel the ALMRs, and a portion of the stockpile of LWR spent fuel could also be processed and transmuted during this period. The new ALMRs would be primarily to treat new spent fuel emerging from LWRs after the first spent-fuel processing plant is built, around 2010 or earlier (Goldner et al., 1991). However, if the ALMRs are available and economically competitive, they could be applied to process and transmute LWR spent fuel that would otherwise be emplaced in the first U.S. geologic repository for high-level waste (HLW). Therefore, if the TRU-burning ALMR program were successful on the schedule expected by DOE and its contractors, there would be no need to load the first U.S. HLW repository with spent fuel (Griffith, 1991). This would call for considerable change in the repository design, in repository safety analysis, in the qualification of waste forms, and in licensing. Although DOE has not defined performance requirements for the ALMR, the committee finds that the ALMR fuel reprocessing development underway at ANL is aimed at achieving decontamination factors of about 1,000. Therefore, for the purpose of the current study, a 5 A 1991 report from ANL (1991) states that fission products are also to be transmuted in the ALMR. However, no information has been provided that shows that the ALMR program is planning to transmute the long-lived fission products that are key radionuclides requiring geologic disposal.
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Nuclear Wastes: Technologies for Separations and Transmutation decontamination factor is adopted as a performance goal, and evaluations are made of what overall system performance can be achieved with the various ALMR designs. Transmutation Performance While the reactor is transmuting the various isotopes in the fuel assemblies, it is producing higher mass TRUs and fission products in the fuel and blanket assemblies. Calculated annual production and destruction rates of various TRUs in the ALMR are calculated by GE (Thompson et al., 1991) assuming an average fuel burn-up of 150,000 MW·d/Mg, corresponding to a fuel residence time of about 6.5 years.6 At the end of the life of an ALMR at constant power, the core and reprocessing inventory would be essentially intact. These could be transferred to subsequent replacement ALMRs to continue power generation and the transmutation. In assessing the performance of the ALMR concept to consume TRUs that would otherwise appear in a geologic repository, it is important to consider the untransmuted inventory in the reactor and in reprocessing. It is also important to compare the total inventory of TRUs in the waste and in the fuel cycle to the amount that would be present for the same power generation from alternative sources, such as from additional LWRs. Calculations presented in the later section, Comparison of Reactor and Accelerator Approaches show that several hundred years would be required to reduce the total inventory of TRUs by even a factor of 10 by deploying ALMRs, and thousands of years would be required for a hundredfold reduction, assuming a constant level of nuclear power in the future. The time for a hundredfold reduction could be reduced to a few hundred years if nuclear power were to be terminated as rapidly as possible, consistent with operating a progressively smaller number of ALMRs at constant neutron flux as the reactor and fuel-cycle inventories of the system of several initial ALMRs become transmuted. Facilities and Support Requirements To implement the TRU-burning program as described by DOE (Griffith, 1990; Goldner et al., 1991), commercial-scale facilities to reprocess LWR spent fuel would be required. According to the GE ALMR reports, this facility could be an aqueous PUREX/TRUEX process or could be based on ANL's proposal to develop a pyrochemical separation process for LWR spent fuel. GE specifies a reprocessing capacity of 2,700 Mg of LWR spent fuel, more than twice the capacity of any single commercial reprocessing facility. Currently the largest worldwide commercial facility, near completion in the United Kingdom, is the THORP plant, with a throughput of 900 Mg heavy metal. Based on reprocessing/fabrication times estimated by the ALMR project (Taylor et al., 1991), the new commercial-scale LWR fuel reprocessing facility would have to be operational for about 2 years before the commercial ALMRs are 6 Fuel burn-up expressed in terms of MW·d/Mg is the average thermal exposure, in megawatt days of thermal energy per megagram (metric ton) of actinides (heavy metal) in the fresh fuel supplied to the reactor.
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Nuclear Wastes: Technologies for Separations and Transmutation introduced. Thus, the new large-scale reprocessing facility would have to commence operation by 2003 according to one DOE schedule (Griffith, 1991) or 2010 according to another (Thompson and Gonzaga, 1994). Argonne National Laboratory's IFR concept of the ALMR specifies that each ALMR power-generating facility would have its own integrated facility for reprocessing discharged ALMR fuel and for fabricating recycle fuel. For a breeding ALMR, the throughput of driver and blanket fuel discharged from an ALMR would be about 21 Mg/yr of uranium and TRUs for each 1,395 GWe ALMR (Johnson et al., 1990). The throughput of ALMR discharge fuel would decrease to about 13 Mg/yr for an ALMR with no uranium blanket. The reprocessing capacity quoted here is that necessary to process the backlog, following DOE's proposed schedule of deployment. Without large-scale reprocessing deployment in the United States, separation and transmutation is an impossible venture. There must also be facilities for converting the reprocessing waste streams into waste solids suitable for emplacement in a geologic repository or suitable for other waste disposal sites. Necessary processes and facilities, in addition to those described by the ALMR program, include: a process to remove actinides and soluble fission products from the inner surfaces of cladding hulls from LWR fuel and a process to produce a suitable waste solid from the cladding hulls; a process to remove noble-metal fission products from the uranium stream from pyrochemical reprocessing of LWR spent fuel if that uranium is to be sent to surface storage as now specified by the ALMR program, and a process to produce a suitable waste solid from the separated fission products; processes for high-yield (greater than 99.9%) recovery of radioiodine, 14carbon, and technetium if LWR spent fuel is to undergo aqueous reprocessing and processes to produce suitable waste solids from these separated radioelements; and processes to separate radioactive cesium and strontium, and to produce suitable waste forms for these separated radioelements, and processes and facilities for storage of these separated radioelements for 200 to 300 years before disposal as radioactive waste (this is a possible additional separation proposed by ANL [Chang, 1992b] and by DOE [Young, 1991] to reduce heat generation in a geologic repository and to increase the amount of waste that could be emplaced in a given repository area). Qualifying the several different fuel types and compositions that will arise from TRU recycling in the ALMR will be a formidable task. The ALMR project points out that removing most of the actinides from the HLW would result in a lower rate of heat generation by radioactive decay. This would allow an increase in the total waste loading of a geologic repository of a given area, if that waste loading is limited by the total rate of heat generation. The ALMR program also proposes that the heat-generating fission products 90Sr and 137Cs be separated when spent fuel is reprocessed. By storing the separated strontium and cesium on the surface for about 300 years, the first repository could accommodate more waste than now planned and could defer the time when a second repository
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Nuclear Wastes: Technologies for Separations and Transmutation is needed (Till and Chang, 1989; Young, 1991). However, Ramspott et al. (1992) point out that the current limits on waste loading in the proposed first repository are not based on heat generation rate. They point out that if the loading were limited by heat generation rate, other techniques could increase waste loading without the complexity of reprocessing and transmutation. If this option is to be pursued, it will be necessary to develop additional separations processes, means of storing the strontium and cesium for hundreds of years, and new waste forms for satisfactory containment of the strontium and cesium for later emplacement in a geologic repository. Containment of the 3-million-year 135Cs is particularly challenging. Development of the ALMR concept, the new technology for reprocessing LWR spent fuel and for reprocessing and fabricating ALMR fuel, and the additional processes for separating fission products and for preparing special waste forms would require extensive facilities for research, development, testing, and operational demonstration, including prototypes. Risk and Safety Issues The National Research Council's (1992) study of nuclear power options for the future summarizes safety features of the ALMR, which is designed only for power generation, and also summarizes risk and safety issues that must be resolved. Two transient safety tests at the EBR-II with metal fuel successfully demonstrated the safety potential of the ALMR concept. The study notes that: The properties of the metallic fuel and the large thermal inertia of the sodium pool are key to achieving reactor shutdown passively (i.e., without relying on operation intervention, active components such as control rods, pumps, valves, or the use of balance-of-plant for heat removal) while keeping temperatures low. However, the report cautions that the presence of a positive sodium void coefficient of reactivity in the ALMR/PRISM design is a continuing concern. It may require revisions to the present design such as a larger diameter core with a lower height. For ALMRs as TRU burners, GE's approach to eliminating 238U blankets in the ALMR designs of lower breeding ratio requires shortening the core height, with minor adjustment in fuel composition but without changes in fuel-rod diameter or in the number of fuel rods. The diameter of the reactor vessel would remain constant and is limited by the maximum allowable power density or by the linear heat generation rate of the fuel rods so that safety characteristics are not compromised. With reduced fissile breeding and reduced heavy metal inventory, the burner designs also result in increased reactivity swing over a fuel cycle. This then requires larger control rod worths and hence entails potentially larger positive reactivity insertions and degraded performance in transient overpower events. As the breeding ratio decreases, the positive sodium void coefficient of reactivity decreases, and the Doppler coefficient of reactivity remains negative but becomes smaller in magnitude.
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Nuclear Wastes: Technologies for Separations and Transmutation Transmutation Performance Requirements As noted previously, the stated objective for such a scheme is "to reduce the time required for storage of nuclear wastes." Although not stating it explicitly as a goal for Phoenix, the BNL proponents have noted that partitioning losses of 10-5 or less should be a minimum objective for any partitioning/transmutation scheme if it is to succeed in the United States. The rationale for this position is that if partitioning losses of 10-5 or less can be achieved for each of the "problem isotopes" (defined as 237Np, 241Am, 243Am, 244Cm, 90Sr, 99Tc, 129I, and 137Cs), then NRC standard 40CFR191 (that specifies how many curies per ton of heavy metal of these isotopes can be allowed to escape the repository in the first 10,000 years following disposal) could be met without containment that ensures isolation of the waste. Moreover, in theory, HLW could be converted to a waste meeting radionuclide specifications of Class C waste according to standard 10CFR61 if this same level of portioning loss could be met (although it is recognized that standard 10CFR61 does not apply to HLW). Net partitioning losses of 10-5 require process decontamination factors of 106 or better on some elements (e.g., americium and curium) to achieve the specification for Class C waste with a 10% transmutation machine. This puts severe demands on the separations side of the scheme. Facilities and Support Requirements The scheme envisioned would require the following components to support one or more Phoenix irradiation facilities: front-end reprocessing of LWR spent fuel, probably by a combination of PUREX and TRUEX, to separate waste into the requisite six streams with the process decontamination factors stated above (as noted above, this would have to operate with process decontamination factors of 106 or better, and; as in other schemes, effective decontamination of cladding hulls and process equipment would also be needed); remote target fabrication for the MA and iodine targets for Phoenix and remote fuel fabrication for the plutonium and technetium recycle into power reactors; intermediate reprocessing of Phoenix targets to separate plutonium and fission products from the MAs and iodine from the xenon products; again process decontamination factors of 106 would be required if the "minimum objective" for a transmuter is to be met by Phoenix; and an interim storage facility for cesium and strontium, if these isotopes are to be separated from the other fission product wastes. Residual Waste Management and Environmental Impact The primary benefit of this scheme is to minimize the inventory of species (MAs, iodine, and technetium) in the waste repository that contribute most significantly to the current
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Nuclear Wastes: Technologies for Separations and Transmutation estimates of long-term risk. If separation and isolation of cesium and strontium isotopes are also accomplished, their removal from the waste repository would substantially reduce the near-term heat load on the repository and would permit a higher concentration of waste to be disposed of; this would postpone the need for repository expansion or addition. If the "minimum objective" is met, the partitioning and transmutation scheme would obviate the need for a HLW repository at all. Risk and Safety Analysis To date, no detailed risk and safety analyses have been made. Hence, at this point only some generic issues for transmuters in general, and accelerator-based transmuters in particular, can be anticipated. SAFETY The promoters point to the speed at which accelerator-driven sources can be "turned off" and to the inherent safety of a subcritical assembly (i.e., that supercriticality accidents would not be design-basis accidents). However, the large reactivity swings anticipated for a 2-year cycle (Δk/k ≈ 0.25) and steps to counteract this (increased leakage, poisons) could introduce substantial changes in reactivity coefficients (e.g., sodium void worth, Doppler coefficients, etc.) that will have to be addressed; moreover, there may be safety issues associated with a failed beam raster (e.g., burning holes in targets with the attendant radioactivity release). In addition, decay heat will still be an issue, even with a rapid removal of the proton source. A liquid-sodium-cooled system provides an opportunity to configure the system for passive cooling under accident conditions, but this has not yet been addressed specifically. Safety considerations for the balance of the plant are probably similar to those for LMRs in general (sodium fires, sodium–water reactions, etc.). Of somewhat unique concern is the potential for Na-D2O reactions between the sodium-cooled MA targets and the D2O-cooled and moderated iodine blanket. LICENSING Licensing a Phoenix facility, a combination of a large accelerator and a near-critical fast reactor, can draw on issues and procedures developed for the component facilities to date, but new safety issues associated with operating the combination will have to be identified and resolved. Each of the support facilities (reprocessing, remote fuel fabrication, interim storage) has its own set of licensing issues that are not unlike those associated with these facilities for other transmutation schemes. The operational assumptions for this scheme are that the following activities are acceptable to the public and can be carried out in a manner acceptable to regulatory agencies:
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Nuclear Wastes: Technologies for Separations and Transmutation centralized fuel reprocessing; plutonium recycle in LWRs; the operation of a Phoenix-type waste transmuter; the operation of one or more reprocessing facilities for Phoenix waste; surface storage of separated radioisotopes for long periods (mainly cesium and strontium); and substantially increased cost of electricity (see below) to pay for (perhaps only perceived) long-term risk reduction in nuclear waste management. RISK Risk analyses will have to include: decreased risk from uranium-mining/milling as a result of plutonium recycle and energy generation from actinide burning; increased short-term risks associated with transmutation and separations operations and surface storage versus disposal of radioisotopes; and reduced long-term risks associated with inventory reductions of long-lived and water-soluble isotopes. State of Technology The Phoenix concept is still at a very preliminary stage. BNL has had no dedicated funding to support the development of Phoenix. Work has primarily proceeded using limited internal funding and a lot of voluntary effort. While a team of competent people have examined the first-order issues, feasibility studies are still required to proceed. Outstanding technical issues include the following: Accelerator performance: As noted earlier, the development group appears to have made reasonable estimates of what can be achieved in accelerator technology based on current knowledge, but considerable engineering improvements will be required to actually achieve this. These issues are common to ATW, although the accelerator requirements are less severe than for ATW-1. MA target performance: Again, almost no analysis of target performance has been made, other than to assume that by designing a target environment similar to FFTF, no "show stopper" materials issues will be introduced. Under these circumstances, 2-year fuel lifetimes are probably reasonable. However, there is little experience with purely MA-oxide materials in such an environment, and the integral performance of an MA-oxide fuel in this environment remains to be demonstrated. Moreover, the combined effects of proton and neutron irradiation on structural materials will have to be addressed (although this is probably minor). Issues of the influence of transmutants introduced by high energy neutrons, along with the "usual" list of
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Nuclear Wastes: Technologies for Separations and Transmutation radiation-damaged induced property and dimensional changes (hardening, embrittlement, accelerated creep, void swelling), will have to be addressed for the specific materials and environments envisioned for the Phoenix targets. Iodine transmutation: Little has been done to assess the iodine transmutation, other than to assume this can be done in a blanket surrounding the target. Further conceptual design is required; however, this is probably not a critical issue. Separations: As noted, the work to date has relied on the CURE study for guidance in what might be feasible from a separations point of view. Hence, all of the outstanding technical issues for CURE apply here as well. These may be exacerbated, however, if process decontamination factors of 10-6 are required in full-scale reprocessing plants to meet the "minimum objective." Additional waste: Additional waste generated by systems (contamination of fuel cladding hulls and separations equipment, radioactivation of Phoenix system components, etc.) has not been addressed. First-order estimates are that this will be small compared with the main sources of inventory. However, the impact of all activities on all of the waste streams needs to be addressed in any comprehensive incremental (decremental) risk analysis, as discussed above. Costs DEVELOPMENT COSTS Based on numbers provided by BNL and figures from similar proposals, a development effort is estimated to consist of the following: Phase I: a conceptual design study to address the issues listed above; the study would take at least 2 years at a funding level of about $12.5 million per year; Phase II: an engineering design study combined with the development of an engineering database for critical components (accelerator, target, balance of plant); this would be about an 8- to 10-year study at a level of about $100 million per year; Phase III: construction and operation of a demonstration plant; the cost may be near $2 billion and would take about 5 to 8 years to construct and demonstrate operations. Phases I, II, and III would overlap. SYSTEMS COSTS AND ECONOMICS Again, the estimates here are based on preliminary input. The estimated cost of the accelerator is about $2 billion (1992 dollars). The estimated cost of the subcritical target, sodium loops, and power plant is $5 billion (for a large power plant). The operation and maintenance costs are expected to be double that of a large power plant.
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Nuclear Wastes: Technologies for Separations and Transmutation As noted above, the effective efficiency of the plant would be about 24%. Coupling this with the 40% increase in the capital cost and a doubling of the operation and maintenance costs would mean that electricity would cost roughly twice as much to produce in Phoenix as in a large reactor. Hence, the Phoenix operator would either have to find a buyer for this expensive electricity or, more likely, charge the nuclear industry the difference in the name of improving waste disposal. If the average cost of production were about 7 cents/kW·h, this would mean charging the industry about $420 million per year in addition to separations and disposal costs to annually burn 2,600 kg of MAs (and 580 kg of iodine), or about $132,000 kg of MA and iodine burned. This cost must be compared with that of burning MAs and iodine in LWRs or ALMRs, and evaluated against the "intrinsic safety" of such a machine and the speed of inventory reduction over the reactor-based systems. REFERENCES Argonne National Laboratory. 1991. Research Highlights. 1990-1991. p. 14. Arthur, E. D. 1992a. The Los Alamos accelerator transmutation of nuclear waste (ATW) concept. Vu-graph presented to STATS Symposium, Washington, D.C., January 13. Arthur, E. D. 1992b. The Los Alamos Accelerator Transmutation of Nuclear Waste (ATW) Concept. ATW-92-60. Presented to the STATS Subcommittee on Transmutation, Los Alamos, N. Mex., April 15-16, 1992. Baetsle, L. H. 1993. Role and influence of partitioning and transmutation on the management of nuclear waste systems. Pp. 1235-1241 in Proceedings of the Symposium on Waste Management, Tucson, Ariz., February 28-March 4, 1993. R.G. Post, ed. Tucson: Arizona Board of Regents. Bairiot, H. 1984. Laying the foundation for plutonium recycle light water reactor. Nuclear Engineering International. 29(350):27-33. Bairiot, H., and C. Vandenberg. 1989. Nuclear Fuel Cycle in the 1990s and Beyond the Century: Some Trends and Foreseeable Problems. IAEA Tech. Report Series No. 305. pp. 65-69. Vienna: International Atomic Energy Agency. Benedict, M., and T. Pigford. 1981. Nuclear Chemical Engineering. New York: McGraw-Hill. Berglund, R. C., Y. I. Chang, S. Rosen, and C. E. Weber. 1990. Actinide recycle in advanced liquid metal reactors. Korean Atomic Industrial Forum. April 16.
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Representative terms from entire chapter: