APPENDIX G
EFFECTS ON REPOSITORY

BACKGROUND INFORMATION ON REPOSITORY PERFORMANCE

Repository Design and Operation

Nuclear waste repositories have been investigated since the late 1960s. Site evaluation and conceptual designs have been developed in many countries, including the United States, Sweden, and Germany. All concepts are largely based on disposal in repositories approximately 200 to 1,000 meters below the surface. The host rock for repositories varies. In the United States, volcanic tuff in the unsaturated zone at Yucca Mountain is now the only rock type being investigated. In Germany, a salt dome is being characterized, while Sweden is concentrating on granite under saturated groundwater conditions as the primary focus.

An underground configuration for the proposed Yucca Mountain repository, shown in Figure G-1 (reference: Yucca Mountain Site Characterization Plan), is representative of that in most repositories. Waste packages containing either spent fuel or solidified waste forms from reprocessing are emplaced in the floor of the mine drifts or, in some cases, directly in the drifts. In the United States, the wastes will consist of spent fuel removed from the reactors and some solidified high-level waste from U.S. defense activities. Each package will contain a few metric tons of spent fuel or an amount of reprocessing waste approximately equal in radioactivity and thermal power. Approximately 40,000 waste packages would be distributed over an area ranging from 1,000 to 2,000 acres. In the United States, each waste package is expected to include a high-integrity container capable of providing up to 1,000 years of total containment. The current design for the Yucca Mountain repository allows for a total of 70,000 metric tons uranium (MTU) with 63,000 MTU of spent fuel. Higher loading densities or use of additional area would allow an even greater capacity.

The operations at a repository consist of a major facility for receipt and packaging of the waste and underground excavation for waste emplacement. After waste is prepared in special canisters, it is moved underground and emplaced in a manner to preclude radiation exposure to operating personnel. The annual capacity of the repository is determined by the ability to receive, package, and emplace the waste. At Yucca Mountain, an annual capacity of 3,400 MTU per year is anticipated after an initial startup period of a few years. The start-up operations are planned for 2010 and should continue for about 22 years.

The density of waste per unit area of a repository is primarily determined by the heat produced by the waste. Other factors that contribute are the allowable excavation patterns, permissible canister spacing, and the need to avoid certain geologic features. Typical waste forms produce approximately 1 kW of heat per MTU for fuel 10 years out of the reactor. However, by the time of the earliest possible repository operation in the United States,



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Nuclear Wastes: Technologies for Separations and Transmutation APPENDIX G EFFECTS ON REPOSITORY BACKGROUND INFORMATION ON REPOSITORY PERFORMANCE Repository Design and Operation Nuclear waste repositories have been investigated since the late 1960s. Site evaluation and conceptual designs have been developed in many countries, including the United States, Sweden, and Germany. All concepts are largely based on disposal in repositories approximately 200 to 1,000 meters below the surface. The host rock for repositories varies. In the United States, volcanic tuff in the unsaturated zone at Yucca Mountain is now the only rock type being investigated. In Germany, a salt dome is being characterized, while Sweden is concentrating on granite under saturated groundwater conditions as the primary focus. An underground configuration for the proposed Yucca Mountain repository, shown in Figure G-1 (reference: Yucca Mountain Site Characterization Plan), is representative of that in most repositories. Waste packages containing either spent fuel or solidified waste forms from reprocessing are emplaced in the floor of the mine drifts or, in some cases, directly in the drifts. In the United States, the wastes will consist of spent fuel removed from the reactors and some solidified high-level waste from U.S. defense activities. Each package will contain a few metric tons of spent fuel or an amount of reprocessing waste approximately equal in radioactivity and thermal power. Approximately 40,000 waste packages would be distributed over an area ranging from 1,000 to 2,000 acres. In the United States, each waste package is expected to include a high-integrity container capable of providing up to 1,000 years of total containment. The current design for the Yucca Mountain repository allows for a total of 70,000 metric tons uranium (MTU) with 63,000 MTU of spent fuel. Higher loading densities or use of additional area would allow an even greater capacity. The operations at a repository consist of a major facility for receipt and packaging of the waste and underground excavation for waste emplacement. After waste is prepared in special canisters, it is moved underground and emplaced in a manner to preclude radiation exposure to operating personnel. The annual capacity of the repository is determined by the ability to receive, package, and emplace the waste. At Yucca Mountain, an annual capacity of 3,400 MTU per year is anticipated after an initial startup period of a few years. The start-up operations are planned for 2010 and should continue for about 22 years. The density of waste per unit area of a repository is primarily determined by the heat produced by the waste. Other factors that contribute are the allowable excavation patterns, permissible canister spacing, and the need to avoid certain geologic features. Typical waste forms produce approximately 1 kW of heat per MTU for fuel 10 years out of the reactor. However, by the time of the earliest possible repository operation in the United States,

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-1 An underground configuration for a nuclear waste repository. SOURCE: Hunter et al. (1989)

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Nuclear Wastes: Technologies for Separations and Transmutation approximately 60,000 tons of spent fuel will have been discharged with an average age of approximately 16 years. At least 10,000 MTU will have an average age greater than 30 years. The conceptual designs for repositories determine a reference value for thermal loading density by evaluating temperatures expected for the waste packages, the excavation drifts, and surrounding rock units. These predictions of temperature are compared with the expected operational and long-term performance concerns, and a loading density is established. At Yucca Mountain, approximately 57 kW/acre has been the reference value, although both higher and lower values have been considered. Waste emplacement configurations and schedules can be designed to achieve a wide variety of heat loading patterns, especially if underground ventilation is used to remove heat during operation. Long-Term Performance Assessment BASIS FOR EVALUATION The standards for the necessary resolution of radioactivity in nuclear waste repositories are still being developed; however, all proposed standards consider limits on the potential dose to individuals, total population doses, or maximum radioactivity releasable over periods of time 10,000 years and longer. Repository developers must evaluate all possible mechanisms that might result in releases of radioactivity to the environment and must determine the risk to humans that is associated with this release. Three measures that have been adopted or proposed in various countries to evaluate potential public health risk from geological disposal of radioactive wastes are: annual radiation dose to an individual,1 which is usually applied for expected releases or high-probability events; annual risk to an individual from radiation exposure, which takes into account the probability of exposure; and collective radiation dose to populations, integrated over all population exposed and over time from the beginning of geologic disposal. Possible benefits of partitioning and transmutation relative to each of these performance measures are discussed herein and are based on published analyses of the performance of conceptual repositories in various geologic media. 1   It is the practice in some countries to calculate the dose to the maximally exposed individual, a person whose entire intake of water is contaminated by radionuclides released from the repository and whose entire intake of food is grown in or nurtured by the contaminated water. Other analyses focus on the doses to an individual in the critical population group, as recommended by the International Committee on Radiation Protection.

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Nuclear Wastes: Technologies for Separations and Transmutation APPROACH TO EVALUATE PERFORMANCE The U.S. Environmental Protection Agency (EPA) has established a derived set of limits on the cumulative amount of radionuclides that could be released to the environment over 10,000 years (EPA, 1985). These limits were derived from the EPA goal of no more than 1,000 health effects in 10,000 years for all the population exposed. The derivation assumed (1) that the radionuclides released are distributed throughout the surface waters of the world, (2) a constant population of 10 billion people, and (3) eating and drinking habits remain the same as today. Thus, the EPA limits are derived from a 10,000-year limit on collective dose, assuming that all repositories release radionuclides to the same environment as the EPA ''generic" assumptions. The U.S. Congress decided in 1992 that the current EPA standard shall not apply to the proposed repository at Yucca Mountain. A committee of the National Research Council is charged with evaluating and proposing a technical basis for a new standard for geologic disposal of high-level waste (HLW), such as a limit on individual doses, to be considered subsequently by EPA in promulgating a new standard. It is possible that a new standard may preserve some form of the EPA release limits, so the effects of separation and transmutation (S&T) on cumulative releases are included in the following discussion. Long-term performance assessments describe two broad types of scenarios to assess the probability and quantities of radioactivity that might be released. The first includes evaluating all those processes that are reasonably expected to occur in the region of the repository. Principal among these are the "dissolution-and-migration" scenarios in which groundwater eventually penetrates the waste packages and slowly moves radioactivity to the accessible environment. In the unsaturated zone, it is also necessary to consider the movement of air in the host rock and the potential to transport gaseous radionuclides (e.g. 14C in carbon dioxide. In most repository settings, these scenarios typically consider groundwater travel times to the accessible environment that exceed 1,000 years and most often 10,000 years. Further, they include an evaluation of the waste package release and migration potential of individual radioactivity in various chemical forms. The second type of scenario includes "disturbed" conditions that might be expected to accompany undesirable geologic events such as earthquakes, volcanos, and abrupt changes in local or regional hydrologic conditions. These scenarios also include the effect of "human intrusion," in which people in future generations unknowingly penetrate into a repository and release a portion of its contaminants into the earth's surface or the groundwater system. All performance assessments begin with assumptions about the form, characteristics, and radioactivity content of the wastes. The initial inventory for spent fuel consists of actinides, fission products, and activation products. The radioactivity content is distributed between the major isotopes, as shown in Figure G-2. Initially the short-lived fission products (90Sr, 137Cs) and a few short-lived transuranic (TRU) elements (241Pu, 258Pu, 244Cm) dominate the radioactivity. However, for longer-term scenarios (greater than 1,000 years), only long-lived fission products (129I, 135Cs, 126Sn, 99Tc, and 79Se) and certain actinides (243Am, 239Pu, 257Np, 241Am, 240Pu, and 234U) have a potential to contribute to releases. For gaseous releases, 14C is expected to be the primary contributor to release.

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-2 Contributions to total radioactivity content of spent fuel (Ci/MTU). SOURCE: Roddy et al. (1986).

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Nuclear Wastes: Technologies for Separations and Transmutation The predicted releases from dissolution-and-migration scenarios will be determined by radionuclides with long half-lives, high solubility in groundwater, and low sorption along the transport pathway. In unsaturated conditions like Yucca Mountain, the actinides have low solubility and high sorption, and thus less mobility, compared with the fission products; consequently, as shown in Figure G-3, only 129I and 99Tc contribute significantly to the releases to the environment. Further, these releases are typically shown to be much less than allowed by the original EPA standard. This is reinforced by the fact that these two radionuclides are only present in the initial inventory at concentrations below or equivalent to that allowed to be released under the EPA standard. The most notable exception to this general rule is 14C, which contributes many times more to the release than any transport through groundwater. The unique nature of 14C for an unsaturated repository deserves special attention for standards and regulatory development. For "disturbed" scenarios, principally those involving human intrusion, radioactivity can be transported directly to the environment via the breaching or drilling operations. These events are assumed to occur over the history of the repository including some at very early times (less than 1,000 years). Thus sorption and half-life and, for the most part, solubility are not major factors in determining the potential release. Figure G-4 shows a typical distribution of contributions to the result of a human intrusion that involves release of radioactivity to the surface. In this case, the dominant contribution is from TRU elements (240Pu, 239Pu, 241Am) with some contributions from fission products (137Cs). Again, the total releases are small compared with those allowed by EPA standards. IMPACT OF TRANSMUTATION Thermal Effects Partitioning transmutation significantly alters the radionuclide composition of the wastes emplaced in the repository. As a result, all temperature-dependent aspects of the repository are affected to varying degrees. This section describes and evaluates these effects. The analysis will focus on the effects resulting from S&T of the light water reactor (LWR) spent fuel (hereafter spent fuel) that is scheduled for emplacement in the first repository. There is essentially no difference between the various cases being considered because only one evaluation is required to characterize the evaluation cases considered in this appendix. Consideration of HLW from various fuel types would not be expected to change the qualitative aspects of the analysis or the conclusions, because the HLW essentially contains only fission products, which vary little in the amount generated per unit electricity generated. The discussion in this section emphasizes the Yucca Mountain study site as the potential location for the first repository because of the wealth of information already available on this site. However, the consequences of thermal effects on a saturated repository site (such as granite) are also discussed.

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-3 Contributions of individual radionuclides to releases calculated using composite-porosity flow model. SOURCE: Barnard et al. (1992)

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-4 Contribution of individual radionuclides to releases on the surface from human intrusion. SOURCE: Barnard et al. (1993)

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Nuclear Wastes: Technologies for Separations and Transmutation EFFECT OF S&T ON THERMAL POWER OF WASTE The decay of any radionuclide results in the emission of radioactive particles. These emanations are absorbed in surrounding materials and manifest themselves as heat. In the case of spent LWR fuels, the emanations are sufficient to require engineering measures (e.g., limits on the size of packages, active cooling measures, minimum package spacing) to keep the fuel within prescribed temperature limits. The thermal-power profiles of spent LWR fuel and its various major constituent groups are shown in Figure G-5. After about 30 or 40 years, the thermal power of spent fuel comes from the short-lived radionuclides (90Sr and 137Cs) and the actinides (primarily americium and plutonium). The former decay with a half-life of about 30 years. The latter decay with a thermal-power half-life ranging from 400 to 500 years. The contribution of the actinides to thermal power equals that of the fission products after about 70 years out of the reactor and exceeds that of the fission products thereafter. There are two S&T options that might significantly alter the thermal-power profile of high-level wastes (HLW) generated by processing spent fuel. Option 1 involves removal and destruction of the TRU actinides (uranium does not contribute significantly to thermal power), which results in the generation of actinide-free HLW. The amount by which the thermal power is reduced is shown in Table G-1 for selected decay times. TABLE G-1 Relative Thermal Power of Actinide-Free HLW and Spent Fuel Decay Time (years) 10 30 100 300 1,000 Thermal Power (percent) Ratioa 83 70 34 0.8 0.001 a Thermal power of spent fuel without actinides to unprocessed spent fuel. Option 2 involves removal of the TRU actinides plus 90Sr and 137Cs. In this case, the thermal power is reduced to low levels (about 1% of the spent-fuel thermal power at 10 years) and declines steadily thereafter. CONSEQUENCES OF ALTERED WASTE THERMAL ATTRIBUTES TO REPOSITORY CAPACITY Removal of major constituent groups from spent fuel as a part of processing could significantly increase the amount of waste that can be emplaced in a unit area of a repository

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-5 Thermal power profiles of spent LWR fuel and its major constituents groups.

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Nuclear Wastes: Technologies for Separations and Transmutation site. This increase is directly related to the total capacity of a finite repository site, and thus, to the time during which it can receive and emplace HLW. To the extent that these consequences have been studied (mostly scoping studies), they appear to be qualitatively the same for both saturated and unsaturated repository sites. As a result, the following discussion does not attempt any differentiation between these sites. Option 1. Actinide-free HLW differs in thermal attributes from spent fuel in two distinct ways that may increase the capacity of the repository. The first is that removal of the actinides immediately reduces the thermal power of the waste from a unit amount of power generation by about 20%. Because the spacing of waste packages is limited by the temperature limits of the waste form (e.g., spent-fuel cladding, borosilicate glass centerline), the reduction in thermal power would allow about 20% more of a given waste form to be emplaced in a given area. The second change in thermal attributes is that the actinide-free HLW thermal power declines much more rapidly than for spent fuel. It is possible to design waste-emplacement scenarios in which the waste is initially emplaced relatively far from its nearest neighbor and meets applicable limits. Additional emplacement drifts are constructed (currently, drift spacing is limited by thermal considerations), and subsequent HLW packages are emplaced in these drifts. After about 60 years of emplacement, continuation would require that the next package be emplaced between the first two packages emplaced in the repository. Because of the short half-life of strontium and cesium, the thermal power of the first two packages would only be 25% of the initial thermal power, and this should allow applicable limits to be met. This approach would continue until new packages were emplaced between the early-emplaced, widely spaced packages. Scoping calculations indicate that, taken together, the above effects would allow the capacity of a unit amount of repository area to be increased by factor of 4 to 5 (assuming about the same receipt rates as those projected for the Yucca Mountain study site) as compared with a relatively aggressive spent-fuel emplacement case specified by Johnson (1991). It should be noted that this conclusion would not apply to the earlier concepts of ATW cases or other cases in which the thermal efficiency deviates substantially from that of LWRs and liquid-metal reactors (LMRs). Although the same technique can be employed to a limited extent with spent fuel, the benefits are thought to be minimal (perhaps a 20% increase). This occurs because the thermal power declines much more slowly with actinides present, and other temperature limits specified in the site characterization plan of the Yucca Mountain site are soon encountered. In the case of both spent fuel and HLW, a number of other measures might be taken to increase the capacity of a unit amount of repository (Ramspott, 1991). These might include fuel/waste aging, devices to enhance heat transfer from the waste package into the emplacement drift, and reevaluation of thermal criteria. In general, these measures would be complementary to the capacity benefits of actinide removal. Such an approach is not without penalities, which include a more complex emplacement scenario, the need to maintain mined openings for very long times, and the increased and continuing need for ventilation to remove the heat. These disadvantages must be weighed

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Nuclear Wastes: Technologies for Separations and Transmutation TABLE G-5 Relative Doses to Future Individuals From a Repository in Unsaturated Tuff: Effects of Transmutation, Increased Repository Loading, and Separation of Strontium and Cesium   Relative Peak Dose Ratea Radionuclide Spent Fuelb Waste From Reprocessing and Transmutationc Waste From Reprocessing, Transmutation, and Sr-Cs Separationd 129I 1 9×10-1 2×101 99Tc 7×10-1 4×10-1 7 237Npe 1×10-5 2×10-6 3×10-5 242Pu 7×10-19 1×10-19 1×10-18 a Maximum dose rate from the listed radionuclide relative to the maximum dose rate from 129I from spent fuel. The individual doses are calculated for dissolution of radionuclides from waste solids and their hydrogeologic transport to the 5-km environment. b Waste from LWR spent fuel, loaded to design limit of decay-heat rate per unit repository area at time of emplacement. c Waste from pyrochemical reprocessing, loaded to the same areal-heat-loading design limit as for spent fuel. d Waste from pyrochemical reprocessing and separation of strontium and cesium for surface storage, loaded to the same areal-heat-loading design limit as for spent fuel. e The relative peak doses from 237Np in unsaturated tuff would increase and could exceed those from 129I if the more conservative solubility data for neptunium, discussed earlier, were used. SOURCE: Hirschfelder et al. (1991, 1992). EFFECT OF TRANSMUTATION ON MEETING THE TECHNICAL CONTAINMENT LIMITS OF THE EPA STANDARD 40CFR191 Barnard (private communication, 1993) has calculated the effect of reprocessing and actinide transmutation on the probabilistic distribution of curie releases by aqueous pathways for a conceptual repository in unsaturated tuff. The results are shown in Figure G-8, expressed as the complementary cumulative probability as a function of the EPA sum. Three curves are shown: one for unreprocessed spent fuel, another for waste from aqueous reprocessing of spent fuel and transmutation in ALMRs with pyrochemical reprocessing of ALMR spent fuel, and a third for pyrochemical reprocessing of both LWR and ALMR spent fuel. All curves are based on the same generation of electrical energy. The calculations were made on the basis of the composite-porosity model of hydrogeologic transport in the unsaturated zone. The EPA aqueous-pathway sums for reprocessing and transmutation are about the same as for unreprocessed LWR spent fuel for probabilities of 0.1 and greater. However, for a

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Nuclear Wastes: Technologies for Separations and Transmutation probability of 0.001 the EPA sum is about fivefold smaller for reprocessing and transmutation. The calculated differences in EPA sums for aqueous and pyroprocessing of LWR spent fuel are small. According to these calculations, the greatest benefit from reprocessing LWR spent fuel results from the elimination of gaseous 14C, as may be seen by comparing Figures G-6 and H-8. Although gaseous 14C is emitted to the atmosphere from conventional aqueous reprocessing plants, it is assumed for this discussion that the proposed new high-recovery processes proposed by transmutation proponents, whether aqueous or pyrochemical, would include recovery of 14C into a new waste form of low volatility, suitable for geologic disposal. This 14C benefit results entirely from reprocessing. It would occur whether or not transmutation were also carried out. Barnard and Lee (1992) calculated the curves of complementary cumulative probability versus the EPA sum for human intrusion, as affected by reprocessing and transmutation, for a conceptual repository in unsaturated tuff. The results for surface releases, based on the composite-porosity model of hydrodynamics in unsaturated tuff and equal electrical energy from reprocessing and nonreprocessing options, are shown in Figure G-9. Three curves are shown: one for unreprocessed spent fuel, another for waste from aqueous reprocessing of spent fuel and transmutation in ALMRs with pyrochemical reprocessing of ALMR spent fuel, and a third for pyrochemical reprocessing of both LWR and ALMR spent fuel. None of the curves would exceed the EPA standard. For probabilities greater than about 10-4, the magnitudes of the releases are lower for repositories containing waste packages from either of the reprocessing options. For probabilities of about 10-4 and lower, where the largest predicted releases occur, the predicted releases are about the same for the three scenarios. Comparing Figures G-8 and G-9, surface releases from human intrusion would be greater, for a given probability, than aqueous-pathway releases for unreprocessed spent fuel, but less for waste from reprocessing and transmutation. For probabilities greater than about 10-4, the EPA sums for surface releases from human intrusion would be lower for transmutation wastes, but the EPA sums would then become dominated by releases from the aqueous pathways. Thus, the reduction in overall EPA sums due to transmutation reduction in actinide inventory is small, less than an order of magnitude. This assumes the composite-porosity hydrodynamic model of the unsaturated zone. For the more rapid hydrodynamic transport in conductive fractures, without local equilibrium with the rock matrix (see Figure G-7), releases via the aqueous pathways are predicted to be more important than those from human intrusion, and separation and transmutation would not be expected to result in as large a reduction in releases as for the composite-porosity model. However, in either of the predictions using the two hydrodynamic flow models, a significant reduction in calculated EPA ratios is predicted due to elimination of gaseous 14C. This is a consequence of the improved waste forms for 14C, a result only of fuel reprocessing. Additional reductions in EPA sums would occur if soluble 99Tc and 129I were also transmuted. In summary, the calculations by Barnard and Lee (1992) for composite-porosity hydrodynamics show that aqueous-pathway curie releases, as measured by the EPA sums, approach the EPA limit most closely for a probability of 0.001, where aqueous-pathway EPA sums are predicted to about fivefold smaller for transmutation waste. A larger reduction would

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-8 Comparison of aqueous releases for LWR spent fuel and S&T waste in unsaturated tuff. SOURCE: Barnard and Lee (1992)

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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE G-9 Comparison of human-intrusion surface releases for S&T waste and LWR spent fuel in unsaturated tuff (equal generation of electrical energy). Source: Barnard and Lee (1992).

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Nuclear Wastes: Technologies for Separations and Transmutation arise only from reprocessing, which could result in a better waste form for retention of 14C that could otherwise be a gaseous release from spent-fuel waste. SUMMARY AND CONCLUSIONS Implementation of S&T would have two direct effects on the repository: substantial reduction of the amounts of certain radionuclides in the repository wastes, with the identity of the radionuclides and degree of removal being dependent on the spent-fuel reprocessing technology employed; and alteration of the waste form from spent nuclear fuel (rods of actinides and fission products encased in metal) to a monolithic HLW form designed for repository conditions and contained in an unirradiated metal canister including waste forms tailored for containment of specific spent-fuel constituents (e.g., 14C). These direct effects have diverse impacts on the postclosure behavior of the repository. While these impacts are not yet fully elucidated, they are well enough understood to support the following conclusions: Implementation of S&T would significantly affect the thermal attributes of waste emplaced in a repository. Removal of actinides only would reduce near-term thermal power by about 20% and reduce the total heat released to the geology surrounding the repository over the long-term by a factor of 4. The removal of the actinides from the material being emplaced in the repository allows 4 to 5 times more waste to be emplaced in a given area of the repository. While increases in emplacement density can be achieved by other engineering measures, the increase resulting from S&T is additive to these other measures. The increase in emplacement efficiency could extend the life of the repository by four-to fivefold, thus deferring the need to undertake contentious and expensive activities associated with a second repository until about the twenty-second century. In addition, it offers a mechanism to lower the temperature of the repository rapidly (relative to spent fuel) if a ''cool" repository concept should be deemed desirable. However, it would also eliminate the ability to establish a "hot" repository that might be employed to keep a repository constructed in unsaturated rock dry for an extended time, as has been proposed for Yucca Mountain. Removal of actinides and strontium and cesium may increase the capacity of the repository by factors ranging from 10 to 40. This would greatly postpone the need to site, license, and pay for another repository. While operations are simplified, the repository would have to remain open over extremely long time periods. Most important, a MRS-license facility must be sited, built, and operated to store strontium and cesium for hundreds of years. The incremental advantages of removing strontium and cesium in addition to the actinides (i.e., no thermal pulse, very large capacity) will probably not justify the incremental disadvantages (i.e., the need for a very-long term MRS-like facility, the need to separate and

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Nuclear Wastes: Technologies for Separations and Transmutation process strontium and cesium waste), although additional quantification of benefits and penalties is required to substantiate this. The impact of S&T on the long-term risk from the repository is heavily dependent on (1) the specific repository being considered, (2) the types of radionuclides that are removed during reprocessing, and (3) the release scenario being considered. A generalized summary of the effects of S&T on long-term repository risk is given in Table G-6. It should be noted that the contents of Table G-6 are somewhat dependent on the repository host medium being considered (dry versus wet, oxidizing versus reducing) and the measure of risk used (individual versus population). It is to be emphasized that a S&T scenario in which the need for a repository is eliminated is considered to be highly unlikely if not absolutely impossible. All current approaches to S&T result in a waste containing significant amounts of radionuclides that are extremely difficult to separate (i.e., require isotopic separation) or that are not amenable to transmutation. Elimination of a repository would require an extremely diverse and sophisticated combination of chemical and isotopic separation technologies in concert with both transmutation and alternative radionuclide disposal technologies such as extraterrestrial disposal. The benefits of S&T (which is defined as enhanced reprocessing to recover essentially all radionuclides that would otherwise report to repository wastes) to long-term repository risk can largely, if not totally, be achieved by employing basic reprocessing (i.e., recovery of about 99% of the uranium and plutonium in the spent fuel). This is because the major benefits are reducing heat generation and reducing the toxic actinide content (which can be largely achieved by removing most of the plutonium) and improving the waste form for the residual (which results from reprocessing, irrespective of actinide decontamination levels). The committee recommendations pursuant to the above conclusions are as follows: The Department of Energy should consider the removal of actinides as one option in its broader systemic evaluation of the thermal strategy for Yucca Mountain. Pursuit of a HLW repository should be continued. The benefits of S&T should continue to be studied as part of the continuing evaluation of repository performance. This should include explicit consideration of the optimum recovery of various radionuclides. S&T technology should continue to be developed in an orderly manner, and by the turn of the century it should be brought to the point where preferred technologies could be selected and demonstration projects initiated if deemed appropriate. This development should be closely coordinated with continued development of the ALMR and its attendant nuclear fuel cycle. The design of the repository should incorporate features that would allow spent fuel to be readily retrieved and reprocessed and the resulting HLW to be emplaced at a higher effective density.

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Nuclear Wastes: Technologies for Separations and Transmutation TABLE G-6 Summary of the Effect of Partitioning-Transmutation on Repository Risk   Release Mechanism Waste Component Released Dissolution and Migration Human Intrusion Gaseous Release Actinidesa Small reduction in expected risk; Substantial reduction of already small risk from low-probability, high-consequence release scenarios; Significant reduction in individual dose Reduction of risk approximately proportional to reduction in actinide concentration No significant effect Iodine and Technetium Significant reduction in an already-small risk Small reduction in expected risk No significant effect Carbon-14 No significant effect No significant effect Major reduction in potentially limiting species that poses a very small individual risk All Species High-level waste form is expected to be more resistant to degradation in the repository than spent fuel, thus reducing radionuclide release rates and consequential risk. The magnitude of the benefit depends on the specific radionuclide, with more soluble species being benefitted more. a The relative peak doses from 237Np in unsaturated tuff would increase and could exceed those from 129I if the more conservative solubility data for neptunium, discussed earlier, were used.

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