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Nuclear Wastes: Technologies for Separations and Transmutation (1996)

Chapter: G EFFECTS ON REPOSITORY

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Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

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,

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

FIGURE G-1 An underground configuration for a nuclear waste repository.

SOURCE: Hunter et al. (1989)

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

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:

  1. annual radiation dose to an individual,1 which is usually applied for expected releases or high-probability events;

  2. annual risk to an individual from radiation exposure, which takes into account the probability of exposure; and

  3. 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.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×
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.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

FIGURE G-2 Contributions to total radioactivity content of spent fuel (Ci/MTU).

SOURCE: Roddy et al. (1986).

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

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.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

FIGURE G-3 Contributions of individual radionuclides to releases calculated using composite-porosity flow model.

SOURCE: Barnard et al. (1992)

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

FIGURE G-4 Contribution of individual radionuclides to releases on the surface from human intrusion.

SOURCE: Barnard et al. (1993)

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×
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

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

FIGURE G-5 Thermal power profiles of spent LWR fuel and its major constituents groups.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

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

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

against a capacity increase that would greatly delay the need for another repository and the controversial and expensive activities that accompany it.

Option 2. In this option, 90Sr, 137Cs, and the actinides are removed and the thermal power of the waste is reduced to very low levels as compared with spent fuel. This means that the waste form temperature limits of the waste become essentially irrelevant, which yields the following benefits if the waste continues to be emplaced according to the concept specified in the site characterization plan (e.g., in vertical boreholes): (1) the diameter of a waste package can be much larger, which saves in package and handling costs; and (2) the waste packages can be spaced as close together as rock integrity considerations will allow.

The amount by which repository capacity might be increased in this option has not been calculated, but a factor of 5 should be approximately correct.

If alternative emplacement concepts are adopted, much larger capacity increases may be possible. The thermal-power level should be low enough to allow the wastes to be closely stacked in the drifts in lightly shielded packages or by shielded emplacement vehicles. Again, the magnitude of the capacity increase cannot be precisely quantified. However, factors of 10 to 40 appear to bound the possibilities.

This option entails one major penalty: disposition of the recovered strontium and cesium. The nuclear properties of 90Sr and 137Cs make them very difficult, if not impossible, to transmute. Because of the considerable emission of penetrating radiation and heat, they are clearly unacceptable for near-surface disposal. Thus, this material would presumably have to be stored in an engineered surface facility until it reached concentrations approximating the levels that come under the category of Class C LLW defined in the Nuclear Regulatory Commission (NRC) standard 10CFR161. This time ranges from about 200 to 500 years, depending on the initial concentration of the contained radionuclides. The radiocesium also contains a small amount of long-lived (3-million-year half-life) 135Cs. The concentration of 135Cs appears to be less than the limit for Class C LLW at all times, although definitive regulations have not yet been established. To be more precise, the penalty is the need to site, license, build, operate/monitor over an extended time and decontamination and decommissioning a major storage facility that is likely to be contentious. The facility might be much like the currently proposed monitored retrievable system (MRS) in most ways.

CONSEQUENCES OF ALTERED WASTE THERMAL ATTRIBUTES TO LONG-TERM REPOSITORY PERFORMANCE

Amount and duration of the thermal pulse from HLW as compared with that from spent fuel has a major effect on the physicochemical processes that occur after the repository is closed. Elucidation of these effects and their consequences is a very complex task that requires an extensive base of assumptions (e.g., emplacement scenario and waste characteristics), computational results (e.g., determination of temperature as a function of space and time), and site response functions (e.g., behavior of natural and engineered materials as a function of

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

temperature and time). Because this information base is not yet available, the following discussion is based on broader generalities.

Saturated Sites. At typical repository depths (300 to 1,000 m), the water pressure is sufficiently high that the repository horizon will resaturate and remain saturated in perpetuity. After the repository is completely closed, it is typically assumed that it will resaturate over times that could range from immediately to several centuries. At the same time, the host rock is heated to a temperature much higher than ambient. Temperature maxima within waste packages (very near field) are expected within a few to several years after closure, based on heat conduction to surrounding rock, although this time might be extended if boiling occurred during repressurization. Maxima within meters of the waste package (near field) are expected within decades, and within centuries at distances of 50 to 100 m from the emplacement level (the far field). Thereafter, temperatures slowly fall as the thermal power of the waste declines exponentially.

This temperature pulse can potentially have a number of deleterious effects on the performance of repository components. Some of the most important impacts are as follows:

  • The rate at which reactions (e.g., corrosion) occur in engineered materials typically increases exponentially with temperature.

  • Radionuclide solubilities typically increase as temperature increases, although some of the actinides may exhibit retrograde solubility under some geochemical circumstances.

  • Increased temperature can irreversibly degrade the ability of natural ion exchange materials (e.g., zeolites) to retard migrating radionuclides.

  • The thermomechanical stresses induced in the surrounding rock can provide additional or more-transmissive pathways for water to enter repository, degrade waste packages, dissolve radionuclides, and transport them to the environment.

  • Temperature differences can provide a driving force for movement of water through the repository horizon because of buoyancy effects.

As a result, in saturated repository, performance is expected to be affected as the magnitude of the temperature pulse increases. It then follows that any option that might reduce the temperature pulse should be beneficial to repository performance unless radionuclides exhibiting retrograde solubility were found to dominate repository risk, which is unlikely.

In the case of Option 1, the duration of the temperature pulse would be substantially reduced by removing the actinides. In approximate terms, removal of the actinides reduces the total amount of heat emitted by the waste by about a factor of 4 over the long term, allowing the duration of the elevated temperatures in the repository to be reduced. In the case of 30-year-old fuel, the time during which the repository center temperature is above 100° C is reduced from about 1,500 years to zero. Reduction of the maximum temperature could also be achieved but at the expense of repository capacity increases.

In the case of Option 2, which involves removing both actinides and strontium, the temperature pulse is essentially nonexistent, and thermal impacts would be expected to be essentially nil.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

Unsaturated Sites. In an unsaturated site, the waste is emplaced above the water table. This is projected to be the case at the proposed Yucca Mountain study site. In such a situation, all of the disadvantageous effects of the thermal pulse, which are listed above, still apply.

However, in the case of unsaturated sites, there may be major advantageous effects of the thermal pulse. To understand this, one must first understand that the primary (although not only) radionuclide release mechanisms require that groundwater contact the waste package, degrade the package to expose radionuclides, dissolve the radionuclides, and transport them to the biosphere. The advantage of the thermal pulse is a result of designing the repository to prevent water from contacting the packages by keeping them hotter than the boiling point of water for an extended time. Although design calculations are not complete, it appears that it may be possible to maintain a dry environment at the surface of waste containers for many thousands of years. Ramspott (1991) has proposed an ''extended dry" concept whereby the thermal pulse could be extended to maintain a dry environment for over 10,000 years.

The thesis that a "hot," unsaturated repository is preferable is not universally accepted. Some investigators (Eriksson, 1991) have concluded that the disadvantages of a larger temperature pulse outweigh the advantages and that a "cold" repository is preferable. This is based on the notion that a hot repository and significant changes in repository temperatures result in complex processes and events (human-induced perturbations) that are difficult to measure and predict and that are associated with significant uncertainty in safety assessments.

At this juncture, the committee endorses the National Waste Technical Review Board's (1992) views that "… a technical basis for the DOE's [Department of Energy's] current thermal loading strategy for the Yucca Mountain site does not exist" and that "… DOE needs to thoroughly investigate alternative thermal loading strategies. …" The primary reasons for this are the lack of a system-wide evaluation of the benefits and penalties of various thermal loading strategies, especially as they affect long-term repository performance and the fact that the current reference strategy is based on calculations that still need to be experimentally validated.

Assuming, for the sake of discussion, that a "hot," unsaturated repository is preferable, then removal of the actinides (Option 1) would clearly diminish the benefits of this approach, and removal of the actinides and strontium and cesium would completely eliminate these benefits.

Long Term Performance Effects
IMPORTANT RADIONUCLIDES OF THE REFERENCE ONCE-THROUGH URANIUM FUEL CYCLE IMPACTING WATER PATHWAYS AND INDIVIDUAL DOSE

To evaluate possible benefits from transmuting various radionuclides, it is necessary to first identify those species that are calculated to be the main contributors to long-term individual doses from geologic disposal. For this purpose it is sufficient to focus on the relative values of maximum annual dose to individuals from each radionuclide. This can be calculated by using relative values of annual doses and published information from other countries that present future dose rates from assumed failure of a single waste container. Relative values for the once-

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

through uranium fuel cycle, resulting from dissolution in groundwater and transport to the biosphere and calculated from published dose estimates for conceptual repositories in granite, are listed in Table G-2. The relative dose rate is the ratio of the maximum dose rate for a future individual and a given radionuclide to the maximum dose rate from the radionuclide that is the greatest contributor to future doses to individuals. Each maximum dose rate is calculated at the future time when that maximum occurs.

Estimated relative doses to future individuals from conceptual repositories in unsaturated tuff, loaded with spent fuel from LWRs, differ from those listed in Table G-2. The groundwater in unsaturated tuff is expected to be saturated with oxygen, resulting in an oxidizing atmosphere surrounding the waste containers. Containers are expected to fail by localized penetrations and cracks, allowing water to leak in and dissolve the waste solid. If the water within the waste container is as oxidizing as that in the surroundings, the solubilities of many important radioelements are expected to be greater than those that would occur in the reducing environment of a granite repository. The fission product 99Tc would become much more soluble and is calculated to result in peak doses in the same range as those calculated for 129I (Eslinger et al., 1993; Hirschfelder et al., 1991, 1992; Wilson et al., 1994; Andrews et al., 1994; Duguid et al., 1994). In a granite repository, the reducing environment provides sufficiently low solubility that the 99Tc peak dose is well below that of 129I, as shown in Table G-2.

However, it is not clear that the conditions within a waste container in unsaturated tuff will be oxidizing. If the bulk of the container remains intact after penetration, the container metal can be expected to promote a reducing environment within the container, particularly if the container is composed of thick metallic iron. If so, 129I would remain the main contributor to peak dose, as shown in Table G-3.

The solubility of the actinide neptunium is particularly sensitive to the oxidizing potential, as well as to pH. Recent performance assessments for waste in unsaturated tuff (Wilson et al., 1994; Andrews et al., 1994; Duguid et al., 1994) have conservatively adopted a log-normal or log-beta distribution of neptunium solubilities ranging from 10-8 to 10-2 M. The assumed high solubilities result from the assumption that sufficient air-saturated groundwater enters through penetrations in failed waste packages in sufficient quantity to control the oxidation potential within the waste package. The effect of the metallic structure of the waste container to cause a reducing environment within the waste package has been neglected.

With such assumed distributions of neptunium solubility that emphasize the high end of the solubility range, the resulting peak doses of 237Np can become greater than that of 129I and 99Tc. The peak doses of 237Np are predicted to occur in the era of about 800,000 years. If the effective solubility of neptunium within the waste package does turn out to be near the high end of that assumed range, then either transmutation of 237Np or improved waste forms containing separated 237Np could improve the performance of a tuff repository.

The high individual doses appearing in the recent studies of a repository in unsaturated tuff are a consequence, in part, of the simplifying assumption that a failed waste container would present no barrier to the release of radionuclides from a waste package. If the container fails by localized small penetrations, as is expected by the designers, the small penetrations can present a significant impedance to the transport of groundwater through the container wall, thus reducing the effect of aerated groundwater on the oxidation potential within the container.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

TABLE G-2 Relative Doses to Future Individuals From a Granite Repository Loaded With Unreprocessed LWR Spent Fuel

 

 

Relative Dose Ratea

Radionuclide

Half-Life, (years)

Sweden-SKIb

Sweden-SKBc

Finlandd

United Kingdom-NRPBe

129I

1.57×107

1

1

1

1

99Tc

2.11×105

1×10-5

 

6×10-5

6×10-1

135Cs

2.3×106

3×10-2

1×10-2

1×10-2

1×10-1

14C

5.73×103

3×10-5

 

6×10-2

4×10-12

36Cl

3.01×105

 

 

2×10-3

2×10-9

79Se

6.5×104

3×10-3

 

1×10-2

2×10-11

126Sn

1.0×105

 

1×10-2

6×10-4

2×10-14

226Ra

1.6×103

 

1×10-3

2×10-2

 

231Pa

3.28×104

2×10-2

3×10-2

4×10-1

 

234U

2.45×105

1×10-8

 

6×10-5

 

238U

4.47×109

1×10-8

 

1×10-5

 

237Np

2.14×106

3×10-5

 

4×10-4

 

239Pu

2.41×104

3×10-5

 

2×10-3

 

240Pu

6.56×103

1×10-8

 

 

 

242Pu

3.73×105

3×10-5

 

4×10-2

 

243Am

7.38×103

 

 

4×10-4

 

a Maximum dose rate from the listed radionuclide divided by the maximum dose rate from the radionuclide that contributes the largest dose. The maximum values occur at different times.

b SKI - Swedish Nuclear Power Inspectorate (1991).

c SKI - Swedish Nuclear Fuel and Waste Management Co. (1992).

d Vieno et al. (1991).

e Mobbs et al. (1991) United Kingdom - National Radiation Protection Board (UK-NRPB). Calculated doses from technetium and actinides are not included because solubility limits were ignored. Solubility limits are included in all other safety analyses reviewed here.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

TABLE G-3 Relative Doses to Future Individuals From a Repository in Unsaturated Tuff, Loaded with Unreprocessed LWR Spent Fuel

 

 

Relative Dose Ratea

 

 

 

 

University of California.d

Radionuclide

Half-Life (years)

Pacific Northwest Labb

Hirschfelder, et al.c

TW = 103ae

TW = 105a

129I

1.57×107

4×10-1

1

4×10-1

8×10-1

99Tc

2.11×105

1

8×10-1

1

1

135Cs

2.3×106

 

 

2×10-1

3×10-7

226Ra

1.6×103

 

4×10-12

 

 

234U

2.45×105

 

1×10-9

3×10-6

1×10-6

238U

4.47×109

 

6×10-7

1×10-6

1×10-6

237Np

2.14×106

 

1×10-5

4×10-5

3×10-5

239Pu

2.41×104

 

 

1×10-3

10-40

240Pu

6.56×103

 

 

2×10-4

<10-100

242Pu

3.73×105

 

7×10-19

2×10-5

1×10-7

243Am

7.38×103

 

 

3×10-5

<10-100

a Maximum dose rate from the listed radionuclide divided by the maximum dose rate from the radionuclide that contributes the largest dose. The maximum values occur at different times.

b Eslinger et al. (1993). 129I and 99Tc are the principal contributors, with maximum doses about a hundred-fold greater than doses from any other radionuclide.

c Hirschfelder et al. (1991, 1992).

d Pigford (1990).

e TW is the assumed time for ground water to flow from the buried waste to the environment.

f The maximum dose rate from any of these radionuclides is less than 1% of the maximum dose rate from 99Tc.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

Similarly, the small penetrations can present a significant impedance to the transport of dissolved radionuclides to the surrounding rock, thereby lowering the calculated individual doses.

Further implications to recycle and transmutation are discussed below.

129Iodine. 129Iodine is the main dose contributor from spent fuel in granite. It is either the most important or second-most important contributor in an unsaturated tuff repository. Iodine is soluble in groundwater. Some iodine will have been released from the UO2 matrix during reactor operation and will be released to groundwater from defective fuel cladding soon after water enters a defective waste package. In the analyses reported here, the UO2 matrix is assumed to restructure within a few thousand to a few ten-thousand years, releasing the remaining iodine into groundwater.2 Iodine forms negative ions in solution and is not expected to be retarded by sorption on the rock. The half-life of 129I is long enough (1.57 × 107 years) that eventually most of it will reach the environment. To reduce the potential individual doses from dissolution and transport in a geologic repository, either the 129I must be transmuted or the reprocessing operations must produce waste forms that release iodine far more slowly than is calculated here for unreprocessed spent fuel. Only the accelerator transmutation of waste (ATW) and the LWR transmutors propose to transmute 129I.

99Technetium. 99Technetium, a fission product, is one of the two (the second one 129I) most important contributors to individual doses in unsaturated tuff. Technetium is released by essentially the same mechanisms as iodine. It is not solubility limited in an oxidizing environment. However, a granite repository is expected to resaturate after sealing, and ferrous iron dissolved in the groundwater is expected to restore a reducing environment that promotes low-solubility forms of technetium that would precipitate in or near the waste packages. Consequently, the dissolution source term for technetium is reduced in granite repositories. The calculations by Sweden and Finland take into account the limited solubility of technetium and predict that technetium doses will be several orders of magnitude below the iodine dose from granitic groundwater.

To reduce the potential individual doses that result from dissolution and transport in a geologic repository, either the 99Tc must be transmuted or the reprocessing operations must produce waste forms that release technetium far more slowly than is calculated here for unreprocessed spent fuel. Only the ATW and the LWR transmutors propose to transmute 99Tc.

2  

In a repository in unsaturated tuff, the spent-fuel matrix may eventually oxidize if sufficient air or air-saturated groundwater can leak through penetrations in failed containers. In the Swedish analyses for granite repositories, the spent fuel is conservatively assumed to oxidize and release its soluble fission products as a result of alpha radiolysis. In the Canadian repository program, it is expected that alpha radiolysis will not be sufficient to oxidize and restructure the spent uranium fuel, so release of iodine from the fuel matrix is expected to be very slow, limited by the solubility-limited net dissolution of the uranium fuel matrix. Thus, in the Canadian estimates for a granite repository, releases of most of the radionuclides from spent fuel to ground water are expected to be controlled by the slow rate of net dissolution of the unaltered fuel matrix, so the doses predicted for iodine and many other radionuclides are much smaller than predicted in the analyses by Sweden and Finland.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

135Cesium. 135Cesium is predicted by Sweden and Finland to be the second-most important contributor in a granite repository. It is also important in the predictions by the United Kingdom's National Radiation Protection Board. Dissolution of cesium is not expected to be limited by cesium solubility. In unsaturated tuff 135Cs is predicted to be an important contributor for pathways that have water travel times to the environment of a few thousand years, third in importance after iodine and technetium. For the longer water travel times (about 104 to 105 years) predicted from the calculated infiltration rates at the proposed Yucca Mountain repository. Cesium sorption and retardation may delay the transport of cesium to the environment sufficiently for most of the 135Cs to decay in transit. Transmutation of 135Cs would be particularly difficult because of the intense radiation field from 137Cs and because stable 133Cs would absorb neutrons to produce more 135Cs. Separation of the cesium isotopes prior to transmutation would avoid the problem of generating more 133Cs, but isotope separation in the intense radiation field 137Cs would be difficult.

To reduce the potential individual doses from 135Cs in a geologic repository that does not have sufficient water travel time and retardation for 135Cs decay, there appear to be three options: (1) solve the difficult problem of isotope separation and transmutation of 135Cs, (2) produce waste forms that release cesium far more slowly than is calculated here for unreprocessed spent fuel, or (3) store separated cesium for several hundred years until 137Cs has decayed, then proceed with isotope separation and transmutation. Only the ATW considers isotope separation and transmutation of 135Cs as an option to be pursued.

231Protactinium. The data in Tables G-2 and G-3 show that there would be relatively little reduction in the doses to individuals by transmuting only the actinides. However, if the more significant problems of the long-lived soluble fission products are solved, as discussed above, reducing the potential individual doses from dissolution and transport of the actinides would become important. In the calculations for granite repositories, the main actinide contributor is 231Pa, a decay daughter of 235U. It grows in slowly in separated uranium, with a time constant of about 50,000 years. The long-term doses from 231Pa have not been addressed in the calculations for a tuff repository. There is no reason to expect that the importance of 231Pa in tuff would be significantly less than calculated for disposal of spent fuel in granite.

There would not be enough 231Pa in spent fuel to justify its transmutation, but transmutation of its precursors 235U, 239Pu, and 243Am would remove the potential problem of 231Pa. However, transmutation requires reprocessing. Transmutation of the separated TRUs would eliminate 239Pu and 243Am. By recycling the separated uranium to facilities that enrich uranium to fuel LWRs, much of the 235U could eventually be transmuted in the LWRs. This would require continuing use of LWRs while constructing and operating transmutor systems. About 30% of the 235U recycled would appear in the depleted uranium from isotope separation. Long-term build-up of 231Pa in the resulting depleted uranium would continue as a significant environmental issue. In the time scale of a few hundred thousand years and longer, it could present more significant hazards from surface or near-surface dissolution and transport than is expected from the mill tailings from producing new uranium fuel for LWRs.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

However, none of the transmutation proposals contemplate recycling the recovered uranium. Instead, the flowsheets proposed by the transmutation proponents to reprocess LWR spent fuel, whether pyrochemical or aqueous, would separate uranium from the other actinides and store it above ground. If that is to be the ultimate fate of the uranium, the long-term individual doses resulting from growth of 231Pa and its ultimate dissolution in surface water could surely be greater than the doses calculated for that same amount of 231Pa that would form in deep geologic disposal in the reference once-through fuel cycle. In this sense, implementing the transmutation proposals that would leave uranium from spent fuel on the earth's surface indefinitely could reduce the overall safety of waste disposal. Even if the separated uranium is recycled to isotope separation, the hazard from the long-term build-up of 231Pa in the resulting depleted uranium, if stored on the surface, may exceed the hazard from 231Pa in spent fuel in a geologic repository. The relative hazards can be and should be calculated.

Other Transuranic Actinides. All of the transmutation proposals reviewed herein would transmute the isotopes of neptunium, plutonium, americium, and curium that appear in LWR spent fuel. The advanced liquid-metal reactor (ALMR) transmutor would transmute only these TRUs; the ATW-1, -2, and -33 and the LWR would transmute the TRUs and selected fission products. Transmuting only the TRUs would result in relatively little reduction in the maximum individual radiation dose from dissolution and groundwater transport of waste buried in granite or other reducing rock, because these doses would continue to be dominated by the soluble fission products. Transmuting neptunium could be of benefit to a repository in unsaturated tuff, if the effective solubility of neptunium within failed waste containers is as high as assumed in recent studies (see discussion in section Long-Term Performance Effects).

Assuming that transmutation of TRUs could be justified, the most important TRUs whose transmutation could benefit the water-transport pathway in all proposed repositories are the precursors of 226Ra and 231Pa. Transmutation of the 238Pu in plutonium recovered from LWR spent fuel would eliminate the major source of 226Ra. If the uranium separated from LWR spent fuel were buried with the reprocessing waste, the long-term dose from 226Ra4 would be about seventyfold less than from spent fuel. If the separated uranium were stored on the surface, as proposed in the reprocessing flowsheets, and if the 238Pu were transmuted, the amount of 226Ra that could be formed would be seventyfold less than in LWR spent fuel because 238Pu would not be present. However, if this uranium were not recycled but left on the earth's surface, the lack of geologic isolation could result in a greater long-term dose from 226Ra than calculated for the geologic disposal of spent fuel.

3  

The ATW-4 transmutor would operate as a self-sustaining thorium breeder. It would not be fueled with TRUs from LWR spent fuel. It would transmute the thorium and all transthorium actinides.

4  

This assumes that uranium is dissolved in groundwater and transports to the environment within a few hundred thousand years, so that the main source of 226Ra would be from decay of the 234U in spent fuel. If the time for uranium transport to the environment is of the order of a million years or more, transmuting the 238Pu would be of little benefit in reducing the dose from 226Ra. The dose from 226Ra reaching the environment would then be governed by the transport of 4.47 × 109-year 238U.

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

No transmutation projects have proposed transmutation of 235U, the precursor of 231Pa. However, transmutation of 235U in spent fuel, to avoid the build-up and transport of 231Pa, could benefit the long-term performance of geologic repositories.

237Np would be next in importance to dissolution-transport in granite or other reducing rock. It could be more important in unsaturated tuff, as discussed above. Transmutation of neptunium alone would result in only about a fourfold reduction in the long-term dose from 237Np. Decay of untransmuted 241Am would produce additional 237Np that could transport to the environment.

Of the longer-lived plutonium isotopes, transmutation of 239Pu, 240 Pu, and 242Pu would be next most important. The predicted doses from these isotopes are of concern only for very short groundwater travel times, such as considered in the Finnish analysis (Table G-2) and in the analysis for 1,000-year water travel time in unsaturated tuff.

243Am and the curium isotopes are not predicted to contribute significantly to doses from TRUs from the dissolution-transport scenarios.

Waste-Form Benefits From Chemical Reprocessing. New waste forms specifically tailored to retain these fission products much longer within the repository offer the potential for improved performance. Benefits could result regardless of transmutation. Candidate waste forms suggested in earlier studies (Pigford et al., 1983) are silver iodide; compounds of separated technetium and separated neptunium, each in thick ferrous containers; and pollucite for cesium. Radiation degradation would need to be investigated. However, the cost of chemical processing solely for waste improvement (i.e., with no recycle of actinides or fission products to reactors) would be staggering. If a U.S. commercial reprocessing plant were to operate with a unit cost of $2,000/kgHM (mostly uranium) in the spent fuel (compare Chapter 7), with no return for recycling uranium and plutonium fuel, the total cost for reprocessing the 63,000 Mg of spent fuel destined for the first U.S. repository would be $126 billion.

IMPORTANT RADIONUCLIDES CONTRIBUTING TO LONG-TERM INDIVIDUAL DOSES FROM INTRUSION SCENARIOS FOR THE REFERENCE ONCE-THROUGH FUEL CYCLE

Intrusion scenarios, such as inadvertent exploratory drilling, magmatic intrusion from volcanic activity, etc., could bring portions of the buried waste packages to the surface. Also, intrusion by drilling into a repository in unsaturated rock could result in waste cuttings falling through an empty borehole to a lower aquifer, with a correspondingly shorter time of transport of dissolved species to the environment. Here the important radionuclides that could contribute to individual doses are those discussed for the dissolution-transport scenario.

Individual doses to people from waste cuttings brought to the surface could be quite large (Eslinger et al., 1993). If the probabilities of intrusion are low, a risk criterion that considers both dose and probability is appropriate. The radionuclides of highest radioactivity are generally considered to be those of greatest risk. The most important radionuclides for early intrusion, within a few hundred years after emplacement, are 90Sr and 137Cs. The isotopes of

Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
×

plutonium and americium would be important thereafter, until a few hundred thousand years when 226Ra and 231Pa dominate the radiotoxicity of the spent-fuel waste.

Calculated individual doses and risks for assumed intrusion into a granite repository loaded with LWR spent fuel, at 100 years and 105 years after emplacement, are listed in Table G-4 (Mobbs et al., 1991). The calculated risk from intrusion is 1.8 × 10-7/yr for intrusion at 100 years and 2.1 × 10-9/yr for intrusion at 105 years. The same study calculates a maximum risk of 5 × 10-5/yr from the long-term dissolution of spent-fuel waste in ground water and the long-term transport of dissolved radionuclides to the environment. Thus, Mobbs et al. conclude that dissolution-transport is the more important mechanism to consider in assuring long-term public health and safety from geologic disposal. If so, emphasis in transmutation would be directed to the long-lived, soluble fission products, as discussed above.

THE REFERENCE ONCE-THROUGH URANIUM FUEL CYCLE: PRELIMINARY ESTIMATES OF CONFORMANCE WITH TECHNICAL CONTAINMENT LIMITS OF THE EPA STANDARD 40CFR191

Even though the present EPA standard 40CFR191 no longer applies as such to the proposed Yucca Mountain repository, it is useful to examine how a conceptual repository in unsaturated tuff containing unreprocessed spent fuel would perform under the technical criteria used in the standard. Barnard et al. (1992) have estimated the probabilities of releasing radionuclides to the accessible environment by a large variety of scenarios, including gaseous releases, aqueous pathways, human intrusion, and others. For a given release mechanism and probability, the curie releases of each radionuclide is calculated and ratioed to the allowable curie release of each radionuclide, if released alone, as specified by EPA. The sum of these radionuclide ratios for a release scenario of given probability is called the "EPA sum." For each scenario, a complementary cumulative probability distribution function (CCDF) is constructed, consisting of a plot of the plot of CCDF as a function of the EPA ratio. The result is compared with a limiting CCDF specified by EPA.

Results based on a composite-porosity model of flow and transport in unsaturated tuff5 are shown in Figure G-6. The region that would equal or exceed the EPA limit is shown by the shaded area in the upper right corner. Included are curves for gaseous releases (mainly 14C), releases by dissolution and hydrogeologic transport (the "aqueous" release scenario), releases by human intrusion (i.e., "drilling"), and releases by volcanism. Human intrusion includes three different scenarios, all from exploratory drilling:

  1. drilling that intersects one or more waste packages and brings radioactive cuttings to the surface;

5  

The unsaturated tuff of the proposed Yucca Mountain repository consists of porous fractured rock. In the composite-porosity model, it is assumed that there is local equilibrium of water between the pores and fractures, resulting in a long predicted time for water to move from the zone of waste emplacement down to saturated aquifers several hundred meters below.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    TABLE G-4 Individual Doses and Risks From Intrusion Scenarios, Granite Repository, Unreprocessed Spent Fuelab

     

     

    100 years

    100,000 years

    Radionuclide

    Half-Life (years)

    Dose Rate (mrem/yr)

    Risk (yr-1)

    Dose Rate (mrem/yr)

    Risk (yr-1)

    99Tc

    2.11×105

    5.0×10-1

    1.5×10-14

    1.9×10-4

    2.5×10-14

    129I

    1.57×107

    1.8×10-1

    5.4×10-16

    3.2×10-5

    4.2×10-15

    135Cs

    2.3×106

    2.4×10-1

    7.0×10-15

    1.3×10-2

    1.7×10-12

    137Cs

    3.01×101

    1.5×105

    1.8×10-7

    -

    -

    234U

    2.45×105

    7.3×101

    2.1×10-12

    5.0×10-1

    6.5×10-11

    235U

    7.04×108

    1.8

    5.1×10-14

    1.4×10-1

    1.9×10-11

    236U

    2.34×107

    1.9×101

    5.4×10-13

    2.8×10-2

    3.7×10-12

    238U

    4.47×109

    2.0×101

    5.9×10-13

    5.9×10-2

    7.7×10-12

    237Np

    2.14×106

    1.1×102

    3.1×10-12

    3.6×10-1

    4.7×10-11

    238Pu

    8.77×101

    3.6×105

    1.8×10-7

    5.1×10-1

    6.7×10-11

    239Pu

    2.41×104

    1.0×105

    1.8×10-7

    1.2×101

    1.5×10-9

    240Pu

    6.56×103

    1.5×105

    1.8×10-7

    3.1×10-2

    4.0×10-12

    241Pu

    1.44×101

    1.0×106

    1.8×10-7

    9.2×10-1

    1.2×10-10

    242Pu

    3.73×105

    5.4×102

    1.6×10-11

    8.9×10-1

    1.2×10-10

    241Am

    4.33×102

    2.0×105

    1.6×10-7

    1.8×10-1

    2.4×10-11

    243Am

    7.38×103

    6.4×103

    1.9×10-10

    3.4×10-1

    4.4×10-11

    244Cm

    1.81×101

    8.7×103

    2.5×10-10

    4.7×10-4

    6.××10-14

    245Cm

    8.50×103

    4.8×101

    1.4×10-12

    7.3×10-4

    9.6×10-14

    Total

     

    2.1×106

    1.8×10-7

    1.6×101

    2.1×10-9

    a Repository contains 18,000 Mg unreprocessed spent fuel.

    b Dose-risk conversion factor = 1.65×10-4rem (1.65×10-2/Sv) recommended by ICRP (1977).

    SOURCE: Mobbs et al. (1991).

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×
    1. drilling that penetrates to a lower aquifer in tuff, with waste cuttings falling through the open drill open to the lower aquifer; and

    2. similar drilling but to a deeper lower aquifer in carbonate rock.

    The CCDF total for all scenarios is dominated by release of gaseous 14C6. Next in importance, for the hydrodynamic model of Figure G-6, is releases by drilling. The principal radionuclides of importance that can be released to the surface and to the lower carbonate aquifer are the isotopes of plutonium and americium. However, for human-intrusion releases to the tuff aquifer the most important radionuclides, as measured by contributions to the EPA ratio, are 14C and 237Np. In the tuff aquifer the groundwater velocity is low enough and sorption high enough that little of the plutonium and americium reach the environment.

    Barnard et al. (1992) have also constructed similar CCDF plots for a different model of hydrologic flow in unsaturated tuff, wherein the groundwater is assumed to move rapidly through a few conductive fractures, without local equilibrium with the porosity of the bulk rock, as shown in Figure G-7. Here hydrogeologic transport of dissolved radionuclides is predicted to be a more important contributor to the overall CCDF than human intrusion. Determining which of the flow models best represents unsaturated tuff at Yucca Mountain must await data from further experiments and from site characterization.

    The EPA standard 40CFR191, in specifying a generic limit on allowable cumulative releases from a repository, assumes a ''generic repository," whereby all radionuclides released to the local environment appear in a hypothetical river. It assumes that a constant future human population of 10 billion people use a certain fraction of that contaminated water for growing food and for potable water. That fraction is assumed to be the same as the fraction of the total surface waters of the northern hemisphere that are used for potable water and irrigation (Pigford, 1981). A calculated collective dose and risk on this basis for the proposed Yucca Mountain repository would be too tenuous to be useful.

    The apparent problem of gaseous 14C in dominating the EPA ratio for spent fuel in unsaturated tuff suggests that special attention be given to potential collective doses from gaseous 14C for this type of waste and repository rock. As noted above, individual doses from 14C are predicted to be very low and do not appear to be a problem. None of the proposed transmutation systems are designed to transmute 14C, and such transmutation would be impossible because of its very small neutron cross section. 14C transmutation seems unnecessary because the spent fuel must first be reprocessed to transmute other species. 14C in reprocessing waste, whether from aqueous or pyrochemical reprocessing, is expected to be in a form not easily converted to gas. However, the waste forms expected to contain 14C from reprocessing have not been specified or have not been sufficiently analyzed for 14C retention.

    6  

    Although there is not a large curie inventory of 14C in spent fuel, essentially all of it is assumed to be released as a gas during a few thousand years. This may be due in part to conservative estimates of the oxidation of the UO2 in spent fuel to a solid of higher oxidation state, such as U3O8.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    FIGURE G-6 Overall CCDF for releases assuming the composite model for aqueous transport, spent fuel in unsaturated tuff.

    SOURCE: Barnard et al. (1992)

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    FIGURE G-7 Overall CCDF for releases assuming the conductive-fracture model for aqueous transport, spent fuel in unsaturated tuff.

    SOURCE: Barnard et al. (1991)

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×
    INDIVIDUAL DOSES FROM TRANSMUTATION FUEL CYCLES

    Benefits from Reprocessing. The simplest transmutation fuel cycle is the reprocessing of LWR spent fuel to separate plutonium and uranium for recycle to LWRs, as is being done in France and planned in other countries. Potential benefits to reduce release of gaseous 14C in an unsaturated-tuff repository, and to reduce collective dose therefrom, are discussed above. Other benefits could arise from reduction in plutonium inventory and from potentially better waste forms,7 as discussed below.

    Benefits from Aqueous Reprocessing and Plutonium Recycle. A reprocessing fuel cycle using aqueous reprocessing could benefit disposal in unsaturated tuff because the borosilicate-glass waste is expected to release soluble fission products more slowly than spent fuel. Also, radioiodine is recovered separately and could be incorporated in a special waste package as AgI. 99Tc in borosilicate-glass waste would be expected to be a dominant radionuclide contributing to individual dose from a repository in unsaturated tuff. Adding technetium separation could result in an improved waste package and lower individual doses from 99Tc. If the conservatively high solubilities of neptunium prevail, as discussed earlier, deploying the neptunium recovery process already developed for aqueous reprocessing could result in a waste form tailored specifically to neptunium, which could mitigate the individual dose from 237Np. Reprocessing to recover and recycle plutonium is predicted to result in over a hundredfold reduction in the risk from plutonium in the human intrusion scenarios, but the total risk from human intrusion is reduced only about twofold because americium and curium are not recycled. The risk from dissolution and migration is still predicted to be the major risk from geologic disposal, about fortyfold greater than the risk from human intrusion. If the waste is to be emplaced in unsaturataed tuff and if the conservative solubility distributions for neptunium, discussed earlier, prevail, adding neptunium separation to aqueous reprocessing could result in a better waste form that would be expected to improve repository performance. Technology for neptunium separation was developed during the 1970s.

    Benefits From Reprocessing and Transmutation in ALMRs. A 1991 evaluation of TRU burning in ALMRs by the Electric Power Research Institute (Thompson and Taylor, 1991; Wilems and Danna, 1991) included analysis of the new waste forms that would result from pyroprocessing of spent fuel from LWRs and from TRU-burning ALMRs. Lee and Choi (1991) revised and updated the Institute's analysis of waste forms, to size waste packages at comparable rates of decay-heat generation. Wastes from both aqueous and pyrochemical reprocessing of LWR spent fuel were considered, as well as waste from pyrochemical reprocessing of ALMR spent fuel. They estimated the rates of dissolution of radionuclides from the new waste forms when the radionuclides are contacted with groundwater in a repository in unsaturated tuff. They

    7  

    Whether reprocessing yields a better waste form than spent fuel depends much on the geochemical environment of the repository. Reprocessing waste, such as borosilicate glass, may be a better waste form for a repository in unsaturated rock, but spent fuel may be superior in a repository in saturated rock, where a reducing environment is expected.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    used release models developed for the Yucca Mountain project and assumed inventories resulting from 0.1% loss of TRUs to waste in each reprocessing cycle. Elemental solubilities developed by the Yucca Mountain project were used, together with solid-solid alteration rates for spent fuel and borosilicate glass waste. For entirely new waste forms, such as the proposed copper-matrix waste from pyrochemical reprocessing, release rates were based in part on oxidation rates reported in the literature. Release rates of soluble species from waste packages would be further attenuated by the assumed slow rate of intrusion of groundwater into a waste container, based on the "wet-drip" water contact mode proposed by the Lawrence Livermore National Laboratory.

    From these studies, Lee and Choi (1991) conclude that dissolution rates from reprocessing waste packages are approximately the same as from LWR spent fuel. Even though TRUs have been reduced in inventory by over two orders of magnitude, solubilities rather than inventories control the predicted release rates of low-solubility species from a given waste package.8 Thus, for the amount of waste that has resulted from a given generation of thermal energy, the predicted doses and risks from dissolution and transport of solubility-limited species to the environment would not be significantly lower for the ALMR wastes that contain lower inventory of TRUs, as compared to spent LWR fuel waste. However, the lower inventories of TRUs in the waste packages would benefit the intrusion scenarios.

    Using the dissolution models and parameters developed by Lee and Choi (1991), Hirschfelder et al. (1992) predicted long-term individual doses from using contaminated water from a lower aquifer for a repository in unsaturated tuff. Hydrogeologic transport was based on composite porosity models of groundwater flow, calculated to a distance of 5 km from the emplaced waste. In this analysis, waste from pyrochemical reprocessing was assumed. Maximum doses to future individuals from using contaminated groundwater9 were calculated, extending to 1 million years. Calculated individual doses for ALMR-transmutation fuel cycles were compared with doses for a reference repository containing LWR spent fuel. It was assumed that temperature limits of the waste package and nearby rock allowed no more than 57 kW/acre of decay heat. Consequently, for the transmutation fuel cycles, the smaller inventory of TRUs and the correspondingly smaller decay heat at the time of emplacement allowed waste from the generation of a larger amount of electrical energy to be emplaced in the same repository area. An additional scenario was calculated assuming that strontium and cesium fission products are separated in reprocessing and stored for a few hundred years. Removing the intense decay-heat source from 90Sr and 137Cs would allow a far greater amount of waste to be emplaced in a given repository area for the same areal heat loading. The stored cesium would eventually have to be emplaced in a geologic repository because of the inventory of long-lived

    8  

    Ultimately, if the inventory of a given species in a waste solid is reduced sufficiently, the rate of release from the waste matrix will control its dissolution rate, rather than the rate of dissolution of a precipitate of that species. However, the alteration and dissolution models predict that for 0.001 percent reprocessing loss, the rate of dissolution of uranium, neptunium, plutonium, and americium will be controlled by solubility (Pigford, 1990a).

    9  

    At the site of the proposed repository at Yucca Mountain, there is no surface water into which groundwater can discharge.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    135Cs. The 135Cs is evidently below NRC Class C limits for storage as Class C LLW. However, as discussed earlier, the dissolution and transport of 135Cs can contribute such significant individual doses when emplaced in deep geologic storage. Therefore, it is questionable that storing that amount of 135Cs as LLW could be acceptable if subjected to similar safety analyses.

    Relative values of the maximum dose rates for these three scenarios are listed in Table G-5. The maximum dose rates for the reprocessing-transmutation fuel cycle of the ALMR, without strontium-cesium separation, are near or slightly below those for unreprocessed spent fuel, reflecting benefits from improved waste forms. Adding strontium-cesium separation considerably increases the inventory of fission products in the repository and increases the number of waste packages. Even the doses from long-lived TRUs such as 242Pu and 237Np are greater than those for the reference LWR fuel cycle, even though the inventories are lower. Release of plutonium and neptunium are solubility limited. The release of 242Pu and 237Np increase with increased total loading because of the greater number of waste packages, the greater total surface area of waste packages for mass transfer to ground water, and the greater effective volumeric flow rate of ground water into which plutonium and neptunium can dissolve. The largest increases are from the long-lived soluble fission products 129I and 99Tc. The maximum dose rate for 129I in reprocessing waste with strontium-cesium separation is predicted to be about twentyfold greater than for the reference fuel cycle. The relative peak dose 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.

    These calculations show that there could be a penalty in increased individual dose if transmutation were employed so that waste from a greater amount of electrical energy generation could be loaded in the same emplacement area of the proposed Yucca Mountain repository. Collective dose from these increased loadings would also increase. However, this same increase in collective dose would occur if the increased loading were emplaced instead in a second repository with the same characteristics of the proposed Yucca Mountain repository.

    The present approach of the Yucca Mountain Project is to use decay heat to maintain a dry atmosphere at the surfaces of waste containers for thousands of years, with loadings of unreprocessed spent fuel. If that same amount of spent fuel were loaded into a smaller area, to increase the time for a dry repository, individual and collective doses would not necessarily increase. However, if the remaining repository area now designated for Yucca Mountain and the additional area that could be developed were loaded with additional waste produced by power-generating transmutors, individual and collective doses from some radionuclides could increase. Thus, proposals to delay the need for a second repository by implementing transmutation and burying transmutation waste in Yucca Mountain could make it more difficult for the proposed repository to meet a dose or risk limit. However, individual and collective doses would increase if the remaining repository area and the available area were loaded with additional spent fuel, thereby delaying the need for a second repository.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    FIGURE G-8 Comparison of aqueous releases for LWR spent fuel and S&T waste in unsaturated tuff.

    SOURCE: Barnard and Lee (1992)

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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).

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    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.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    REFERENCES

    Andrews, R. W., T. F. Dale, and J. A. McNeish. 1994. Total System Performance Assessment - 1993: An Evaluation of the Potential Yucca Mountain Repository. Intera Report B00000000-01717-2200-00099-Rev. 01. Albuquerque, N. Mex.: Sandia National Laboratories.


    Barnard, R. W., and W. W.-L. Lee. 1992. Performance-assessment comparisons for a repository containing LWR spent fuel or partitioned/transmuted nuclear waste. Pp. 1397-1403 in High-Level Radioactive Waste Management, Third International Conference. La Grange Park, Ill.: American Nuclear Society.

    Barnard, R. W. 1993. Radionuclide releases from geologic repositories. Paper presented to National Academy of Sciences Committee on Technical Basis for Yucca Mountain Standards, Sandia National Laboratories, July 1993.

    Barnard, R. W., M. L. Wilson, H. A. Dockery, J. H. Gauthier, P. G. Kaplan, R. R. Eaton, F. W. Bingham, and T. H. Robey. 1992. TSPA 1991: An Initial Total-System Performance Assessment for Yucca Mountain. SAND91-2795. Albuquerque, N. Mex.: Sandia National Laboratories.

    Bergstrom, U., and S. Nordlinder. 1991. Uncertainties related to dose assessments for high level waste disposal. Nuclear Safety 32(3):391-402.


    Duguid, J. O., R. W. Andrews, E. Brandstetter, T. F. Dale, and M. Reeves. 1994. Calculations Supporting Evaluation of Potential Environmental Standards for Yucca Mountain. Intera Report B00000000-01717-2200-00094-Rev. 01. Albuquerque, N. Mex.: Sandia National Laboratories.


    Environmental Protection Agency. 1985. Environmental radiation protection standards for management and disposal of spent nuclear fuel, high-level and transuranic radioactive wastes. Code of Federal Regulations, Title 40, Part 191.

    Eriksson, L. G. 1991. The MD Design - A cool concept geologic disposal of radioactive waste. Pp. 1569-1584 in Proceedings of the Second Annual High-Level Radioactive Waste Management Conference. La Grange Park, Ill.: American Nuclear Society.

    Eslinger, P. W., L. A. Doremus, D. W. Engel, T. B. Miley, M. T. Murphy, W. E. Nichols, M. D. White, D. W. Langford, and S. J. Ouderkirk. 1993. Preliminary Total-System Analysis of a Potential High-Level Nuclear Waste Repository at Yucca Mountain. PNL-8444. Richland, Wash.: Pacific Northwest National Laboratory.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    Hirschfelder, J., P. L. Chambre, W. W.-L. Lee, T. H. Pigford, and M. M. Sadeghi. 1991. Effects of actinide burning on waste disposal at Yucca Mountain. Trans. American Nuclear Society 64(November):111-113.

    Hirschfelder, J., W. W.-L. Lee, and T. H. Pigford. 1992. Effects of actinide burning on risk from geologic repositories. In Proceedings of the American Institute of Chemical Engineers 1992 Summer National Conference. LBL-32061. Berkeley, Calif.: Lawrence Berkeley Laboratory.

    Hunter, T. O., J. R. Tillerson, and A. L. Stevenson. 1989. A conceptual design for a nuclear waste repository at the Yucca Mountain site. Radioactive Waste Management and Nuclear Fuel Cycle 13:93-104.


    International Commission on Radiological Protection (ICRP). 1977. Recommendations of the International Commission on Radiological Protection. ICRP Publication 26. Ann. ICRP 1 (3). London: Pergamon.


    Johnson, G. L. 1991. Thermal Performance of a Buried Nuclear Waste Storage Container Storing a Hybrid Mix of PWR and BWR Spent Fuel Rods. UCID-21414 Rev 1. Berkeley, CA: Lawrence Livermore National Laboratory.


    Lee, W. W. L., and J. S. Choi. 1991. Release Rates from Partitioning and Transmutation Waste Packages. LBL-31255. Berkeley, Calif.: Lawrence Berkeley Laboratory.


    Mobbs, S. F., M. P. Harvey, J. S. Martin, A. Mayall, and M. E. Jones. 1991. Comparison of the Waste Management Aspects of Spent Fuel Disposal and Reprocessing: Post-Disposal Radiological Impact. EUR 13561 EN. Harwell, United Kingdom: NRPB.


    National Waste Technical Review Board. 1992. Fifth Report to the U.S. Congress and the U.S. Secretary of Energy. Washington, D.C.: National Waste Technical Review Board.


    Pigford, T. H. 1981. Derivation of Release Limits in EPA's Proposed Standard for Geologic Disposal or Radioactive Waste. Board on Radioactive Waste Management, National Research Council.

    Pigford, T. H., J. O. Blomeke, T. L. Brekke, G. A. Cowan, W. E. Falconer, N. J. Grant, J. R. Johnson, J. M. Matuszek, R. R. Parizek, R. L. Pigford, and D. E. White. 1983. A Study of the Isolation System for Geologic Disposal of Radioactive Wastes. Washington, D.C.: National Academy Press.

    Pigford, T. H. 1990. Reprocessing Incentives for Waste Disposal. UCB-NE-4171. Berkeley: University of California.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
    ×

    Pigford, T. H. 1990. Actinide burning and waste disposal. In Proceedings of the M.I.T. International Conference on the Next Generation of Nuclear Power Technology. UCB-NE-4176. Berkely: University of California.

    Ramspott, L. D. 1991. The constructive use of heat in an unsaturated tuff repository. Pp. 1602-1607 in Proceedings of the Second Annual High-Level Radioactive Waste Management Conference. La Grange Park, Ill.: American Nuclear Society.

    Roddy, J. W., H. C. Claiborne, R. C. Ashline, P. J. Johnson, and B. S. Ryhne. 1986. Physical and Decay Characteristics of Commercial LWR Spent Fuel. ORNL/TM-9591. Oak Ridge, Tenn.: Oak Ridge National Laboratory.


    Statens Kärnkraftinspektion (Swedish Nuclear Power Inspectorate). 1991. SKI Project-90. SKI Technical Report 91:23. Vol. 2. Stockholm: Statens Kärnkraftinspektion.

    Svensk Kärnbränslehantering AB (Swedish Nuclear Fuel and Waste Management Co.). 1992. Final Disposal of Spent Nuclear Fuel. Importance of the Bedrock for Safety. SKB Technical Report 92-20. Stockholm: Svensk Kärnbränslehantering AB.


    Thompson, M. L., and I. N. Taylor. 1991. Projected Waste Packages Resulting from Alternative Spent-Fuel Separation Processes. EPRI NP-7262. Palo Alto, Calif.: Electric Power Research Institute.


    Vieno, T., A. Hautojarvi, L. Koskinen, and H. Nordman. 1991. TVO-92: Safety Analysis of Spent Fuel Disposal. YJT-92-33E. Helsinki: Nuclear Waste Commission of Finnish Power Companies.


    Wilems, R. E., and J. G. Danna. 1991. The Effects of Transuranic Separation on Waste Disposal. EPRI NP-7263. Palo Alto, Calif.: Electric Power Research Institute.

    Wilson, M. L., J. H. Gauthier, R. W. Barnard, G. E. Barr, H. A. Dockery, E. Dunn, R. R. Eaton, D. C. Guerin, N. Lu, M. J. Martinez, R. Nelson, C. A. Rautman, T. H. Robey, B. Ross, E. E. Ryder, A. R. Schenker, S. A. Shannon, L. H. Skinner, W. G. Halsey, J. D. Gansemer, L. C. Lewis, A. D. Lamont, I. R. Triay, A. Meijer, and D. E. Morris. 1994. Total-System Performance Assessment for Yucca Mountain, SNL Second Iteration (TSPA-1993). SAND93-2675. Albuquerque, N. Mex.: Sandia National Laboratories.

    Suggested Citation:"G EFFECTS ON REPOSITORY." National Research Council. 1996. Nuclear Wastes: Technologies for Separations and Transmutation. Washington, DC: The National Academies Press. doi: 10.17226/4912.
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    Disposal of radioactive waste from nuclear weapons production and power generation has caused public outcry and political consternation. Nuclear Wastes presents a critical review of some waste management and disposal alternatives to the current national policy of direct disposal of light water reactor spent fuel. The book offers clearcut conclusions for what the nation should do today and what solutions should be explored for tomorrow.

    The committee examines the currently used "once-through" fuel cycle versus different alternatives of separations and transmutation technology systems, by which hazardous radionuclides are converted to nuclides that are either stable or radioactive with short half-lives. The volume provides detailed findings and conclusions about the status and feasibility of plutonium extraction and more advanced separations technologies, as well as three principal transmutation concepts for commercial reactor spent fuel.

    The book discusses nuclear proliferation; the U.S. nuclear regulatory structure; issues of health, safety and transportation; the proposed sale of electrical energy as a means of paying for the transmutation system; and other key issues.

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