R. P. Rechard, M. E. Lord, C. T. Stockman
Nuclear Waste Management Center, Sandia National Laboratories,
Albuquerque, NM 87185-1328

R. D. McCurley
New Mexico Engineering Research Institute, University of New Mexico,
Albuquerque, NM 87110

* This work was supported under contract DE-AC04-94AL85000. Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the U.S. Department of Energy.

The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE-owned spent nuclear fuel and high-level waste (DSNF/DHLW). A desirable option, direct disposal in the potential repository at Yucca Mountain, depends on the final repository acceptance criteria, which will be set by DOE's Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating repository acceptance criteria is to gain an understanding of the types of DOE-owned spent nuclear fuels and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of ⁹⁹Tc and ²³⁷Np from commercial spent fuel (assuming no protection from Zircaloy cladding) exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the repository dose criteria, then DSNF can also meet these criteria. Additionally, these results are supported by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%).

The Office of Environmental Management of the U.S. Department of Energy (designated here as DOE/EM) is responsible for the safe management and disposal of DOE-owned spent nuclear fuel and high-level waste (DSNF/DHLW). A desirable option is direct disposal in the potential repository at Yucca Mountain, which is being designed and characterized by DOE's Office of Civilian Radioactive Waste Management (OCRWM). The potential repository is designed primarily for disposal of commercial spent fuel, although a portion is reserved for DSNF/DHLW. Thus the final repository acceptance criteria for the potential repository, which will be set by the OCRWM, are expected to accommodate characteristics of commercial spent fuel.

The viability of direct disposal of DSNF/DHLW, then, depends on the final repository acceptance criteria. However, evolving regulations make it difficult to determine what these criteria will be. Currently, the technical criteria as promulgated by the U.S. Nuclear Regulatory Commission (NRC) in 10 CFR 60 (which must be revised per the Energy Policy Act of 1992, Pub. L. 102-486) establish stringent minimum requirements for disposal subsystems. Because these criteria are already spelled out, their requirements would appear to be the logical starting point for establishing repository acceptance criteria. However, recent proposed amendments to the Nuclear Waste Policy Act [1], along with guidance from the NAS [2,3], emphasize an evaluation of the whole disposal system, rather than regulations per system component.

At the same time, DOE/EM contractors that manage storage sites and are responsible for preparing the DSNF for disposal would prefer definite procedures for characterizing and treating DSNF-procedures that could be followed now, prior to completion of the lengthy licensing procedures and even before the establishment of the final repository acceptance criteria. Although such arbitrary limits are appealing, because they would permit a straightforward course of action, they have the potential to be extremely costly by overcompensating for the unknown, and would themselves be subject to analysis in the later stages of regulatory licensing to ensure that they provide safe conditions.

An alternative to arbitrary limits is to gain an understanding of the DOE-owned fuel types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent nuclear fuel.¹ Given that the final repository acceptance criteria will accommodate the behavior of commercial spent fuel, a preliminary study that indicates whether the DSNF/DHLW performs better (or worse) than commercial fuel is a strong indicator of the level of characterization and treatment that will be necessary before disposal.

The purpose of the study reported here, referred to as the 1997 DSNF/DHLW performance assessment (or 1997 PA), is to assist the DOE/EM by providing information about its fuel so that it is well acquainted with the issues and options available as site-specific regulations for the potential Yucca Mountain repository emerge. In the 1997 PA, the behavior of the DOE-owned fuel was compared to commercial spent fuel after disposal in the potential Yucca Mountain repository. The radioisotope inventory-75,410 Mg of heavy metal-included a large proportion of commercial spent nuclear fuel (83.7%), with only 12.5% of the mass of heavy metal from DHLW, and 3.8% from DSNF. Also, the DSNF represents only 1.5% of the activity (curies), because this fuel was usually "burned" for only a short time. Thus, the DSNF's small activity (1.5%) and mass of heavy metal (3.8%) is very likely within the error band of the commercial spent fuel's characteristics, and so the possibility of the DSNF influencing the overall performance of a geologic system is remote. However, some types of DSNF have characteristics that might seem to differ substantially from commercial fuel. Of particular interest are characteristics related to 10 CFR 60 criteria, specifically, the prohibition of combustibility, as in the graphite blocks of Fort St. Vrain fuel, and pyrophoric material, such as metallic uranium spent fuel. In the 1997 PA, these characteristics were analyzed to confirm the supposition that DSNF would not influence disposal system performance. Detailed knowledge of the post-disposal behavior of DSNF is an asset for developing acceptance criteria that are prudent, cost-effective, and time-saving.

An example can help to clarify the purpose of the 1997 PA. The pyrophoric nature of N-Reactor spent fuel, because of its metallic uranium, might be expected to adversely affect conditions in the repository. However, O₂ must be present for the pyrophoric fuel to pose a danger. This study suggests that as corrosion progresses on the massive amounts of steel in the waste containers, the amount of O₂ in the repository will deplete; thus, the repository environment may greatly diminish the need for concern about the metallic uranium. Consequently, this type of fuel might be accepted without restrictions, which would also eliminate the need for the DOE/EM to reprocess or otherwise condition the fuel, at additional expense. The OCRWM may also benefit from this detailed analysis (O₂ transport is not currently considered in overall repository performance) because the corrosion results are relevant to the life of waste containers and so could aid the OCRWM in decisions about further modeling refinements for licensing of the repository.

The first goal of this analysis was to demonstrate the minor influence of the DSNF in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project's (YMP) Managing and Operating (M&O) Contractor, TRW, when they conduct their Total System Performance Assessment (TSPA) for the OCRWM. A second goal, closely related to the first, was to identify the most sensitive parameters through a preliminary analysis of the results to determine which DSNF characteristics should be carefully estimated or measured and which could be neglected, after demonstrating their minor influence. A third goal was to continue development and implementation of modeling features to examine the availability of O₂ in the repository.

In general, this analysis used the methodology of two related studies (1993 and 1994 PAs [4,5]). The methodology was originally established by Sandia National Laboratories for assessing the long-term performance of the Waste Isolation Pilot Plant (WIPP) geologic disposal system [6,7]. The 1997 PA is an important step in examining the unique characteristics of DSNF. The specific characteristics of concern are (1)_the potential for pyrophoric behavior of the uranium metallic fuels in up to 89% of the heavy metal mass (e.g., Category 1, N-Reactor), and (2) the potential to go critical in the near or far field in up to 19% of the heavy metal mass. Studying these specific characteristics required developing detailed process models for the degradation of the containers and fuel matrices within the repository.

Modeling conditions for the analysis were coordinated with those to be used in the YMP model for the potential Yucca Mountain repository. As a means of comparison and benchmarking, Sandia's method of analysis provided the complex model to YMP's more simplified, uncoupled model [8], i.e.,

• The YMP performance assessment team used simple analytic expressions developed from abstractions of complicated analyses performed earlier in the program.
• YMP ran a deterministic simulation, then varied parameter values one at a time to determine sensitivity.
• Sandia simulated the complex details of contaminant transport in heated, unsaturated tuff directly in its analysis to complement the development of abstractions used by the TRW performance assessment team.
• Sandia improved the unsaturated zone modeling by including transport of O₂ in order to evaluate whether reactants such as O₂ and H₂O are limited significantly, thus reducing the rate of oxidation of container or waste.
• Sandia adjusted its modeling assumptions to match current TSPAs, such as

• Sandia ran a probabilistic simulation varying 36 parameters.

• used a simple process source term model but tightly coupled it with the two-phase flow model to more adequately model radioisotope releases.
• chose a more accurate representation of the behavior of DSNF to more definitively identify DSNF fuel characteristics that have the most influence on performance.
• included in the source-term model both oxic and anoxic corrosion rates and solubilities. (However this feature was not important in the preliminary work because carbon steel corrosion rates were chosen to be high.)

The main performance criteria used in the 1997 PA are based on 40 CFR 191 [9], guidance from the NAS in 1995 regarding a proposed future standard for repository licensing [3], and a proposed 1997 amendment to the Nuclear Waste Policy Act of 1982 [1]. The criteria examined are the probabilistic maximum doses received by an individual; the time period for analysis is generally 50,000 yr (the deterministic simulations with sampled parameters set at their mean or median value were run out to 100,000 yr).

Although the absolute position of the results with relation to the total system criteria of 40 CFR 191 (or replacement regulations) is crucial when a site applies for a license, the more important results here are those that can be used for project guidance with regard to the potential repository at Yucca Mountain. These are (1) the relative position of the various fuels in relation to the commercial spent nuclear fuel and (2) the relative importance of model parameters in explaining the value of a particular performance measure.

Final results from the 1997 PA should be available in spring of 1998. Preliminary findings from this performance assessment are presented here and are grouped into four areas: specific findings about the waste form, the waste containers and repository, and the geologic barrier, and then general findings about the system.

Two categories of commercial spent fuel were modeled: 21-PWR (Pressurized Water Reactor) assemblies and 44-BWR (Boiling Water Reactor) assemblies (Categories 14 and 15) (Table I). Their combined mass of heavy metal was 63,170 MTHM. The commercial spent nuclear fuel will potentially be sent from 110 different reactors; however, the radioisotope inventory supplied for this analysis was an average recently estimated by the OCRWM for its performance assessment, TSPA-VA, to be completed in 1998. The DSNF under study includes ~2831 Mg heavy metal, codisposed with ~9409 Mg heavy metal of DHLW, for a total of 12,240 Mg heavy metal. Of the total 75,410 Mg of heavy metal, only 12.5% of the mass of heavy metal is from DHLW, and 3.8% from DSNF. (For the purpose of this analysis, the DSNF does not include the Navy propulsion reactor spent nuclear fuel.) Within the total repository, 2.9% (75.3% of the 3.8%) of the heavy metal mass is N-Reactor spent nuclear fuel, which was grouped as Category 1 for the analysis. The remaining 1.0% (24.7% of the 3.8%) of waste represents over 200 types of experimental fuel, and was grouped into 12 categories, for a total of 13 DSNF categories (see Table I). In the 1997 PA, the 12 DSNF categories representing 1.0% of the heavy metal mass in the repository were emplaced in large disposal containers together with between three and five handling containers of DHLW in borosilicate glass.

Table I. Spent Nuclear Fuel and High-Level Waste Categories

When using the source term code, CST², the preliminary comparison of commercial spent fuel with DSNF/DHLW showed that releases of ⁹⁹Tc and ²³⁷Np³ from commercial spent nuclear fuel exceed those of DSNF/DHLW both per mass and per package (Fig. 1). Thus, if commercial spent fuel can meet regulatory dose criteria, then the DSNF/DHLW can also meet the criteria. In large part, this result is caused by the lower burnup values for DSNF and by its small percentage of total activity in the repository (1.5%) and regulatory mass (4%). These results also assume that cladding of the commercial spent fuel provides no protection (i.e., has failed) and so, in contrast to an earlier study (1994 PA), the type and integrity of cladding were not important. However, temperature-sensitive alteration rates of the fuel matrix appeared to be especially important because the corrosion rates of the containers were assumed to be so rapid that failures occurred when the repository was still hot. Releases of ⁹⁹Tc were limited only by the amount of waste that was exposed. Releases of ²³⁷Np were limited by the amount of waste exposed and by its solubility. In this PA and the new YMP TSPA [8], ²³⁷Np solubility was reduced two orders of magnitude from that used in the previous DSNF PA [5] and YMP TSPA [10]. The reduction was the result of reevaluating previous experiments and excluding those by Nitsche et al. [11], which were not representative of expected repository conditions.

In the 1997 PA, the current YMP plans for the disposal containers and codisposal configurations for the DSNF, DHLW, and commercial fuel were incorporated, although the variety of dimensions and configurations were reduced to ease the modeling burden. For the 1997 PA, the disposal region of the repository was sized to accommodate ~75,410 MTHM of spent fuel or equivalent high-level waste, with DOE-owned and commercial spent fuel comingled.

The DSNF and DHLW were modeled as being placed in both handling and disposal containers. Category 1 (N-Reactor fuel) was modeled in a multi-canister overpack (MCO), which consisted of a 0.95-cm-thick, 61-cm outer diameter stainless steel shell. DSNF Categories 2 through 13 were modeled in a 6.35-mm 304L stainless steel handling container. The DHLW handling container was modeled as the 0.95-cm-thick, 61-cm-diameter standard DOE canister.

The disposal container for all waste packages included a 10-cm-thick outer carbon steel layer and a 2-cm-thick inner Inconel 625 layer. For Waste Package 1, four MCOs were placed in a disposal container; for Waste Packages 2 through 13, DSNF/DHLW were codisposed in various configurations (Table II); for Waste Packages 14 and 15, 21-PWR and 44-BWR assemblies were placed in disposal containers, respectively.

The orientation of waste packages was center in-drift on pedestal emplacement [10]. The mine design for center in-drift emplacement is a tunnel and pillar configuration, with the entire repository consisting of 87 tunnels, 1236-m-long and bored by machine. The tunnels are surrounded and connected by access drifts, 7.62 m in diameter, which lead to ramps that connect to the surface. The design goal in the 1997 PA was to use the hot repository concept. Hence, a center-to-center spacing of 28.6 m between the 5.3-m-wide tunnels was designed to maintain a heat load equivalent of 85 MTHM/acre for commercial fuel. The tunnels were assumed to be lined with a 0.2-m-thick precast concrete liner. A thick segment on the base of the tunnel forms an invert into which a minimum of 3 piers can be set for container support. Only the ramps to the disposal area were assumed to be backfilled and effectively sealed.

The repository tunnels in which the waste packages are to be emplaced include an air gap between the waste package and the intact host rock. During the regulatory period, rubble from the back (top) and ribs (sides) of the tunnels was assumed to eliminate the air gap, to some extent. If rubble were present (absence or presence of rubble was a sampled parameter) and if the H₂O content of the tuff were high enough, movement of groundwater from the partially saturated tuff through the rubble to the waste packages was assumed. For the 1997 PA, four container groups were defined that specified (a)_whether the inner Incoloy layer of the disposal container failed at the time of emplacement (yes/no) and (b) whether there was an air gap between the waste package and the intact host rock (yes/no). The corrosion submodel used in the 1997 PA included (a) general and localized corrosion and (b) oxic (humid and wet) and anoxic corrosion reactions.

Also, the effect of infiltration in the repository from climate change was modeled by means of a cosinusoidal function to simulate potential variation of precipitation. Minimum infiltration, maximum infiltration, and time period were sampled. The resulting average infiltration over the model grid was varied and, in addition to this regional variation, a sample factor focused infiltration at one cell over the repository near the crest of Yucca Mountain between 1 and 10 times that over the rest of the repository.

Results from BRAGFLO_T⁴ showed that the thermal loading from combining the commercial spent fuel with the codisposed DSNF/DHLW was sufficient to boil away all of the water near the repository center. Maximum temperatures near the repository center were between 125· C and 145· C. Peak temperatures occurred in the first 30 to 300 yr (Fig. 2). At the repository edge, some runs (about one third) produced superheated steam, while others maintained the water in the liquid phase; hence protection from drying out the tuff was not consistent at the edge of the repository (Fig. 2).

Fig. 2. Temperature at (a) center and (b) edge of potential repository over 50,000 yr. Fifty runs are shown, with 36 model parameters varied using Latin Hypercube Sampling. (Preliminary results from 1997 DSNF/DHLW PA.)

Water vaporization during early times displaced the noncondensible gases (e.g., O₂) within the repository. The oxygen mass fraction was at or near zero from approximately 40 to 200 yr. At later times, the water vapor recondensed and the oxygen migrated back to the repository (Fig. 3).

When the anoxic corrosion rate of the carbon steel was set high, then failure of 72% of the containers occurred in the first 3000 yr. Hence, the alteration of the fuel matrix also was rapid because it occurred when the repository was still hot.

Containers at the edges of the repository breached earliest, followed by the containers under the high infiltration zone artificially specified near the crest of Yucca Mountain. Containers near the edge breached earliest because (1) they were below the boiling point longer than other containers at early times and (2) the oxygen mass fraction at the edge was not suppressed as low or as long as at the center of the repository. Away from the edges and the high infiltration zone, containers with an intact air gap and with a good Inconel layer never breached.

Technetium is very soluble so its releases were dominated by advective fluid flow. Neptunium is much more insoluble, and so its advective and diffusive releases were more similar. The time histories of releases were quite different across the repository because of differences in temperature and water flux, which affected the time of radionuclide exposure and the rate at which they were removed from the containers.

The geologic barrier of the tuff disposal system comprises the sequence of tuff that isolates the repository from the accessible environment, i.e., the surface and any location elsewhere that is beyond 5 km from the repository [9]. Only information from the unsaturated zone is presented here.

The stratigraphy of the tuff disposal system was idealized as a series of constant thickness hydrologic modeling units ("pancakes") with a dip of 4.6· . The thirteen units of the unsaturated zone consisted of consecutive tuff layers with degrees of welding and porosity that were similar. Welded lithologies were assumed to be more densely fractured than nonwelded. The repository was assumed to be located in the unsaturated zone about 260 m below the surface and 380 m above an aquifer. Two faults, Solitario Canyon and Ghost Dance, were modeled as having an effect on flow, with Solitario Canyon assumed as a lower permeability zone and Ghost Dance as a higher permeability zone. Model layers were offset across the Solitario Fault but not across the Ghost Dance Fault. The cross-section modeled was parallel to the 4.6· dip.

Two-phase flow (liquid and gas) with heat conduction, convection, and phase changes were included in the two-dimensional model. In the unsaturated zone, the composite porosity simplification (also called equivalent continuum) of the dual permeability model was used for fluid flow, i.e., the fractures and matrix were assumed to be in local thermodynamic equilibrium. Thus a single saturation and permeability function was used to represent the fracture and matrix media. For transport in the unsaturated zone, the model used a similar formulation.

The gas phase flow in the repository was driven by the heating of the repository at early times and subsequent cooling at later times. Gas flow was circulatory, with gas flowing along the sides of the repository, especially down from the surface through the more permeable Ghost Dance fault. Gas then moved vertically up through the repository and returned to the surface. As the repository cooled in time, the gas flow weakened.

The radionuclide transport showed that the highly soluble technetium demonstrated a high release at early time with a resulting slug of high concentration that was transported via water flow to the vicinity of the water table with maximum concentrations at approximately 22,000 yr (Fig. 4a). The transport of the less soluble neptunium from the failed containers (mean solubility = 5.818 ´ 10⁶ KG/M³) was more gradual because precipitation occurred within the waste packages (Fig. 4b). Advection and diffusion of ²³⁷Np from the packages brought the fluids in the repository floor up to the solubility limit soon after container breach. Continued release of neptunium maintained the grid blocks below the repository at the solubility limit well past the calculation times. At 100,000 yr, the concentration at the water table (3 ´ 10⁶ kg/m³) had almost reached the solubility limitt of ²³⁷Np (Fig. 4b).

For the preliminary calculations, the maximum annual effective dose equivalent was evaluated for a maximally exposed individual who received the peak dose by consuming two liters of drinking water per day for 365 days. Time histories of the committed dose were generated for a 100,000-yr period for two deterministic runs with parameters set at mean and median values. A biosphere transport code, GENII-A, was used to calculate a 50-yr committed dose equivalent based on a 1-yr exposure, i.e., the 50-yr committed dose from consuming contaminated water and food over a 1-yr period.

For the calculations, the water that was consumed and was contaminated with radionuclides originated at (a) the center and edge of the repository, (b) the region in the repository beneath the area of increased infiltration, (c) a maximum concentration region just above the water table, and (d) a point in the underlying aquifer 5_km from the repository. The pathway to humans was from a well, located at one of the sources listed above (with no dilution), to a farm family that drank the water directly and subsisted on crops and animal products grown on the farm using the same contaminated water.

The doses from water taken at the repository location beneath the area of increased infiltration showed the highest peaks of the three repository locations. Early spikes that occurred for sources inside the repository (at times much earlier than 1000 yr) were due to ⁹⁹Tc and ¹²⁹I, whose concentrations then fell off rapidly. Although ²³⁷Np concentrations also rose early, they did not rise above their plateau, at about 17 rem, and there was little change out to 100,000 yr.

As noted above, the doses from the locations inside the repository for this analysis were based on unrealistic assumptions. However, these calculations demonstrate the attenuation of peak concentration (and dose) as the radioisotopes moved through the geologic barrier. Very significant attenuation of concentrations of ¹²⁹I and ⁹⁹Tc occurred during the transport to the water table from the repository through the unsaturated zone, for the doses were substantially reduced (by several orders of magnitude). The doses from ²³⁷Np in the underlying aquifer at an arbitrary boundary 5 km from the repository at 100,000 yr were 400 and 350 mrem for the mean and median runs, respectively.

The overall dose calculations confirm results from past performance assessments [e.g., 5,10] that the unretarded radioisotopes of technetium, iodine and neptunium (i.e., ⁹⁹Tc, ¹²⁹I, and ²³⁷Np) are the most important radioisotopes for evaluating potential doses. In turn, the specific results about the waste form show that releases of ⁹⁹Tc, ¹²⁹I, and ²³⁷Np from DSNF/DHLW are less than those from commercial spent nuclear fuel on both a per package and per mass of heavy metal basis. Hence, if the commercial spent fuel can meet future regulatory dose criteria, then DSNF can also meet the criteria.⁵

In general, the overall results about the waste form, container, and geologic repository show that the source term model (which includes the waste form and container, and its interaction with the geologic barrier) is very important for evaluating the efficacy of the potential Yucca Mountain repository. (The source term model appears to be much more important here than for other types of repositories, such as the WIPP repository in bedded salt for transuranic nuclear waste [6].) Only a few properties of the DSNF significantly influence the source term however, and thus only these properties should become part of the repository acceptance criteria. First, the inventory of ⁹⁹Tc, ¹²⁹I, and ²³⁷Np is of primary importance. These radioisotopes are a function of the consumption of fissile uranium (²³⁵U) ("burnup"); thus, burnup is an important DSNF characteristic to determine. Because DSNF typically has lower burnup than commercial fuel, the DSNF usually contains fewer of these radioisotopes. It is also important to note that Category 1, consisting predominately of N-Reactor Fuel, makes up ~80% of the DSNF mass of heavy metal. (N-Reactor fuel also accounts for about 18% of the DSNF volume.) Thus, the performance of N-Reactor fuel could be significant with regard to compliance with potential acceptance criteria. Second, the hot repository concept makes temperature-dependent alteration rates of the fuel matrix (and possibly cladding) important information to acquire. Third, the specific results for the geologic barrier show that oxygen is likely to be depleted in the repository when the container fails, and thus the pyrophoric nature and combustibility of DSNF are unlikely to affect overall performance of the repository. Oxygen and water will eventually become available, but very slowly (i.e., the length of time required to oxidize or otherwise alter the fuel matrix deep in the geologic repository would be much shorter than the time it would take in a surface waste processing facility specifically designed to oxidize the waste for eventual disposal).

In summary, because the mass, volume, and activity of the DSNF are modest in relation to the commercial fuel, analysis of the waste forms should continue to demonstrate that DSNF is acceptable in the potential Yucca Mountain repository. Any negative characteristics of DSNF discovered over time are not likely to adversely influence the entire disposal system. Therefore, a likely recommendation, after completing the 1997 PA, is that direct disposal of DSNF remain the primary option considered by the DOE/EM, a recommendation that was supported by the previous performance assessments. This recommendation is predicated on the assumption that analysis to enhance DOE/EM's understanding of its waste will continue so that convincing arguments with regard to repository acceptance criteria can be presented to the NRC regulator. It is believed that preliminary analysis that generates understanding of the real issues is less expensive in the long run than adopting expensive treatment and packaging options in the hope of reducing later analysis costs and gaining faster acceptance during licensing of the repository. Treatment and packaging plans, based on information gained through the waste analysis, can be developed as a secondary option but not necessarily implemented. The advantage of continued preliminary analysis is that it not only informs the decision makers about significant issues but can also contribute to safe, reasonable, and cost-efficient criteria for acceptance of DSNF/DHLW.

Comparison of the behavior of DSNF with commercial spent fuel is a potential method of gauging the level of characterization and treatment necessary before disposal of DSNF. However, it is not without uncertainty, because the results are dependent on the features used in the model. Thus, the addition of features such as credit for cladding for the commercial spent nuclear fuel, which is currently not included in OCRWM's analysis but is planned for the future, could affect the ranking of DSNF performance after disposal. For example, the commercial spent fuel would be greatly improved by credit for cladding, while DSNF performance would show only marginal improvement. Therefore, it may be in the DOE/EM's best interest not only to analyze DSNF under disposal conditions currently envisioned by OCRWM but also to focus on significant parameters, such as the effect of cladding, that are revealed as important through sensitivity analysis.