3.2.1 Commercial Spent Nuclear Fuel: Assigning the Right Waste Package

The characteristics of spent nuclear fuel assemblies (i.e., size, thermal output, and reactivity) will be used to select the appropriate waste package design. The size of assemblies is used to determine the size and configuration of the fuel within the waste package, and to perform structural analyses to evaluate the integrity of the waste package during normal handling and event sequences. The thermal output and reactivity of the fuel is used to determine which waste package can accommodate each given fuel assembly.

3.2.1.1 Physical Characteristics of Commercial Spent Nuclear Fuel

The physical characteristics of commercial spent nuclear fuel include length, cross section, weight, and cladding. Table 3-4 summarizes boiling water reactor assembly dimensions and weights by related groups; Table 3-5 provides similar information for pressurized water reactor assemblies (CRWMS M&O 1999a, Section B.1.1). The information provided in Tables 3-4 and 3-5 represents the full inventory of approximately 292,000 assemblies. Operating and shut down reactors are shown in Figure 1-2 of Section 1.

Based on physical dimensions, approximately 95 percent of the fuel assemblies can be emplaced using two waste package designs (the 21-PWR Absorber Plate and the 44-BWR). The 12-PWR Long is designed to accommodate assemblies from Combustion Engineering and the South Texas Project, which are longer than the others. Because the 12-PWR Long holds fewer assemblies, it can also be used to manage thermal load and criticality concerns about fuel with higher thermal output and reactivity.

3.2.1.2 Thermal Output

Commercial spent nuclear fuel arriving at the repository is expected to have a wide range of thermal outputs. The waste package has been designed to ensure that this anticipated range can be accommodated in a way that supports the range of thermal operating modes being considered for the potential repository (see Section 2).

The key factors used to determine the thermal output of spent nuclear fuel are its age (i.e., number of years out of the reactor), burnup (measured in gigawatt-days per metric ton of uranium), and initial enrichment of fissile material (i.e., uranium-235 or plutonium). To cover anticipated thermal outputs, waste package designers considered the average characteristics provided in Table 3-6. Maximum characteristics were also evaluated to ensure that these fuel assemblies could be placed in the waste package.

Determining which waste package design can accommodate a particular spent nuclear fuel assembly thermally will require calculating the assembly's thermal output at the time it is emplaced in the repository. The appropriate waste package is chosen to ensure that the maximum thermal output limit is not violated. In the higher-temperature operating mode, this limit has been set at 11.8 kW.

Of the five commercial waste package designs, the 21-PWR Absorber Plate is the most limiting in thermal output. The characteristics of spent nuclear fuel assemblies loaded into this waste package type will be carefully chosen to ensure that the thermal output limit is not violated.

The thermal output of the waste package can be reduced, if necessary, to accommodate either a range of thermal operating modes or potential changes in the characteristics of the waste stream. Reduction can be achieved by using one or more of the following waste package loading strategies: (1) fuel blending (i.e., combining low heat output fuel and high heat output fuel within a single waste package); (2) de-rating (i.e., loading fewer assemblies than the waste package is designed to hold); or (3) increasing the use of the 12-PWR Long waste package (i.e., placing high heat output fuel in smaller waste packages). For more information on variables that can be modified to accommodate a range of operating modes, see Section 2.1.

3.2.1.3 Criticality Control

Waste package designs will be evaluated to ensure that subcritical limits can be met, as well as the thermal limits described in the previous section. As Section 3.5.2 describes in detail, loading curves will be developed from commercial spent nuclear fuel parameters to determine the method of loading waste packages. This will ensure that the reactivity of the fuel being loaded is below the level at which criticality could occur.

3.2.2 Commercial Spent Nuclear Fuel Waste Package Designs

All the waste package designs for commercial spent nuclear fuel have similar components that perform multiple functions. Figure 3-2 illustrates a representative waste package design. The general features of waste package design were described in Section 3.1; the internal components of commercial spent nuclear fuel waste package design are described in this section, along with their functions.

3.2.2.1 Internal Basket Design

Baskets are composed of interlocking plates, fuel tubes, thermal shunts, and structural guides. These elements will displace any water that might be present inside a breached waste package, helping to prevent criticality. All four elements and their functions are described below.

Interlocking Plates—The interlocking plates set the pattern for how the fuel assemblies will be arranged inside the waste package. The basket for each design is customized to meet requirements for the size, type, and number of fuel assemblies it can hold, as well as the specific waste form being packaged. The material composition and thickness of the interlocking plates are tailored to provide enough structural strength to maintain fuel geometry during normal handling and design basis events, and to prevent criticality. However, this extra durability is not considered in the structural analyses (CRWMS M&O 2000au). The interlocking plates are made of either Neutronit A 978 (a stainless steel and boron alloy) or SA 516 Grade 70 carbon steel and range between 5 and 10 mm (0.2 and 0.4 in.) in thickness.

The neutron absorber materials that prevent criticality can be placed directly into the plates using Neutronit A 978 or can take the form of separate control rods. Plates that include neutron absorber material will vary in thickness because of the number of plates in the design. For example, the 44-BWR waste package has more plates than the 21-PWR waste package; therefore, the plates in the 44-BWR can be thinner, but the entire waste package will still have about the same mass of neutron absorber material as the 21-PWR.

The corrosion rate for Neutronit A 978 plates is slow, and the plates tend to corrode by pitting. Because of this, the plates remain in place between the fuel assemblies even as they corrode. To avoid processes that could accelerate stress corrosion cracking, there will be no bends or structural welds on the Neutronit A 978 plates.

Fuel Tubes—The fuel tubes are long, square containers that line the insides of the cavities created by the interlocking plates. They support the internal structure created by the interlocking plates while holding the fuel assemblies in place. The fuel tubes provide structural strength for the internal basket during event sequences they also help conduct heat away from the cladding. The fuel tubes for each waste package design are made of SA 516 Grade 70 carbon steel that is 5 mm (0.2 in.) thick (CRWMS M&O 2000au, Section 2.4.1.2).

Thermal Shunts—All the waste package designs for commercial spent nuclear fuel except the 24-BWR require thermal shunts. These shunts, which are made of 5-mm (0.2-in.) thick SB 209 6061 T4 (an aluminum alloy), are placed alongside the interlocking plates (CRWMS M&O 2000au, Section 2.4.1.3). The shunts are added to help transfer heat from the waste form to the walls of the waste package. Adding thermal shunts is a simple and effective method to improve heat conduction between the center of the waste package and the outer edge of the internal basket, providing a reliable means of keeping the temperature of the cladding within design limits. Limiting cladding temperatures helps protect the waste form by minimizing damage to the fuel cladding (CRWMS M&O 2000au, Section 2.4.1.3).

Structural Guides—The structural guides for each waste package are made of 10-mm (0.4-in.) thick SA 516 Grade 70 carbon steel and are placed inside the inner layer of the waste package to hold the basket structure in place. They help maintain fuel geometry, which can prevent criticality during event sequences. The structural guides also help conduct heat from the waste form to the walls of the waste package, where it is radiated to the surrounding drift walls (CRWMS M&O 2000au, Section 2.4.1.4).

3.2.2.2 Control Rods

Control rods similar to those used in reactors will be placed in waste packages that need additional long-term criticality control, such as those containing highly reactive fuel assemblies from pressurized water reactors. Control rods are made of boron carbide and have Zircaloy cladding. Because this is the same material used in most fuel rod cladding, it will have similar corrosion properties and longevity.

3.2.3 Preliminary Engineering Specifications for the Commercial Spent Nuclear Fuel Waste Package Designs

The preliminary engineering specifications for the waste package design include the waste form characteristics, the physical dimensions of the waste package, and material specifications. Tables 3-7 and 3-8 provide preliminary engineering specifications for the waste package designs for commercial spent nuclear fuel based on the physical dimensions, thermal output, and reactivity of the fuel. Table 3-9 shows the material specifications of the waste package components. These engineering specifications were developed to meet the performance specifications given in Table 3-1.

3.3 U.S. DEPARTMENT OF ENERGY SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND IMMOBILIZED PLUTONIUM

Ten types of canisters of DOE spent nuclear fuel and high-level radioactive waste may be received at the potential repository (CRWMS M&O 2000az, Sections 4.2 and 4.3):

1. Naval spent nuclear fuel canisters, short

2. Naval spent nuclear fuel canisters, long

3. DOE spent nuclear fuel canisters, short

4. DOE spent nuclear fuel canisters, long

5. Larger-diameter DOE spent nuclear fuel canisters, short

6. Larger-diameter DOE spent nuclear fuel canisters, long

7. Solidified high-level radioactive waste canisters, short

8. Solidified high-level radioactive waste canisters, long

9. Solidified high-level radioactive waste canisters containing immobilized plutonium cans, short

10. Multicanister overpacks containing spent nuclear fuel from the Hanford N Reactor.

The number of canisters of solidified high-level radioactive waste will greatly exceed the number of canisters of DOE spent nuclear fuel. Therefore, the DOE has developed an efficient arrangement for packing them together (CRWMS M&O 2000az, Section 4.2). This mixing of DOE spent nuclear fuel and high-level radioactive waste is called "codisposal." Codisposal also helps maintain criticality control for DOE spent nuclear fuel that contains highly enriched uranium. Naval spent nuclear fuel canisters, which are larger in diameter, will not be placed in codisposal waste packages; they will be placed one canister per waste package. Because the waste package designs being considered will contain both DOE spent nuclear fuel and high-level radioactive waste, the following section describes both waste forms, as well as the appropriate waste package designs.

3.3.1 U.S. Department of Energy Spent Nuclear Fuel

DOE spent nuclear fuel has a wide variety of physical, chemical, and nuclear characteristics and represents an inventory of approximately 2,500 MTHM; 2,333 MTHM of this is included in the waste allocation for disposal in the first repository (DOE 1999d, Section 8.1). The waste packages designed for DOE spent nuclear fuel will accept fuel irradiated at DOE facilities, naval spent nuclear fuel, and certain types of material irradiated at commercial nuclear reactors, including debris from the Three Mile Island-2 reactor and fuel from the Fort Saint Vrain reactor. All DOE waste canisters will be sealed before they are transported to the potential repository.

The largest single component of the DOE spent nuclear fuel inventory by weight is uranium metal fuel, at approximately 2,130 MTHM (DOE 1999d, Appendix C, Section 5.1, Table 1). Fuel from the N Reactor at Hanford, Washington, accounts for 2,100 MTHM of this inventory. During its 20-year life, the N Reactor produced nuclear isotopes for defense purposes. N Reactor fuel has an initial enrichment of less than 2 percent uranium-235. It will be placed in multicanister overpacks that will both store the waste onsite and transport it to the potential repository. The multicanister overpack is a stainless steel container that is slightly wider at the top than at the bottom (DOE 1999d, Appendix C, Section 5.1, Table 1). Although N Reactor fuel is the largest portion of the DOE spent nuclear fuel inventory by weight, it will be emplaced in the repository in only one percent of the waste packages (Table 3-3).

Approximately 184 MTHM of the DOE inventory is low-enriched uranium oxide, some of which is standard commercial spent nuclear fuel used for testing. Some is the fuel debris from the damaged reactor core at Three Mile Island-2, which is already stored in small canisters that can be placed inside a standard DOE canister. The DOE canister can then be inserted into a transportation cask and transported to the potential repository (DOE 1999d, Appendix C, Section 5.1, Table 1).

Approximately 125 MTHM of the DOE inventory includes uranium enriched initially to more than 20 percent uranium-235, uranium enriched initially to between 5 and 20 percent uranium-235, and thorium- and plutonium-based fuels (DOE 1999d, Appendix C, Section 5.1, Table 1).

Naval nuclear fuel is designed to operate in a high-temperature, high-pressure environment for many years. Naval fuel is highly enriched. To ensure it can withstand battle-shock loads, naval fuel is surrounded by large amounts of structural material made of Zircaloy. There are two canister designs for naval fuel; both use similar materials and mechanical arrangements. The DOE plans to emplace about 65 MTHM of naval spent nuclear fuel in the potential repository. This fuel will be contained within about 300 sealed canisters, which will be transferred directly from transport casks into waste packages (DOE 1999d, Appendix C, Section 5.1, Table 1).

3.3.1.1 Physical Characteristics

The canisters for DOE spent nuclear fuel will be standardized to efficiently utilize the waste package design (CRWMS M&O 2000aw). Table 3-10 gives the preliminary canister dimensions.

3.3.1.2 Thermal Output

DOE spent nuclear fuel has a low thermal output, with naval spent nuclear fuel being the hottest. The maximum thermal output of a naval spent nuclear fuel canister is 8.01 kW, which is well below the maximum limit of 11.8 kW (CRWMS M&O 2000ax, Section 2.5.4.2).

3.3.1.3 Criticality Control

The controlling factors in disposal criticality analyses of DOE spent nuclear fuel are fuel matrix, primary fissile isotope, geometry, and enrichment (DOE 1999d, Section 5.2). The 250 types of DOE spent nuclear fuel have been divided into groups based on these four factors to perform criticality analyses.

DOE non-naval spent nuclear fuel from each of the four groups will be analyzed in the appropriate configuration, and data important to preventing criticality—such as fissile loading, enrichment, initial configuration of the basket and spent nuclear fuel, and neutron absorber loading in the canister—will be identified. From these results, waste acceptance criteria will be developed for each group. The canister design will also control criticality by limiting the amount of fissile material in each waste package. If required, neutron absorbers would be added into a canister for further criticality control.

A separate analysis will be performed for naval spent nuclear fuel to demonstrate that criticality would be precluded for all credible event sequence conditions during handling at the repository (Mowbray 1999).

3.3.2 High-Level Radioactive Waste and Immobilized Plutonium

About 22,000 canisters of high-level radioactive waste will be generated by 2035 (DOE 1997b, Section 1.5.4). Approximately 1.5 percent will come from reprocessed commercial nuclear fuel; the rest will come from treatment of materials from the defense nuclear program. The estimated number of high-level radioactive waste canisters to be emplaced in the first repository is approximately 8,300, based on the total inventory limit in the NWPA.

Liquid high-level radioactive waste will undergo a process at its current site that yields a solid leach-resistant material, typically a borosilicate glass. While still liquid, the glass is poured into stainless steel canisters. After the glass cools and solidifies, the canisters are sealed. The potential repository would accept solid high-level radioactive waste generated from activities at DOE's Savannah River, South Carolina, and Hanford, Washington, sites, as well as from the Idaho National Environmental and Engineering Laboratory. The waste will arrive in presealed canisters. The potential repository would also receive, subject to the execution of a disposal contract between the DOE and the state of New York, commercial high-level radioactive waste from the West Valley Demonstration Project in New York.

Up to 17 metric tons of surplus plutonium that is excess to national defense needs will be immobilized within ceramic discs that will have neutron absorber material evenly distributed throughout their matrix (65 FR 1608). The ceramic will resist the leaching of plutonium. Section 3.2 describes the additional 33 metric tons of surplus plutonium that will be converted into mixed-oxide fuel.

3.3.2.1 Physical Characteristics

The canisters containing high-level radioactive waste will be standardized to accommodate the waste package design and to reduce manufacturing costs. Table 3-10 gives the canister dimensions.

3.3.2.2 Thermal Output

DOE high-level radioactive waste has a low thermal output. The total heat generation rate will not exceed 1.5 kW per 3-m (9.8-ft) canister or 1.97 kW per 4.5-m (15-ft) canister (DOE 1999c, Section 4.2.3.1). The maximum thermal output of the hottest waste package, the 5-DHLW/DOE short, is 9.16 kW—well below the maximum limit of 11.8 kW (CRWMS M&O 2000aw, Section 2.5.4.2).

3.3.2.3 Criticality Control

With the exception of high-level radioactive waste canisters containing immobilized plutonium, evaluations have indicated that DOE high-level radioactive waste will not contain enough fissile material to pose a criticality risk.

A principal criticality control measure for the immobilized plutonium is the incorporation of neutron absorbing materials (i.e., gadolinium and hafnium) into the waste form. These materials are an effective criticality control measure for both the preclosure and postclosure phases. The planned loading strategy for immobilized plutonium is to transfer it into a codisposal waste package containing five high-level radioactive waste canisters but no DOE spent nuclear fuel canister in the center. Detailed criticality analyses (CRWMS M&O 2000ba) have shown that preclosure and postclosure criticality for a waste package that contains five plutonium-loaded canisters, but that does not have a center DOE spent nuclear fuel canister, is below the subcritical limit for criticality to occur. See Section 3.5.2.4 for a brief discussion of the criticality potential of the immobilized plutonium waste form.

3.3.3 U.S. Department of Energy Waste Package Designs

Three waste package design configurations have been developed for codisposal of DOE non-naval spent nuclear fuel and high-level radioactive waste. In two designs that differ only in length, the typical arrangement places a DOE spent nuclear fuel canister in the center of a ring of five high-level radioactive waste canisters. The exception occurs for high-level radioactive waste canisters containing immobilized plutonium, which will be packaged without a DOE spent nuclear fuel canister in the center. Structural guides will provide support to ensure that the DOE spent nuclear fuel canister is not damaged by an impact to the waste package. These guides also facilitate waste package loading. Such support is not needed for high-level radioactive waste glass, which maintains its own structural integrity (CRWMS M&O 2000aw).

The third waste package design will accept multicanister overpacks, which will have a diameter larger than those of DOE spent nuclear fuel or high-level radioactive waste canisters. To prevent criticality, no more than two multicanister overpacks will be put into a waste package. To improve packaging efficiency without compromising criticality prevention, two long high-level radioactive waste canisters will be codisposed with the two multicanister overpacks. A DOE study determined that codisposing two multicanister overpacks with two long DOE high-level radioactive waste canisters would be the most efficient arrangement (CRWMS M&O 2000aw).

Naval spent nuclear fuel will arrive at the potential repository in canisters suitable for long-term disposal. The canisters will fit one to a waste package. Because the naval fuel will arrive in canisters of two sizes (one short and one long), the DOE has devised two waste package designs for it. The larger of these two types will be the heaviest and longest of all the waste packages. No additional features would be necessary for structural support, heat transfer, and criticality control, since these are provided by the naval spent nuclear fuel or the canister (CRWMS M&O 2000ax).

3.3.4 Preliminary Engineering Specifications

The preliminary engineering specifications for the waste package include the waste form, the physical dimensions of the waste package, and material specifications. Tables 3-10 and 3-11 present preliminary engineering specifications for waste package designs for DOE spent nuclear fuel and high-level radioactive waste, which are based on the physical dimensions, thermal output, and criticality potential of the fuel. Table 3-9 shows the material specifications of the waste package components. These engineering specifications were developed to meet the performance specifications given in Table 3-1.

3.4 SELECTING MATERIALS AND FABRICATING WASTE PACKAGES

The selection of materials from which reliable waste packages could be fabricated followed a multistep analysis and design process. It began by analyzing the critical functions of a particular waste package and its various components. In selecting a material for a component, the designers considered both the material's availability and the critical functions the component would serve as part of the waste package. They identified eight major components and eight performance criteria for selecting materials to fabricate them (CRWMS M&O 1997c, Section 3). The eight major components are:

• Structural shell

• Corrosion-resistant barrier

• Fill gas

• Interlocking plates for commercial designs

• Fuel tubes for commercial designs

• Structural guides for commercial designs

• Guide tube for codisposal designs

• Thermal shunts for commercial designs.

Not every waste package design requires all of these components; it varies according to the waste form each will hold. However, all eight of these components cover the major requirements of all ten waste package designs.

The eight criteria that contribute to performance are:

• Mechanical performance (strength)

• Chemical performance (resistance to corrosion and microbial attack)

• Predictability of performance (understanding the behavior of materials)

• Compatibility with materials of the waste package and waste form

• Ease of fabrication using the material

• Previous experience (proven performance record)

• Thermal performance (heat distribution characteristics)

• Neutronic performance (criticality and shielding).

Reasonableness of cost was considered as a discriminator.

3.4.1 Material Selection

The first step in selecting the waste package materials was identifying the functional requirements for each component. Next, the characteristics of materials that would help meet the requirements were selected. Candidate materials were chosen from commonly available materials (or, in the case of fill gas, from common gases). The materials were then analyzed in terms of how they would perform their intended functions. Once the candidate materials and alternates were selected, they were tested; the results of these tests are summarized in Section 4.2.4.

Table 3-9 lists the component materials selected after testing. The following sections explain the material selection process in more detail.

3.4.1.1 Waste Package Materials: Contributing to Containment

Corrosion-Resistant Materials—Corrosion performance has been determined to be the most important criterion for a long waste package lifetime. Essential performance qualities therefore include a material's resistance to general and localized corrosion, stress corrosion cracking, and hydrogen-assisted cracking and embrittlement. The effects of long-term thermal aging are also important. To address recommendations provided by the Nuclear Waste Technical Review Board, the DOE has initiated studies to gain a better understanding of the processes involved in predicting the rate of waste package material corrosion over the 10,000-year regulatory period.

Combinations and arrangements of materials as containment barriers were carefully considered from several perspectives. In the process, analysts considered such criteria as (1) material compatibility (e.g., galvanic/crevice corrosion effects); (2) the material's ability to contribute to defense in depth (e.g., because it has a different failure mode from other barriers); (3) the material's ease of fabrication; and (4) the potential impact of thin, corrosion-resistant materials used as containment barriers on a repository's essential operations, such as waste package loading, handling, and emplacement.

The major objectives centered on understanding the temperature and humidity conditions that would exist at different times for a range of thermal operating modes in a particular unsaturated zone, then designing the waste packages accordingly. Since the properties of any material selected for a corrosion barrier would inevitably be influenced by the temperature and humidity conditions in a repository of a particular design at a particular site, selecting the right corrosion-resistant material became one of the most important priorities.

After assessing potential materials available for waste package corrosion barriers, analysts selected nickel- and titanium-based alloys as the most promising candidate materials for corrosion resistance in an oxidizing environment such as Yucca Mountain. Using a corrosion-resistant material as the outer barrier of the waste package will significantly lower the risk of waste package failure from corrosion. Alloy 22 was selected as the preferred material for the outer barrier because it has excellent resistance to corrosion in the environment expected at Yucca Mountain; it is easier to weld than titanium; and it has a better thermal expansion coefficient match to Stainless Steel Type 316NG than titanium. A structurally strong material (stainless steel) was chosen for the inner layer of the waste package (CRWMS M&O 2000av, Section 7.6).

Alloy 22 also offers benefits in the areas of program and operating flexibility. It is extremely corrosion-resistant under conditions of high temperature and low humidity, such as those that would prevail for hundreds to thousands of years in a repository designed to allow a relatively high thermal output from the waste packages. At low temperatures, Alloy 22 is extremely corrosion-resistant in either low or high humidity. Thus, the selection of Alloy 22 supports the flexibility to operate the potential repository over a range of thermal modes (CRWMS M&O 2000av). Uncertainty about the waste package corrosion rate may be reduced by avoiding the conservatively defined window of corrosion susceptibility for Alloy 22, which can be accomplished by keeping waste package temperatures below 85°C (185°F) or maintaining the in-drift relative humidity below 50 percent (Dunn et al. 1999, p. xvi; CRWMS M&O 2000n, Section 3.1.3.1). Table 3-12 presents the chemical composition of Alloy 22.

Structural Materials—The major functional requirement of the structural material for the inner layer of the waste package is to support the corrosion-resistant outer material. The DOE chose Stainless Steel Type 316NG for the structural layer (CRWMS M&O 2000av, Section 5.2). This material provides the required strength; has a better compatibility with Alloy 22 than carbon steel; and provides an economical solution to functional requirements. Table 3-13 presents the chemical composition of Stainless Steel Type 316NG.

3.4.1.2 Waste Package Materials: Internal Components

The designs for commercial spent nuclear fuel and DOE codisposal waste packages include internal components (i.e., structural guides, interlocking plates, fuel tubes, and thermal shunts) that must be able to sustain the mechanical loads created by handling, emplacement, and, if necessary, retrieval. Thus, mechanical performance was a major selection criterion. Thermal performance was also an important selection criterion because these components provide an additional path for conducting heat from the waste form to the walls of the waste package. The fuel tubes contact both the waste form and the basket plates. If the material selected for the tubes causes the waste form to degrade, release rates could be increased; if it causes the plates to degrade, criticality control could be compromised. Therefore, compatibility with other materials was an important criterion. The waste package design does not rely on these components for postclosure performance, so corrosion-resistant materials are not needed. Two grades of carbon steel (SA 516 Grades 55 and 70) were found to be the best choices for these internal components, based on the criteria; the designers chose to use Grade 70 (CRWMS M&O 2000bd, Section 4).

Neutron Absorber Interlocking Plates—The most important function of the neutron absorber is to reduce the potential for criticality. The neutron absorber material is typically an additive to a carrier material (e.g., stainless steel alloyed with a boron compound). The neutron absorber is used in the interlocking plates in the internal basket. Corrosion behavior is important in keeping the neutron absorber material in place and effective long after emplacement, so chemical performance in a variety of environments was an important selection criterion. Mechanical performance was an evaluation factor because the interlocking plates must be able to sustain the mechanical loads created by handling, emplacement, and, if necessary, retrieval. Compatibility with other materials was considered, since the plates must not cause the waste form to degrade. The plates also provide an important path for conducting heat from the waste form to the walls of the waste package, so thermal performance was considered. The material of choice was Neutronit A 978. Its selection was based on its corrosion performance compared to the other candidate materials, as well as its available boron concentration. The composition of Neutronit is similar to SA 240 Stainless Steel Type 316 but with 1.6 percent boron added (CRWMS M&O 2000be, Section 3.1.3).

Thermal Shunts—The thermal shunts provide another important path for conducting heat from the waste form (in this case, commercial spent nuclear fuel) to the walls of the waste package. The thermal conductivity of the material is very important. The thermal shunts would be in contact with the waste form, so compatibility with spent nuclear fuel was an important evaluation criterion. The thermal shunts are only needed during the early period of repository performance, when the decay heat from spent nuclear fuel would be relatively high. The material selected does not need a high degree of corrosion resistance. The thermal shunts must have enough structural strength to withstand handling, emplacement, and possible retrieval operations. However, these service loads are not very large, so mechanical performance was not selected as an evaluation criterion. Aluminum alloys 6061 and 6063 were selected over copper because of concerns that, should a waste package be breached and water enter, copper may react with the chloride ions in the water. This could result in accelerated degradation of the Zircaloy cladding on the spent nuclear fuel, which would eventually release radionuclides from the waste (CRWMS M&O 2000be, Section 3.2.3).

3.4.1.3 Fill Gas

The fill gas can be a significant conductor of heat from the waste form to the internal basket, so thermal performance was deemed one of the most important criteria in choosing a gas. The fill gas should not degrade other components of the waste package, so compatibility with other materials was another important criterion. Helium is routinely used as the fill gas for fuel rods, which indicates that helium would have an excellent compatibility with spent nuclear fuel. Based on a review of data on thermal conductivity and the fact that helium is chemically inert, it was chosen over other candidate gases, such as nitrogen, argon, and krypton (CRWMS M&O 2000be, Sections 3.3.1 through 3.3.3).

3.4.2 Waste Package Fabrication Process

This section describes the fabrication process for the waste package, which is shown schematically in Figures 3-7 and 3-8. The fabrication process was based on both the design criteria and the physical characteristics of the selected materials.

The waste package will be fabricated, welded, and inspected in accordance with those portions of the ASME Boiler and Pressure Vessel Code, Section III, Division I, Subsection NB (Class 1 Components) (ASME 1995) that will ensure the waste package will perform in accordance with the design basis. Because the largest number of waste packages will be manufactured for commercial spent nuclear fuel, this type will serve to illustrate the basic fabrication process for all waste package types. The other types would be fabricated in a similar way, though some dimensions would vary.

The commercial fuel waste package uses Neutronit A 978 and carbon steel interlocking plates with carbon steel tubes (CRWMS M&O 2000bf, Section 8.1). All waste package fabrication would take place offsite, including the welding of the bottom lids. The top lids, although fabricated offsite, would be welded on in the Waste Handling Building after the waste package had been loaded.

3.4.2.1 Outer Cylinder Fabrication

Forming the outer cylinder of rolled and welded Alloy 22 plate requires two half-length cylinders (see Table 3-7 for completed waste package lengths) because of the limitations of most rolling fabricators. Initially, the plate would be approximately 5,080 mm (200 in.) long by 2,540 mm (100 in.) wide. The thickness would permit machining (for rounding) after welding. After being received by the fabricator, the plate would be inspected, laid out to establish the developed length, and thermally cut to size. The plate would then be rolled (CRWMS M&O 2000bf, Section 8.1.1).

The cylinder would then be adjusted to meet the required diameter and the inner circumference, taking into consideration the subsequent weld shrinkage of the longitudinal seam. The long seam weld preparations would be machined and prepared for welding. The cylinder would be braced to minimize the weld distortion and welded. The braces would be removed, and the weld seam would be prepared for nondestructive examination. One end of the cylinder would be prepared for circumferential seam welding. In parallel, a second cylinder would be prepared the same way (CRWMS M&O 2000bf, Section 8.1.1).

The two cylinders would then be joined and circumferentially welded, with subsequent nondestructive examination testing performed on the circumferential seam. The outer cylinder would be inspected to verify that the inside diameter is within tolerance. The inside of the cylinder would then be machined (CRWMS M&O 2000bf, Section 8.1.1).

3.4.2.2 Inner Cylinder Fabrication

To form the stainless steel inner cylinder, workers would start with two plates of Stainless Steel Type 316NG large enough to make half the inner cylinder length (see Table 3-7 for completed waste package lengths). The full cylinder length would require two plates, each approximately 4,980 mm (196 in.) wide by 2,540 mm (100 in.) long. Each plate would be cut or machined for size and longitudinal weld preparations. The plates would be roll-formed to make two half-cylinders and welded. Inspectors would use nondestructive techniques to examine both the cylinders and the welds. The weld preparations for the circumferential seam would be machined, then the cylinder would be assembled and circumferentially welded. Nondestructive examination inspections would again be performed on the circumferential weld, and the cylinder would then be machined (CRWMS M&O 2000bf, Section 8.1.2).

3.4.2.3 Lid Fabrication

The outer and flat closure lids would be fabricated from Alloy 22 plates approximately 1,900 mm (75 in.) wide and 3,800 mm (150 in.) long. The plates would then be cut to the correct diameter and the edges machine cleaned to prepare for the weld (CRWMS M&O 2000bf, Section 8.1.3).

The inner lids would be fabricated from a Stainless Steel Type 316NG plate approximately 1,800 mm (71 in.) wide and 3,600 mm (142 in.) long. The plate would be laid out for the cutting of two circles. The plates would then be thermally cut to the correct diameter, and the edges would be machine cleaned to prepare for the weld (CRWMS M&O 2000bf, Section 8.1.3).

3.4.2.4 Assembly of Support Ring

A support ring attached to the inner diameter of the outer cylinder is required to hold the inner cylinder in place. This ring would be made from 20-mm (0.8-in.) thick Alloy 22 plate. A piece would be cut 100 mm (4 in.) wide and rolled into a ring, then weld preparations would be machined. The ring would be fit to the inside of the outer cylinder near the bottom end, welded, and inspected (CRWMS M&O 2000bf, Section 8.1.4).

3.4.2.5 Assembly of Lid to Cylinder

Once the inner and outer cylinders had been completed, the inner and outer bottom lids would be welded in place. The structures would be set in a vertical position and the lids assembled to each cylinder. The welding could then be done in the flat position. After the welding had been completed, radiographic examination, ultrasonic examination, and liquid penetrant examination would be performed on the inner and outer lid seams. Radiographic and ultrasonic inspection would ensure that all detectable flaws, regardless of their orientation, were identified. Liquid penetrant inspection would ensure that surface indications were identified (CRWMS M&O 2000bf, Section 8.1.5).

3.4.2.6 Annealing of Outer Cylinder

Annealing is a process in which a material is subjected to a controlled heating and cooling cycle to affect material properties, for example, to relieve residual stress. Residual stresses are a common by-product of fabrication processes, such as forming, machining, and welding. Stress mitigation techniques, such as annealing, will be applied to the outer cylinder to minimize the potential for stress corrosion cracking. Since the closure lids will be welded shut after the waste has been loaded, different mitigation techniques may be employed. The parameters for the annealing operation are still being developed, but the DOE expects that the cylinder assembly will be heated in a furnace and then quenched by water (CRWMS M&O 2000bf, Section 8.1.7).

3.4.2.7 Assembly of Commercial Spent Nuclear Fuel Waste Package

The objective of machining is to produce a gap ranging from 0 to 4 mm (0 to 0.16 in.) between the inner and outer cylinders. Over time, the stainless steel inner cylinder will expand in response to the heat emitted by the radioactive decay of its contents. Even with the cylinders touching, there is enough allowance for the inner cylinder to heat up and expand without putting excessive stress on the Alloy 22 outer cylinder. Static loads in the outer barrier shell will not produce tensile stresses above 10 percent of the yield strength of the outer barrier material (CRWMS M&O 2000au, Section 1.2.1.23).

After both the outer and inner reinforcement cylinders had been machined, they would be fitted together. The outer cylinder would be heated to about 370°C (700°F) to allow the inner cylinder to be lowered inside. The heat would then be removed and the cylinders allowed to cool (CRWMS M&O 2000bf, Section 8.1.8).

3.4.2.8 Basket and Internal Components

Once the inner and outer cylinders and the bottom lids had been assembled, the container would be ready for the addition of the internal components. The cylinder would be laid out to establish the location of the internal corner guide assemblies and the internal side guides. The corner guide assemblies and the side guides would be put in place and welded, using manual gas tungsten arc welding.

The bottom set of plates would be installed in an interlocking fashion, followed by three additional sets. The tubes would be inserted and the tube tops stitch-welded together, if required.

The waste package would then be cleaned, wrapped, and protected for shipment to the repository surface facility and storage until it was ready to be loaded and sealed (CRWMS M&O 2000bf, Section 8.1.10). The final fabrication step, performed at the repository surface facility, would be the annealing of the closure weld area to mitigate the stresses that may have been induced during the welding of the outer closure lid. For more details on the process of loading the waste package, refer to Section 2.2.4.

3.5 WASTE PACKAGE DESIGN EVALUATIONS

The waste package must satisfy defined performance specifications to protect the public and workers and to meet the performance objectives of a repository. An example of a performance specification is the ability of a waste package to withstand a tipover event without breaching. Performance specifications are discussed in the following sections, where they are categorized by relevant engineering discipline (i.e., thermal, criticality, structural, and shielding). Detailed discussions of performance specifications are available in System Description Documents (e.g., CRWMS M&O 2000au).

Some of the performance specifications and supporting evaluations depend on temperature. In these cases, the evaluation is based on the higher-temperature operating mode. Further evaluations of lower-temperature operating modes are part of ongoing engineering studies.

To show that waste packages can be successfully developed for the various waste forms expected to be received at the repository, a sensitivity analysis was performed to determine which waste package designs best represent the widest array of design configurations and waste forms (CRWMS M&O 2000az). This selection was based on the number of waste package types expected to be needed for the repository—for instance, the two commercial spent nuclear fuel waste package designs chosen for analysis represent over 95 percent of the total required—and the relationship of a waste package type to a limiting performance specification (e.g., a heavier waste package would be more susceptible to a drop event). Detailed design work was performed for four waste package types:

• 21-PWR Absorber Plate
• 44-BWR
• 5-DHLW/DOE SNF Short
• Naval SNF Long.

3.5.1 Thermal Evaluations Performed on the Waste Package Design

Thermal analyses have been performed to demonstrate that waste form temperatures will not exceed levels established to maintain waste form integrity. The thermal specification for commercial spent nuclear fuel ensures that the cladding temperature will not compromise the integrity of the cladding, a barrier to radionuclide release. A specification for DOE high-level radioactive waste ensures that the glass does not reach a transition temperature that would cause significant changes in its phase structure or composition. Such an alteration would increase the solubility of the glass and reduce the time required for movement of the radionuclides embedded inside.

With respect to these thermal functions, two performance specifications were selected for evaluation: (1) Zircaloy commercial spent nuclear fuel cladding must be maintained below 350°C (660°F) under normal conditions (CRWMS M&O 2000au, Section 1.2.1.6), and (2) the temperature of DOE high-level radioactive waste must be maintained below 400°C (750°F) under normal conditions (CRWMS M&O 2000aw, Section 1.2.1.6).

3.5.1.1 Spent Nuclear Fuel Cladding Temperature

To calculate the cladding temperature for commercial spent nuclear fuel, time-dependent heat generation rates of the waste packages were adjusted to ensure that the average heat generation rate of an emplacement drift segment was the same as that for the repository as a whole, as discussed in Section 2.3.1. The calculation used a representative section of a drift containing one 21-PWR Absorber Plate, one 44-BWR, and one 5-DHLW/DOE SNF waste package, arranged as shown in Figure 3-3. The 21-PWR Absorber Plate serves as the design basis waste package, with a maximum heat generation rate of 11.8 kW. The second waste package, the 44-BWR, is based on an average heat generation rate of 7.0 kW. The 5-DHLW/DOE SNF waste package serves as a balancing package in which the heat generation rate is varied to ensure the average in the drift segment is the same as that in the repository as a whole. The time-dependent waste package surface temperatures calculated were used to perform a two-dimensional analysis of the internal components of the waste package. This calculation gives the peak cladding temperature for the design basis waste package (CRWMS M&O 2000au). The peak cladding temperature for commercial spent nuclear fuel was calculated to be 282°C (542°F), with the peak occurring 35 years after emplacement. The waste package spacing for the calculation was modeled as 0.1 m (0.3 ft), with a 25-year ventilation period (CRWMS M&O 2000au, Section 2.5.1.6). Active ventilation for 25 years would provide heat removal, limiting the heat-up of the waste package during preclosure.

Peak cladding temperature is not an issue for DOE spent nuclear fuel because no credit is taken for the fuel cladding in performance assessment. A canister of naval spent nuclear fuel will have lower heat generation than a waste package containing commercial spent nuclear fuel. Thermal analysis indicates that thermal limits associated with naval spent nuclear fuel will not be exceeded in the repository (CRWMS M&O 2000ax, Section 2.5.4.2).

3.5.1.2 High-Level Radioactive Waste Canister Temperatures

The vitrified high-level radioactive waste form could undergo devitrification at temperatures above 400°C (750°F). A maximum peak value of 214.5°C (418°F) occurs in the glass under normal conditions, which is well below the 400°C (750°F) threshold (CRWMS M&O 2000aw).

3.5.2 Criticality Evaluations Performed on Waste Package Designs

3.5.2.1 Preclosure Evaluations—Commercial Spent Nuclear Fuel

This section describes the performance specifications for criticality of commercial spent nuclear fuel. Further details on preclosure criticality analysis are available in the Preclosure Criticality Analysis Process Report (CRWMS M&O 1999i).

Each specific fuel assembly or material received by the repository has an associated fuel reactivity, or capability of contributing to a self-sustained nuclear fissioning process (criticality). The major properties of fuel reactivity in available commercial spent nuclear fuel are the initial enrichment and the burnup of the fuel when it was discharged from the reactor. Enrichment is the weight percentage of a fissile isotope compared to the total amount of uranium in the fuel assembly. Burnup is a measure of the amount of energy produced by the assembly while it was in operation. The higher the initial enrichment of the fuel, the more it can contribute to a critical system. The higher the burnup, for domestic commercial nuclear fuel, the less it can contribute to a critical system. A detailed description of the physics of criticality is provided in Section 4.3.3.2.1.

During the preclosure period, criticality is prevented in waste packages by ensuring, before loading, that the reactivity for each fuel assembly is below the level required for criticality. This is accomplished through the use of loading curves. Loading curves show the minimum allowed burnup as a function of initial enrichment and provide a simple go/no-go check, using the available information, on whether a fuel assembly can be loaded unaltered into a standard waste package without any concern for criticality. The presence of a moderator will be controlled as necessary to ensure subcriticality.

There are two main aspects to loading curve development: the determination of the value of reactivity considered limiting and the determination of the burnup/enrichment pairs that correspond to this value.

The loading curve methodology summarized below is described in detail in the Preclosure Criticality Analysis Process Report (CRWMS M&O 1999i, Section A.1). The effective neutron multiplication factor (k_eff)of a system is the measure of criticality; when k_eff is greater than or equal to 1.0, criticality occurs. If the effective neutron multiplication is less than 1.0, the system is considered subcritical. When designing a system (e.g., a waste package) to be subcritical, there must be a means for ensuring that a k_eff value conservatively represents the true k_eff of the system. This is accomplished by choosing a value below 1.0 at which nuclear criticality is assumed to occur. This value is known as the critical limit. The critical limit is determined by accounting for and bounding the bias in the calculational method used, and the uncertainty in the experiments used, to validate the method of calculation. For the calculations performed thus far, the critical limit for k_eff is 0.98. Because ensuring subcriticality is important to the health and safety of the repository workers, the calculations add an administrative safety margin of 5 percent for preclosure to provide additional conservatism. The loading curve is therefore based on a k_eff value of 0.93, referred to as the administrative limit. Any criticality calculation that uses the same tools on which the critical limit was based and shows a k_eff less than or equal to 0.93 provides high confidence that the system will not become critical (CRWMS M&O 2000bg, Section 6.2.1).

Once the criticality limit is established, many calculations are performed to determine, for a specific value of initial enrichment, what burnup results in a k_eff that equals the administrative limit (Figure 3-9). These calculations are performed for a point where the waste package is fully loaded with the same fuel assembly. Criticality calculations are performed in increments over the range of initial enrichments and burnups for each commercial fuel type to be received at the repository. The points where the different burnup/enrichment pairs equal the administrative limit can then be represented as a single curve of initial enrichment versus burnup, referred to as a loading curve.

Figure 3-9 shows an example of a loading curve. Loading curves will be developed for each waste package type containing commercial spent nuclear fuel. Any fuel that has a burnup/enrichment pair equal to or above the loading curve line can be loaded unaltered into that waste package type, and criticality will not occur in that waste package during preclosure. Some assemblies that fall below the curve will have to be loaded into the waste package with additional neutron absorber material so they will be above the line. Another option is to load the fuel assembly into a type of waste package that contains fewer fuel assemblies (e.g., the 12-PWR and 24-BWR), making the waste packages subcritical.

In the loading curve evaluation, margin to criticality is shown assuming the highly unlikely (beyond design basis), but essential to criticality, event of a waste package that is completely filled with water. A loading curve evaluation has been performed for the 21-PWR Absorber Plate model, as documented in the 21 PWR Waste Package Loading Curve Evaluation (CRWMS M&O 2000bg, Section 6). The 21-PWR waste package design was chosen because more information is available about it than about the other designs. The 21-PWR waste package is also considered to be the design with the highest inherent reactivity (CRWMS M&O 2000az, Table 8). The 21-PWR loading curve evaluation is representative of calculations that will be performed for the other waste package types.

The waste package will be designed so that during preclosure, nuclear criticality cannot be possible unless at least two unlikely, independent, and concurrent or sequential changes occur in the conditions essential to nuclear criticality safety.

3.5.2.2 Postclosure Criticality Evaluation: Commercial Spent Nuclear Fuel

The methodology that has been developed to evaluate potential for criticality and to ensure that significant impacts are prevented during the postclosure period is described in detail in Disposal Criticality Analysis Methodology Topical Report (YMP 2000c, Section 3). Section 4.3.3.2 summarizes this methodology and the analyses conducted using it.

Evaluations show that there is a very small probability that a potential critical configuration might result under the repository and fuel conditions described in this report, but that if a postclosure criticality were to occur, the effects would not compromise the ability of the repository to protect public health or meet design objectives or regulatory limits. The evaluations were performed assuming the most reactive combinations of such factors as the amount of fissile uranium and plutonium, the physical arrangement of the fuel, and the presence and amount of moderator in the waste package. The results indicate that all applicable limits can be met and public health protected using the current waste package designs.

3.5.2.3 Evaluations of Criticality Potential of U.S. Department of Energy Spent Nuclear Fuel

Analyses to demonstrate the viability of disposal have been performed or are in process for seven groups of DOE non-naval spent nuclear fuel. A separate analysis is being performed for naval spent nuclear fuel to demonstrate that criticality will be prevented for all credible event sequence conditions in the repository (Mowbray 1999).

In general, the amount of DOE spent nuclear fuel allowed per canister is a function of the physical size and weight limitations of the canister. The limitation on the amount of fissile material per canister provides criticality control. However, insoluble neutron absorbers (gadolinium compounds and alloys) are required for criticality control within the canister for some DOE spent nuclear fuel groups. To date, several items have been identified as important to criticality (DOE 1999d, Section 5.2). The performance and distribution of the neutron absorber material is important in preventing criticality.

The canister shell is also important in preventing criticality because it initially confines the fissile elements and neutron absorber material so they cannot be separated. The canister baskets developed for the representative fuel types are particularly important in cases where they provide the distribution mechanism for neutron absorber material.

The configurations evaluated for each fuel type include varying degrees of degradation, resulting in many different geometric configurations and fissile distributions. These degraded configurations also bound the other types of fuels in a group as long as the limits on fissile mass, linear fissile loading, and enrichment are not exceeded (DOE 1999d, Section 5.2). Further details on DOE spent nuclear fuel criticality analysis are available in completed viability evaluations (CRWMS M&O 2000bh; CRWMS M&O 2000bi; CRWMS M&O 1999j; CRWMS M&O 2000bj; CRWMS M&O 2000bk).

3.5.2.4 Evaluation of Criticality Potential of the Immobilized Plutonium Waste Package

The criticality potential of the immobilized plutonium ceramic discs is effectively controlled by neutron absorbers (i.e., gadolinium and hafnium), which are fabricated into the waste form itself. These materials are an effective criticality control measure. During postclosure, about 50 percent of the hafnium and about 1 percent of the gadolinium is needed to maintain subcriticality under all dilution situations. The DOE has exhaustively examined the physical and chemical processes that might be able to cause a separation, or removal of the neutron absorbers from the waste package entirely, and concluded they are insufficient to warrant further consideration. Detailed criticality analyses (CRWMS M&O 2000ba, p. xii) have shown that five plutonium-loaded canisters, without a center DOE spent nuclear fuel canister, can be placed in the same waste package.

3.5.3 Structural Evaluations Performed on Waste Package Designs

A performance specification for the waste package requires that it not breach during normal operations and event sequences. To address the term "breach" in a quantified manner, threshold limits for failure from the American Society of Mechanical Engineers code will be used. The waste package is designed to meet American Society of Mechanical Engineers code requirements. For event sequences, breach is assumed to have occurred when 90 percent of the ultimate tensile strength has been exceeded. To demonstrate the design adequacy of the waste package with respect to these structural functions, several performance specifications were selected, as documented in the Waste Package Design Sensitivity Report (CRWMS M&O 2000az, Section 7). These include:

• Internal pressurization
  
• Retrieval
  
• Rockfall
  
• Vertical drop
  
• Tipover
  
• Missile impact (i.e., from an accidental airborne projectile).

3.5.3.1 Internal Pressurization

Pressurization of a commercial spent nuclear fuel waste package could be caused by the rupture of all the fuel rods. Evaluations have been performed over uniform waste package temperatures, ranging from 20° to 600°C (68° to 1,100°F). The peak stresses at the junction of the waste package shell and lid were then compared to the ultimate tensile stress of the waste package materials.

The 21-PWR Absorber Plate waste package was evaluated because it would have the maximum internal pressure due to fuel rod failure. Detailed calculations show that the resulting stresses for all components of the waste package are less than 90 percent of the ultimate tensile strength of the materials; therefore, the waste package will not breach as a result of pressurization, and the criterion is met (CRWMS M&O 2000au, Section 2.5.2.10).

3.5.3.2 Retrieval

The waste package is being designed to allow retrieval up to 300 years after emplacement. The Naval SNF Long waste package is the heaviest waste package for retrieval. The ability of the waste package and pallet to be lifted together as a single unit was calculated. The results show that the maximum stress intensities from lifting among the emplacement pallet Alloy 22 and Stainless Steel Type 316L components are 96 MPa (14,000 psi) and 39 MPa (5,700 psi), respectively. These stress intensity magnitudes are less than one-third of the yield strength and one-fifth of the tensile strength for each of the corresponding materials.

Atmospheric corrosion penetration rates for Alloy 22 and Stainless Steel Type 316 are 0.0093 µm/yr and 0.025 µm/yr, respectively. The calculated cumulative decrease of thickness of the structural members of the emplacement pallet over 300 years of preclosure emplacement is 2.8 µm for Alloy 22 and 7.5 m for Stainless Steel Type 316. This negligible level of corrosion renders the waste package retrieval calculation unnecessary, since the consequential change of the results presented in this calculation would be insignificant (CRWMS M&O 2000ax, Section 2).

3.5.3.3 Rockfall

The 21-PWR Absorber Plate waste package design was selected for this evaluation because it would be the most vulnerable to rockfall. Its thinner walls in the outer barrier and its greater sensitivity to internal basket deformation cause greater vulnerability. It is also the most common waste package, and hence the most likely to actually suffer a rockfall impact. This calculation modeled a representative rockfall from the roof of the drift onto the unprotected waste package during preclosure. A height of 3.1 m (10 ft) and a rock size of 13 metric tons (14 tons), which were determined in the event sequence hazards analysis, were modeled (CRWMS M&O 2000au, Section 2.5.2.1). The calculated results, presented in Table 3-14, indicate the survivability of a 21-PWR Absorber Plate waste package in a rockfall event (CRWMS M&O 2000au, Section 2.5.2.1). For event sequences such as rockfall, breach has occurred analytically when 90 percent of the ultimate tensile strength has been exceeded.

3.5.3.4 Vertical Drop

The vertical drop evaluation was performed using the Naval SNF Long waste package because it is the heaviest design and has the highest internal load; therefore, it will have the highest stresses in the lids during a vertical drop. This calculation modeled a waste package being dropped from a distance of 2 m (6.6 ft).

Detailed calculations demonstrate the survivability of a Naval SNF Long waste package in a vertical drop. The results show that the maximum stresses among the waste package (made of Alloy 22, except for the lower trunnion collar sleeve) and Stainless Steel Type 316NG components are 433 MPa (63,000 psi) and 275 MPa (40,000 psi), respectively. Since these stress intensities are less than 90 percent of the ultimate tensile strength for each of the corresponding materials, the performance specifications are met. The maximum stress in the lower trunnion collar sleeve contacting the unyielding surface is 799 MPa (116,000 psi), which exceeds the tensile strength of Alloy 22. However, this does not constitute failure because the collar sleeve is designed to protect the waste package by acting as a crush zone at the point of impact (CRWMS M&O 2000ax, Section 2.5.2.3).

3.5.3.5 Tipover

A waste package might tip over because of a vertical drop or a seismic event. The 21-PWR Absorber Plate waste package was selected for the preclosure tipover evaluation because of its thinner walls in the outer barrier (made of Alloy 22) and its greater sensitivity to internal basket deformation. It is also the most common waste package. The tipover analysis was simulated in a detailed calculation. Table 3-15 shows the results. The stresses for all components that make up the waste package are less than 90 percent of the ultimate tensile strength of those materials (CRWMS M&O 2000au, Section 2.5.2.6).

3.5.3.6 Missile Impact

A potential internal missile event sequence could take the form of a valve stem being ejected from equipment operating at high pressures. The valve was estimated to have a mass of 0.5 kg (1.1 lb), a diameter of 1.0 cm (0.39 in.), and a velocity of 5.7 m/s (19 ft/s). The missile impact evaluation was performed for the 21-PWR, 44-BWR, 5-DHLW/DOE SNF, and Naval SNF waste package designs.

The calculated minimum velocity is significantly less that would be required to compromise the integrity of the waste package (CRWMS M&O 2000au, Section 2.5.2.8).

3.5.4 Shielding Evaluations Performed on the Waste Package Design

Shielding analyses evaluate the effects of ionizing radiation on personnel, equipment, and materials. The primary sources for waste package radiation are gamma rays and neutrons emitted from spent nuclear fuel and high-level radioactive waste. Loading, handling, and transporting of waste packages would be carried out remotely to keep personnel exposure as low as is reasonably achievable (e.g., having the human operators behind radiation shield walls, using remote manipulators, viewing operations with video cameras). Shielding analyses were performed for waste package designs to assess the effects of radiation on material and equipment. These analyses provide information used by both the subsurface and surface design programs to determine shielding requirements in the surface facility, on the waste package transporter, and in the subsurface facility.

Because they were designed to contain the waste forms for thousands of years, the waste packages must reduce radiation levels at their surfaces so that radiolytically enhanced corrosion under aqueous conditions is negligible. The shielding analyses determined radiation exposure rates on the surface of the waste package and evaluated whether radiolytically induced corrosion would be a contributing factor to the overall degradation of the waste package.

Shielding analyses were also performed on equipment to determine radiation exposure during the welding of the waste package closure lids. Various pieces of monitoring and control equipment, such as the welding heads and camera, would be close to radiation sources. The results of the shielding analyses will be used to quantify the shielding necessary for a piece of equipment to function properly at a given location for a required period of time.

In emergency situations, which could occur during the transport of waste packages from the surface facilities to the emplacement drifts or during the emplacement of waste packages in the drifts, personnel may have to enter areas near waste packages. Shielding analyses provide an evaluation of the radiation environment surrounding the waste packages so that worker safety can be ensured.

3.5.4.1 Source Term

Engineering calculations were performed to generate source terms, which are used to evaluate an upper limit for the surface dose rate of waste package. The source terms calculated for both pressurized water reactor and boiling water reactor spent nuclear fuel have the following characteristics: 5.5 percent (by weight) initial uranium-235, 75.0 GWd/MTU, and a 5-year decay time for the active fuel region; and 0.711 percent (by weight) initial uranium-235, 75.0 GWd/MTU burnup, and a 5-year decay time for the hardware regions of the assembly. The rationale for these assumptions is discussed in PWR Source Term Generation and Evaluation (CRWMS M&O 1999k) and BWR Source Term Generation and Evaluation (CRWMS M&O 1999l).

3.5.4.2 Results

The calculated maximum dose rate at the external surface of a 21-PWR Absorber Plate waste package is 1,130 rem/hr (+/-60 rem/hr) (CRWMS M&O 2000bl). The calculated maximum dose rate at the external surface of a 44-BWR waste package is 1,409 rem/hr (+/-32 rem/hr) (CRWMS M&O 2000bl, Section 6.2.3). Radiolytically enhanced corrosion is expected to be insignificant because the gamma dose on the surface of the waste package will not affect the corrosion properties of the waste package (see Sections 4.2.3.1.4 and 4.2.4.3.3).

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