Metallic Waste Forms

5.20

Metallic Waste Forms

W. L. Ebert Argonne National Laboratory, Argonne, IL, USA

Published by Elsevier Ltd.

5.20.1

Introduction

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5.20.2 5.20.2.1 5.20.2.2 5.20.2.3 5.20.2.4 5.20.3 5.20.4 5.20.4.1 5.20.4.2 5.20.5 5.20.5.1 5.20.5.2 5.20.5.3 5.20.6 5.20.6.1 5.20.6.2 5.20.6.3 5.20.7 5.20.7.1 5.20.7.2 5.20.7.3 5.20.8 5.20.8.1 5.20.8.2 5.20.8.3 5.20.8.4 5.20.8.5 5.20.8.6 5.20.8.7 5.20.9 References

Waste Streams Electrometallurgical Wastes Aqueous Processing Wastes for UREX + Flowsheets Hulls and Hardware Processing Additives Alloy Compositions Processing Methods Alloy Production Waste Conditioning Testing Objectives Matrix Degradation and Radionuclide Release Waste Form Consistency Through Process Control Waste Form Acceptance and Regulatory Requirements Modeling Matrix Corrosion and Radionuclide Release Mechanisms Performance in Disposal System Processing Control Test Methods Electrochemical Test Methods Corrosion Test Methods Service Condition Test Methods Tests with INL MWF Formulation and Phase Compositions Radionuclide Distribution Electrochemical Tests Corrosion Tests Corrosion Models Repository Model Metallic Waste Form Product Consistency Summary

509 509 510 512 512 513 514 514 515 515 515 516 517 517 518 520 520 521 522 523 524 525 525 526 528 529 531 533 534 535 536

Abbreviations DOE EBR-II EBS FCR&D FPEX GTCC HLW MWF

US Department of Energy Experimental breeder reactor Engineered barrier system Fuel Cycle Research and Development Fission product extraction Greater-than-Class C High-level radioactive waste Metallic waste form

NIST

US National Institute of Science and Technology NL(i ) Normalized mass loss (based on element i) RCRA Resource conservation and recovery act SEM Scanning electron microscopy TALSPEAK Trivalent actinide–lanthanide separations by phosphorous reagent extraction from aqueous complexes

505

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Metallic Waste Forms

TEM TRUEX UDS UREX VHT

Transmission electron microscopy Transuranic element extraction Undissolved solids Uranium extraction Vapor hydration test

5.20.1 Introduction This chapter provides an overview of current strategies and approaches for formulating, processing, testing, and developing performance models for the metallic waste forms that are being designed to immobilize high-level radioactive wastes. Some aspects have been completed and demonstrated, while others are planned approaches based on previous experience that have not yet been fully implemented. Issues that must be addressed when designing a waste form are conveniently grouped as performance, processing, and waste acceptance issues, although these are interrelated. Waste acceptance issues include regulatory requirements that the waste form and disposal system as a whole must meet, identification of waste form properties that can be measured to demonstrate that the waste form is acceptable, and specifications for performing those measurements. Performance issues include determining the capacity of the waste form to restrict the release of radionuclides to rates at which the engineered repository can meet regulatory requirements (this is the basis for waste form acceptability), and, importantly, the ability to predict the release rates over very long durations under the anticipated range of environmental conditions with a mathematical model. Processing issues include the ability to produce waste forms from the anticipated range of waste stream compositions and additives that have consistent, predictable, and acceptable chemical, radiological, and physical properties by maintaining a consistent phase assemblage and distribution of radionuclides. Although the primary role of a waste form is to retard the release of the radionuclides it contains to the surrounding environment, this should occur in a predictable manner that can be related to both long-term performance and waste form production controls. The link between production and performance will provide confidence to regulators that the performance of any waste form produced within the anticipated range of waste stream compositions is adequately represented in the performance assessments used to license the disposal facility. The radionuclide release rates and the degradation rate of the host waste form matrix are used as

source term models in performance assessment calculations to predict the capacity of the disposal system to meet regulatory limits, such as groundwater dose limits or contamination levels at or beyond the system boundary. Assessments must address very long time periods and both anticipated environmental conditions and possible extreme conditions, including highly unlikely scenarios. The unprecedented challenge of predicting material performance for waste management over a geologic time scale has led to highly conservative bounds being used in risk and safety assessments. For example, the assessment calculations for the proposed Yucca Mountain repository include igneous intrusion (a magma plume from volcanic activity) and eruption scenarios. The inclusion of catastrophic events that have extremely low probabilities of occurring in performance assessments may be necessary for identifying conditions under which a repository may fail to meet regulatory requirements. Although those conditions will likely be far outside the anticipated range of service conditions, the possible impacts on the performance of the waste form should be considered when developing testing programs. In the case of metallic waste forms, this may call for an understanding of the impacts of long-term heating on the phase composition and retention of radionuclides of the host phases. The term ‘phase composition’ is used to denote the assemblage of component phases in the waste form, including solid solution and intermetallic phases, the chemical composition of each phase, and the distribution of radionuclides. The matrix material used to immobilize radioactive waste must be compatible with both the waste stream and the disposal environment. Although the specific disposal site may not be identified at the time waste forms are being designed, general site characteristics may be known, for example, if the site will be located in clay, salt, tuff, granite, or other medium, if it will be in a reducing or oxidizing environment, in a hydrologically saturated or unsaturated horizon, etc. If not even the generic disposal conditions are known, then testing programs must consider waste form compatibility in all potential environments. In most cases, the general compatibility of a waste stream and potential host matrix can be anticipated from known materials characteristics, but it must be evaluated experimentally to determine details of the phase composition and microstructure. The radionuclide(s) may be incorporated into the host matrix or sequestered in separate phases that are encapsulated by the matrix. For example, radionuclide-bearing oxide phases may be encapsulated within a metal

Metallic Waste Forms

matrix. In that case, compatibility may simply mean that the waste form can be processed to encapsulate the radionuclide-bearing phases within the matrix. Some waste forms may simply serve as a diffusion barrier to the ingress of water or the diffusive release of the radionuclides, but degradation of metallic waste forms will likely be required prior to radionuclide release. Note that ‘degradation’ is used to refer to alteration that results in the release of radionuclides, whereas ‘corrosion’ is used to refer to alteration that does not necessarily result in release. Regardless of the mode, the radionuclide release rate from each waste package must be predictable over the service life of the disposal system as a source term in performance assessment calculations. This usually means that waste form products must have consistent composition, degradation behaviors, radionuclide inventories, and size (exposed surface area). Consistent does not mean identical in the present context. Rather, it means within the range deemed to ensure acceptable performance. The anticipated range of waste stream compositions, which may vary due to variations in the fuels being treated and efficiencies of separations processes, and the capacity to blend waste streams and adjust additives, will affect the composition range of the waste forms. The impact of the anticipated range of waste form compositions on radionuclide release must be determined and taken into account in the source term model and preliminary performance assessment calculations. The impact on the performance of the disposal system can then be used to establish the acceptable composition range. This may require reformulation of the waste form, such as lowering the waste loading or changing the additives, to accommodate the range of waste stream compositions. Some radioactive waste streams are better immobilized in a metallic matrix than in glass or other matrices because the waste materials are metallic, because the radionuclides in the waste can be more efficiently retained during waste form processing if they are in the metallic state, or because the radionuclides are more effectively retained in the waste form if they are in the metallic state. Technetium is an important example of a radionuclide that is best processed and immobilized as a metal. The predominant species in most aqueous waste streams is the pertechnetate ion TcO
4 , but Tc(VII) is neither readily processed due to volatility of species such as CsTcO4 nor strongly incorporated into the structure of a glass or other matrix. The most stable form in an oxidizing disposal environment, TcO2, is sparingly

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soluble in water, but sublimes at 900  C and Tc(IV) disproportionates to metallic Tc and gaseous Tc2O7 at about 1100  C.1 In contrast, metallic Tc can be processed at high temperatures under reducing conditions and alloyed in a metallic waste form without the loss of volatile species. Waste streams with high concentrations of Tc and other transition metals can also be immobilized in a metallic waste form at higher waste loadings than in a glass or other waste form matrix. Specific waste streams that are expected to be amenable to metallic waste forms are discussed in Section 5.20.2. In a metallic waste form, most radionuclides are dissolved or alloyed in a host metal with other waste constituents. Formulations of alloy waste forms must identify host metals and mixtures that can be processed at practicable temperatures and form durable phases to incorporate all radioactive components. Many metallic components in waste streams have very high-melting points and must be reactively dissolved into molten metals before they can be alloyed. Available binary and ternary phase diagrams provide insights into likely melting temperatures and phase compositions for processing complex mixtures of wastes and added metals. For example, significant amounts of Zircaloy scraps from cladding hulls in a waste stream can be dissolved into molten copper or iron below 1600  C, even though Zircaloy itself (Zircaloy-2 and Zircaloy-4) melts at about 1850  C. Other metals present in spent nuclear fuel and expected to be present in metallic waste steams either as pure metals or in alloys include Mo (2623  C), Ru (2334  C), Tc (2204  C), Rh (1963  C), and Pd (1555  C). The five-metal alloy formed from these components during irradiation of oxide fuels is expected to have a melting point near 2000  C (see Kaye et al.2). These high-melting wastes can be processed into a metal waste form at temperatures lower than their melting points by utilizing eutectic mixtures. The binary phase diagrams give insights into the solubilities of waste components in iron and identify intermetallic phases likely to form in the two-component systems. Elemental substitutions can also be predicted based on metallurgical data. For example, each component in the five-metal alloy is expected to be dissolved by molten iron at a temperature well below 1600  C, and so is the five-metal alloy. The ability to produce a waste form is an obvious initial requirement in waste form design. Formulation of metal alloy waste forms is discussed in Section 5.20.3. Contaminants will be present in the waste stream (e.g., oxides and sulfates) and absorbed by the molten

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Metallic Waste Forms

metal from the crucible and furnace atmosphere (trace O2, CO, etc.) during processing. Slags (and dross) typically form on the surfaces of molten metals from impurities that are not incorporated into the alloys. Silica, calcium, and other oxidizable materials are commonly added to molten metals in industrial processes to help sequester impurities in slag layers that can be removed from the surface of the molten metal and discarded. The addition of silica usually results in the formation of a glassy phase in the slag that encapsulates other phases of metal oxide, carbide, sulfate, etc., and may itself be quite durable. Whereas slags are removed and discarded as waste from metal and alloy products, they can serve to immobilize contaminants excluded from the alloy phases of a metal waste form. The use of slag phases as a component of metallic waste forms is being studied.3 Processing methods are discussed in Section 5.20.4. The durability of a metallic waste form will depend on the durabilities of the component phases and their interactions with other phases in the waste form and with engineering materials, primarily through galvanic couples. The phases that are present and potential couples will affect the capacity of the waste form to retain radionuclides, and long-term performance models must take both into account. Clearly, a constant phase assemblage and a constant distribution of radionuclides between those phases will simplify performance models. Although a mechanistic understanding of waste form corrosion and radionuclide release adds confidence to predictions of long-term waste form performance, a fully mechanistic model for waste form degradation may be impractical due to budget limitations or unachievable due to an incomplete understanding. Many aspects can be adequately captured with empirical models, such as pH and temperatures effects, but confidence in empirical models based solely on experimental observations is usually limited to the duration of the longest experiments. A mechanistic understanding of processes that establish the time-dependent aspects of corrosion should be the top priority of a testing program. This could include understanding how passivating layers form, mechanisms for their breakdown, and the kinetics of their healing. The term ‘passivating’ is used in the very general sense of slowing the corrosion, regardless of the controlling factor. Key testing objectives are summarized in Section 5.20.4, including performance modeling, process control, and waste acceptance. Modeling is discussed in Section 5.20.5 and testing to develop a mechanistic degradation model is discussed in Section 5.20.6.

Consistency is another important requirement for metallic waste forms related to the need to predict waste form performance (particularly radionuclide release) over a very long disposal system service life. The performance of a metallic waste form in a given disposal system will, in general, depend on the phase assemblage that comprises the waste form and the radionuclide inventory. The preferred approach for producing consistent waste forms is control of the feed (waste streams and additives) and the processing conditions. Controlling the feed controls the gross composition of the waste form, and controlling the composition and processing condition controls assemblage of phases that form, and presumably, maintains a consistent distribution of radionuclides among the component phases. A key objective of waste form design and testing is to determine the relationships between (1) the process conditions and the phase composition and (2) the phase composition and waste form performance. These relationships link the radionuclide release rates calculated using the waste form degradation model with the controlled production of the waste form. These two sets of relationships should take precedence in waste form development and their importance should not be underestimated. Whereas both high waste loadings and high chemical durabilities are key objectives of waste form design, the acceptance by regulators for disposal will likely require evidence that the waste forms will perform to the level modeled in performance assessment calculations. Because it is not practical to subject samples of each waste form product to the suit of tests needed to characterize and model performance in the disposal system, the performance of each waste form product that is made can instead be related to the performance measured for the representative material based on the phase composition and microstructure. That relationship must be established through the waste form composition and processing conditions as a part of waste form development. Therefore, it is important to determine and experimentally demonstrate the ranges of waste form compositions and processing conditions that are represented by the model. That range of compositions is the basis for waste form acceptance. For borosilicate waste glasses, the relationship between performance and process control is demonstrated using the Product Consistency Test (ASTM standard test method C 1285), which is a partial dissolution test conducted under specified test conditions.4 An extensive database was generated to show the relationship between glass composition and test response,5 and the test response

Metallic Waste Forms

can be related to the modeled glass degradation and radionuclide release behavior.6 An analogous test method and database must be developed to establish the link between production control and performance for metallic waste forms. A metallic waste form was designed and developed to immobilize metallic wastes generated during the electrometallurgical treatment of spent sodiumbonded nuclear fuel from the Experimental Breeder Reactor-II (EBR-II).7
9 That waste stream consists of Type 316 stainless steel anode baskets, steel cladding hulls, and residual metals from the fuel that were not oxidized during the electrorefining procedure (primarily Zr). The primary radionuclide in the metallic waste from the fuel is 99Tc, but the waste stream also includes a small amount of entrained salt that contains actinides (primarily U and Pu). The work done to formulate this waste form, measure the corrosion behavior and radionuclide release, develop a performance model, and develop an approach for tracking waste form consistency provides a valuable example of the general approach for developing a metallic waste form and generating the database to support acceptance for disposal. To distinguish the specific metallic waste form alloy developed for electrometallurgically treated EBR-II fuel from other metallic waste forms, it will be referred to herein as the EBR-II metallic waste form (MWF). The work done to develop, test, and model the EBR-II MWF is summarized in Section 5.20.8. All testing, modeling, and production activities related to waste form development and qualification will likely be governed by Quality Assurance requirements. Most of the work done to formulate compositions, determine degradation mechanisms, develop models, and evaluate processing effects can probably be performed at less stringently controlled research and development levels. Other aspects, such as parameterizing the degradation models and establishing processing and consistency test limits, must likely meet higher Quality Assurance standards. These typically involve the use of documented and approved testing and analytical methods, certified and calibrated standards and equipment, fully documented and technically reviewed data and analyses, etc. Additional requirements apply to processing equipment and the actual waste form products.

5.20.2 Waste Streams Waste streams amenable to a metallic waste form either include a significant amount of metal waste

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components or include radionuclides that are best immobilized in the metallic state. Waste streams that are currently being immobilized in a metal waste form10 or being evaluated for immobilization in a metallic waste form3,11 are discussed in the following sections. 5.20.2.1

Electrometallurgical Wastes

An electrometallurgical process has been developed under the auspices of the US DOE Office of Nuclear Energy to treat the spent sodium-bonded fuel inventory from the Argonne Experimental Breeder Rector-II (EBR-II) at the Idaho National Laboratory.9,12 The inventory includes 1.1 MT of what are referred to as ‘driver fuel rods,’ which contain the majority of the fission products, and 22 MT of blanket fuel rods.10 The driver rods contain an enriched U
10Zr fuel alloy (63% 235U) and the blanket rods contain depleted U. Both fuel types are fixed in the rods with metallic sodium, which provides a thermal bond between the fuel and cladding. Treatment is necessary to separate waste radionuclides from the metallic sodium prior to their disposal. The electrometallurgical process recovers the uranium from the fuel and generates two waste streams: sodium and most fission products dissolve in the waste salt, whereas steel cladding hulls and metals not oxidized under the refining conditions remain as metallic waste. A ceramic waste form has been developed to immobilize the salt wastes and a metallic waste form has been developed to immobilize the metallic wastes. The electrometallurgical process may be used to treat other DOE-owned spent sodium-bonded fuel in the future, for example, 34 MT of sodium-bonded blanket fuel from the Fermi 1 reactor,10 but that decision remains to be made. Electrometallurgical treatment of other spent fuels that would generate metallic waste streams is being considered. Scoping work was conducted to evaluate the use of electrorefining on oxide fuels that have been reduced as an alternative to aqueous reprocessing of oxide fuels, but that option is not currently being pursued. The more likely use of electrorefining will be treatment of the used metallic fuels being developed as part of a closed fuel cycle within the DOE Fuel Cycle Research and Development program (FCR&D). New metallic waste forms will be needed for these wastes due to the planned used of Zircaloy cladding for the new fuels. The alloy developed for the steel-clad sodium-bonded EBR-II fuel is not appropriate for Zircaloy-clad fuel. However, it should be noted that Zircaloy cladding could

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Metallic Waste Forms

be recycled in the future and the waste stream compositions might not be dominated by Zircaloy cladding hulls. About 90% of the EBR-II inventory is steel-clad fuel and the rest is Zircaloy-clad fuel. The blanket fuel cladding is Type 304L stainless steels and the driver fuel claddings include Type 316, Type D9, and Type HT9 stainless steel. The waste form made with steel cladding was the primary focus during waste form development research, and only few scoping studies were carried out to evaluate potential waste form compositions for waste streams dominated by Zircaloy cladding.8,13
17 The composition and the microstructure of the waste form are affected primarily by the relative amounts of stainless steel fuel cladding and Zr from the driver or blanket fuel in the mixture used to make the alloy. The cladding supplies Fe, Cr, Ni, Mo, Mn, Co, Cu, V, and Si to the metal waste stream, plus trace amounts of Sn, C, and S. The chemical composition of the waste stream is dominated by Fe and Zr. The low-melting eutectic composition Fe
15Zr was selected as the target composition of the waste form to allow processing at about 1600  C. This requires adding Zr to the mixture. The predominant phases that are formed are those predicted by the Fe
Zr binary phase diagram, namely, similar amounts of an iron solid solution and an Fe2Zr intermetallic phase. Radionuclides and other components in the steel (e.g., Ni and Cr) are distributed between these phases and minor amounts of other phases (e.g., Fe23Zr6). This is discussed in detail in Section 5.20.8.1. The primary radionuclide is 99Tc, but the waste contains a small amount of carry-over salt with actinides that are retained as contaminants in the waste form. A small amount of depleted uranium is added to some waste mixtures to down-blend the 235U content in the waste form to below 20 mass%. Most waste forms are expected to contain about 1 mass% U, but some could contain as much as 11 mass% U. Table 1

The radionuclides expected to be immobilized that provide the highest activities are listed in Table 1, expressed as the activity in a disposal canister containing two waste form ingots (see Appendix A in Ebert.)18 The 18 radionuclides listed in Table 1 provide 99.6% of the total 3  1012 Bq expected in a canister. For comparison, the total radionuclide activity in an average canister of high-level radioactive waste (HLW) glass is estimated to be more than 2.8  1015 Bq. 5.20.2.2 Aqueous Processing Wastes for UREX + Flowsheets The US Department of Energy is evaluating methods to reprocess commercial spent fuel from light water reactors through the FCR&D program. The technologies developed to partition fuel elements and recycle actinides are intended to expand the nuclear power capacity to meet the global future energy needs while minimizing the amount of high-level radioactive waste requiring geologic disposal. The near-term goal is to develop the technologies needed for a proliferation-resistant nuclear fuel cycle. To this end, various operational flowsheets are being developed and tested to recover uranium and actinide elements from spent oxide fuel for recycle and separate long-lived fission products from other fuel wastes for separate immobilization and disposal. An important potential benefit of the processing operation is minimizing the heat load of wastes destined for geologic disposal. The partitioning of short-lived heatgenerating radionuclides, such as 137Cs and 90Sr, from short-lived (e.g., lanthanides) and long-lived radionuclides that do not generate significant heat loads, such as the 129I, 99Tc, and 79Se, provides the opportunity to tailor waste forms to the properties of particular waste streams and dispose these waste forms according to risk.3,11 Although current US regulations classify all waste streams from reprocessed fuel

Estimated radionuclide activities in EBR-II metallic waste form

Nuclide

Bq/canister

Nuclide

Bq/canister

Nuclide

Bq/canister

14

2.63  10 1.01  1011 3.24  1010 1.68  1010 6.59  1010 1.93  1012

234m

5.99  10 2.04  109 3.30  109 2.42  109 1.73  1010 6.22  109

99

7.84  1011 8.07  108 5.99  108 4.74  109 1.55  108 5.99  108

C Co 93m Nb 94 Nb 59 Ni 63 Ni 60

10

Pa Pu 125 Sb 126 Sb 126m Sb 126 Sn 239

8

Tc Te 234 Th 234 U 235 U 238 U 125m

Data from Ebert, W. L. Testing to Evaluate the Suitability of Waste Forms Developed for Electrometallurgically Treated Spent SodiumBonded Nuclear Fuel for Disposal in the Yucca Mountain Repository, Argonne National Laboratory Report ANL-05/43; Argonne National Laboratory: Argonne, IL, 2005.

Metallic Waste Forms

as high-level radioactive waste, safety analyses might indicate some waste streams/waste forms as benign and the risks low enough that they could be regulated as Class C or Greater-Than-Class C (GTCC)waste rather than high-level waste. The FCR&D program is currently studying the application of aqueous separation technologies to spent oxide fuels. Prior to dissolution, the cut fuel is heated in a voloxidation step intended to remove gaseous and volatile fuel components, including 3H, 14 C, and 129I. These are volatilized at increasing temperatures and are captured (separately) from the offgas for immobilization and disposal. Dissolution of these fuels (typically in hot HNO3 with HF) is incomplete, with residues including five-metal particles and Zr from the fuel and minute scraps of Zircaloy cladding generated when the cladding was cut. (Additional gas release occurs during dissolution.) The undissolved solids (UDS) are removed from the dissolved fuel solution by filtration. The UDS waste stream is suitable for immobilization in a metallic waste form. The dissolved fuel solution is subjected to a uranium extraction (UREX) operation that separates the U and Tc in the solvent phase and the other fuel components in the raffinate. The Tc (as pertechnetate ion) is removed from the U
Tc solution using an anion exchange change column. The U is recovered for recycle. The pertechnetate is later eluted from the column and can be recovered as either TcO2 or tetra n-butyl pertechnetate using one of the two methods that are being developed. The TcO2 can be reduced to Tc metal by heating in a reducing atmosphere (e.g., CO or Ar/H2 gas) and the tetra n-butyl pertechnetate can be reduced to Tc metal by steam reformation in a separate operation. The recovered metallic Tc is suitable for immobilization in a metallic waste form. The UREX raffinate can be treated to separate Rb, Sr, Cs, and Ba from the rest of the dissolved fuel with the fission product extraction (FPEX) operation. The 137 Cs and the 90Sr are the major heat-generating constituents of the fuel waste. Both have half-lives of about 30 years, decaying to stable 137Ba and 90Zr, respectively. The Cs/Sr/Ba/Rb stream will likely be immobilized in a glass waste form. The transuranic elements and lanthanides are next separated from the fuel solution by a transuranic element extraction (TRUEX) operation. The TRUEX raffinate contains dissolved transition metals from the fuel, but is dominated by ferrous sulfamate added to improve the efficiency of the TRUEX separation for Np and Pu. The sulfur must be removed from the TRUEX raffinate before it can be immobilized in either a glass or a metallic

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waste form. The dissolved transition metal fission products can be recovered by increasing the pH to coprecipitate them with iron as hydroxides and then reduced for incorporation into a metallic waste form at higher waste loadings. Finally, the lanthanides are separated from the actinides using TALSPEAK (trivalent actinide–lanthanide separations by phosphorous reagent extraction from aqueous complexes). The actinides are recycled (as fuel) and the lanthanides immobilized in a glass waste form with the Cs/Sr/Ba/Rb stream. For a typical light water reactor fuel, this scheme would result in about 37% of the waste immobilized in a metallic waste form, about 46% immobilized in glass, and about 17% immobilized as captured offgas in another waste form. The compositions of the waste streams will depend on the dissolution conditions, separation efficiencies, fuel burn-up (weakly), and fuel aging prior to reprocessing. The compositions of waste stream suitable for immobilization in metallic waste forms 51 GWd/ MTHM (gigawatt days per metric ton metal) fuel stored 20 years before reprocessing are given in Table 2.19

Table 2 Estimated metallic waste stream compositions for 20-year-old 51 GWd/MTHM fuel

S Fe Se Rb Zr Mo Tc Ru Rh Pd Ag Cd Sn Sb Te Total

UDS (kg/ MTHM)

Recovered TRUEX Recombined Tc with (kg/MTHM) ferrous sulfamate (kg/MTHM) kg/ Mass MTHM %



0.00463 0.114 1.86 4.75 0.0275 1.64 0.0238 0.0684 0.0517 0.0513

0.166 9.876







0.875







0.875

13.5 23.5 0.0394
3.78 0.357
1.85 0.373 1.67 0.0631 0.146 0.141 0.0279 0.589 32.536


23.5 0.0857 0.114 5.64 5.11 1.15 3.49 0.611 2.35 0.115 0.197 0.141 0.0279 0.755 43.288


54.29 0.20 0.26 13.03 11.80 2.66 8.06 1.41 5.44 0.27 0.46 0.33 0.06 1.74

Data from Ebert, W. L. Immobilizing GNEP Wastes in Pyrochemical Process Waste Forms, US Department of Energy Report GNEP-WAST-PMO-MI-DV-2008-000150; Idaho National Laboratory: Idaho Falls, ID, 2008.

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Current waste form research under the FCR&D program is focused on two options for these waste streams: combining the UDS, recovered Tc, and pretreated TRUEX raffinate wastes in a single alloyed waste form or combining the UDS and recovered Tc in a metallic waste form and immobilizing the TRUEX raffinate wastes separately or in glass. The composition of the three combined streams is dominated by Fe from the TRUEX raffinate, which accounts for 54% of the total mass. The next most abundant elements are Zr, Mo, and Ru. Those elements dominate the combined UDS and recovered Tc streams. The amount of ferrous sulfamate assumed in Table 2 is conservatively high, but Fe is expected to still dominate the waste stream composition after the amount of ferrous sulfamate is optimized. Research is in progress to replace ferrous sulfamate with an alternative reductant. However, iron could be added to the TRUEX raffinate to recover the dissolved metals by coprecipitation. The gross composition of the recombined waste streams is similar to the composition of the EBR-II SS
15Zr metallic waste form considering that Mo, Ru, and other components will dissolve in iron to produce a solid solution similar to the steel phase in the SS
15Zr alloy. 5.20.2.3

Hulls and Hardware

The metallic wastes from the electrometallurgical treatment of EBR-II spent sodium-bonded fuel are dominated by the cladding hulls, and the metallic fuel wastes (primarily Zr) are only a small component of the waste stream. In the aqueous reprocessing flow sheets for oxide fuels, the fuel can be mechanically separated from the cladding hulls after voloxidation (if that operation is used) or after fuel dissolution. The hulls can be acid washed to remove adhering TRU contamination and disposed of separately; the hulls will likely retain enough residual contamination and activation products in the bulk metal to require disposal as GTCC waste. Some of the Zircaloy hulls could be used as a source of Zr added to alloy other waste streams. The remaining hulls could be alloyed separately, combined with the separations waste streams, or compacted with the assembly hardware for separate disposal. Preliminary studies17 indicate that Zircaloy hulls can be alloyed with waste components if significant amounts of metal additives are added to lower the processing temperature to practicable levels. However, the amount of added metal needed to lower the processing temperature is by itself adequate to alloy the wastes (i.e., without the hulls). Because there will be about 10 times

more cladding hulls than metallic fuel wastes, compacting the hulls as a separate waste form will probably be more economical than alloying them with the wastes. Neutron activation products generated by components in the various hardware materials used in fuel assemblies will include 14C, 55Fe, 59Ni, 60Co, 63Ni, 93 Zr, 94Nb, and 99Tc. The activity contribution of each can be calculated from the elemental concentration in the metal, the neutron flux, the abundance and thermal neutron capture cross-section of the reacting isotope, and the specific activity of the product radionuclide. The last three are constants and their product provides a measure of the efficiency of each neutron activation reaction; that product is referred to herein as the ‘efficiency factor’ and provides insight into the likely contribution of that radionuclide to the overall radioactivity. Activation of 98Mo leads to the production of 99Tc with an efficiency factor 2  109 Bq barns g
1. Small amounts of Mo (2
3 mass%) are present in Type 316 and other stainless steels, Inconel-718, and in trace amounts in the Zircaloys. Activation of nitrogen, carbon, and oxygen in the fuel and hardware produces 14 C with efficiency 1.27  1012, 1.70  108, and 1.48  108 Bq barns g
1, respectively. Activation of 92Zr in Zircaloy (only) produces 93Zr, which has a very long half-life, but the efficiency factor for 73Zr is only 5.55  108 Bq barns g
1. Activation of Ni and Nb produces 59Ni, 63Ni, and 94Nb with efficiency factors 8.99  1011, 9.16  1013, and 7.96  1011 Bq barns g
1, respectively. All hardware materials contain Ni, but 94Nb will only be produced in assemblies with Inconel components. Generation of 63Ni will dominate the initial radioactivity, but it will decay faster than 59Ni and 94Nb. Both 55Fe and 60Co are short lived and neither will be a significant long-term dose contributor. The efficiency factors are 1.09  1015 and 8.14  1015 Bq barns g
1, respectively. 5.20.2.4

Processing Additives

Metals can be added to the waste streams to tailor the phases comprising the waste form or serve a role in the processing operation. The steel anode baskets used in the electrorefiner are a source of Fe, Cr, and Ni to the waste form. Zirconium is added to provide a neareutectic composition to facilitate processing and generate a consistent phase composition. Some of the metal components used in waste separations and containment can be selected based on how they would impact the waste form. One method for recovering 99 Tc that is being investigated within the FCR&D

Metallic Waste Forms

program is reductive deposition on steel wool. The steel wool and the canister in which it is housed could provide the Fe and other components used to make the waste form. Many components of steels are added to improve the durability. Some of the waste components could serve the same role in the steel-like phases formed in Fe-based alloy waste form. The waste components might also increase the durability of intermetallic phases that form. The approach taken when formulating metallic waste form compositions is to minimize the amounts of additives (and thereby minimizing the mass and volume of waste) while optimizing the processing conditions, waste form durability, and consistency of the phase composition and distribution of radionuclides. The addition of metal to the waste streams allows for better control of the waste form phase composition. Variances in the amounts or compositions of the waste streams from different processed fuels can be compensated for by adjusting the amounts of additives to result in a consistent waste form. The need for a consistent phase composition (because that results in consistent performance) will likely take precedence over the desire to minimize the amount of waste forms produced.

5.20.3 Alloy Compositions Metallic waste forms are formulated to incorporate high concentrations of radionuclides into durable phases at low processing temperatures. The very high-melting temperatures of many metallic waste components require reaction with a lower-melting metal to be dissolved. An example of this is the Fe
15Zr composition used for the EBR-II MWF (see Section 5.20.8.1). The melting temperature of Zr is 1855  C, but reaction with molten Fe allows Zr to be dissolved at temperatures as low as 1337  C at the 84.9Fe
15.1Zr wt% eutectic and at 928  C at the 16.2Fe
83.8Zr wt% eutectic. Eutectics permit waste processing at practical temperatures. Compositions slightly off the eutectic can be used to promote a particular phase assemblage. In the case of the EBR-II MWF, the fact that the actinide elements report to the intermetallic phase demands that an adequate amount of that phase be present in the alloy to avoid forming another phase, which would probably change the waste form performance. Using an alloy composition that is slightly Zr-rich relative to the eutectic will generate a sufficient amount of the intermetallic phase to sequester all the actinides.

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On the other hand, a sufficient amount of a solid solution phase may be needed to encapsulate the intermetallic phases in a physically robust monolith. Binary and ternary phase diagrams are useful for estimating the solubilities of waste components in a host metal, the intermetallic phases that may form, and the required processing temperature. In complex waste mixtures, competition between solutes (e.g., for interstitial sites in intermetallic phases) will usually lower the solubilities and could affect the phase assemblage and waste form performance. This is most readily determined experimentally. Multiphase alloy waste forms having a simple phase composition are preferred over more complex compositions due to the importance of waste form product consistency. The need to take into account only a few phases will benefit performance calculations, waste form analyses, and waste form qualification. Host matrices that can accommodate a wide range of waste elements within a few phases balance the desire to minimize the amount of waste form produced with the benefits of a simple phase composition that provides enough flexibility to accommodate the anticipated range of waste stream compositions. The austenite and the ferrite solid solution phases form in the SS
15Zr alloy developed for EBR-II wastes provide compositional flexibility for the formation of intermetallic phases having a narrower range of stoichiometries by accommodating the excesses. In a multiphase alloy waste form, the corrosion potential of one phase will necessarily be lower than those of the other phases, and that phase will corrode preferentially and mitigate corrosion of the other phases. The dissolving phase serves as the anode in the oxidation
reduction reactions of the other phases, which act as cathodes. Such a phase is often engineered into systems to serve as a sacrificial anode to protect other metals. The waste form should be designed so that the phases containing radionuclides are not the least durable phases. Although a sacrificial phase could be added as a surface coating similar to zinc-coated iron, it could instead be distributed throughout the bulk of the waste form. This would have the advantage of not being scratched off the waste form by abrasion, but the disadvantage that corrosion of that phase would leave a porous waste form with a high specific surface area. Pits and crevices usually provide microcells that promote corrosion and should be avoided. Steel is galvanized by dipping into a bath of molten zinc at about 455  C. Although steel melts at about 1510  C, an oxide-free steel surface will react

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Metallic Waste Forms

with molten zinc to form various Fe
Zn intermetallic phases (referred to collectively as the ‘alloy layer’) between the steel and the pure zinc outer surface within minutes in a batch process. It may not be necessary to coat the entire surface, since steel corrosion commonly occurs at point defects at the surface and where microcells develop.20

5.20.4 Processing Methods Requirements for processing facilities will depend on the processing temperature and atmospheric conditions needed to melt the wastes and added metals into an assemblage of durable phases in a physically robust waste form. Small prototype alloys have been made using laboratory-scale furnaces under vacuum and under flowing inert gases. Reducing gases such as Ar/H2 and CO could also be used to control the furnace atmosphere. It may be necessary to contain the furnace in a radioactive hot cell. 5.20.4.1

Alloy Production

Alloys can be produced in resistance or induction furnaces using single-batch or continuous casting methods. Batch melts provide a convenient means of controlling the composition through the relative amounts of wastes and additives placed in the crucible and may be better suited for the required process controls. Continuous casting commonly provides energy and total cost savings, but can result in products having microstructures significantly different than those made with a batch method and a different assemblage of phases. Continuous casting usually generates coarser and more elongated phases, although the microstructure can be modified by annealing. The need to contain the furnace, such as in a hot cell, may dictate which process is used. The EBR-II MWF to be made at INL will be processed using an induction furnace and cast from batch melts as ingots in a crucible. The planned method is summarized here as an example of designing a batch processing method for a specific waste stream. The EBR-II metallic waste stream will be composed of 0.025-m (1-in.) sections of chopped cladding hulls with adherent metallic fuel wastes, primarily Zr, and a small amount of entrained salt. These are removed from the steel anode basket and loaded into a crucible with appropriate amounts of added Zr metal and depleted U. A 2-piece crucible assembly is used to facilitate loading the hulls and

accommodating the 60% volume reduction that occurs upon melting. The crucible assembly includes a crucible funnel that drains into a casting crucible. Both are made of graphite coated with an yttria refractory. The crucible funnel is also lined with steel to protect the refractory-coated graphite funnel during loading. The wastes are loaded into the funnel, which channels the metal into the receiving crucible as it melts. The steel liner melts during processing and drains into the casting crucible with the waste and becomes incorporated into the waste form. Typical processing conditions ramp to a maximum processing temperature of 1600  C over about 2.5 h and hold the mixture at that temperature for about 2 h. A vacuum of about 133 Pa (1 torr) is applied while the temperature increases to about 1300  C to distill the entrained salt. The distilled salt is condensed and recovered outside the furnace; it can be recycled back to the electrorefiner or added to the salt waste stream. The furnace is then backfilled with argon gas to atmospheric pressure (or slightly higher) before the furnace temperature is further increased. Melting begins when the temperature exceeds about 1350  C, but processing at 1600  C provides a 250  C overtemperature that helps dissolve waste metals into the molten iron and form an intimate mixture of iron solid solutions and intermetallic phases as the mixture is allowed to cool. Laboratory tests have shown the importance of locating the lower-melting Fe waste on top to flow over the higher-melting Zr waste. Cooling rates for prototype waste forms have been about 10  C min
1, but fullsized waste forms may cool more slowly. Low cooling rates facilitate solid-state diffusion impacting the formation of the thermodynamically preferred phases, but phase changes predicted to occur at below about 1000  C are not expected to be complete because of slow diffusion. Subsequent heat treatments of prototype waste forms have shown that modifications to the phase composition do occur, such as increasing amounts of intermetallics.21 Ferritic and, to a lesser extent, austenitic steels are susceptible to accelerated intergranular corrosion, primarily due to depletion of Cr near the interface, and this may be exacerbated by heat treatments. The concentrations of some waste components may have a similar effect on the severity of intergranular corrosion in metallic waste forms. The aggressive environment created by molten metal dictates the use of durable materials. Yttrialined graphite crucibles have been used to successfully cast alloys during the development and

Metallic Waste Forms

demonstration phases of electrometallurgical treatment. These are sufficiently durable in the furnace environment, but fragile. Developing better crucible materials is an on-going effort. Waste forms can be cast directly in these crucibles, but it may not be practical to pour or drain the molten metal into separate molds. A steel rod can be inserted into radioactive cast alloys to simplify both removal of the waste form from the furnace and loading it into a disposal canister. 5.20.4.2

Waste Conditioning

Some waste streams contain nonmetallic components that cannot be immobilized in an alloy. Small amounts of these contaminants are expected to be sequestered within slag phases that form on or near the surface of the waste form, but large amounts will probably need to be removed prior to alloying. As described earlier, contaminant chloride salts are removed from EBR-II metals by distillation within the furnace. Other waste streams to be alloyed will likely contain contaminants from the fuel, from separations operations, from storage containers and transfer lines, etc. It may be possible to remove some contaminants within the processing furnace, as is done with entrained EBR-II salt, whereas other contaminants will need to be removed beforehand. It may not be possible or worthwhile to remove low levels of contaminants from the feed. The presence of some nonmetallic materials may even be desirable to help immobilize other contaminants, and waste forms could be formulated to utilize phases to sequester contaminants. Small amounts of slag were formed in surface layers 1
10-mm thick on all SS
15Zr materials made for testing during the development of the EBR-II MWF. These were commonly enriched in C, O, and, in some cases, N. For example, examination of the slag that formed during the production of a SS
15Zr material in a graphite casting mold with scanning electron microscopy (SEM) revealed the presence of ZrC and ZrO2 inclusions within the alloy. The alloy associated with the slag was also enriched in Zr relative to the underlying intermetallic phases. Slags were not detected on Zr
8SS materials. This may be due to the higher solubilities of C, N, and O in Zr(b) than in Fe(g), which are the metal phases that form initially when the Zr
8SS and SS
15Zr materials are produced. These convert to Zr(a) and Fe(a) as the alloy cools. Contaminants are less soluble in these phases, but cannot diffuse out during the phase transition.

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Slags were formed intentionally on some test materials by oxidizing the steel (in air) before melting. These were used in tests to measure the durability of the slag, which was found to be similar to the durability of the alloy. Whereas slags can simply be removed from industrial metal products, they must remain a part of the waste form, and their role in immobilizing radionuclides must be taken into account in performance models. The presence of a slag layer will not necessarily be deleterious to the waste form and could be used to an advantage. Although metallic radioactive waste forms are likely to be produced with induction or resistance melters, blast furnace slags provide insight useful for tailoring slag phases in waste form alloys. Blast furnace slag consists of nonmetallic contaminants from the ore and from additives (e.g., coke and limestone) that are not incorporated in molten metals but combine to form slag. The slag is less dense and floats on the surface of the molten iron where it can be physically separated from the metal product. Materials are often added to the mixture to facilitate the formation and removal of slag. Slag is readily skimmed from the surface of the molten metal or the metal can be drained from beneath the slag, but slag can adhere strongly to the cooled metal and be very difficult to remove. It may be possible to control the composition of the slag phase to influence both its physical nature and its capacity to retain radionuclides. The role and benefits of a slag layer remain to be studied and evaluated.

5.20.5 Testing Objectives The laboratory tests conducted with radioactive waste forms can be grouped according to three primary objectives to be addressed by testing: (1) waste form performance, (2) waste form consistency, and (3) regulatory acceptance for disposal. These objectives are not independent, since both performance and consistency must be deemed adequate for disposal, etc., but listing information needs in each group can help guide the planning and interpretation of laboratory and field tests conducted. 5.20.5.1 Matrix Degradation and Radionuclide Release The key measure of waste form performance is the rate at which radionuclides are released under the anticipated range of environmental conditions. Tests

516

Metallic Waste Forms

are conducted to determine the modes and rates of waste form degradation and radionuclide release, to develop a mechanistic model that takes into account the effects of key variables in the disposal environment, to quantify the effects of environmental variables on the waste form degradation and radionuclide release kinetics, and to parameterize the model. Several test methods are needed to highlight particular processes. The initial tests are conducted to determine and understand the processes that control the release of radionuclides from the host phase and the degradation of the host phase. The test methods and testing conditions are selected based on conceptual models for degradation and release. For metallic phases, oxidation of the metal is usually the first step followed by dissolution of the metal oxide. The oxidation of some radionuclides, such as Tc, may result in the formation of a species with diffusion-controlled release, such as TcO
4 . Therefore, the intent of the initial tests should be to determine whether there is an effect of the electrochemical potential (Eh) or an imposed potential on the generation and release of soluble species. A separate set of tests is needed to determine the effects of environmental variables on the degradation and release rates. These could include temperature, solution pH, solution or atmospheric Eh, dissolved halide concentrations, groundwater composition, radiolysis products, etc. The sensitivity of the rates to these variables should be measured under conditions spanning the anticipated values of the disposal system to ensure that the model bounds the modeled environment. How the dissolution step is coupled with the oxidation step, if this is indeed the mechanism, should also be determined. Although a fully mechanistic model is desired, it is usually not possible to fully describe the waste form degradation and radionuclide release based on atomistic and theoretical understanding. Instead, empirical models are usually required to describe some aspects of the observed corrosion and degradation behavior. For example, although the effect of Cl
on steel corrosion is well known, quantifying the effect of the Cl
concentration on the measured release of Tc will probably require an empirical model supported by a series of laboratory tests. To the extent possible, the tests used to measure the dependence should be based on an understanding of the effect so as to highlight its impact on the test response. Some degree of coupling in the measured parameter values is unavoidable, such as temperature and pH. Separating the effects of temperature and pH neglects the possible dependence of the activation energy on the pH.

After the model has been developed, it should be validated using tests other than those used to develop and parameterize the model. Validation is intended to show that the model adequately represents the process determined to control waste form degradation and radionuclide release under relevant conditions. Separate confirmation tests may be needed to demonstrate that the mechanism that is being modeled is appropriate for the integrated disposal system. The computational requirements for atomistic and detailed mechanistic models usually exceed the capabilities of repository assessment calculations. Simplified (abstracted) models that represent the effects of environmental conditions and the degradation and release rates are used instead. Confidence in the predictions using simplified models is based on the underlying mechanistic models and, usually, conservative aspects built into the simplified model. Performance assessment calculations are usually based on reactive transport models in which radionuclide release due to waste form degradation is treated as a source term that is coupled with expressions for groundwater flow, diffusion, sorption, and other relevant transport processes (vapor transport, diffusion through thin films of water, etc.) in equations expressing conservation of mass and conservation of energy. These are used to calculate contaminant transport over time under a range of conditions to determine dose to individuals under various scenarios such that the impact of the radionuclide release on the calculated dose depends on many other factors. Depending on the approach taken for risk assessment, a specific performance requirement might not be assigned for radionuclide release or waste form degradation rates. Specific performance requirements were not assigned to waste forms for the Yucca Mountain total system performance assessment, but are specified for lowactivity waste glasses in the performance assessments for the Hanford disposal system, which includes glass and grouted waste forms. 5.20.5.2 Waste Form Consistency Through Process Control Consistent waste form behavior is of primary importance for waste form acceptance. Performance assessment calculations will relate regulatory dose limits to radionuclide release rates, which will be related to waste form composition (and phase composition) by the source term model(s). Confidence in the performance assessment relies on confidence in the waste form composition, which must be controlled during

Metallic Waste Forms

production. Important aspects of a metallic waste form composition might include the domain sizes of the component phases, the degree of mixing, the fraction of slag phases, etc. Process controls might include the amounts and chemical compositions of the feed materials, both the wastes and additives, and particular processing conditions, including the mixing and loading of the feed materials, the heating schedule (temperature, time at temperature, cooling rate, heat treatment, etc.), furnace atmosphere, and crucible material. The operating range of each processing variable that results in consistent waste products must be determined based on its direct impact on the component phases that form and the distributions of radionuclides (and the indirect impact on performance). The relationships that must be established by testing and modeling are summarized in eqn [1]. Production control $ Phase composition control $ Matrix degradation rate control $ Radionuclide release control $ Acceptable performance

½1

Ultimately, the processing limits must be related to the performance of the disposal system, usually through the release rates of radionuclides contributing to the regulated dose. In the performance assessment calculations, transport through natural and engineered barriers can decrease the dose contributions of some radionuclides at the point of interest so that the radionuclide that is released the fastest or to the greatest extent may not have the greatest impact. For example, a low solubility limit and strong adsorption to mineral phases retard the transport of plutonium isotopes and their impact on total dose at a regulated point. The relative impact of Tc may be greater regardless of the relative amounts of Pu and Tc released from the waste form due to the high solubility and negligible transport retardation of the pertechnetate ion in an oxidizing environment.

5.20.5.3 Waste Form Acceptance and Regulatory Requirements In the United States, regulatory requirements include compliance with Nuclear Waste Policy Act, as amended, and dose limits specified by the US Environmental Protection Agency for disposal systems. Disposed waste is also subject to hazardous waste regulations, including the Resource Conservation

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and Recovery Act (RCRA). Additional safeguard and security requirements may be added in the future. Although waste form acceptance ultimately occurs through a US Nuclear Regulatory Commission license for a facility to receive and dispose waste, separate acceptance criteria may be established between various DOE agencies responsible for producing, characterizing, packaging, licensing, and disposing the waste forms. In general, the acceptance criteria are intended to provide assurance that the chemical, physical, and radiological performances of the waste form are such that the disposal system will meet regulatory requirements over its service life as predicted by performance assessment calculations. These are obviously closely related to the degradation behavior of the waste form (and radionuclide release), the radionuclide inventory, and the disposal environment. For example, the DOE Office of Civilian Radioactive Waste Management has provided the Waste Acceptance System-Requirements Document22 to address the acceptance of commercial spent nuclear fuel, DOE-owned spent nuclear fuel, naval spent nuclear fuel, and high-level radioactive waste for disposal in the planned Yucca Mountain repository. Requirements are described in that document for the waste form materials, such as phase composition, radionuclide content, product consistency, and chemical durability; for the canistered waste forms, such as criticality and thermal outputs; and for the canisters themselves, such as material, dimensions, maximum weight, labeling, handling fixtures, required levels of surface decontamination, etc. Although most of these requirements are not regulatory requirements, they provide insight into additional issues that should be considered when designing a testing program.

5.20.6 Modeling Modeling activities should be integrated with testing activities at the earliest stages of developing test plans. The initial model will probably be a conceptual model based on the identity of the waste stream and matrix material, the expected material alteration modes based on literature surveys, anticipated environmental conditions for the disposal system, and insights from the behaviors of available analog materials. Testing and modeling should evolve iteratively as test results are collected and the model is modified. When the model progresses to the point of predicting degradation behavior, validation tests should be conducted for direct comparisons with model

518

Metallic Waste Forms

predictions. Other tests are needed to confirm that the modeled process is operational under disposal conditions. Eventually, the validated and confirmed model is adapted for use in performance assessment calculations. An important aspect of the performance model is correlation of the phase composition to performance. For homogeneous glass waste forms, the chemical composition can be directly related to performance through the glass dissolution rate. For multiphase alloyed waste forms, performance will probably depend on the assemblage of alloyed phases that form and the distribution of radionuclides. The more complicated corrosion mechanism of alloys is an added challenge to modeling and to demonstrating process control of waste form performance. 5.20.6.1 Matrix Corrosion and Radionuclide Release Mechanisms A mathematical model is required to represent the release of radionuclides from the waste form to become available for transport through the disposal system as a function of time. Release may occur by diffusion of radionuclides out of the waste form or by degradation of the waste form matrix and phase(s) that contain the radionuclide. A mechanistic model that takes into account the key processes in waste form degradation and radionuclide release will provide maximum confidence to long-term predictions and performance assessments. Rate equations have been developed for various release models, including constant dissolution rate models, concentration-dependent models, reaction affinity models, and diffusion-controlled release models. Empirical rate expressions have been developed based on detailed measurements. An approach for developing a source term model for waste forms is provided in ASTM Standard C 1174 Prediction of the Long-Term Behavior of Materials, Including Waste Forms, used in Engineered Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste.4 This standard provides a philosophy and logic that can be followed to develop a performance model by integrating testing and modeling activities. It stresses the importance of a clear definition of the problem, characterizing the disposal environment, maintaining interfaces between testing and modeling to evolve the model based on test results and design tests based on model predictions, using analog materials and systems, and both validating the fidelity of the model to the mechanism and confirming the appropriateness of the model for the disposal system.

A variety of test methods is needed to understand the degradation mechanism of a waste form material and how environmental and engineering factors will affect the releases of important radionuclides, and to develop and parameterize a performance model. Many of the information needs related to waste form processing and waste form acceptance require an understanding of the corrosion mechanism. It is usually wise to address aspects of the mechanistic, environmental, and engineering system early in a testing program to gain confidence that all the important factors are being taken into account in the model. As the mechanistic understanding develops, other aspects of waste form behavior that must be investigated usually become apparent. The general information needs for developing a performance model are summarized as follows. Identify the radionuclide release mechanism: The objective is to determine whether each radionuclide is released stoichiometrically as the matrix degrades or whether separate models are required. In most cases, these tests will serve to confirm the release mode based on an understanding of the matrix material and how the radionuclide is incorporated. For the various waste forms being developed for reprocessing wastes, the release of radionuclides may be controlled by diffusion (leaching), congruent dissolution of the matrix, or degradation of the matrix to expose the sequestering phase that contains the radionuclide, which could then dissolve, be leached of radionuclides, be released as a colloid, etc. Release may require prior oxidation of the sequestering phase and/or the radionuclide itself, as is expected for Tc in a metallic waste form. Characterize the matrix degradation mechanism: It is expected that the release of radionuclides will be affected physically or chemically by the degradation of the waste form matrix. Degradation of the matrix may be required before a radionuclide can be released; it may simply need to be physically or chemically altered rather than dissolved. For some of the multiphase waste forms considered for reprocessing wastes, dissolution of an encapsulating phase may be required before water can react with the phase bearing the radionuclide. Radionuclide release will be affected by both the radionuclide-bearing phase and the host matrix. These could degrade by different mechanisms, for example, in the case of an oxide phase encapsulated in a metal matrix. Characterize the effects of environmental variables: The release rates of radionuclides and the degradation rates of matrix materials will probably be affected

Metallic Waste Forms

directly by environmental variables such as temperature and the groundwater chemistry (pH, Eh, dissolved component concentrations, etc.) and indirectly by sorption onto colloids, minerals, and engineering materials in the disposal system. The groundwater chemistry may change significantly due to interactions with engineering materials and waste form degradation. Tests generally need to be conducted under conditions exceeding the expected range of values for environmental variables to initially determine and model the dependency and then to verify that the mechanism remains operative over the anticipated range of disposal conditions. A variable that affects the degradation of a waste form but is not tracked in performance assessment calculations can be captured using another variable that is tracked or through the uncertainty ranges of coefficients in the rate equation. Develop a degradation model that accounts for environmental variables: If possible, the dependence of the degradation rate or radionuclide release on each environmental variable that affects the rate should be modeled mathematically according to the mechanism. It may be beneficial (or necessary) to combine the effects of two or more variables in a bounding semiempirical model rather than modeling each explicitly. This may be necessary if factors cannot be distinguished in tests or practically differentiated by modeling. Residuals from fitting the test data using separate dependencies will include the effects of cross terms, and the interdependence will be taken into account through both the regressed values and the uncertainties. Consider waste package interactions: The effects of the waste package materials on the groundwater chemistry, transport of released radionuclides, etc. must be considered in performance assessment calculations. Although these are taken into account separately from waste form degradation in performance assessments, the interactions that affect the groundwater prior to contacting the waste form should be considered when defining the range of environmental variables to be represented in tests. Standardized test methods routinely call for control tests and/or blank tests in which interactions of test solutions, dissolved components, and test specimens with vessels and supports are measured and taken into account in analyses of the test data. Interactions between metallic waste forms, waste packages, and engineering materials need to be evaluated, such as Galvanic coupling, and taken into account in models. Alloyed waste forms should not serve as sacrificial anodes for other metallic components in the disposal system.

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Measure model coefficients: The same test methods used to characterize the effects of environmental variables often provide data that can be used to determine the values of model coefficients. To measure model coefficient values, the effects of variables other than the one being quantified must be held constant or reliably taken into account when evaluating the results. Typically, this involves conducting a series of tests in which the values of all variables are held constant, except the variable of interest. Such tests can be conducted at constant temperature, in pH-buffered solutions, with spiked leachant solutions, in controlled atmospheres, etc. It is important that the measured coefficients are appropriate over the anticipated range of conditions in the disposal system, and it is possible that different coefficient values could be required for specific ranges if the degradation mechanism changes. Model validation: The intent of model validation is to demonstrate that the mathematical model adequately represents the particular process of interest, be it oxidation, diffusion, radionuclide release, or matrix degradation. A model is considered validated if it predicts responses consistent with the results of separate tests that were not used for model development or parameterization with an acceptable range. The validation test should focus on the specific process being modeled and the measured response should be sensitive to that process. For example, if diffusion-controlled release from a solid is being modeled, the validation test should be conducted under conditions that avoid saturation in the contacting fluid. The effects of other processes are considered in model confirmation tests. Model confirmation: The intent of model confirmation is to demonstrate that the process represented in the degradation model controls the waste form behavior under the range of conditions and other interactions expected in the disposal system. Confirmation tests generally couple the modeled process with other system processes that could affect waste form behavior. Model confirmation cannot be completed until the disposal system is identified and characterized. However, aspects of potential disposal systems likely to affect waste form behavior should be taken into account when developing the model, including the groundwater chemistry and flow characteristics, interactions with engineering materials and the host geology, radiation fields, etc. In most cases, the laboratory tests used to determine waste form corrosion behavior must subject the test specimen to more aggressive conditions than those

520

Metallic Waste Forms

expected in a disposal system to generate a measurable response. For example, the dissolution of a waste form due to contact by groundwater may have been modeled using the results of static immersion tests, whereas the disposal system is hydrologically unsaturated. A possible model confirmation test would be to expose the waste form to intermittent contact by dripping water to see how well the model represents degradation under these conditions. Other test methods that could be used include vapor hydration tests to simulate contact by vapor and unsaturated flow tests to simulate intermittent contact with flowing water. The responses of model confirmation tests will usually be too complicated for use in determining mechanisms for specific processes. This is because they will be affected by several processes acting simultaneously. They provide data that are responses to several interacting processes that must be taken into account in the model either directly through model variables or the ranges of model coefficient values. Nevertheless, such tests can provide confidence in the testing process that the correct process is being modeled early for a particular environment. 5.20.6.2

Performance in Disposal System

As with the waste form degradation and radionuclide release model, performance assessment models based on a mechanistic understanding of processes affecting the spread of radionuclides through engineered and natural barriers would provide the highest confidence. This is usually not possible because of the complex pathways and interactions affecting transport. Instead, transport models use effective distribution coefficients to model complex interactions with various minerals in the soil, including reversible and irreversible sorption to fixed and mobile substrates, effective flow rates for advection through fractures and along grain boundaries, diffusion and dispersion, etc. Small variations in temperature and water content spatially and temporally are neglected, but large variations can be modeled by separating the regions of interest into several cells that can be assigned different characteristic values.23 The solute transport models that have been developed for environmental impact analyses generally include equations to account for water infiltration rates through the system, source term release rates for important nuclides, and solute transport rates through the geologic media. Based on the conservation of mass for a transported radionuclide, the

dependence of the change in concentration C with time on these factors is written as @C @C @2C X Qn þ r ¼
n þ D 2 þ @t @x @x n

½2

Reactive transport models usually only consider transport in the direction of flow (x in this example), except for dispersive effects.23,24 The first term on the right-hand side represents advective transport, where n is the groundwater velocity, and the second term represents dispersive transport. The dispersion coefficient, D, is sometimes written as the sum of an effective diffusion coefficient for the geologic medium and Pthe coefficient of mechanical dispersion. The term n Qn represents the sum of the effects from each of the n processes that affect the solution concentration C, which may include waste form degradation, leading to radionuclide release (the source term), sorption, colloidal and secondary phase formation, and several other processes depending on the species of interest and the disposal system. The value of C is usually constrained in models by a speciesspecific solubility limit. The fourth term in eqn [2], r, takes into account the in-growth and decay of the radionuclide of interest. Most transport calculations for contamination assessments include transport through hydrologically unsaturated (vadose) zones. A dimensionless volumetric moisture content y can be included to represent the fraction of the total pore volume available in the medium that is occupied by water, and C can be replaced in eqn [2] by the product Cy. This simply scales the solution volume by the moisture content. The radionuclide release model provides the source term Q used in the transport model. Depending on the form of the rate equation for radionuclide release from a metallic waste form, its integration into a reactive transport expression may require additional terms to represent electrochemical processes. 5.20.6.3

Processing Control

The ability to maintain acceptable waste form performance (the retention of radionuclides) using processing controls must be demonstrated by establishing a correlation between the assemblage and compositions of alloy phases in the waste form and waste form performance measured in laboratory tests. The correlation should have a mechanistic basis, and, ideally, the dependence of the performance on the phase composition should be quantified and

Metallic Waste Forms

parameterized in the degradation model. This may not be possible because waste form performance cannot be measured with a single test method. Instead, a test method having a response that is correlated with the performance of the anticipated range of waste form compositions can be used to represent the relative performance. In the case of high-level waste glasses, the Product Consistency Test (ASTM method C 1285)4 is used to correlate performance with composition; the response in that test is used by DOE to identify acceptable waste glasses.22 The use of an analogous approach for metallic waste forms will require the identification of a test method and development of a correlation model that can be related to the performance model. These will then allow processing limits to be established for producing acceptable waste forms, such as waste loadings, additives, and processing temperature.

5.20.7 Test Methods A variety of test methods is required to understand the corrosion behavior of metallic waste forms, measure the kinetics of various processes, and accelerate corrosion to simulate aspects of long-term degradation. The degradation rates of well-designed waste forms should be difficult to measure under anticipated service conditions if they effectively retain radionuclides. Test conditions that differ significantly from the service conditions are often required to produce a measurable test response or measure the effect of a variable. Relating the corrosion occurring under extreme test conditions to service conditions usually requires an understanding of the corrosion mechanism, even though an understanding of the mechanism is often the objective of the tests. Testing and modeling to determine the corrosion mechanism should be coordinated and iterative activities. The selection of initial test methods should be based on a conceptual model and hypothesized mechanism, both of which should be developed from literature reviews and considerations of possible analog materials. In the case of metal waste forms, experience with steel corrosion provides a logical starting point. The initial tests should be conducted to evaluate the hypothesis and conceptual model, and both the conceptual model and the suite of tests should evolve as new data are collected. The degradation of a metal waste form and subsequent release of radionuclides is hypothesized to involve oxidation reactions to convert the metal

521

components to oxides and hydrolysis reactions to dissolve the oxides. Diffusion of oxygen through the evolving oxide layer is another important process required to continue the corrosion. It is likely that the oxidation, dissolution, and diffusion processes may each control the radionuclide release rate at some point and under some reaction conditions. Components might be released as quickly as the metal surface is oxidized early in the reaction but slow as an oxide layer develops and slowly dissolves. The oxide layer will likely serve as a diffusion barrier to oxygen, such that diffusion rate controls the continued corrosion rate. Passivation of the surface (of each component phase) may occur to prevent further corrosion, or the layer could become thick enough that the mismatch between the metal surface and the crystalline oxide layer causes the layer to spall off the metal surface and corrosion resumes. To characterize this hypothetical corrosion mechanism, separate test methods would be needed to highlight the oxidation reactions, the dissolution reactions, and the long-term stability of a passivating layer. The majority of test methods and techniques developed by industry for metal corrosion address shortterm behavior (surface oxidation, pitting, passivation, etc.) related to the service objectives, whereas the release of radionuclides to the solution over very long corrosion times is of primary interest for alloys used as waste forms. The oxidation of metal components will have no effect on waste form performance until transportable radionuclides are released to solution, either by dissolution or colloid formation. Many of the tests used to study the degradation of glass, minerals, and ceramics address the effects of the solution chemistry on the dissolution reaction. These materials dissolve by an affinity-controlled mechanism that is sensitive to the accumulation of components in the solution. The effect of the solution chemistry on the degradation of metals is not expected to be as significant because the most abundant components (e.g., Fe and Zr) are sparingly soluble in most solutions and will have a very limited dissolved concentration range. Nevertheless, similar test methods are needed to study dissolution of the oxide layers resulting in radionuclide release and diffusion (e.g., of oxygen) through the layers. Passivity is usually investigated by studying the nature of the films that form and the kinetics of the electrochemical processes. Corrosion tests in which specimens are immersed in solution or contacted by steam can be used to measure the release of components to solution and generate oxide layers for detailed examination. Electrochemical tests can be used to

522

Metallic Waste Forms

measure the kinetics of processes taking place on clean surfaces. Several useful electrochemical and corrosion test methods are discussed in the following sections. 5.20.7.1

Electrochemical Test Methods

Metal corrosion by oxidation reactions is conveniently and efficiently measured with traditional electrochemical techniques.25,26 These include construction of potentiostatic curves (anodic and cathodic charge curves), studying potential shift curves and double-layer capacity, electrochemical impedance spectroscopy, etc. Measurements can be conducted at a range of temperatures and in various solutions to determine and quantify the effects of many environmental parameters on the oxidation rate, such as pH and the presence of various solute species, and combined effects such as the Cl
concentration and critical pitting temperature for various alloy formulations. Electrochemical techniques used to determine corrosion rates utilize the corrosion current to calculate the corrosion current density, corrosion intensity, and corrosion penetration rate. The current that is passed upon application of an external potential can be related to the mass of material reacted using Faraday’s law and then expressed as a rate using the reaction time and specimen area. This gives the reaction rate in terms of the current density. The polarization resistance can be determined from a plot of the applied potential against the current and can be related to the current density by the anodic and cathodic Tafel constants for the material. The corrosion rate of the material can be calculated from the current density and the density and atomic weight of the metal (or the equivalent weight of an alloy). This gives the rate for uniform corrosion (oxidation) of a bare metal or alloy surface. If the corrosion is localized rather than uniform, the penetration rates at particular sites can be orders of magnitude higher than the average value. Some of the reacted specimens should be examined to look for areas of localized corrosion (pitting). Because the data analysis techniques require extrapolating measured curves, the accuracy of the experimental measurements is crucial to the calculated rates and their sensitivity to the compositions of the alloy and the solution. Standardized procedures should be followed whenever possible. The ASTM procedure Standard Test Method for Conducting Potentiodynamic Polarization Resistance Measurements (G 59) provides a standardized method for measuring corrosion potentials and potentiodynamic polarization resistance for determining the

general corrosion rates of metals.27 A small potential scan near the corrosion potential of the metal is applied to a sample and the resulting currents are measured. Tests conditions can be varied to study the effects of temperature, pH, dissolved O2, solution composition (e.g., oxidizers), etc. Potentiodynamic measurements (in general) provide a convenient comparison of the oxidation behaviors of various metals and alloys, and a rapid and economical approach for studying the oxidation step of the corrosion mechanism. Interpretation of the measurements requires assumptions regarding Tafel slopes and equivalent weights of alloys. A galvanic cell is formed when two (or more) metals or alloys having different polarization resistances are coupled and in contact with an electrolyte solution, including the separate phases in a multiphase alloy. In such a system, the least stable material will act as the anode material and corrode in preference to the other materials, which will act as cathodes. Galvanic couples can also form between waste form phases and engineering materials such as waste canisters. Obviously, radionuclide-containing alloy phases should not serve as the anode of a galvanic couple; rather, such phases should be protected by phases without hazardous components. Galvanic couples with sacrificial anode materials are used extensively to protect other metals, such as steels galvanized with zinc and iron anodes in residential water heaters. It may be possible to add such a phase as part of the waste form to protect radionuclidebearing phases, but it is crucial that radionuclidebearing phases do not act on the anode. The ASTM procedure Standard Guide for Conducting and Evaluating Galvanic Corrosion Tests in Electrolytes (G 71) addresses conducting and evaluating galvanic corrosion tests.27 The standard addresses selection of materials and preparation of test specimens, selection of test conditions (electrolyte, materials coupling, atmosphere, duration, etc.), and the evaluation of test results. Galvanic corrosion measurements provide insights into likely materials interactions, including interactions between different phases within an alloy and between an alloyed waste form and the waste canister material. Microscopic examination of the specimen before and after analysis can provide confirmatory evidence of the relative durabilities of the components phases, although test results are sensitive to specimen preparation and electrical contact. Several other standardized test methods, practices, and guides are available for conducting and

Metallic Waste Forms

evaluating electrochemical measurements, and consensus standards should be followed when available and appropriate. Besides providing the benefits of experience, standardized methods often provide reference test results for comparisons and quantify testing bias. For example, the ASTM standards G 1, G 3, G 5, G 15, G 50, G 60, and G 102 provide helpful insights into preparing specimens, conducting electrochemical tests and measurements, and interpreting test results.27 5.20.7.2

Corrosion Test Methods

Corrosion tests are used to study the release of waste form components into solution during waste form degradation. Many corrosion test methods have been developed to study the dissolution of oxide materials, such as glasses, that do not require a preceding oxidation step. These methods can be applied to a metal waste form, but the results may be complicated by the coupled oxidation reactions and depend on whether the metal test specimen is grounded or insulated. The ASTM procedure Standard Practice for Laboratory Immersion Corrosion Testing of Metals (G 31) provides a standardized method for directly measuring mass loss in laboratory immersion tests to determine the general corrosion rates of metals.27 Conditions that may be controlled include test duration, temperature, pH, oxygen concentration, solution flow, solution composition, and solution concentration. This procedure addresses factors important to sample preparation, test conditions, methods of cleaning specimens, and cautions regarding the interpretation of test results that are relevant to any test method. The relative contributions of generalized corrosion cannot be distinguished from localized or intergranular corrosion based on the mass loss alone. Although the mass can increase due to the formation of oxide surface layers or decrease as the layers dissolve, the corrosion products are removed from reacted specimens prior to weighing in the ASTM G 31 procedure. Simply tracking the mass change in test specimens will not be sufficient to understand the corrosion mechanism, but other analyses can be conducted to gain insight into the corrosion behavior, such as microscopic examinations of reacted specimens to detect pitting and identify corrosion products. The ASTM Test Method for Static Leaching of Monolithic Waste Forms for Disposal of Radioactive Waste (C 1220) provides a standardized method for measuring the mass of a monolithic test specimen

523

that has dissolved into solution under static conditions.4 The test results are sensitive to the corrosion behavior of the waste form with little feedback from dissolved components occurring in short-term tests. The geometric surface area of the specimen can be measured to allow accurate calculation of the specific release rate of soluble components. The test is easy to run, can be conducted under a wide range of conditions, provides sufficient solution volumes for analysis, and is economical. Only small volumes of waste solution are generated. Short-term tests can be used to measure the effects of temperature, pH, and components in the leachant on the release rates. Responses may become affected by the chemical affinity of the solution in long-term tests. Aspects of ASTM G 31 regarding specimen preparation and cleaning should be followed when conducting ASTM C 1220 tests with metallic specimens. Two important variations of ASTM C 1220 involve interrupting the test intermittently to recover solution for analysis: the partial replacement test and the solution exchange test. In the partial replacement test, a small portion of the solution is removed for analysis and replaced with fresh leachant solution. This allows the extent of reaction to be tracked in an otherwise static test, but dilutes the remaining solution slightly. The impact of diluting the solution is usually negligible in tests with metallic specimens because dissolved components do not affect the continued corrosion to a significant degree. The entire volume of solution is replaced with fresh leachant in a solution exchange test. Solution exchange can have a significant effect on metal corrosion due to opening the vessel to refresh the air and oxygen content. Replacing the entire solution volume will also replace all the dissolved oxygen (assuming that the added solution is air saturated), whereas partial placement will only replace a small fraction of the dissolved oxygen. This will be important for oxidation reactions forming oxide layers. In addition, solution exchange will re-establish the initial pH, but partial replacement will allow the pH to drift as the specimen corrodes. ASTM Standard Test Method for Diffusive Releases from Solidified Waste and a Computer Program to Model Diffusive, Fractional Leaching from Cylindrical Waste Forms (C 1308) provides a standardized method for solution replacement tests that should be consulted prior to running a modified ASTM C 1220 test.4 The test conditions specified in the standard are better suited to rapid release from grouted waste forms than slow release expected from

524

Metallic Waste Forms

metallic waste forms, although long-term tests at elevated temperatures could lead to measurable test responses. The vapor hydration test (VHT) is a static test in which a monolithic specimen is suspended in a sealed vessel with a small amount of water.28 When heated (typically in the range 125
200  C), the vapor phase becomes saturated and a thin film of water condenses on the specimen. The specimen temperature rises more slowly than that of the test vessel and provides the coolest surface when the test is initiated. As the specimen corrodes, the condensed solution usually becomes hygroscopic, so the thin film of water remains on the test specimen even after thermal equilibrium. In the standard VHT, the amount of water added to the vessel is carefully controlled so that a condensed layer forms on the specimen but no liquid water remains in the vessel to establish a reflux cycle. Alteration of the reacted sample can be analyzed and thickness of the altered surface layer measured on a cross-sectioned specimen. This method has been used to promote the formation of oxide layers on metallic specimens for microscopic investigations. Neither the actual acceleration factor nor the chemistry and the volume of the solution contacting the specimen during reaction are known in VHTs. The test response is sensitive to the volume of water that condenses on the specimen, which cannot be controlled accurately or measured during the test. The extent of corrosion can be estimated based on the amounts of alteration phases that formed. However, the VHT is not well suited for quantifying corrosion rates, because the precision of the test (i.e., the amounts of alteration phases formed) is usually poor. The consistency of the corrosion mechanisms at the relatively high temperatures at which most VHTs are performed and at temperatures relevant to disposal environments must be established to extrapolate VHT results to service conditions. In a modification of the VHT, enough water is added to promote refluxing during the test to flush components released from the specimen to the bottom of the vessel. The test is interrupted intermittently to recover and analyze the solution and track the release of waste form constituents. This modification is similar to a Soxhlet test,29 except that water vapor condenses on the sample itself rather than in a separate condenser and maintains an adhering layer of water on the sample rather than filling a sample boat. The modified VHT method has been used to

measure the release of U from metallic specimens (see Section V.C.3 in Ebert).18 5.20.7.3

Service Condition Test Methods

The electrochemical and corrosion tests discussed earlier are usually conducted to understand the corrosion mechanism and to develop and parameterize a degradation model. The test conditions that are used are typically very different than the environment that will exist in a disposal system (referred to as the ‘service conditions’). Most laboratory tests are conducted by contacting a small metal surface area with a large volume of water to highlight alteration processes or at elevated temperatures to promote corrosion and facilitate measurements. The service conditions of most disposal sites will be restricted to small volumes of water and thin films contacting small waste form surface areas. Metallic waste forms will likely not be thermally hot due to the low activities expected in amenable waste streams (e.g., those given in Table 2). The effects of limited groundwater volumes on waste form degradation have been studied in laboratory tests in which groundwater is periodically dripped onto crushed material or a monolithic specimen and in tests where a mixture of water and air is passed through crushed material. The unsaturated test was developed to simulate small volumes of groundwater dripping through a breached waste container onto spent fuel or HLW glass.30 The water collects on and reacts with the specimen, and occasionally drips off the specimen into the bottom of the test vessel. The accumulated solution is collected and analyzed periodically to track the release of radionuclides. The pressurized unsaturated flow test was developed to simulate disposal in a hydrologically unsaturated environment.31,32 A mixture of water and air is pumped through a column of crushed materials that may include soil, rock, waste form, and backfill materials to simulate a disposal system. Water interacts with these materials as it flows through the column, and the effluent chemistry is tracked during the test. Changes in the solids due to dissolution and precipitation of secondary phases can be measured at the end of the test. Column tests are commonly used to simulate groundwater interactions with soils and minerals to measure dissolution rates, distribution coefficients of contaminants, transport parameters, etc.33,34 Column tests can be conducted in the laboratory to control

Metallic Waste Forms

test conditions, such as temperature, water content, and flow rates, and to simplify sample collections. Column tests are also conducted in the field using lysimeters.35
38 Lysimeters are simply columns placed in the ground and filled with soil (sand, rock, etc.) of interest. Containers and probes can be placed at the bottom and various intermediate locations to analyze and collect groundwater that passes through the column. The top of the lysimeter can be left open to allow meteoric water (rain and snow) to enter and flow through the lysimeter or intermittently covered to regulate the amount of water entering the lysimeter. Waste form corrosion can be coupled with transport by emplacing one or more waste form specimens within the soil. The waste form specimens and adjacent soil can be recovered for analysis after testing. Finally, materials can be simply buried in the soil and allowed to corrode naturally with no attempt to control, collect, or analyze the groundwater. Examples of this are underground corrosion studies conducted by US National Institute of Standards and Technology (NIST), formerly the National Bureau of Standards, such as the corrosion of stainless steels buried in various soils in the early 1970s.20 These studies were focused on the corrosion effects on the specimens, such as etching, blistering, cracking, and pitting, which are important for steels used as canisters (see Sullivan39), but less so for steels used to immobilize radionuclides that are components of the alloy or solid solution.

5.20.8 Tests with INL MWF Electrometallurgical treatment is being used to condition used sodium-bonded fuel from the EBR-II reactor for disposal as high-level radioactive waste.9,12,40 Waste forms have been developed for the salt and metallic wastes resulting from the electrorefining process. The development of the waste form for metallic wastes is summarized in the following sections as an example of formulating and testing an alloy to address performance and processing issues, and development of a degradation model. A few key results are provided here; a more extensive summary of the test results and full bibliography are provided in Appendix B of Ebert.18 Detailed analyses of the alloy structure and phase composition provide a crucial link between the consistency of the metallic waste form maintained by process controls and the performance of the waste form in the disposal site as

525

modeled in assessment calculations and utilized for waste acceptance. Of course, the waste form performance reflects the combined behavior and interactions between all the component phases, which are not modeled individually for the MWF. 5.20.8.1 Formulation and Phase Compositions Two alloy compositions were identified as potential waste forms to immobilize waste streams that were dominated by stainless steel or Zircaloy cladding hulls. These target compositions were selected to maintain consistency in the phase compositions for the fairly narrow range of waste stream compositions and simplify the waste form qualification process. The binary Fe
Zr phase diagram41 was used as a guide for waste forms made by alloying stainless steel and Zircaloy claddings with the metallic wastes.17 An eutectic exists at about 84.9 mass% Fe and 15.1 mass% Zr with a melting temperature of 1337  C, whereat Zr6Fe23 and g-Fe are predicted to form. A maximum processing temperature of 1600  C was desired to allow use resistance or inductively heated furnaces. Alloys having various proportions of steel and Zr were made for analysis, and the microstructures, likely distribution of radionuclides, and durabilities of these materials were studied. Steel-dominated compositions from SS
5Zr to SS
42Zr and zirconiumdominated compositions from Zr
50SS to Zr
8SS were studied. The analog of the Fe
15Zr alloy was selected as target composition for waste streams dominated by steel cladding hulls based on phase composition, chemical and physical durability, and processing considerations. The analog of the Zr
8Fe alloy was identified as a possible target composition for waste streams dominated by Zircaloy cladding hulls. Both mixtures result in alloys having similar amounts of metallic and intermetallic phases but with one important difference: the metallic waste components in the fuel (Zr, Mo, Ru, Rh, Pd, Tc, etc.) and other metals in the cladding (e.g., Cr and Ni) form solid solutions in iron but are essentially insoluble in zirconium. Whereas the Fe-dominated alloys include an iron solid solution phase that serves to encapsulate the intermetallic phases in a physically durable monolithic waste form, the Zr-dominated alloys comprise loosely aggregated intermetallic phases without a binding metal phase. Because the initial focus was on steel-clad fuels, only the SS
15Zr mixed alloy

526

Metallic Waste Forms

was developed as a waste form for EBR-II wastes. However, these initial results suggest that a Zrdominated mixed alloy that is not well consolidated will not be useful as a waste form. 5.20.8.2

Steel

Radionuclide Distribution

The likely distributions of many elements in the component phases of the mixed alloy can be inferred from well-established binary phase diagrams, but the distribution of the radionuclides from the fuel in these phases must be measured because they are less studied and more important to performance than other waste components. Several SS
15Zr materials were made to study the dispositions of Tc and other radionuclides in the component alloy phases. The behavior of these radionuclides during corrosion was also measured for some of these materials. The results of tests and analyses with the materials made during development of EBR-II MWF alloys (both with and without added radionuclides) provide valuable insight into potential waste forms for the other waste streams that could be generated in the aqueous and pyrochemical processes. Samples of SS
15Zr were prepared with various amounts of added U, Np, Pd, Ru, and Tc. The capacity of the MWF to incorporate U is of interest, whether U is present as a contaminant or is added intentionally to down-blend 235U to below 20 mass %42 to meet criticality limits. The backscattered electron SEM image of a SS
15Zr
11U mixed alloy in Figure 1 shows U to be concentrated nonuniformly in the intermetallic phase.43 The darkest phase is the iron solid solution, the gray phase is the intermetallic phase without U, and the white phase is the intermetallic with U. Note the intimate mixing of the iron and intermetallic phases on a 100-mm scale. The polytypes detected in MWF materials during analyses are noted later, although the effects on performance are not known. Intermetallic phases with the general formula AB2 are commonly referred to as ‘Laves phases’; the A and B sites are usually occupied by several different elements in the Laves phases formed in the MWF. For example, the Fe sites in ZrFe2 are occupied by various amounts of Fe, Cr, and Ni in the MWF materials that have been studied by McDeavitt and coworkers.17 Some of the intermetallic phases can exist in cubic, hexagonal, and dihexagonal structures depending on the packing arrangement of identical (or very similar) layers. These related phases are referred to as polytypes.

U-rich intermetallic

U-poor intermetallic

200 µm

Figure 1 Backscattered electron scanning electron microscopy image of SS
15Zr
11U showing iron solution phase (black), U-poor intermetallic phase (gray), and U-rich intermetallic phase (white). Modified from Figure 4 of Keiser, D. D. Jr.; Abraham, D. P.; Sinkler, W.; Richardson, J. W. Jr.; McDeavitt, S. M. J. Nucl. Mater. 2000, 279, 234
244, used with permission. License number 2287090615691.

The compositions of the (mostly) ferrite and intermetallic phases of three MWF materials are summarized in Table 3.43 The value reported for each element has an absolute uncertainty of 2 at.%. Actinides were only detected in the ZrFe2-type intermetallic phase of SS
15Zr. The U-rich and U-poor designations indicate analyses of the brighter and the darker regions of the intermetallic phases, respectively. (The U-rich areas are brighter due to the higher electron density of U.) Differences in the relative concentrations of Zr and U in the U-rich and U-poor zones and the fairly similar Fe concentrations indicate that U substitutes at Zr sites rather than Fe sites in the intermetallic phase. Comparison of the SS
15Zr
5U alloy with the SS
15Zr alloy (see Table 4 in Keiser et al.18) showed relative increases of 5, 5, and 15 vol.% in the amounts of ferrite, ZrFe2 (cubic), and Zr6Fe23, and decreases of 4 and 20 vol.% in the amounts of austenite and ZrFe2 (dihexagonal). Although Zr6Fe23 could not be distinguished from ZrFe2 in the SEM analyses of most Ucontaining MWF materials, neutron diffraction revealed its presence in trace amounts and indicated that the lattice was contracted about 2%.43 Analysis of Zr6Fe23 that could be distinguished in the SS
15Zr
11U
0.1Pd
0.6Ru
0.3Tc material indicated that phase accommodates less U than the ZrFe2 phase. This is consistent with the preferred accommodation of U in cubic structures.

Metallic Waste Forms

Table 3

527

Phase compositions in SS
15Zr MWF materials

Additions to SS
15Zr

Composition (at.%)a

Phase

5U

2U
1Nb
1Pd
1Rh
1Ru
1Tc

11U
0.1Pd
0.6Ru
0.3Tc

Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves Zr6Fe23

Fe

Cr

Ni

Zr

U

67.5 44.9 49.1 67.6 43.2 46.5 65.9 49.3 53.3 58.1

23.2 3.3 6 22.3 3.3 4.1 26.8 3.1 6.5 11

5 25.7 18 3.5 22.5 17.9 3.1 18 12.1 9.5

Negligible 7.6 20.6 Negligible 10.9 19.1 Negligible 8 21.9 17.2

Negligible 17.2 1.5 Negligible 12.2 3 Negligible 19.3 2.7 1.7

a Absolute uncertainty of each value estimated to be 2 at.%. Data from Keiser, D. D. Jr.; Abraham, D. P.; Sinkler, W.; Richardson, J. W. Jr.; McDeavitt, S. M. J. Nucl. Mater. 2000, 279, 234
244, used with permission. License number 2287090615691.

Table 4

Phase compositions for actinide-bearing SS
15Zr alloys

Alloy

SS
15Zr
2U
2Pu

SS
15Zr
6Pu

SS
15Zr
10Pu

SS
15Zr
2Np

SS
15Zr
6Pu
2Np

Phase

Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves Ferrite U-rich Laves U-poor Laves

Composition (at.%)a Fe

Cr

Ni

Zr

U

Pu

Np

74 49 57 71 38 55 70 35 55 71 42 58 70 46 60

21 3 7 24 3 8 25 2 7 23 3 7 24 3 8

3 22 13 4 34 14 3 34 14 5 34 21 5 30 17

1 21 22 1 8 22 1 7 22 0.4 6 13 0.5 6 13

0.2 9 0.5













0.2 5 0.5 0.2 17 1 0.5 20 2


0.4 10 1










0.5 15 0.5 0.1 5 0.5

a Absolute uncertainty of each value estimated to be 2 at.%. Data from Keiser, D. D. Jr.; Abraham, D. P.; Sinkler, W.; Richardson, J. W. Jr.; McDeavitt, S. M. J. Nucl. Mater. 2000, 279, 234
244, used with permission. License number 2287090615691.

Table 4 shows the compositions of phases in SS
15Zr materials made with added U, Np, and Pu.43 Like U, both Np and Pu are contained almost entirely (and in the same vicinity) within the intermetallic. Selected area electron diffraction and EDS in an analytical transmission electron microscopy study indicated that the U was accommodated primarily in the cubic structure and segregated to the edges of the intermetallic adjacent to ferrite domains. The higher magnification transmission electron microscopy image in Figure 2 of a SS
15Zr
5U

material shows the actinides to be segregated at grain boundaries of the intermetallic phase.44 The distribution of U and other radionuclides within the intermetallic phase is not uniform on a millimeter scale.45 The measured concentrations of contaminants in the phases formed in the SS
15Zr
2Tc materials are summarized in Table 5.46 Similar amounts of cubic and dihexagonal ZrFe2 were detected in neutron diffraction analysis (19 vol.% dihexagonal and 22 vol.% cubic), but these were not distinguishable in

528

Metallic Waste Forms

SEM analysis. Compared with SS
15Zr, the SS
15Zr
2Tc material had more cubic ZrFe2, less dihexagonal ZrFe2, and more Zr6Fe23. Slightly lower lattice parameter values were measured for both ZrFe2 structures in the SS
15Zr
2Tc material than in SS
15Zr, which suggests that Tc was substituted at the Zr sites in that lattice. Small increases were seen in both the ferrite (0.104%) and the austenite (0.056%) lattices. Small amounts of Cr, Mn, Ni, and Si were detected in all phases, and Mo was detected in all but perhaps the Zr6Fe23 intermetallic phase. The assessed (theoretical) Tc
Fe binary phase diagram47 indicates that Tc is soluble in Fe(d) and Fe(g) (up to about 30%), but is essentially insoluble

Intermetallic grains

Steel

in Fe(a), which is the thermodynamically preferred phase below about 912  C. Discrete Tc-bearing phases were not detected in any of the samples that were examined. However, similar amounts of Tc were detected in the ferrite, austenite, and ZrFe2-type and Zr6Fe23-type intermetallic phases formed in a SS
15Zr
2Tc material.43 The Tc was probably dissolved in the Fe(d) formed initially at high temperatures and not excluded when it transformed to Fe(g) and then partially to Fe(a) as the material cooled. That is, the Fe(a) is probably supersaturated with Tc. Although austenite could be converted to ferrite by aggressive heat treatment, austenite was usually seen to remain as a metastable phase in as-cast MWFs made with austenitic steel.43 The authors suggested that ‘the solubility of technetium in the austenite and ferrite may be due to the presence of chromium and nickel in these iron solid solution phases.’ The alloy that was studied only contained about 1.7 wt% Tc. The capacity of the iron solid solution in the MWF for Tc is not known, but the assessed Fe
Tc phase diagram suggests that up to 66 wt% Tc can be accommodated in an FeTc1 + x intermetallic phase.

Actinide-enriched

5.20.8.3

10 µm

Figure 2 Scanning electron microscopy backscattered electron images of SS
15Zr
5U alloy. The dark gray areas are steel and the lighter areas are intermetallic phases. Several small white areas seen between some of the intermetallic grains (located by the arrows) are enriched in actinides. Modified from Figure 2b of Janney, D. E. J. Nucl. Mater. 2003, 323, 81
92, used with permission. License number 2287090487208.

Table 5 Host phase

Ferrite Austenite ZrFe2b Zr6Fe23 a

Electrochemical Tests

The initial corrosion rates (of bare metal surfaces) were calculated from linear polarization resistance measurements of several alloy compositions in various solutions. The results of tests in simulated tuff groundwater solutions and groundwaters spiked with NaCl at 20  C are summarized in Table 6 (data from Snyder et al.48). The simulated tuff groundwater solution SJ-13 had a pH of 9 and contains 109 kg m
3
3
3 HCO
Si, 18 kg m
3 3 , 51 kg m , Na, 34 kg m 2

3
SO4 , and 4.3 kg m Cl , and the concentrated
3 SJ-13 contained 12 700 kg m
3 HCO
3 , 5300 kg m
3
3 2
Na, 30 kg m Si, 22 kg m SO4 , and 727 kg m
3 Cl
and had a pH of 8.2. In general, the MWF alloys

Elemental compositions of phases formed in SS
15Zr
2Tc alloy Elemental concentration (at.%)a Cr

Fe

Mn

Mo

Ni

Si

Tc

Zr

23.5 17.8 5.7 8.4

66.2 66.4 45.4 51.7

1.6 1.9 1.6 1.0

1.2 0.7 0.7 Negligible

5.3 11.3 23.9 19.1

0.4 0.6 1.4 1.3

1.6 1.3 1.0 0.9

Negligible Negligible 20.2 17.6

Absolute uncertainty of each value estimated to be 2 at.%. Includes cubic and dihexagonal polytypes. Data from Keiser, D. D. Jr.; Abraham, D. P.; Richardson, J. W. Jr. J. Nucl. Mater. 2000, 277, 333
338, used with permission. License number 2287090753236. b

Metallic Waste Forms

529

Table 6 Corrosion rates measured in a simulated tuff groundwater solution SJ-13, SJ-13 spiked with NaCl, and concentrated SJ-13 at 20  C, in micrometer per year Material

SJ-13

SJ-13 þ 1000 kg m
3 Cl


SJ-13 þ 10 000 kg m
3 Cl


Concentrated SJ-13

SS
5Zr
2Nb
1Ru
1Pd SS
15Zr
1Nb
1Ru
1Pd
1Rh SS
20Zr
2Nb
1Ru
1Pd Type 316 stainless steel Alloy C22 AISI 1018

0.11  0.05 0.20  0.02 0.19  0.03 0.42  0.15 0.17  0.09 16.9  6.60

0.70  0.56 0.52  0.17 0.99  0.59 1.70  0.65 0.56  0.04 105  24

0.75  0.86 1.53  1.89 2.12  1.62 2.31  1.41 0.81  0.56 176  14

1.25  1.00 2.18  2.02 1.80  0.91 2.18  1.40 0.88  0.38 2.20  0.19

Data from Snyder, C. T.; Barnes, L. A.; Fink, J. K. Metal Waste Form Corrosion Release Data from Immersion Tests, Argonne National Laboratory Report ANL-04/15; Argonne National Laboratory: Argonne, IL, 2004.

are more durable than Type 316 stainless steel and became more reactive as the dissolved Cl
concentration was increased in both the spiked SJ-13 and the concentrated SJ-13 solutions. However, the corrosion rates increased by less than a factor of 12. The results in Table 6 are based on at least three linear polarization measurements, and the standard deviations indicate large uncertainties for some values. The effect of pH on the corrosion rates was likewise small, less than a factor of 20 for rates measured at pH 2, 4, 7, and 10.17,49 These test responses reflect the oxidation rate; the effects of the solution chemistry on the dissolution step may be different. A galvanic cell is formed when alloys having different corrosion potentials are coupled by direct contact or through an electrolyte solution. Possible galvanic couplings between the phases in the EBR-11 MWF and Alloy C22 were studied. Alloy C22 was expected to be the most stable metal in the waste package for the proposed Yucca Mountain disposal system and to act as the cathode in all galvanic couples. Tests were conducted to measure the coupling of Alloy C22 with SS
15Zr and SS
15Zr
1Nb
1Pd
1Rh
1Ru waste form compositions.49 The SJ-13 solution (pH 9) and the SJ-13 solution adjusted to pH 2 were used as electrolytes. Both materials were found to be electrochemically noble when coupled with Alloy C22. Companion tests coupling AISI 1018 steel with Alloy C22 showed the steel to be electrochemically active and preferentially oxidized. The galvanic potentials were about
200 and 62 mV for the SS
15Zr and SS
15Zr
1Nb
1Pd
1Rh
1Ru specimens and
600 mV for AISI 1018 steel. Similar results were obtained in the pH 2 solution, except that all potentials were shifted to higher values and the currents were slightly higher. The currents measured in tests with each MWF specimen were very small (but nonzero) and positive, indicating that both alloys are noble

relative to Alloy C22. Enhanced corrosion of the EBR-II MWF is not expected due to galvanic coupling with Alloy C22 or any other material (e.g., carbon steel) that is part of the waste package. 5.20.8.4

Corrosion Tests

Various test methods have been used to study the degradation behavior of alloys and the release of radionuclides. Initial tests indicated that degradation of the alloy occurs through a two-step mechanism of oxidative dissolution in which the metal atoms at the surface are first oxidized to form an oxide layer, and then the oxide layer dissolves. The oxidation step was studied with electrochemical tests and the combined oxidative dissolution processes were studied with static immersion tests, solution-exchange tests, and vapor hydration tests. The release of waste components as the steel and intermetallic phases corrode has been measured in immersion tests under a range of conditions. The metal waste form was subjected to test methods commonly applied to glass waste forms for direct comparison. Composition differences prevented direct comparisons of particular elements, but the releases of all components from the EBR-II MWF were well below the releases of elements from glasses.50 Other tests were conducted with saw and drill shavings of EBR-II MWF material for comparison with tests conducted with crushed glass to assess the possible use of ASTM C 1285 to track metallic waste form consistency.51 Particles reacted under those conditions for 2 years were examined with transmission electron microscopy (TEM) and found to generate oxide layers similar to those formed on alloys reacted at 200  C for shorter durations.52 This suggests that temperature can be used to accelerate corrosion progress.

Metallic Waste Forms

0.5 U Tc

0.4 NL(i) (g m–2)

In most tests with EBR-II MWF materials, the solutions in which the test specimens were immersed were either partially or completely replaced periodically to detect and minimize any effects that released components accumulating in the solution might have on the release rate. For example, the concentration of dissolved Zr may affect the dissolution of the intermetallic phase. Dissolution of EBR-II MWF materials was not stoichiometric, and U was found to be released to the greatest extent under most test conditions. The fractional release of U from SS
15Zrbased materials was typically about 10 times greater than the fractional release of Tc (i.e., when normalized to the amounts of U and Tc in the materials). The results of tests conducted with EBR-II MWF materials containing different amounts of U and Tc are shown in Figure 3.53 The normalized mass loss, NL(i ), is calculated as the mass of an element i that is released into solution divided by the mass fraction of element i in the corroding solid and by the surface area of the solid. This allows the releases of different elements in the same and different tests to be compared directly. The release rates of both U and Tc are high initially (i.e., during the first 50 days) but decrease and then become nearly constant over time. The initial decrease in the release rate is interpreted to correspond to the formation of oxide layers over the steel and intermetallic phases, which limit the continued corrosion (oxidation) of those metal phases. The negligible release at longer sampling times is probably due to the very slow dissolution of the oxide layers. To characterize the oxide layers, relatively thick layers were generated on the surfaces of exposed steel and intermetallic phases by reacting a SS
15Zr material in steam at 200  C for 91 days.54,55 Oxide material at the solution interface appears less compacted and has a higher oxygen content than oxide at the metal interface. The difference is probably due to dissolution of oxide at the outer surface, while corrosion of the metal generates oxide at the oxide layer/ metal interface. Oxygen must diffuse through the oxide layer to oxidize the metals and grow the inner layer, while the outer region of the oxide layer dissolves slowly into solution. As seen in the photomicrographs in Figure 4, both oxide layers adhere to the underlying metallic phase.56 Solids analyses indicate that both Tc and U are retained in the oxide layers.56 This implies that Tc is oxidized to insoluble TcO2 that is immobilized within the oxide layer and that further oxidation is required to release the soluble species TcO
4 . Immersion tests with various Tc-doped EBR-II MWF materials indicate that Tc is

0.3

0.2

0.1

0

0

50

(a)

100 150 200 250 300 350 400 Time (days)

0.4 U Tc

0.3 NL(i) (g m–2)

530

0.2

0.1

0 0 (b)

200

400

600 800 Time (days)

1000

1200

Figure 3 Releases of U and Tc measured in a long-term solution exchange tests with (a) SS
15Zr
4NM
2U
1Tc and (b) SS
15Zr
11U
0.6Ru
0.3Tc. Data from Johnson, S. G.; Noy, M.; DiSanto, T.; Keiser, D. D. Jr. In Proceedings of the DOE Spent Nuclear Fuel and Fissile Materials Management Meeting, Charleston, SC, Sept 17
20, 2002; American Nuclear Society: La Grange Park, IL, 2002; Waste Form Testing session.

released initially (presumably as TcO
4 ) when the surface is still active, but that the release decreases with time as the oxide layers form on the underlying steel and intermetallic phases. This is probably due to the slow dissolution of the oxide layers and their roles as diffusion barriers to oxygen infiltration. Additional evidence is needed to demonstrate that the passivating effect of the oxide layers will be maintained over geological time scales. A series of immersion tests was conducted to characterize the effects of temperature, pH, and Cl
concentration on the release of U from a SS
15Zr
10U alloy.57 A modification of the ASTM C 1220 method was used wherein monolithic test specimens were immersed in a pH buffer solution spiked with 0, 1000, or 10 000 kg m
3 NaCl at a specimen surface

Metallic Waste Forms

531

1

Oxide layers

1

2

2

3

Steel

Oxide layers

100 nm

(a)

(b)

Intermetallic

100 nm

Figure 4 Transmission electron microscopy photomicrographs of oxide layers formed over (a) steel phase and (b) intermetallic phase of MWF reacted at 200  C for 91 days. Modified from Dietz, N. L. Transmission Electron Microscopy Analysis of Corroded EBR-II Metallic Waste Forms; Argonne National Laboratory Report ANL-05/09; Argonne National Laboratory: Argonne, IL, 2005.

area-to-solution volume ratio of 10 m
1 at 50, 70, and 90  C. The entire solution was removed for analysis and replaced with fresh buffer solution after 14, 28, and 70 days. Although the test conditions are not representative of disposal conditions, the tests provide a convenient means to measure the effects of temperature, pH, and Cl
concentration on the release of U as the oxide layer forms. These dependencies can be incorporated into a more realistic model that more accurately represents the disposal environment. Figure 5(a) and 5(b) shows the results of tests conducted at 50  C in pH 4 and pH 8 solutions spiked with 0, 1000, and 10 000 kg m
3 NaCl, where the normalized mass losses based on the cumulative mass of U released to solution are plotted against the reaction time.57 Note that each test under a particular set of pH/Cl/temperature conditions was conducted with a single specimen to eliminate the variance in U concentration at the surface. These results show significant effects of pH, Cl
concentration, and time on the amount of U released. The release slows with time such that release during the initial 14 days accounts for about half the cumulative release over 70 days. The release of U is of primary interest because U is the most rapidly released radionuclide and is used to provide a conservative bound for the release rates of all radionuclides in performance assessments. The average release rates based on the cumulative values over 70 days are plotted against the final pH in Figure 5(c) for all tests at 50  C, where NR(U) ¼ NL (U)/70 days.57 The variance in the results of replicate tests may reflect the nonuniform distribution of U in the alloy and between test specimens. The release of U is faster than most other elements in most tests. Figure 5(c) shows that the release rates do not have a simple pH dependence.

5.20.8.5

Corrosion Models

The corrosion, degradation, and release of radionuclides from the EBR-II MWF is hypothesized to occur through an oxidation
dissolution mechanism in which metallic components exposed at the surface are first oxidized and form an oxide layer and then the outer layer dissolves. Formation of oxide layers on the exposed surfaces of the steel and intermetallic phases slows the release of all components to solution. The release rates of all radionuclides were modeled to equal that of the most efficiently released constituent in each particular test regardless of whether the component was released from the steel or intermetallic phase. In this regard, the waste form was modeled to be a homogeneous single phase that dissolved stoichiometrically. An empirical model for EBR-II MWF degradation was developed based on the results of electrochemical and dissolution tests with several materials.48,58,59 That model incorporated the dependencies of the release rate on temperature, pH, and Cl
concentration that were measured in test environments ranging from pH 2 to 12, 25 to 90  C, and about 0 to 10 000 kg m
3 Cl
. The model is based on the general concept that the initially bare alloy surface becomes covered with an oxide layer that slows the releases of radionuclides and matrix components to solution. The layer is modeled to protect the EBR-II MWF surface from continued corrosion, but credit is only taken for the length of time this was observed in the tests that were used to measure the dependencies. At the end of that period, the passivating effect of the layer is modeled to vanish and then redevelop at the same rate during the next period. That is, the layer is modeled to periodically spall from the underlying metal and

532

Metallic Waste Forms

pH 4

10

1

0.1

0

20 40 60 Test duration (days)

(a)

pH 8

1

NL(U) (g m–2)

NL(U) (g m–2)

100

0.1

0.01

0.001 0

80

20 40 60 Test duration (days)

(b)

80

NR(U) (g m–2 days–1)

0.1

0.01

0.001

0.0001 2

4

6

(c)

8 pH

10

12

14

▪

Figure 5 Results of static tests at 50  C in solutions with (●) 10 000 kg m
3 Cl
, ( ) 1000 kg m
3 Cl
, and (⧫) without added NaCl: NL(U) at (a) pH 4 and (b) pH 8, and (c) average release rates NR(U) through 70 days for tests at various pH values in solutions with 1000 kg m
3 Cl
. Data from Ebert, W. L.; Lewis, M. A.; Barber, T. L.; DiSanto, T.; Johnson, S. G. Static Leach Tests with the EBR II Metallic Waste Form; Argonne National Laboratory Report ANL-03/29; Argonne National Laboratory: Argonne, IL, 2003.

then reform. This moderation of the passivating effect is due to the absence of direct evidence regarding the long-term stability of the oxide layers. The model presumes a common time dependence of the oxidation and dissolution reactions for releasing constituents to solution and growing the oxide layer. If the corrosion and release rates decrease exponentially due to growth of the oxide layers, then the cumulative release of all constituents should follow the logarithmic form: Cumulative constitute release ¼ a  lnð1 þ bt Þ ½3 where a and b are fitting parameters and t is time. For sparingly soluble oxides, the thickness of the layers should increase following the same equation. The fitting parameters a and b have the following significance59: the product a  b gives the initial release rate prior to formation of the layer, 1/b gives the

characteristic time required to passivate the surface (i.e., until the release rate becomes negligible compared to experimental uncertainty), and a represents the extent of corrosion necessary for the layer to significantly slow the release to solution. The metallic waste form does not dissolve stoichiometrically in laboratory tests because of differences in the oxidation rates of individual elements, their solubility limits, and their sequestration in alteration phases. Although there is no experimental evidence that the slowing effect of the oxide layers will not persist over long times, for example due to the layers sloughing off, application of the model to the EBR-II MWF was conservatively limited to 1-year periods, which is the longest duration over which most test series were conducted. The average release rate over the time interval Te that the oxide layer remains an effective barrier is:

Metallic Waste Forms

Release rate ¼

amax  lnð1 þ bt Þ Te

½4

The term amax is used to indicate that the element released the fastest under particular conditions was used to fit the model. Values of the model parameters amax and b capture the dependencies on temperature ( C), pH, and the Cl
concentration (kg m
3). Expressions were determined by assuming a simple linear or quadratic dependence on these variables, using a least-squares fit of the experimental data, and minimizing the number of free parameter combinations. The dependencies, which were determined by fitting experimental results,48,53 are given in eqns [5] and [6] (see Bauer et al.58,59): ln ðb  amax Þ ¼
0:10105 þ ð0:015112 þ 5:8201  10
6  ½Cl
Þ  T
0:69848  pH ½5 and ln amax ¼ 7:9812 þ ð2:3938  10
4  ½Cl
Þ
1:2273  pH

½6

The fitted dependencies given in eqns [5] and [6] are not unique relationships, but they provide physically sensible fits for conditions relevant to a geological disposal site. The modeled rate increases with increasing temperature and Cl
concentration, and decreases with increasing pH. Equation [4] can be rewritten in terms of the fitted dependencies as   amax  ln 1 þ b amaxamax t ½7 Bounding release rate ¼ Te where the cumulative release over the interval Te can be used to provide a conservative time-independent rate that can be compared with the rates calculated for other waste forms. Note that the rate equation given in eqn [7] represents release from a metallic waste form having the specific phase composition and chemical composition for treated EBR-II spent fuel. Rate equations for other compositions would require conducting separate suites of tests to determine parameter values for each composition. 5.20.8.6

Repository Model

The EBR-II waste forms were not included in the performance assessment calculations conducted in support of the Yucca Mountain repository license application for construction. To evaluate the likely

533

acceptability of the EBR-II MWF, the source term for metal waste form degradation using eqn [7] was compared to that for HLW glass.60 The radionuclide release rates used in performance assessment calculations are calculated as the product of the release rate from the waste form degradation model, the radionuclide inventory, and the reacted surface area.61 The products of the degradation rate, inventory, and surface area for each waste form can be compared to estimate the impact of replacing some of the HLW glass with EBR-II MWF on repository performance. For both the EBR-II MWF and the HLW glass, the release rates of all radionuclides are modeled to be equal to the waste form degradation rate. Solubility and transport limits are based on individual radionuclides and independent of the source. The effects of temperature and pH on the degradation rate of the EBR-II metal waste form can be compared to the temperature and pH dependence of the defense HLW glass degradation model developed for use in the Yucca Mountain performance assessment calculations. Because the glass model does not include a term for the effects of Cl
(HLW glass dissolution is not affected by dissolved Cl
), the metal waste form model rates were calculated using the highest anticipated Cl
concentration for comparison with the glass model. The Cl
concentration is treated as a constant rather than a variable for the purpose of comparison with the glass model. The concept for disposing EBR-II wastes is to codispose ceramic waste forms with metal waste forms in the same canister to distribute the weight of the metal waste forms among several canisters and waste packages. Degradation of the ceramic waste form would provide Cl
to groundwater contacting the metal waste form. It was estimated that a maximum of 620 kg m
3 Cl
could be dissolved in the water inside a breached canister based on the volume and surface area of ceramic waste form available to infiltrating water. The model predictions can also be compared with the rates measured in modified ASTM C 1220 tests conducted with leachants spiked with NaCl (these were discussed in Section 5.20.8.4). The dissolution rates of a SS
15Zr
10U alloy were measured in tests conducted at 50, 70, and 90  C over a range of pH values using buffer solutions spiked with NaCl to attain 1000 kg m
3 Cl
. The cumulative amount of U released over 70 days was used to calculate the average dissolution rate for comparison with the HLW glass and EBR-II metal waste form models.60 The rates from tests conducted at 50 and 90  C are

534

Metallic Waste Forms

plotted in Figure 6 along with the lines showing the maximum rates from the defense HLW glass models at these temperatures over the full pH range. The dashed lines in Figure 6 show the rates calculated with the empirical metal waste form model given in eqn [6] calculated at 50 and 90  C with 620 kg m
3 Cl
and a time interval Te ¼ 1 year; the rates calculated with 1000 kg m
3 Cl
are only slightly higher. The EBR-II MWF model is representative of the measured rates in acidic and neutral solutions, but underestimates the rates measured in alkaline solutions by more than an order of magnitude. The poor fit in alkaline solutions calls into question the simple pH dependence that is used in the EBR-II MWF model. The key finding demonstrated in Figure 6 is that the rates calculated with the defense HLW glass model bound both the rates calculated with the EBR-II MWF model and the rates measured in the modified ASTM C 1220 tests (solution exchange tests) over the entire pH range, including the rates in alkaline solutions. One exception is the rate measured in the test at 50  C and pH 9, which was slightly higher than the glass model. It is important to note that the MWF model pessimistically ignores the likely long-term stability of the oxide layers that will probably protect the MWF surface throughout the service life of the repository. The oxide layers

3

log rate g (m–2 days–1)

2

MWF immersion 90 ⬚C MWF immersion 50 ⬚C

HLW glass model at 90 ⬚C

1

HLW glass model at 50 ⬚C

0 –1

5.20.8.7 Metallic Waste Form Product Consistency

EBR-II MWF

–2 model at 90 ⬚C –3 –4 0

EBR-II MWF model at 50 ⬚C

2

4

likewise have a minor effect on the rates determined from the short-term modified ASTM C 1220 tests. In effect, the EBR-II MWF model assumes that the oxide layers disappear and reform on an annual basis, such that the average degradation rate from the model becomes increasingly conservative over time. The surface area of HLW glasses available for degradation in performance assessment calculations is based on the dimensions of the glass pour canisters and a cracking factor. The exposed surface area of an average HLW glass log is taken to be 30 m2 for comparison with the EBR-II MWF, which will be cast as ingots having a right cylinder geometry 0.35
0.41 m (14
16 in.) in diameter and 0.05
0.13 m (2
5 in.) thick. One or two ingots will likely be codisposed with two ceramic waste form monoliths in a disposal canister. The ingots are not expected to fracture due to cooling or impact, so the geometric surface area represents the maximum surface area that can be exposed to water. The surface area of a representative ingot 0.4 m (16 in.) in diameter and 0.1 m (4 in.) thick is about 0.38 m2, so the surface area of two EBR-II MWF ingots in a breached canister is about 0.76 m2. This is about 2.5% of the glass surface area. The total inventory of radionuclides in the EBR-II MWF is only about 0.1% of the total inventory of HLW glass, on a per canister basis. The predominant radionuclides are 1.92  1012 Bq 63Ni, 7.77  1011 Bq 99 Tc, 9.99  1010 Bq 60Co, and 6.66  1010 Bq 59Ni. (The total inventory of the EBR-II ceramic waste form is about 16% of the HLW glass inventory, on a per canister basis.) Based on comparison of the combined release rate, inventory, and surface area, the performance of the EBR-II MWF is conservatively bounded by the performance of HLW glass in total system performance assessments.

6

8

10

12

14

pH Figure 6 Comparison of measured MWF degradation rates (data points), rates from empirical MWF model (dashed lines), and rates from HLW glass degradation model (solid lines) at 50 and 90  C. Data from Ebert, W. L. Testing to Evaluate the Suitability of Waste Forms Developed for Electrometallurgically Treated Spent SodiumBonded Nuclear Fuel for Disposal in the Yucca Mountain Repository; Argonne National Laboratory Report ANL-05/ 43; Argonne National Laboratory: Argonne, IL, 2005.

The objective of tracking the consistency of waste form products is to link the controls applied to waste form production with predictable performance in a disposal system (see Section 5.20.5.2). The narrow range of waste stream compositions anticipated from the EBR-II spent fuel inventory simplified the task of tracking waste form product consistency. The control limits placed on the feed compositions are 1
11 mass% total U and 5
20 mass% Zr with targets of 10 mass% total U and 15 mass% Zr. The composition Fe
15Zr results in nearly equal amounts of the steel and Fe2Zr intermetallic phases

Metallic Waste Forms

being formed, and variations in the amount of Zr in the alloy will change the relative amounts of each phase. Acceptable performance requires that the waste form contains at least the minimum amount of the Fe2Zr intermetallic phase required to sequester the actinides present in the waste. It was determined that 5 mass% Zr in the alloy is adequate to produce enough intermetallic phase to accommodate the maximum amount of actinides that could be present in the waste. Most waste streams will contain <5 mass % Zr, and additional Zr must be added to generate an adequate volume of intermetallic phase. The addition of Zr is a critical step in waste processing. Because Zr is not soluble in the steel phase and the actinides substitute for Zr in the intermetallic,44 the Zr content of the waste form provides a reliable measure of the amount of intermetallic phase that will form. It is expected that waste streams from several electrorefining runs will be blended to control the U enrichment in the EBR-II MWF. The Zr content will be tracked, but not optimized. In practice, 5 mass% Zr can be added to any waste mixture to ensure that the total Zr will exceed the minimum content of 5 mass% by the amount of Zr present in the fuel wastes stream. Measurement of the Zr content can be used to track waste form consistency by verifying enough Zr is present.42 Based on the Fe
Zr binary phase diagram and scoping tests, about 30 mass% Zr can be accommodated in the phase structure of the waste form, although the temperature required to reactively dissolve the Zr will increase with the Zr content of the mixture. However, scoping tests showed that alloys with that much Zr were brittle, so an upper limit of 20 mass% Zr is imposed to provide enough of the steel phase to maintain the physical integrity of the alloy. As part of this approach, it must be demonstrated that the composition analysis of small samples taken from waste forms having the fine-grained microstructure of the SS
15Zr alloy will provide a sufficiently accurate measure of the Zr content. This can be done by comparing the analytical results of various sample sizes to bulk compositions based on known formulations or analysis of larger aliquants.

5.20.9 Summary Metallic waste forms are appropriate for waste streams that include significant amounts of metallic components or contain radionuclides that are most effectively processed and immobilized as metals. The current design of metallic waste forms is based on the

535

compatibility of the waste stream with the targeted assemblage of phases that form from the waste and added metals. Other metals can be added to facilitate melting the waste metals or forming phases that effectively immobilize the waste components in a consistent assemblage of phases. Binary phase diagrams provide useful insights regarding the likely melting temperatures, eutectic melting compositions, likely phase compositions and assemblages, and solubilities for various mixtures of wastes and additives. The degradation behaviors of radionuclidebearing alloy phases must be modeled to predict radionuclide release and migration over very long times in performance assessment calculations. These calculations are used to demonstrate that regulatory requirements will be met over the service life of the disposal system as the waste forms degrade. This requires testing to understand and quantify radionuclide release rates under the range of conditions anticipated in a disposal facility and those possible under extreme conditions. Long-term predictions based on mechanistic models are more reliable than those based on empirical models, but mechanistic models are very difficult to develop. An empirical model with mechanistic underpinnings has been developed for the specific metal waste form composition formulated and designed to immobilize metallic wastes from the electrometallurgical treatment of spent EBR-II fuel. Tests and analyses of the EBR-II MWF provide valuable insight regarding the disposition of radionuclides in component phases and their releases as those phases degrade. They also show the difficulties of quantifying a complicated and coupled corrosion process. Development of a mechanistic model for the degradation of alloyed waste forms is probably the most important research need, especially a term to represent the effect of the phase composition. Qualification of metallic waste forms for disposal will likely require evidence that acceptable waste form performance (i.e., controlling the radionuclide release) can be expected based on how the waste form is produced. That is, controlling the gross composition of waste streams and additives and controlling the processing conditions (processing temperature, furnace atmosphere, etc.) will result in a waste form with predictable phase assemblage and radionuclide disposition that can be related to waste form performance. Because a very large number of waste forms will be made over long times with varying waste stream compositions, controlling waste form performance through processing controls is a very

536

Metallic Waste Forms

important aspect of waste form design. This is necessary to provide confidence that the performances of all waste forms produced are bounded by the radionuclide release rates used in the performance assessment calculations. A variety of test methods is required to provide the range of data needed to determine the radionuclide release mechanism(s), the influence of environmental variables and waste form composition on the release rate (both the chemical and phase composition), develop and parameterize a model for calculating radionuclide release over very long times, and establish process control limits for making consistent products. The electrochemical and immersion test methods used to characterize the EBR-II MWF have provided an extensive database to support its qualification for disposal in the proposed Yucca Mountain repository. Lacking a mechanistic model for radionuclide release, the approach taken was to show that the impact of disposing EBR-II waste forms would be dwarfed by the impact of commercial fuel and HLW glass and bounded by performance assessment calculations. A mechanistic model of radionuclide release is an important goal for advancing the development of metal waste forms. From the conceptual model of sequential oxidation and dissolution steps, new test methods coupling electrochemical techniques with test methods that accelerate corrosion are probably needed to characterize radionuclide release under a range of conditions. The oxidation step may also complicate coupling waste form degradation with the transport models used in performance assessments. For example, the need to take into account galvanic couples within the waste from and between the waste form and other metals may require additional terms in transport models. Methods to accelerate metal corrosion processes are also needed to help understand long-term performance, support waste form modeling, and lead to acceptance for disposal. Testing and modeling completed to date indicates that metallic waste forms provide a preferable option for several existing and anticipated waste streams from fuel treatment and recycling processes. Several challenges remain in development of metallic waste forms for high-level and low-activity radioactive waste streams in the areas of performance and consistency testing, developing waste form degradation and radionuclide release models, and integrating those into performance assessments. Innovative research in electrochemical measurements and modeling is needed to support continued

model development and waste form formulations, and replace the current empirical approaches with sound mechanistic models. (See also Chapter 3.01, Metal Fuel; Chapter 5.14, Spent Fuel Dissolution and Reprocessing Processes; Chapter 5.16, Spent Fuel as Waste Material; Chapter 5.18, Waste Glass and Chapter 5.19, Ceramic Waste Forms).

Acknowledgments The submitted manuscript has been created by UChicago Argonne, LLC, Operator of Argonne National Laboratory (‘‘Argonne’’). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02-06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.

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Goff, K. M.; Teske, G. M.; Howden, K. L.; Johnson, T. A. Electrometallurgical treatment of EBR-II spent fuel. In Waste Management’04 Conference, Tuscon, AZ, Feb 29
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