The following article appears in the journal JOM,
52 (9) (2000), pp. 26-29

Radioactive Waste: Overview

Materials Issues in Nuclear-Waste Management

Man-Sung Yim and K. Linga Murty


In this article, materials issues in the management of nuclear waste, including its generation, processing, storage, transport, and disposal, are examined for low-level and high-level waste, with an emphasis on the aspects of their immobilization and long-term isolation. Selecting materials for low-level and high-level waste form and containers is reviewed, and the long-term performance issues with these materials as barriers to nuclide migration or release are discussed.


Many activities dealing with radioactive materials produce nuclear wastes, including civilian nuclear power programs (nuclear powerplant operations and nuclear fuel-cycle activities), defense nuclear programs (nuclear weapons production, naval nuclear reactor programs, and related R&D), and industrial and institutional activities (scientific research, medical operations, and other industrial uses of radioisotopic sources or radiochemicals). To minimize the potential adverse health impacts to people during the entire lifetime of the radionuclides involved, nuclear waste must be carefully and properly managed. The scope of nuclear-waste management encompasses generation, processing (treatment and packaging), storage, transport, and disposal.

To avoid complexity in covering a wide variety of materials, nuclear waste is classified into several categories. According to the U.S. classification system, these are high-level waste (HLW), transuranic (TRU) waste, uranium mill tailings, and low-level waste (LLW). HLW includes the spent nuclear fuels withdrawn from nuclear power reactors following utilization for power generation and the highly radioactive liquid/solid waste resulting from the reprocessing of spent fuel. LLW includes all radioactive wastes other than spent fuel, HLW, TRU waste, and uranium mill tailings. HLW and LLW represent the two most important types of nuclear waste. More than 99% of the total radioactivity in nuclear waste is contained in HLW,1 while, in terms of volume, LLW takes up the biggest share (about 85% of the entire nuclear waste generated).

Materials issues are important during the entire process of nuclear-waste management— performance of the materials used in nuclear waste management determines its safety/hazards. Since the safety of nuclear-waste management relies mainly on the immobilization of radioactive constituents and long-term isolation of these from the biosphere, materials issues with the immobilization and long-term isolation of nuclear waste are particularly important. As the requirements for materials performance are dependent upon the type of nuclear waste, this article concentrates on LLW and HLW.


The proper selection of materials or quality control of materials manufacturing can lead to a reduction in radioactivity in nuclear waste. For example, reducing the amount of nitrogen impurities in nuclear fuels, both in the sintered UO2 pellets and Zircaloy (zirconium alloyed mainly with tin and iron) cladding tube material, will lead to the reduction of C-14 activity in the spent fuel HLW. Use of Inconel-690 over Inconel-600 as a steam-generator tube material in pressurized water reactors can reduce the radioactivity in LLW from nuclear power plants2 due to the lower content of cobalt in Inconel-690. The use of low-cobalt-impurity, iron-based, hard-facing alloys instead of the Stellite alloy can have a similar effect.2

Figure 1

Figure 1. A spent fuel shipping cask (IF-300).3

Spent nuclear fuels can be reprocessed for recycling uranium or plutonium. This is achieved through the dissolution of fuels in nitric acids and selective solvent extraction using special organic solvent. HLW from the spent-fuel reprocessing can be treated with calcination or vitrification. In nuclear power plants, LLWs are produced through various treatment processes for radioactive liquids and gases prior to discharge to the environment;3 these processes include demineralization, filtration, or evaporation. Thus produced waste can be dewatered; compacted; supercompacted; incinerated; solidified; or manipulated through other special treatment processes, such as vitrification, thermal decomposition (molten-metal treatment or steam reforming), or chemical decomposition (supercritical water oxidation). The systems used for the processing and/or treatment of nuclear waste should be designed with careful consideration of materials in order to minimize the hazards during processing and to ensure the compatibility and the durability of the systems.

In transporting HLW, special transportation casks (e.g., Figure 1)3 are designed to provide physical containment, radiation shielding, heat removal, criticality protection, and theft protection. Materials must be carefully selected to provide the required performance.


A key consideration in nuclear-waste management is the development of a highly durable waste package (including the waste form and the surrounding container barriers) that ensures the long-term stability of materials and the isolation of radioactivity. Use of durable waste packages is also important in the interim storage of nuclear waste. In the waste package, the waste form represents the first and foremost barrier to the release of radionuclides from nuclear waste.

The functions of a waste form are to provide physically, chemically, and thermally stable form; immobilize the radio-active materials (slow release when contacted with water); and resist leaching, powdering, cracking, and other modes of degradation. The waste form into which the radionuclides are incorporated has a considerable impact on the manner and degree to which they are retained in the waste form.

In the case of LLW, typical waste forms include cement, polymers, glass, metals, sorbent materials, and various unsolidified waste. Cement is inexpensive, simple to fabricate, and stable in most environments. Various synthetic organic polymers (e.g., vinyl ester styrene) can be applicable at slightly higher fabrication cost when cement solidification is difficult (e.g., with ion-exchange resins). Glass, as the end-product of waste vitrification, is a much more stable waste form compared to cement or polymers. Vitrification can handle a wide variety of waste feeds with large volume reductions. However, some key radionuclides can be volatilized and released instead of being immobilized in the process.

In the United States, a large fraction of commercial LLW is disposed of without solidification. This is partly due to economic penalties in the disposalrate structure for increased waste volume caused by solidification. Vitrification is not widely practiced with its potentially higher cost involved. The U.S. Nuclear Regulatory Commission requirements for LLW form4 stress the structural stability of waste form, but do not necessarily promote the use of technology to enhance the long-term durability of waste forms. The majority of radionuclide inventory in disposed LLW currently exists in activated metals, dewatered ion-exchange resins, dewatered filters, and mixed trash (plastics and papers). In this case, the necessary stabilization or isolation of waste is mainly achieved by the use of waste containers.

In the case of HLW, activity contents (and their average half-life) are much greater than the LLW, thus requiring much more stringent control on waste-form performance. Selection of a waste form for HLW has been the subject of research for many years. The two primary HLW forms in the United States are spent fuel and borosilicate glass. Other major waste forms that have been considered include Synroc and tailored ceramics.5 Unreprocessed spent fuel can serve as a final waste form6 since the ceramic UO2 matrix retains the nonvolatile fission products. The Zircaloy cladding, if intact, provides an additional barrier. Spent fuels are not vitrified in the United States.

Glass currently represents a smaller volume of HLW as compared to spent fuel in the United States.; however, it is the majority of waste in United Kingdom, France, Germany, Japan, and presumably Russia.7,8 The emphasis on the clean-up of the defense sites has also placed a premium on glass technology that is being used to immobilize contaminated materials and sites. The waste loading determines the volume, radiation dose, and thermal history of the waste form. The borosilicate glass is based on the Na2O-B2O3-SiO2 ternary system, with the waste loading ranging up to 30%. The oxide of boron is used to reduce the melting temperature without a large sacrifice in leachability.9 A wide range of fission products can be accommodated due to the geometrical flexibility afforded by disordered amorphous structure. Worldwide efforts have been made since the 1950s for the preparation of a practical, waste-immobilization scheme in this glass, including the building and operation of fully engineered smelters for the remote production of nuclear-waste glass.10 Thus, there is tremendous momentum behind borosilicate-glass technology in many countries.

Synroc is a titanium-based, polyphase, ceramic material made of specific natural minerals, including TiO2, ZrO2, Al2O3, BaO, and CaO.11 These minerals have the capacity to incorporate a wide range of radionuclides present in HLW into their crystal structures as solid solutions with approximately 20% waste loading. Tailored ceramics refer to multiphase, crystalline, ceramic nuclear-waste forms, which are similar to Synroc but vary depending on the compositions of the given waste (hence, tailored). Tailored ceramics are known to have slightly lower waste loading than Synroc (5– 15% waste).

The waste form is subject to various corrosive and degradation mechanisms that can decrease the effectiveness of the material to act as a barrier to nuclide migration. Corrosion and dissolution, as well as radiation damage, can greatly reduce the desired material properties of the waste form. For the spent-fuel waste form, corrosion is a result of oxidation and hydriding of Zircaloy clad surrounding the spent-fuel pellets.12 As a result, the thermal conductivity of the clad decreases, increasing the thermal load on the waste form. Also, a decrease in the strength of the clad wall occurs due to the thinning of the clad.

Radiation hardening is also a key player in the degradation of spent fuel through creation of point, line, surface, and volumetric defects that inhibit dislocation movement. This can be very detrimental to spent-fuel integrity when the resulting loss of clad ductility is combined with effects such as fuel-pellet swelling or pellet-cladding mechanical interaction (PCI) of the clad with highly reactive fission products, such as iodine, cadmium, cesium, and molybdenum.13

Other issues of importance in long-term repository performance assessment are the inhomogeneous distribution of fission products in the spent fuel and long-term dissolution of the UO2 fuel matrix.14 Most of the release of soluble radionuclides occurs very early in the exposure to water from gap and grains,6 and their inventory is a function of fuel burn-up.15 The long-term release of most actinides is controlled by dissolution of the UO2 matrix. Under reducing conditions, solubility controls the dissolution rate. Under oxidizing conditions (i.e., with water radiolysis), solubility becomes high and the presence of oxidizing species and complexing agents mainly control the kinetics of oxidative dissolution. While the integrity of the spent fuel cladding is of concern during dry storage,16 the cladding barrier is not counted in the repository performance assessment; this provides further conservatism in the repository-performance assessment. In addition, with the current trend to attain higher burn-ups, the performance of spent fuel as a barrier will be worse.

Glasses suffer from high-temperature thermal environment in a waste package. The extremely high temperatures in waste forms can cause devitrification and/or dissolution of glasses, which in turn result in the release of nuclides from the glass or crystal matrix. However, most (if not all) national waste-management programs7 specify extended cooling periods for the spent fuel in order to reduce the temperature excursion within a repository during post-emplacement periods. Other concerns with the glass waste form are the radiation stability and long-term leaching through corrosion. Radiation interactions in glass lead to atomic displacements (by heavy particle radiations) or chemical effects (from beta or gamma rays) enhancing the corrosion rate through disordering or radiolytic processes. Atomic displacements lead to volume/density changes; stored energy change and/or its release; and associated crystallization, hardening, and fracture. The long-term release of glass constituents is controlled by the combined effects of the diffusive limited transport of dissolved species through the surface alteration layer and the crystallization of secondary phases formed on the surface of the glass.14,17


Waste containers provide protective barriers against physical and chemical stresses during transportation, interim storage, and disposal. Since some of the radionuclides in LLWs are short-lived, LLWs are classified into different subcategories (depending upon the activity contents and half-life of the nuclides in the waste), and different stabilization requirements are used for different subcategories accordingly.3 In the case of HLW, separate casks are used for transportation, and special emphasis is given to the disposal containers to ensure long-term isolation of the waste. The key performance parameter is the resistance to environmental attack (chemical performance). Mechanical performance, thermal/ neutronic performance, compatibility with other materials, fabricability, and previous experience, as well as cost, are also taken into account.18,19

Figure 2

Figure 2. A schematic of a cast for disposing of LLW.3

Figure 3

Figure 3. A schematic of a waste container for HLW.19

Waste containers used in the burial of LLW include carbon-steel drums, liners, and boxes and high-integrity containers (HICs). They are placed in a disposal facility with either soil or cement backfills. Carbon-steel containers are inexpensive, but can undergo both uniform corrosion and pitting corrosion within the soil and cemented systems. The life-time of carbon-steel containers in a disposal system is expected to be short (few years or longer); thus, steel is used primarily for the disposal of short-lived nuclides. HICs represent a more durable LLW container and are used for the disposal of long-lived high-activity waste. HICs can be made from corrosion-resistant metal alloys, reinforced concrete, high-density polyethylene (HDPE), or polymer-coated metals. Carbon-steel drums, boxes, or HICs can also be used with concrete modules to improve the long-term integrity of the package, as shown in Figure 2.3 The type of HIC expected to be widely used in the future is a combination of both an HDPE and a concrete overpack. This type of HIC is expected to fail eventually by degradation of the concrete casing and creep of the high-density polymers. HICs are required to have a minimum lifetime of 300 years by current Nuclear Regulatory Commission regulations.20

Waste containers for HLW might include a metallic canister and an overpack (Figure 3).19 A canister is the immediate container surrounding the waste form. It serves as an aid to handling and transportation during processing and maintains geometric integrity for cooling; it also works as a short-term barrier against leaching and nuclide migration in disposal. Canisters cannot be considered a good barrier, especially if the hot glass material is put inside, since it is difficult to predict the thermal history of the metal. An overpack is a hermetically sealed, high-integrity protective barrier and provides the main protection for the desired long-term isolation during disposal. For the design of an HLW package, it is recommended to avoid dependence upon barriers in which geometric considerations may play a role in degradation modes (e.g., film growth, crevice corrosion, or flaw distribution) since it is difficult to predict the geometries of failures and nonuniform attacks. An example of an HLW package including multiple containers is shown in Figure 2 of the article by D.B. Bullen et al. in this issue.21

HLW package materials will be subjected to harsh environments and various kinds of physical and chemical stresses. Long-term exposure of materials in a repository could result in significant alterations in materials during the service life. The presence of HLW inventory will lead to elevated temperatures and furnish high levels of radiation. The host media for the repository can be sources of oxygen, water, and other species that can be aggressive in altering the nature of the materials used for containment of the waste. Due to these stresses, various forms of degradation can be expected. One form of degradation is exposure to high-temperature gases that contain oxygen. These gases can cause oxidation of the materials, resulting in a loss of structural integrity, or could encourage future oxidation upon exposure of the container to groundwater. Another form of corrosion is due to the aqueous environment, including uniform corrosion, localized corrosion, galvanic corrosion, intergranular corrosion, and stress-corrosion cracking. Also, radiation effects, such as radiation hardening and embrittlement, enhanced diffusion, and enhanced creep rate, must be taken into account since all materials are susceptible to these phenomena.

Candidate materials for HLW canisters and overpacks are generally metals such as copper, iron, stainless steels, titanium alloys, and nickel-based alloys.18 Certain ceramics or graphitic materials may also be considered. Copper is one of the few metals found in its native state in the geological environment. Studies on native copper deposits and archaeological artifacts indicate very good environmental durability.22 However, it is known to be poor in brine as well as in radiation environment.23 Iron provides good predictability since much is known about the material. It is not very corrosion resistant, but is less prone to catastrophic failures. Both the natural occurrences and archaeological analog exhibit similar low iron corrosion rates. Carbon steel was considered as the structurally strong outer layer for corrosion allowance material in the current U.S. Department of Energy’s (DOE’s) HLW package design.19,24 Titanium alloys are mechanically strong and possess good corrosion resistance. However, they can experience brittle failure with the uptake of hydrogen. Nickel-based alloys, such as Incoloys@ and Hastelloys@ , are similar to titanium in that they are very corrosion resistant. They are easier to weld than titanium, but could be more expensive. Stainless steels have good mechanical properties and are very corrosion resistant, but catastrophic failures are possible through stress-corrosion cracking or intergranular corrosion. Ceramic materials, such as graphite and silicon carbide, have excellent corrosion resistance and are very much abundant, while mechanical strength is a problem with graphite.25

The features of the current Yucca Mountain repository design are given by Bullen et al.21 For defense HLW, the vitrified waste would be contained inside stainless-steel canisters and put into the overpack. As the design process continues, DOE is evaluating various design options that might increase the ability of the engineered barrier system to contain waste and could reduce uncertainty.24


The longevity of manufactured materials in the repository environment over such long periods of time is subject to significant uncertainty. At the same time, the prediction of material performance is essential in the development and use of waste packages (waste forms and waste containers). In the absence of a good mechanistic understanding of a material’s performance and data that span a wide range of the expected performance and physicochemical conditions, extremely conservative assumptions need to be considered. Many of the performance predictions rely on data collected over a relatively limited range of test conditions; thus, extrapolation of these data requires good mechanistic understanding.16,26 Without proper data support, any benefits that the waste forms or container might provide could be ignored; hence, it is highly desirable to improve the predictability of the materials performance. This also requires demonstration of quality control of the product.

Various technical issues must be addressed in the assessment of the long-term performance of the waste package in a geologic repository.27,28 As all components of a waste package may be altered in time within the repository environment, the environment for a waste package (both internal and external) must be well characterized. A demonstrated understanding of factors that might affect long-term service behavior is required for the characterization of materials for the waste-package components. These factors include variations in characteristics such as chemical composition, stress state, microstructure, fabrication or production history, and thermodynamic phase equilibria. Various interactions may be expected from gaseous or aqueous media that are in contact with the materials of the waste package. For metallic containers, various forms of corrosion that result from interactions with water and oxygen are important, as are the effects of hydrogen, which may result from radiolysis of water and vapor or galvanic coupling with borehole liner or container support structures. The environment may produce hydrostatic or lithostatic pressure, which may alter the stress state in waste-package components. Radiation will change the environment and create species with the potential for accelerated degradation of the waste-package components. Microbial species, if they are present in significant quantities, have the potential for interactions with the waste-package materials.29 The service life of the waste package must be determined based on the consideration of these interactions between the environment and the waste-package components, including joints, seals, and welds. For details on the experimental programs specific to Yucca Mountain, refer to Reference 21.

To ensure proper quality standards, materials and product specifications for each component of the waste package are necessary, including the construction and intermediate and final products. Quality control must be exercised for the fabrication of various waste-package components through the use of qualified personnel and equipment, procedures, and in-process product examinations. Inspection criteria are required to determine the acceptability of fabricated components for repository emplacement. Finally, in-situ monitoring needs to be implemented to evaluate the performance of the waste package with time up to its permanent closure. This requires development of advanced non-destructive techniques, such as magnetic resonance,30,31 automated ball indentation (ABI),32,33 and others,34 in addition to standard methods such as x-ray, ultra-sound, and acoustic emission.35


1. N. Tsoulfanidis and R.G. Cochran, “Radioactive Waste Management,” Nuclear Technology, 93 (1991), pp. 263–304.
2. H. Ocken, “Radiation-Field Control Manual—1997 Revision,” EPRI Report TR-107991 (Palo Alto, CA: EPRI, October 1997).
3. Y.S. Tang and J.H. Saling, Radioactive Waste Management (Washington, D.C.: Hemisphere Publishing Corporation, 1990).
4. Technical Position on Waste Form, Revision 1 (Washington, D.C.: Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear Regulatory Commission, January 1991).
5. L.L. Hench, D.E. Clark, and J. Campbell, “High Level Waste Immobilization Forms,” Nuclear and Chemical Waste Management, 5 (1984), pp. 149–173.
6. L.H. Johnson and D.W. Shoesmith, “Spent Fuel,” Radioactive Waste Forms for the Future, ed. W. Lutze and R.C. Ewing (Amsterdam: North-Holland, 1988).
7. R.A. Knief, Nuclear Engineering (Washington, D.C.: Hemisphere Publishing Corporation, 1992).
8. The Management of Radioactive Waste (London: The Uranium Institute, August 1991).
9. W. Lutze, “Silicate Glasses,” in Ref. 6.
10. R.C. Ewing, W.J. Weber, and F.W. Clinard, “Radiation Effects in Nuclear Waste Forms for High-Level Radioactive Waste,” Progress in Nuclear Energy, 29 (2) (1995), pp. 63–127.
11. A.E. Ringwood et al., “Synroc,” in Ref. 6.
12. For example, K.L. Murty, “Texture-Based Physical, Mechanical and Corrosion Characteristics of Zirconium Alloys,” Textures in Materials Research, ed. R.K. Ray and A.K. Singh (New Delhi, India: Oxford & IBH Publishing Co., Ltd., 1999), pp. 113–160.
13. K.L. Murty, “Stress Corrosion Cracking and Pellet Cladding Mechanical Interaction of Zircaloys—Application to LWRs,” Emerging Trends in Corrosion Control—Evaluation, Monitoring and Solutions, ed. A.S. Khanna, K.S. Sharma, and A.K. Sinha (New Delhi, India: Akademic Books International, 1999), pp. 702–710.
14. B. Grambow, “Source Terms for Performance Assessment of HLW-Glass and Spent Fuel as Waste Forms,” Scientific Basis for Nuclear Waste Management, Mat. Res. Soc. Symp. Proc. Vol. 506 (Warrendale, PA: MRS, 1998), pp. 141–152.
15. S. Stroes-Gascoyne, J. Nuclear Mater., 190 (1992), pp. 87– 100.
16. K.L. Murty, “Internal Pressurization Creep of Thin-Walled Tubing of Zr-Alloys for Dry Storage Feasibility of Nuclear Spent Fuel.,” in this issue.
17. J.S. Small, D.P. Trivedi, and P.K. Abraitis, “Modeling of Glass Dissolution and Transport with the Code SUGAR,” Scientific Basis for Nuclear Waste Management, Mat. Res. Soc. Symp. Proc. Vol. 506 (Warrendale, PA: MRS, 1998), pp. 253– 260.
18. R.D. McCright et al., “Candidate Container Materials for Yucca Mountain Waste Package Design,” Proc. Nuclear Waste Packaging Focus ’91 Conf. (La Grange Park, IL: American Nuclear Society, 1991).
19. Waste Package Final Update to EIS Engineering File, BBA000000-01717-5705-00019 Rev 01 (Washington, D.C.: TRW Environmental Safety Systems, Inc., U.S. Department of Energy, March 1999).
20. Code of Federal Regulations, Title 10, Chapter 1, Part 61, “Licensing Requirements for Land Disposal of Radioactive Waste,” Federal Register, 47, 57446 (Rockville, MD: U.S. Nuclear Regulatory Commission, 1982).
21. D.B. Bullen et al. (in this issue).
22. W. Miller et al., Natural Analogue Studies in the Geological Disposal of Radioactive Waste (Amsterdam: Elsevier Science, 1994).
23. L. Werme, P. Sellin, and N. Kjellbert, Copper Canisters for Nuclear High Level Waste Disposal. Corrosion Aspects, SKB Technical Report, TR-92-26 (Stockholm, Norway: SKB, 1992).
24. Viability Assessment of a Repository at Yucca Mountain, DOE/RW-0508 (Washington, D.C.: Office of Civilian Radioactive Waste Management, U.S. Department of Energy, December 1998).
25. M.-S. Yim, Evaluation of Waste Forms for Immobilization of C-14 and I-129, EPRI Report TR-110096 (Rockville, MD: EPRI, 1998).
26. K.L. Murty, “Significance and Role of Deformation Micromechanisms in Life-Predictive Modeling of Aging Structures,” Appl. Mech. Rev. 5, (May 1993), pp. 194–200.
27. Code of Federal Regulations, Title 10, Chapter 1, Part 60, “Disposal of High-Level Radioactive Waste in Geologic Repositories,” Federal Register, 48, 120, 28194 (Rockville, MD: U.S. Nuclear Regulatory Commission, 1983).
28. H.K. Manaktala and C. G. Interrante, Technical Considerations for Evaluating Substantially Complete Containment of High-Level Waste Within the Waste Package, NUREG/CR-5638 (Rockville, MD: U.S. Nuclear Regulatory Commission, 1990).
29. S. Stroes-Gascoyne and J.M. West, “An Overview of Microbial Research Related to High-Level Nuclear Waste Disposal with Emphasis on the Canadian Concept for the Disposal of Nuclear Fuel Waste,” Canadian J. of Microbiology, 42 (4) (1996), pp. 349–366.
30. K.L. Murty et al., “In-Situ Nuclear Magnetic Resonance Study of Defect Dynamics During Deformation of Materials,” J. Mater. Sci., 31 (1996), pp. 3289–3297.
31. K. Detemple et al., “In-Situ Nuclear Magnetic Resonance Investigation of Deformation-Generated Vacancies in Aluminum,” Phys. Rev., B52 (1995), pp. 125–133.
32. K.L. Murty and M.D. Mathew, “Condition Monitoring of Structural Materials Using Nondestructive Ball Indentation Technique,” Int. Symp. Mater. Aging and Life Management (ISOMALM 2000).
33. F.M. Haggag and K.L. Murty, “A Novel Stress-Strain Microprobe for Nondestructive Evaluation of Mechanical Properties of Materials,” Nondestructive Evaluation (NDE) and Materials Properties III, ed. P.K. Liaw et al. (Warrendale, PA: TMS, 1997), pp. 101–106.
34. B. Raj, C.V. Subramanian, and T. Jayakumar, Nondestructive Testing of Welds (Narosa Publishing House and Materials Park, OH: ASM International, 2000).

Man-Sung Yim is an assistant professor and K. Linga Murty is a professor in the Department of Nuclear Engineering at North Carolina State University.

For more information, contact M.-S. Yim, North Carolina State University, Department of Nuclear Engineering, Raleigh, North Carolina 27695-7909; (919)515-1466; fax (919)515-5115; e-mail yim@ncsu.edu.

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