Using natural ThO2 and UO2 in which radiation damage has accumulated over geologic periods, we have measured the relative in dissolution rates bicarbonate-carbonate solutions of the actinide isotopes 238U, 234U, 232Th, 230Th, and 228Th. An enhanced dissolution by a factor of 2 to 7 of the short lived (1.9 yr) 228Th relative to that of its structurally incorporated primary isotope 232Th indicates the presence of severe localized atomic displacement damage created by alpha-recoil atoms. In contrast, the relatively small (<20%) isotopic fractionation between the radiogenic 8.0 × 104 yr- 230Th isotope and 232Th, and between the radiogenic 2.5 × 105 yr- 234U isotope and its primary structurally incorporated isotope 238U, suggests the nearly complete annealing of the alpha-recoil damage during a period of less than 105 yr. An estimate of the mean fading time for the leachable damage is 15,000 yr.

The results of an induced fission track mapping study of uranium in 12 pegmatitic monazite samples indicate that as the U concentration in monazite surpasses ca. 300–400 ppm, the microdistribution becomes much less homogeneous resulting in localized areas of very high uranium concentration (hot spots) in a phase undetectable by XRD. Microprobe analysis of three additional pegmatitic monazites shows the occurrence of exsolution of the two other endmembers of the monazite solid-solution series, huttonite and cheralite, as indicated by a normative mineralogy computed from the probe data. The presence of each mineral is confirmed by XRD. Also supporting this exsolution hypothesis is a strong similarity between ratios of the structural elements of monazite (LREE, P) within the Th-rich hot spots as compared to these ratios in the surrounding monazite. Gradational contacts between banded monazite and a Th-rich phase can be observed microscopically as well as through microprobe scans for Th, indicating the close genetic relationship of the two phases. To prevent their enhanced leachabilities, actinide redistribution must be prevented in any single- or multi-phase nuclear waste form.

Chemical changes induced in Palo Duro and Paradox Basin natural rock salts and in synthetic NaCl by heat and gamma radiation were investigated. Heating of unirradiated natural rock salts to 300°C resulted in HCI (most prevalent), SO2, CO2, and H2S evolution, and increased the base content of the remaining salt by not more than 10 microequivalents per gram; whereas, heating of synthetic NaCI gave no product. Gamma irradiation produced sodium colloids and neutral chlorine in amounts similar to the results of Levy and coworkers. When the irradiated salts were heated, three reactions were apparent: (1) radiation-induced defects recombined; (2) neutral chlorine was evolved; and (3) HCl, SO2, CO2, and H2S were evolved, similar to results for unirradiated salts. Because reaction (1) appeared to dominate over reaction (2), it is expected that the influence of radiation damage to salt on the near-field chemical environment will be minor.

The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory is the experimental arm of the Materials Characterization Organization (MCO), which was established by the U.S. Department of Energy (DOE) in FY 1980 to ensure high quality characterization and qualification of waste package materials essential to the reliable performance of DOE nuclear waste management programs. The MCC is responsible for publishing key test methods and data in the Nuclear Waste Materials Handbook (DOE/TIC-11400) after the methods and data have been reviewed and approved by the Materials Review Board (MRB), which is the other major operating arm of the MCO.

The MCC is carrying out its responsibilities in several ways. It sponsors a continuing series of workshops that address materials characterization test method issues. It supplies well-characterized reference and testing materials for use by the DOE nuclear waste management programs. It develops generic test methods and supports the repository waste package projects in developing selected site-specific test methods and performing confirmatory testing of these methods. When these test methods are approved by the MRB they constitute the formal tests to be used by laboratories to test and qualify materials, evaluate waste package components, and assure compliance with standards and/or specifications for the final product.

Some nuclear waste is destined for disposal in deep geological formations. The disposal system for wastes from commercial nuclear activities, and perhaps also for high-level wastes from defense-related activities, is to be designed and operated by the Department of Energy (DOE) and licensed by the Nuclear Regulatory Commission (NRC). The Nuclear Waste Policy Act [1] outlines some of the procedures and schedules that are to be followed by DOE in carrying out its assignment in the disposal of high-level nuclear waste (HLW). The regulations of the NRC that deal with HLW [2] are only partly in place, and amendments (e.g., related to the unsaturated zone) are yet to be approved and issued. The Environmental Protection Agency (EPA) has issued only draft versions of the regulations pertaining to HLW disposal [3], but key features of these drafts are at present in adequate agreement with NRC documents. On the basis of the trends that have become evident in the last few years, the DOE will be required to substantiate performance predictions for all pertinent aspects of a repository, especially the performance of the engineered waste package. The basis for demonstrating that the waste package performance in the repository will be in concert with the requirements is data on the waste package materials. These key materials data must clearly be highly reliable, and DOE will be required to assure this reliability. This paper addresses the organization and functions that have been assembled to aid in establishing the quality of materials data that are important in the licensing of a waste repository.

A major responsibility of the Materials Characterization Center (MCC) is to acquire and distribute reference and testing materials endorsed by DOE, for use in research and development in nuclear waste management. MCC standard materials provide a common basis for data obtained in waste repository development work, thereby assisting in developing a defensible database of known accuracy and precision. Homogeneity of standard materials is of particular importance. MCC emphasizes homogeneity of composition and physical characteristics of the standard materials, using statistical methods in sampling procedures and evaluation of analytical data.

Two classifications of standard materials are available from the MCC. Approved Reference Materials (ARMs) are intended as standard reference and calibration materials [similar to the SRMs of the U.S. National Bureau of Standards (NBS)], for field calibration of MCC standard test methods. Each ARM is characterized by MCC and other laboratories such as NBS to a high level of accuracy and precision. Approved Testing Materials (ATMs) are used as common testing materials, representing specific formulations of typical waste forms. Most ATMs are radioactive materials, for which characterization has been conducted on shielded analytical instruments at PNL. More than ten different ATMs (borosilicate glass and spent fuel) are in the MCC inventory.

These materials are available without charge, in reasonable amounts, to DOE programs. Other users pay pro-rated production costs. The materials normally are supplied in the form in which they were produced, e.g., glass as-cast bars. In some cases the MCC can supply special specimen forms, e.g., crushed powders or shaped monoliths.

The Materials Characterization Center (MCC) provides characterized light-water reactor (LWR) spent fuel approved testing materials (ATMs) for use by experimenters in radionuclide release studies in support of repository licensing efforts. The characterization completed to date is described for MCC ATM-101, a Zircaloy-clad UO2 fuel from the H. B. Robinson, Unit 2, Assembly BO-5. ATM-101 is a moderate burnup, low-releasing (fission products in-reactor) spent fuel typical of the fuel that is expected to initially be placed in a repository.

It is important in nuclear waste repository development that testing be done with materials containing a radionuclide spectrum representative of actual wastes. To meet the need for such materials, the Materials Characterization Center (MCC) has prepared simulated high level waste (HLW) glasses with radionuclides representative of about 10-, 300-, and 1000-year-old waste. A quantity of well characterized spent fuel also has been acquired for the same purpose. Glasses containing 10- and 300-year-old wastes, and the spent fuel specimens, must be fabricated in a hot cell. Hot cell conditions (high radiation field, remote operation, and difficulty of repairs) require that procedures and equipment normally used in materials preparation out-of-cell be modified for hot cell applications.

This paper discusses the fabrication of two glasses, and the preparation of test specimens of these glasses and spent fuel. One of the glasses is a 76–68 composition, which is fully loaded with actual commercial reactor fission product waste. The other glass contains simulated Barnwell Nuclear Fuel Plant waste, doped with different combinations of fission products and actinides. The spent fuel is a 10-year-old PWR material. Special techniques have been used to achieve high quality, well characterized testing materials, including specimens in the form of segments, wafers, cylinders, and powders of these materials.

The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results are presented: The radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach test methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-I approved reference material are presented.

Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transportaion, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance.

The characterization of nuclear waste package behavior is primarily based on the outcome of laboratory tests, where components of a proposed waste package are either individually or simultaneously subjected to simulated repository conditions. At each step of a testing method, both controllable and uncontrollable factors contribute to the overall uncertainty in the final outcome of the test. If not dealt with correctly, these sources of uncertainty could obscure or distort important information that might otherwise be gleaned form the test data. This could result in misleading or erroneous conclusions about the behavior characteristic being studied. It could also preclude estimation of the individual contributions of the major sources of uncertainty to the overall uncertainty. Statistically designed experiments and sampling plans, followed by correctly applied statistical analysis and estimation methods will yield the most information possible for the time and resources spent on experimentation, and they can eliminate the above concerns. Conclusions reached on the basis of such information will be sound and defensible. This presentation is intended to emphasize the importance of correctly applied, theoretically sound statistical methodology in characterizing nuclear waste package behavior.

Experiments based on a modified MCC-3S test method were conducted to investigate the solubility of U(VI) hydroxide (the mineral schoepite) and Np(IV) hydrous oxide at ambient temperature. The solubility of U(VI) hydroxide was investigated in carbonate-free, oxidizing aqueous solutions in the pH range of 3 to 11. The measured concentrations of dissolved U confirm the amphoteric behavior of U(VI) hydroxide solid and the importance of anionic U(VI) hydroxyl complexes in accurately calculating the solubilities of U(VI) solids under basic pH conditions. The effect of fluoride complexation on the solubility of Np(IV) hydrous oxide was studied in carbonatefree, near-neutral to alkaline solutions that contained 100 ppm total F. For solutions in which sodium dithionite (Na2 S2O4) and metallic Fe were used as reducing agents, the addition of 100 ppm F had no measurable effect on the solubility of Np(IV) hydrous oxide.

At the request of the Basalt Waste Isolation Project, the Materials Characterization Center has collected and developed a set of procedures for a waste form compliance test method (MCC-14.4). The purpose of the test is to measure the steady-state concentrations of specified radionuclides in solutions contacting a waste form material. The test method uses a crushed waste form and basalt material suspended in a synthetic basalt groundwater and agitated for up to three months at 150°C under anoxic conditions. Elemental and radioisotopic analyses are made on filtered and unfiltered aliquots of the solution. Replicate experiments are performed and simultaneous tests are conducted with an approved test material (ATM) to help ensure precise and reliable data for the actual waste form material. Various features of the test method, equipment, and test conditions are reviewed. Experimental testing using actinide-doped borosilicate glasses are also discussed.

MCC test methods in support of basalt, salt and tuff repositories are described. The methods consist of three general corrosion tests for basalt, two general corrosion tests and three stress corrosion tests for salt, and two general corrosion tests for tuff. The tests encompass the various conditions that are expected in the different repositories. A summary is provided of experimental activities of the MCC, and detailed results of MCC 105.1 Test Method for Long Term General Corrosion in a Basalt Repository (Static System) for standardized, benchmark conditions are presented. A summary of the statistical treatment of data is provided with examples of confidence intervals and the dependence on number of test vessels. The total relative standard deviation for the mean corrosion penetration rate was 11.4% for a mean corrosion penetration rate of 0.0244 mm/yr for 120 d tests at 100°C.

Some test parameters have become issues in test development either because they are difficult to measure and control or because the significance of their effect on waste form dissolution and alteration is unknown. Issues identified by the MCC are Eh/pH system effects, SA/V effects, temperature effects on sampling, waste package composition effects, colloid formation and stability, radiolysis, and formation kinetics and stability of secondary reaction products. These issues are addressed by the MCC through literature reviews, workshops, and direct investigation with the objective of recommending methods for resolving them. This paper summarizes their status and attempts to identify needs required for their resolution.

In 1943, six storage tanks were built at the Clinton Laboratories [later to become Oak Ridge National Laboratory (ORNL)] to contain wastes generated by wartime research and development operations. During the following years, these tanks became an integral part of the ORNL waste system and accumulated ∼1.5 × 106 L (400,000 gal) of sludge containing radioactive wastes. Recently, over a period of ∼18 months, these tanks were sluiced, the radioactive sludge resuspended, and the resuspended slurry pumped to the ORNL Hydrofracture Facility for underground disposal.

In this paper, a summary of the development work is given, and the process design and constraints are described. The operating difficulties encountered and overcome included grinder blade erosion, malfunctioning instruments, pump suction plugging, and slurry settling. About 90% of the settled sludge (containing ∼715,000 Ci) was removed from the system.

At the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.

Two methods for removing technetium from soluble defense high-level waste are described. In the first method, technetium is precipitated as tetraphenylphosphonium pertechnetate and separated from the decontaminated solution using sintered metal crossflow filters. In the second method, pertechnetate is removed from solution using a strong base anion exchange resin and then eluted from the resin with nitric acid. The nitric acid is recovered by sorption of the pertechnetate on a weak base ion exchange resin. The pertechnetate is eluted from the weak base resin with NaOH and recovered by precipitation as the sulfide or oxide.

Cylindrical uranium slugs are used as irradiation targets in the production reactors at the Savannah River Plant. These slugs are first chemically etched, nickel plated, encased in aluminum, inspected, and individually pressure tested. An improved process was developed to recover the uranium from the acidic etching streams by controlling pH and the PO4 to U ratio so that the precipitation of the uranium as hydrogen uranyl phosphate was maximized. Bench scale tests demonstrated that the recovery of uranium could be increased to greater than 99.9% (vs. the current level of about 95% recovery). The recommended changes involved the addition of process effluent “hold” tanks. The addition of the various process streams to the neutralization/precipitation tank could therefore be controlled to maintain a consistent ratio of uranyl nitrate and phosphoric acid. Also, it was determined that a strong caustic solution (resulting from the dissolution of rejected aluminum slugs) could be utilized to neutralize the nitric and phosphoric acid solutions. The buffering action of the aluminum in the “caustic recovery solution” would reduce the sensitivity of the hydrogen uranyl phosphate precipitation to the phosphate ion concentration.

Powder technology is used to immobilize high level nuclear waste (HLW) in sintered borosilicate glass. By uniaxial in-can hot pressing(temperature 950 K; pressure 1 MPa; heating rate ∼100 K/h; cooling rate 5 K/h), glass products containing simulated HLW(15 wt.%) have been produced in stackable steel cans (≤ 200 mm diameter). High densities, bulk integrities and homogeneities for the waste element distribution are realized. The advantages of powder technology are for example: (i) no segregation due to solid state vitrification, (ii) low evaporation losses and no compatibility problems due to low densification temperatures (sinter temperature ∼0.6 softening temperature of glass), (iii) production in easily arrangeable and interchangeable stacking units at modest pressures adequate especially for the use in hot cells under remote handling conditions, (iv) choosing units with alternatively HLWo r MLWin prescribed sequences to control the heat production of the package. As demonstrated now, the process is appropriate for high level waste (HLW), medium level waste (MLW) and mixed (HLW/ MLW vitrification, and is insensitive to waste modifications.

As the next step the effect of glass modifications was studied. In order to improve the glass leaching resistance and maintain low viscosity, silicon dioxide was doped with titanium dioxide. The mixture (referring to eutectic composition) was produced by a sol gel route and combined finally with 15 w/o HLW-oxides. Hot pressing of the obtained gels at 1273 K and 20 MPa provided highly homogeneous products with high densities.

The volatilization of technetium in the course of high-level waste vitrification may be considerable, depending on process variations. A series of experiments has been performed in order to gain more insight into the vaporization behavior of Tc during HLLW vitrification. No losses have been observed during chemical denitration or drying of the solutions on a drum drier at moderate temperatures. However, substantial volatilization (40–60 %) occurred during high temperature processing which applied either liquid feeding or feeding as a dehydrated product into the glass melter. Details of parameters investigated are described, and consequences for technical process application are discussed.