Programs intended to provide supporting information for the high-level radioactive waste (HLW) repository program must consider the licensing requirements and the technical issues involved with extrapolation of short-term test data to periods of up to 10,000 years. The licensing requirements of the Nuclear Regulatory Commission (NRC), and the issues the NRC staff considers important for the development of predictive methods, are described. Because performance predictions of the geologic repository and particular components of the waste package must largely be based upon inference, a reasonable assurance, on the basis of the record before the Commission, is the general standard that will be required.

Current trends in modelling waste package performance are reviewed mainly from the perspective of the Swedish SKB studies. Examples are given, which illustrate the approaches for modelling different waste forms, i.e. HLW glass and spent nuclear fuel, and candidate canister materials, such as copper and steel. The relative importance of thermodynamics, reaction kinetics and near-field transport are discussed.

A historical review of how engineers contributed to the solution of societal (public) and individual (private) problems is presented, with emphasis on the last thirty years when computers began to alter drastically the practice of engineering.

The concept of a 3-prong representation of the engineering decision making process, i.e., database, analysis, and judgment, is introduced. The importance of generating a suitable body of experimental and actual performance data and of coupling the database with a state-of-the-art analysis and extrapolation methodology for “public engineering” is illustrated in three case studies involving (a) the nuclear reactor safety study of 1975, (b) the Alaska pipeline weld defect assessment study of 1976, and (c) the Welding Research Council-Pressure Vessel Research Committee (PVRC) Round Robin Flaw Detection Program (1966-86) for thick-section steel welds.

For most of the public engineering projects involving long-term prediction of performance from short-term test data, the time-honored approach of supplementing experimental or service data with engineering judgment is frequently open to a public challenge. To remedy the situation, a new approach of using both the personal computers (PC) and the mainframes to (a) accelerate the peer review of existing data, (b) design a strategy for acquiring new data, and (c) facilitate a regional or global exchange of field data with standardized format, is outlined.

Application of the above approach to strengthen the judgment component of public engineering projects such as the management of high-level nuclear wastes is discussed through an example on the fatigue life prediction of stainless steel 304 piping and components.

In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs. anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made.

One of the unique and scientifically most difficult aspects of nuclear waste isolation is the extrapolation ofshot-term laboratory data (hours to years) to the long time periods (103-105 years) required by regulatory agencies for performance assessment. The direct verification of these extrapolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the lay public that there is a demonstrable and reasonable basis for accepting the long-term extrapolations. Natural analogues of both the repository environment (e.g. radionuclide migration at Oklo) and nuclear waste form behavior (e.g. alteration of basaltic glasses and radiation damage in minerals) have been used to demonstrate the long-term behavior of large scale geologic systems and, on a smaller scale, waste form durability. This paper reviews the use of natural analogues to predict the long-term behavior of nuclear waste form glasses. Particular emphasis is placed on the inherent limitations of any conclusions that are based on “proof” by analogy. An example -- corrosion of borosilicate glass -- is discussed in detail with specific attention to the proper and successful use of natural analogues (basaltic glass) in understanding the long-term corrosion behavior of borosilicate glass.

Performance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

The oxidation of CANDUTM fuel by the products of alpha radiolysis of water is being investigated using X-ray Photoelectron Spectroscopy and electrochemical techniques. Possible mechanisms of oxidation of U02by 02 and H202, two of the oxidants formed during radiolysis of water, are discussed. Oxidation by H202depends upon its concentration and the pH of the solution. The oxidation of U02by the products of alpha radiolysis of water, formed by an alpha flux comparable with that expected at the used- fuel surface at the estimated time of waste container failure, caused very little dissolution of UO2fuel. The presence of reductants in groundwater reaching the fuel surface will diminish the fuel oxidation by the radiolysis products.

A model for the radiation induced dissolution of U02has been developed and tested using experiments described in the literature. The model makes use of computer calculations based on a reaction mechanism in which the dissolution is initiated by reactions of H202, OH and H02 radicals with the U02surface. The model gave a fair agreement between calculated and experimental results: For a dose rate of 400 rad/s the calculated yields of oxidation were 2.6 and 2.1 % - h−1· m−2after 1 and 2 h, respectively. The corresponding experimental yields were 1.9 and 1.5 % ·h−1· m−2, when corrected for dissolution without irradiation. The effect of various initial concentrations of H202has been tested, and a reasonable rate constant for the reaction of U02and H202has been found. The requirement for the rate constant is that H202begins to be effective above concentrations of 10−3 mol · dm−3in accordance with experiments by L. Johnson.

Semi-static dissolution tests using pressurized water reactor spent fuel rod segments and NNWSI reference J-13 well water in sealed stainless steel vessels at 85°C are being conducted in support of the Waste Package Task of the NNWSI Project. Test specimens include: bare fuel plus the empty cladding hulls, fuel rod segments with artificially induced cladding defects and water-tight end caps, and undefected fuel rod segments with water-tight end caps. The test conditions approximate those expected in the proposed NNWSI Project repository when the waste package has cooled sufficiently to allow water to enter a breached container and contact the fuel rods, some of which may exhibit various degrees of cladding failure. Periodic solution samples (unfiltered and filtered) were analyzed for most radionuclides for which cumulative release limits are listed by the U.S. Environmental Protection Agency. Results from the first six-month cycle of the 85° C tests are presented and are compared with results from the first cycle of a previous test series run at 25° in fused silica test vessels.

In this paper we analyze the transport of radionuclides through penetrations in nuclear waste containers. Penetrations may result from corrosion or cracks and may occur in the original container material, in degraded or corroded material, or in deposits of corrosion products. We do not consider how these penetrations occur or the characteristics of expected penetrations in waste containers. We are concerned only with the analytical formulation and solutions of equations to predict rates of mass transfer through penetrations of specified size and geometry. Expressions for the diffusive mass transfer rates through apertures are presented. We present numerical illustrations for steady-state mass-transfer rates through a circular hole, including concentration isopleths. The results are extended to multiple holes, including a criterion for hole spacing wherein superposition of single-hole solutions can be used. Results illustrated for holes in thin-walled containers show that significant mass transfer can occur even if a small fraction of the container area is perforated. We also illustrate the case of holes facing a water gap, instead of being in intimate contact with porous rock. In this case the radionuclide flux from many small holes approaches that from a bare waste cylinder.

Spent fuel and unirradiated U02were previously reported [1] to have exhibited a 100-fold difference in the total amount of uranium released during static leach tests in salt brine. Two additional studies designed to investigate that difference have now been completed. The first study involved leach tests with U02pellets that had been doped with plutonium to simulate the alpha activity in spent fuel. The second study consisted of leach tests on spent fuel specimens that originated in the same reactor assembly as those used in previous tests but handled differently following discharge from the reactor. Results from the recently completed tests show that at least 90% of the reported difference between the dissolution behavior of spent fuel and U02can be attributed to an apparent surface oxidation of the spent fuel specimens after their discharge from the reactor. The effect of alpha radiolysis on the leaching of spent fuel and U02in salt brine in the presence of an air atmosphere was found to be relatively small in these tests.

We have measured the solubility 8f amorphous UO2in a wide pH range (2 to 10.5), in 0.5 M NaClO, at T=25 C. The species responsible for the solubility are U(OH) + and U'0H)h, with the stability constants lo:1,3=-1.1± 0.1 and log, 4 =- 5.4_.O. 2, respectively. No evidence for U(OH) is f6und up to pH=10.5. The same model explains satisfactorily the solubility data obtained by Parks and Pohl(5) at T=1000 C., in the pH range 1 to 10. The measured solubility of amorphous UO2 (s) in the pH_ýange 6 to 10.5, under reducing conditions, is rather high,. U(IV) =5 10 mols/l. This is in good agreement with previous data of Gayer and Leider(6) and Galkin and Stepanov(9).

Reacted solids from 200°C, 25 MPa tests with spent light water reactor fuel and other component materials of a high-level nuclear waste repository in basalt were analyzed by scanning electron microscopy, X-ray powder diffractometry, transmission electron diffraction, Auger electron spectrome- try, and microautoradiography. The work complements solutions analysis from the same tests conducted in Dickson-type rocking autoclaves for the Basalt Waste Isolation Project. Spent fuel reacted very slowly in fuel/water tests, exhibiting noticeable grain boundary etching and uranium silicate formation only after six months. Particle abrasion in the rocking autoclave produced fuel fines that remained apparently unaltered. In spent fuel tests with basalt or basalt and steel, iron smectite clay formed at the expense of mesostasis phases in the basalt. Low concentrations of β-emitters, probably Cs and Tc released from grain boundaries in the fuel, were incorporated in the clay. Steel corroded very slowly under the low-oxygen conditions of the tests, forming minor amounts of magnetite, hematite and wustite. The sluggishness of spent fuel/water and steel/water reactions is attributed to limited availability of oxygen under the test conditions.

The Basalt Waste Isolation Project (Rockwell Hanford Operations- BWIP) is investigating the feasibility of building a repository in the Columbia River Basalts for the permanent disposal of high-level nuclear waste. One aspect of this effort is to develop an understanding of the chemical behavior of radionuclides in the near-field environment of the waste container. Such information is needed to determine radionuclide release rates from the waste package and to make long-term projections of repository performance. To accomplish this task, ultrasensitive laser- based techniques, such as laser photoacoustic spectroscopy (LPAS) and laser induced fluorescence (LIF), are being developed as analytical methods for the trace-level detection and speciation of actinides in solutions typical of those encountered in groundwaters near the BWIP repository.

Carbon-14 is produced naturally by cosmic ray neutrons in the upper atmosphere. It is also produced in nuclear reactors, in amounts much smaller than the global inventory. About one-third of this is released directly to the atmosphere, and the other two-thirds remains in the spent fuel. Both the Environmental Protection Agency and the Nuclear Regulatory Commission have established limits on release of the 14C in spent fuel. This is of particular concern for the proposed repository in tuff, because of the unsaturated conditions and the consequent possibility of gaseous tansport of 14C as CO2Existing measurements and calculations of the 14C inventory in spent fuel are reviewed. The physical distribution and Wemical forms of the 14C are discussed. Available data on the release of 14C from spent fuel in aqueous solutions and in gaseous environments of air, nitrogen, and helium are reviewed. Projected N behavior in a tuff repository is described. It is concluded that 14C release measurements from spent fuel into moist air at temperatures both above and below the in situ boiling point of water as well as detailed transport calculations for the tuff geological environment will be needed to determine whether the IOCFR60 and 40CFR191 requirements can be met.

A review is given of the state of understanding of the statistical distributions of pit depths and pit populations in carbon steel, with special reference to underground carbon steel structures, which appear to bear significant similarities to the alloys and conditions expected in basalt and salt high level waste repositories.

The data base on pit depth distributions is reviewed and the various phenomenological equations which have been used to rationalize laboratory and field observations are discussed and classified in terms of the statis- tical models underlying them. The present state of understanding of the physical phenomena underlying these distributions is reviewed, and sugges- tions for appropriate future research are presented.

The Materials Characterization Center (MCC) at Pacific Northwest Lab- oratory is performing three kinds of corrosion tests for the Basalt Waste Isolation Project (BWIP) to establish the interlaboratory reproducibility and uncertainty of corrosion rates of container materials for high-level nuclear waste. The three types of corrosion tests were selected to address two distinct conditions that are expected in a repository constructed in basalt. An air/steam test is designed to address corrosion during the operational period and static pressure vessel and flowby tests are designed to address corrosion under conditions that bound the condi ring the post-closure period of the repository.

The results of tests at reference testing conditions, which were defined to facilitate interlaboratory comparison of data, are presented. Data are reported for the BWIP/MCC-105.5 Air/Steam Test, BWIP/MCC-105.1 Static Pressure Vessel, and BWIP/MC-105.4 Flowby Test. In those cases where data are available from a second laboratory, a statistical analysis of interlaboratory results is reported and expected confidence intervals for mean corrosion rates are given. Other statistical treatment of data include analyses of the effects of vessel-to-vessel variations, test capsule variations for the flowby test, and oven-to-oven variations for air/steam tests.

The paper considers the long term corrosion of carbon-steel containers for heat generating nuclear waste in a granitic repository. Under such conditions carbon steel may exhibit general, localised or passive corrosion behaviour depending on the exact composition and redox potential of the groundwater contacting the containers; localised corrosion being of most concern because it has the fastest propagation rate. It is well established, however, that such localised corrosion is only possible when the environment is sufficiently oxidising to maintain a positive potential gradient between the cathodic surface and the corrosion sites, which requires that species with oxidising potentials greater than water need to be present. This fact provides a basis for estimating the periods during which containers may be subject to localised and subsequently to general corrosion, and hence for making an overall assessment of the metal allowance required for a specified container life. A model for the diffusion transport of oxygen has been developed, and a sensitivity analysis has shown that the period of possible localised attack is strongly dependent on the passive film leakage current, the radiation dose rate and the oxygen diffusion coefficient.

In current conceptual designs, a mild steel (ASTM A?16 Grade WCA) is the relerence container material for use in high level nuclear waste packages intended for emplacement in a salt repository. The resistance of the steel to stress corrosion crackinq (SCC) is being investigated as part of the effort underway to verify the suitability of the material for waste package applications. Static tests (U-bend and bolt-loaded fracture toughness specimens) and dynamic tests (slow strain rate and corrosion fatigue) were conducted on both as-cast and weldment specimens of the material, in both low-Mg and high-Mg halite-saturated brines, in the temperature range of 90 to 200°C. The investigations indicate that the steel is not susceptible to SCC under the test conditions employed.