The potential nuclear waste repository at Yucca Mountain is to be in partially saturated rock. Released radioactive gases such as 14CO2 could have a direct pathway to the biosphere. We study the release of 14C released as 14CO2 from partly failed nuclear waste containers by analyzing the flow of gas into and out of a container. We analyze the transport of released 14CO2 in an unsaturated, fractured, porous medium with gas-phase advection and dispersion. We calculate the inhalation dose to a maximally exposed individual above ground, based on some conservative assumptions about release from containers. For the assumed parameter values, a simple atmospheric diffusion model gives very small doses when compared to background radiation doses.

One of the major concerns associated with waste storage in the vadose zone is that toxic materials may somehow be leached and transported by advecting water down to the water table and reach the accessible environment through either a well or discharge to a river. Consequently, care is taken to provide barriers over and around the storage sites to reduce contact between infiltrating water and the buried waste form. In some cases, it is important to consider the intrusion of water vapor as well as water in the liquid phase. Water vapor diffuses through porous material along vapor pressure gradients. A slightly lower temperature, or the presence of water-soluble components in the waste, favors water condensation resulting in leaching of the waste form and advection of water-soluble components to the water table. A simple analysis is presented that allows one to estimate the rate of vapor condensation as a function of waste composition and backfill material. An example using a waste form surrounded by concrete and gravel layers is presented. The use of thermal gradients to offset condensation effects of water-soluble components in the waste form is discussed. Thermal gradients may be controlled by design factors that alter the atmospheric energy exchange across the soil surface or that interrupt the geothermal heat field.

Work supported in part by U.S. Department of Energy, OHER, Contract DE-AC06-76RL0 1830

Simple diffusion coefficients, D(θ), in most unsaturated media were found to be primarily a function of water content and not material characteristics except where the characteristics affect or determine water content. At high water contents, D(θ) gradually declines as water content decreases, from 10−5 cm2/sec at a volumetric water content of about 50% to 10−7 cm2/sec at a volumetric water content of about 5%, followed by a sharp decline as surface water films become thin and discontinuous and pendular water elements become very small, from 10−7 cm2/sec at a volumetric water content of about 5% to 10−9 cm2/sec at a volumetric water content of about 0.5% to 1%.

There is an immediate need to begin to validate models that can be used for assessing the performance of waste package materials in an unsaturated repository environment. This paper examines available testing information and testing approaches that could support validation of models for engineering barrier system (EBS) radionuclide release. The content is presented in the context of the general methodology that has been proposed for validating performance assessment models [1], Available experimental observations are used to test some of the EBS release rate modeling premises. These observations include evidence of fluid film formation on waste glass surfaces in isothermal humid environments, accelerated waste glass reaction rates under repository service conditions of large glass surface area to water volume ratio, and mobilization of radionuclides as solutes and colloids. It is concluded that some important modeling premises may not be consistent with available experimental Information. However, it is also concluded that future laboratory testing, which simulates the Integrated waste package system (i.e., laboratory analog testing), is needed to evaluate the significance of these Inconsistencies and to test the system level models. A small-scale apparatus which was developed and tested to examine the feasibility of laboratory analog testing for the unsaturated Yucca Mountain repository environment is described.

The design-basis, defense-related, transuranic (TRU) waste to be emplaced in the Waste Isolation Pilot Plant (WIPP) could, if sufficient H2O and nutrients were present, produce as much as 1,500 moles of gas per drum of waste. Gas production could pressurize the repository to lithostatic pressure (150 atm) and perhaps higher.

Anoxic corrosion of Fe and Fe-base alloys and microbial degradation of cellulosics are the processes of greatest concern, but radiolysis of brine could also be important. The proposed backfill additives CaC03, CaO, CuSO4, KOH, and NaOH may remove or prevent the production of some of the expected gases. We describe these processes and present preliminary results of laboratory studies of anoxic corrosion and microbial activity.

This paper summarizes the main gas formation mechanisms in deep radioactive waste repositories. Production rates and overall gas volumes were estimated and showed predominance of hydrogen production by anoxic corrosion and radiolysis for French wastes. Gas evolution in the near field has been modeled. First results issued from a sensitivity analysis showed desaturation of the storage cavities for a wide range of parameter values.

The liberation and generation of gases from rock salt due to heat and gamma irradiation is investigated in order to obtain some of the data needed for the development of the concept for the disposal of high level waste in rock salt.

Our work is concerned with the influence of various parameters on gas production, e.g. salt composition and grain size, total absorbed dose, dose rate, temperature and gas atmosphere. Some of these parameters have not been studied previously in detail.

The original gamma irradiator employing spent fuel elements and capable of exposing samples at temperatures up to 250°C will be described. Experimental results from some 150 experiments will be given. The two major gases found were CO2 and N2O.CO, H2, CH4, Hydrocarbons, CI2, HCl and SO2 were also detected. The dependence of the gas yields on the various parameters will be presented and discussed.

The work reported here is part of a program conducted by the Nuclear Regulatory Commission on the efficacy of proposed plans for radionuclide containment for long-term storage of high-level nuclear waste (HLW). An important element of that program is the review and evaluation of available literature on components of a waste package. A review process and a database have been developed and tailored to provide information quickly to an individual who has a question about a particular material or component of a waste package. The database is uniquely suited to serve as a guide to indicate special areas where data and information needs exist on questions related to radionuclide containment. Additions to the database are made as information becomes available, and this source is as current as the published literature. A description of the review process and the database is given.