Mineralogical Charecteristics of Yucca Mountain Alluvium and Effects on Neptunium (V) Sorption

Saturated alluvium is expected to serve as an important natural barrier to radionuclide transport at Yucca Mountain, the proposed geological repository for disposal of high-level nuclear wastes. {sup 237}Np(V) (half-life = 2.4 x 10{sup 5} years) has been identified as one of the radionuclides that could potentially contribute the greatest dose to humans because of its relatively high solubility and weak adsorption to volcanic tuffs under oxidizing conditions. The previous studies suggested that the mineralogical characteristics of the alluvium play an important role in the interaction between Np(V) and the alluvium. The purpose of this study is to further evaluate the mineralogical basis for Neptunium (V) sorption by saturated alluvium located down-gradient of Yucca Mountain

Uranium and Neptunium Desorption from Yucca Mountain Alluvium

Uranium and neptunium were used as reactive tracers in long-term laboratory desorption studies using saturated alluvium collected from south of Yucca Mountain, Nevada. The objective of these long-term experiments is to make detailed observations of the desorption behavior of uranium and neptunium to provide Yucca Mountain with technical bases for a more realistic and potentially less conservative approach to predicting the transport of adsorbing radionuclides in the saturated alluvium. This paper describes several long-term desorption experiments using a flow-through experimental method and groundwater and alluvium obtained from boreholes along a potential groundwater flow path from the proposed repository site. In the long term desorption experiments, the percentages of uranium and neptunium sorbed as a function of time after different durations of sorption was determined. In addition, the desorbed activity as a function of time was fit using a multi-site, multi-rate model to demonstrate that different desorption rate constants ranging over several orders of magnitude exist for the desorption of uranium from Yucca Mountain saturated alluvium. This information will be used to support the development of a conceptual model that ultimately results in effective K{sub d} values much larger than those currently in use for predicting radionuclide transport at Yucca Mountain

Colloid facilitated transport in fractured rocks : parameter estimation and comparison with experimental data.

Colloid-facilitated migration of plutonium in fractured rock has been implicated in both field and laboratory studies . Other reactive radionuclides may also experience enhanced mobility due to groundwater colloids. Model prediction of this process is necessary for assessment of contaminant boundaries in systems for which radionuclides are already in the groundwater and for performance assessment of potential repositories for radioactive waste. Therefore, a reactive transport model is developed and parameterized using results from controlled laboratory fracture column experiments. Silica, montmorillonite and clinoptilolite colloids are used in the experiments along with plutonium and Tritium . . The goal of the numerical model is to identify and parameterize the physical and chemical processes that affect the colloid-facilitated transport of plutonium in the fractures. The parameters used in this model are similar in form to those that might be used in a field-scale transport model

Predictions of tracer transport in interwell tracer tests at the C-Hole complex. Yucca Mountain site characterization project report milestone 4077

This report presents predictions of tracer transport in interwell tracer tests that are to be conducted at the C-Hole complex at the Nevada Test Site on behalf of the Yucca Mountain Site Characterization Project. The predictions are used to make specific recommendations about the manner in which the tracer test should be conducted to best satisfy the needs of the Project. The objective of he tracer tests is to study flow and species transport under saturated conditions in the fractured tuffs near Yucca Mountain, Nevada, the site of a potential high-level nuclear waste repository. The potential repository will be located in the unsaturated zone within Yucca Mountain. The saturated zone beneath and around the mountain represents the final barrier to transport to the accessible environment that radionuclides will encounter if they breach the engineered barriers within the repository and the barriers to flow and transport provided by the unsaturated zone. Background information on the C-Holes is provided in Section 1.1, and the planned tracer testing program is discussed in Section 1.2

Transport of synthetic colloids through single saturated fractures: A literature review

Colloids having the same surface charge sign as the bulk of the geologic media in a groundwater system may be able to travel through the system faster than soluble species because they will follow fluid streamlines more closely and they should have less tendency to diffuse into pores or dead spaces in the media than soluble species. Synthetic colloids with uniform, controlled properties may be ideal for serving as {open_quotes}worst-case{close_quotes} tracers that provide lower-bound estimates of contaminant travel times in hydrologic systems. This report discusses a review of the literature pertaining to colloid transport in single saturated natural fractures. After a brief background discussion to put the literature review in perspective, the phenomenon of colloid transport in saturated fractures is divided into three major topics, each of which is reviewed in detail: (1) saturated fluid flow through fractures; (2) colloid transport by convection, diffusion, and force fields; and (3) colloid interactions with surfaces. It is suggested that these phenomena be accounted for in colloid transport models by using (1) lubrication theory to describe water flow through fractures, (2) particle tracking methods to describe colloid transport in fractures, and (3) a kinetic boundary layer approximation to describe colloid interactions with fracture walls. These methods offer better computational efficiency and better experimental accessibility to model parameters than rigorously solving the complete governing equations

MASBAL: A computer program for predicting the composition of nuclear waste glass produced by a slurry-fed ceramic melter

This report is a user's manual for the MASBAL computer program. MASBAL's objectives are to predict the composition of nuclear waste glass produced by a slurry-fed ceramic melter based on a knowledge of process conditions; to generate simulated data that can be used to estimate the uncertainty in the predicted glass composition as a function of process uncertainties; and to generate simulated data that can be used to provide a measure of the inherent variability in the glass composition as a function of the inherent variability in the feed composition. These three capabilities are important to nuclear waste glass producers because there are constraints on the range of compositions that can be processed in a ceramic melter and on the range of compositions that will be acceptable for disposal in a geologic repository. MASBAL was developed specifically to simulate the operation of the West Valley Component Test system, a commercial-scale ceramic melter system that will process high-level nuclear wastes currently stored in underground tanks at the site of the Western New York Nuclear Services Center (near West Valley, New York). The program is flexible enough, however, to simulate any slurry-fed ceramic melter system. 4 refs., 16 figs., 5 tabs

A refined approach to estimating effective flow porosity from cross-hole tracer tests in fractured media.

Simulations of flow and transport in two-dimensional representations of heterogeneous fractured media are used to investigate the errors and biases associated with effective flow porosity estimates derived from cross-hole tracer tests. A method is presented for constructing probability distributions of 'correction factors' that can be used to correct apparent flow porosities obtained from tracer tests to obtain 'true' flow porosities in fracture systems. Although only a limited number of the many possible variations in fracture flow system properties is investigated, it is concluded that effective flow porosities derived from cross-hole tracer tests have a strong tendency to overpredict true flow porosities in fracture flow systems. This tendency toward overprediction decreases as the fracture conductivity relative to the background conductivity field decreases and as the orientation of the most conductive fractures becomes better aligned with the two wells. Tracer tests with small amounts of recirculation of water from the production well to the injection well are predicted to result in much better estimates of true flow porosity (on average), and with much less variability in the estimates, than tests with no reclrculation. However, the advantage offered by recirculation decreases as the fracture conductivity relative to the background conductivity decreases

Saturated Zone In-Situ Testing

The purpose of this scientific analysis is to document the results and interpretations of field experiments that have been conducted to test and validate conceptual flow and radionuclide transport models in the saturated zone (SZ) near Yucca Mountain. The test interpretations provide estimates of flow and transport parameters that are used in the development of parameter distributions for Total System Performance Assessment (TSPA) calculations. These parameter distributions are documented in the revisions to the SZ flow model report (BSC 2003 [ 162649]), the SZ transport model report (BSC 2003 [ 162419]), the SZ colloid transport report (BSC 2003 [162729]), and the SZ transport model abstraction report (BSC 2003 [1648701]). Specifically, this scientific analysis report provides the following information that contributes to the assessment of the capability of the SZ to serve as a barrier for waste isolation for the Yucca Mountain repository system: (1) The bases for selection of conceptual flow and transport models in the saturated volcanics and the saturated alluvium located near Yucca Mountain. (2) Results and interpretations of hydraulic and tracer tests conducted in saturated fractured volcanics at the C-wells complex near Yucca Mountain. The test interpretations include estimates of hydraulic conductivities, anisotropy in hydraulic conductivity, storativities, total porosities, effective porosities, longitudinal dispersivities, matrix diffusion mass transfer coefficients, matrix diffusion coefficients, fracture apertures, and colloid transport parameters. (3) Results and interpretations of hydraulic and tracer tests conducted in saturated alluvium at the Alluvium Testing Complex (ATC), which is located at the southwestern corner of the Nevada Test Site (NTS). The test interpretations include estimates of hydraulic conductivities, storativities, total porosities, effective porosities, longitudinal dispersivities, matrix diffusion mass transfer coefficients, and colloid transport parameters. (4) Comparisons of sorption parameter estimates for a reactive solute tracer (lithium ion) derived from both the C-wells field tracer tests and laboratory tests using C-wells core samples. (5) Sorption parameter estimates for lithium ion derived from laboratory tests using alluvium samples from NC-EWDP-19D1 (one of the wells at the ATC) so that a comparison of laboratory- and field-derived sorption parameters can be made in saturated alluvium if cross-hole tracer tests are conducted at the ATC

Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B