Proposed Yucca Mountain Repository Research Articles

Yucca Mountain engineered barrier system corrosion model (EBSCOM)

A revised engineered barrier system model has been developed by the Electric Power Research Institute to predict the time dependence of the failure of the drip shields and waste packages in the proposed Yucca Mountain repository. The revised model is based on new information on various corrosion processes developed by the US Department of Energy and others and for a 20-mm-thick waste package design with a double closure lid system. As with earlier versions of the corrosion model, the new EBSCOM code produces a best-estimate of the failure times of the various barriers. The model predicts that only 15% of waste packages will fail within a period of 1 million years. The times for the first corrosion failures are 40,000 years, 336,000 years, and 375,000 years for the drip shield, waste package, and combination of drip shield and the associated waste package, respectively.

Alteration geochemistry of the Nopal I uranium deposit (Sierra Peña Blanca, Mexico), a natural analogue for a radioactive waste repository in volcanic tuffs

AbstractNatural analogues provide an approach to characterize and test the long‐term modelling of a repository performance. This article presents geochemical information about the alteration conditions of the Nopal I uranium deposit, Mexico, an analogue for the proposed Yucca Mountain radioactive waste repository. Mineralization and hydrothermal alteration of volcanic tuffs are contemporaneous, according to petrographic observations. Trace element geochemistry (U, Th, REE) provides evidence for local mobilization of uranium under oxidizing conditions and further precipitation under reducing conditions. O‐ and H‐isotope geochemistry of kaolinite, smectite, opal and calcite suggests that argillic alteration proceeded at shallow depth with meteoric water at 25–75 °C, a low‐temperature context, unusual for volcanic‐hosted uranium deposits. This temperature range is compatible with some post‐closure evolution models of the proposed Yucca Mountain repository.

Hydrogen Diffusion and Trapping Effects in Low and Medium Carbon Steels for Subsurface Reinforcement in the Proposed Yucca Mountain Repository

The electrochemical hydrogen permeation method was used to investigate hydrogen transport, trapping characteristics of low (0.08 pct C) and medium carbon (0.44 pct C) steels proposed for the Yucca Mountain (YM) repository environment. The presence of relatively high amounts of C, Mn, and S increased the density of trapping sites in medium carbon steel. The measured diffusivity of medium carbon steel was lower than that of the low carbon steel due to increased trapping of hydrogen at irreversible sites in the medium carbon steel. Hydrogen concentration values obtained for low carbon steels in YM ground water electrolytes indicate that increased ionic concentration decreases the uptake of hydrogen. The decrease in hydrogen permeation were due the formation of CaCO3 corrosion products on the surface of steels.

The Proposed Yucca Mountain Repository from a Corrosion Perspective

ABSTRACTIn this paper, the proposed Yucca Mountain Repository is viewed from a corrosion perspective. A major component of the long-term strategy for safe disposal of nuclear waste at the Yucca Mountain Repository is first to completely isolate the radionuclides in the waste packages for long times and to greatly retard the egress and transport of radionuclides from penetrated packages. Therefore, long-lived waste packages are important. The corrosion resistance of the waste package outer canister is reviewed, and a framework for the analysis of localized corrosion processes is presented. An overview is presented of the Materials Performance targeted thrust of the U.S. Department of Energy/Office of Civilian Radioactive Waste Management's Office of Science and Technology and International. The thrust program strives for increased scientific understanding, enhanced process models and advanced technologies for corrosion control.

Neptunium Solubility in the Near-Field Environment of a Proposed Yucca Mountain Repository

ABSTRACTFor representing the source-term of a proposed repository at Yucca Mountain, NV, the performance assessment (PA) approach evaluates the disequilibrium degradation of the waste forms to capture a bounding rate for radionuclide source-term availability and use solubility constraints that are more representative of longer-term, equilibrium processes to limit radionuclide mass transport from the source-term. These solubility limits capture precipitation processes occurring either as the waste forms alter, or in the near-field environment as chemical conditions evolve. A number of alternative models for solubility controls on dissolved neptunium concentrations have been evaluated. These alternatives include idealized models based on precipitation of simple, discrete neptunium phases and more complex considerations of trace amounts of neptunium being incorporated into secondary uranyl phases that form during waste form alteration. Thermodynamic constraints for neptunium under oxidizing conditions indicate that tetravalent neptunium solids (e.g., NpO2) are more stable relative to pentavalent phases (e.g., Np2O5), and therefore have solubilities that set lower dissolved concentrations of neptunyl species. Within solution, the pentavalent neptunyl ion (NpO2+) and its complexes dominate. Data on solids and solutions from slow flow-through (dripping) tests on spent nuclear fuel (SNF) grains indicate that neptunium is tetravalent in the SNF and that, over the ∼9 years duration of the tests, the dissolved neptunium concentrations are near to, or below, calculated NpO2 solubility. Currently, the NpO2 solubility model is applied at the waste forms within waste packages because the degradation of SNF and corrosion of alloys ensures active reduction reactions. The more conservative Np2O5 solubility model is applied in environments distal to the waste forms (i.e., the invert below waste packages) where reduction of dissolved pentavalent neptunium is less certain. Consideration of Np incorporation into secondary uranyl phases suggests that both of these idealized models will provide conservative estimates of neptunium release to the geosphere.

Residual Stress Characterization in Structural Materials by Destructive and Nondestructive Techniques

Transmutation of nuclear waste is currently being considered to transform long-lived isotopes to species with relatively short half-lives and reduced radioactivity through capture and decay of minor actinides and fission products. This process is intended for geologic disposal of spent nuclear fuels for shorter durations in the proposed Yucca Mountain repository. The molten lead-bismuth-eutectic will be used as a target and coolant during transmutation, which will be contained in a subsystem vessel made from materials such as austenitic (304L) and martensitic (EP-823 and HT-9) stainless steels. The structural materials used in this vessel will be subjected to welding operations and plastic deformation during fabrication. The resultant residual stresses cannot be totally eliminated even by stress-relief operations. Destructive and nondestructive techniques have been used to evaluate residual stresses in the welded and cold-worked specimens. Results indicate that tensile residual stresses were generated at the fusion line of the welded specimens made from either austenitic or martensitic stainless steel, with reduced stresses away from this region. The magnitude of residual stress in the cold-worked specimens was enhanced at intermediate cold-reduction levels, showing tensile residual stresses in the austenitic material while exhibiting compressive stresses in the martensitic alloys. Comparative analyses of the resultant data obtained by different techniques revealed consistent stress patterns.

SENSITIVITY ANALYSES OF THE STANDARDS FOR THE PROPOSED YUCCA MOUNTAIN REPOSITORY???A REVIEW, EVALUATION, AND COMMENTARY

The standards and regulations for the proposed Yucca Mountain high level radioactive waste repository, which were developed and promulgated by the U.S. Environmental Protection Agency and the U.S. Nuclear Regulatory Commission, respectively, are complex and challenging. A major reason is that they are divided into three parts, an Individual Protection Standard, a Human Intrusion Standard, and multiple Ground Water Protection Standards. Because the individual parts are not fully integrated, the one that controls under a specific set of circumstances depends on the radionuclide being evaluated, its mechanisms of transport, its avenues of intake, and differences in the specified limits. Although the coefficients in Federal Guidance Report (FGR) No. 11 are being used to estimate the doses, other sources (for example, Title 10, CFR, Part 20, and/or FGR No. 13) may deserve consideration. Since the regulations specify that the reasonably maximally exposed individual is an adult, this leaves unanswered the estimated doses to other age groups, such as infants and adolescents. Summarized in this paper are comparisons of the dose coefficients for different age groups, as well as evaluations of the sensitivity of effective and organ dose estimates for adults, depending on the source of the coefficients. All the latter analyses were based only on the consumption of ground water. While the dose estimates are different, depending on the sources of the coefficients, this was not unexpected. What these evaluations demonstrate is the caution that must be exercised to ensure that a full range of considerations is taken into account in interpreting the outcome of the dose assessments being made with respect to the proposed repository.

Nickel-based Gadolinium Alloy for Neutron Adsorption Application in Ram Packages

The National Spent Nuclear Fuel Program, located at the Idaho National Laboratory (INL), coordinates and integrates national efforts in management and disposal of US Department of Energy (DOE)-owned spent nuclear fuel. These management functions include development of standardised systems for long-term disposal in the proposed Yucca Mountain repository. Nuclear criticality control measures are needed in these systems to avoid restrictive fissile loading limits because of the enrichment and total quantity of fissile material in some types of the DOE spent nuclear fuel. This need is being addressed by development of corrosion-resistant, neutron-absorbing structural alloys for nuclear criticality control. This paper outlines results of a metallurgical development programme that is investigating the alloying of gadolinium into a nickel-chromium-molybdenum alloy matrix. Gadolinium has been chosen as the neutron absorption alloying element due to its high thermal neutron absorption cross section and low solubility in the expected repository environment. The nickel-chromium-molybdenum alloy family was chosen for its known corrosion performance, mechanical properties, and weldability. The workflow of this programme includes chemical composition definition, primary and secondary melting studies, ingot conversion processes, properties testing, and national consensus codes and standards work. The microstructural investigation of these alloys shows that the gadolinium addition is presentmore » in the alloy as a gadolinium-rich second phase. The mechanical strength values are similar to those expected for commercial Ni-Cr-Mo alloys. The alloys have been corrosion tested with acceptable results. The initial results of weldability tests have also been acceptable. Neutronic testing in a moderated critical array has generated favourable results. An American Society for Testing and Materials material specification has been issued for the alloy and a Code Case has been submitted to the American Society of Mechanical Engineers for code qualification.« less

The stress-corrosion-cracking model for high-level radioactive-waste packages

A stress-corrosion cracking (SCC) model has been adapted for performance prediction of high-level radioactive-waste packages to be placed in the proposed Yucca Mountain repository. For these waste packages, the outer barrier material is the highly corrosion-resistant alloy UNS-N06022 (alloy 22). The tensile stress is principally from weld-induced residual stress. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model. A seismic-damage-related SCC crack opening area density model is also briefly described.

Can Spent Nuclear Fuel Decay Heat Prevent Radionuclide Release?

AbstractHeat generated by radioactive decay of spent fuel represents a potentially important barrier to water accumulation on commercial spent nuclear fuel in breached waste packages. In the absence of water, fuel degradation and radionuclide release will be negligible. Thermal models for the proposed Yucca Mountain Repository suggest that, after a period of approximately 1000-4000 years (depending on loadingand ventilation conditions), the repository drift walls may decline to sub-boiling temperatures, thus allowing humidity in the drift to increase. The question thus arises, is the thermal gradient between the fuel and the drift sufficient to prevent water accumulation in a humid drift environment throughout the regulatory period? The answer depends on the balance between processes that oppose water condensation ontothe fuel (decay heat) and those that promote condensation such as the deliquescence of hygroscopic phaseswithin the fuel.Our experimental results indicate that deliquescence could lead to the condensation of water onto spent fuel despite the thermal “self-drying”effect if the following criteria are met: (1) the fission product salt CsI is present in the fuel or in the fuel-cladding gap, (2) the relative humidity in the driftexceeds 80% while temperatures in the waste package are around 90oC. Previous work suggests that these criteria may be met for some fuel pins within the proposed Yucca Mountain Repository. However,experiments that account for the role of U(VI) alteration phases suggest that deliquescence may be a self-limiting process in the sense that deliquescent components (e.g. Cs, Ba, Sr) may be incorporatedinto nondeliquescent U(VI) phases that form from the corrosion of spent fuel.

Environmental Risk and Uncertainty: Insights from Yucca Mountain

Abstract This study examines the effects of the risk from transporting high–level radioactive waste to the proposed Yucca Mountain repository on housing location decisions in Southern Nevada. Using data from a survey of southern Nevada households, we develop a model–based subjective risk estimate for each household. We then explore different factors that may influence the household's location decisions if the proposed transportation route is ultimately chosen for nuclear waste transport. We extend the conventional expected utility model to allow for uncertainty surrounding the actual risks borne by the household. Finally, we examine the impact of federal government compensation on households’ location decisions. The findings indicate that residents currently living near the proposed transportation route express subjective risk estimates much larger than those reported by the Department of Energy. In general, households that are uncertain about the future risks are more likely to relocate than those expressing certainty. When everything is considered, the model predicts that between one and three percent of households living near the transportation route are likely to relocate. Compensation can influence some households to remain at their present location and bear the transport risk.

EPA NOMINEES SNARLED IN SENATE

A NEVADA SENATOR is HOLDing up confirmation of President George W. Bush's nominees for upper management positions at EPA until the agency issues radiation for nuclear waste to be stored at Yucca Mountain. The proposed Yucca Mountain repository, located about 90 miles outside of Las Vegas, would hold some 70,000 metric tons of high-level nuclear waste. EPA is to set radiation release for the site, and the Nuclear Regulatory Commission (NRC) is to determine whether Yucca Mountain will meet those benchmarks. EPA, under the Clinton Administration, sent draft for Yucca Mountain to the White House Office of Management & Budget just before President Bush was inaugurated. They undergoing review and possible revision before they issued. Sen. Harry M. Reid (D-Nev.), a foe of the Yucca Mountain repository, says he is concerned that NRC and the Energy Department are attempting to weaken the provisions of the EPA standards during the White ...

Geomechanical and Thermal Effects on Moisture Flow at the Proposed Yucca Mountain Nuclear Waste Repository

The percolation flux through the unsaturated zone at the proposed Yucca Mountain repository for high-level nuclear waste can potentially affect (a) the occurrence and magnitude of water influx into the emplacement drifts, (b) the onset and rates of waste-package corrosion, (c) the mobilization of waste into aqueous states, and (d) the transport of radionuclides to the saturated zone. The magnitude and spatial and temporal variations of percolation flux depend on the infiltration rate but may be significantly influenced by (a) lateral diversion of flow at stratigraphic interfaces between nonwelded and welded tuffs above the repository horizon, (b) focusing of flow within or near steeply dipping fault zones, and (c) lateral diversion of flow within thermal-mechanical altered zones. Results from numerical modeling are presented to argue that (a) areas of the repository located close to and on the up-dip side of faults that intersect the Paintbrush nonwelded Tuff (PTn) would experience elevated percolation flux, irrespective of whether the faults act as flow barriers or conduits; (b) mechanical response of the rock mass to waste-generated heat will likely cause the development of laterally discontinuous zones characterized by dilation of horizontal fractures and net dilation or closure of vertical fractures; (c) areas of the repository located on the downstream side of the thermal-mechanical altered zones would experience elevated percolation flux; and (d) repository areas subjected to elevated percolation flux would experience faster rewetting of dryout zones and, thus, longer periods of wetness and elevated humidity. These results indicate that models used to predict the occurrence and magnitudes of water influx into emplacement drifts and the variations of relative humidity within the drifts need to consider the location of the drifts relative to faults that intersect the PTn and the development, geometry, and hydrological characteristics of thermal-mechanical altered zones.

A fracture mechanics-based model for assessing the mechanical failure of nuclear fuel rods due to rock fall

One of the potential failure mechanisms of nuclear fuel rods that will be disposed in the proposed Yucca Mountain repository is mechanical failure of the degraded cladding tubes caused by rock fall impacting on waste packages. In this paper, the development of a fracture mechanics-based model for treating rock fall-induced failure of nuclear fuel cladding tubes is described. The rock fall model has been developed by treating the forces acting on the fuel-rod cladding due to an impacting rock in terms of either a point or distributed load. The stress intensity factors of cracks in the cladding rods assembled in a 17×17 array are analyzed by performing appropriate elastic bending analyses. The critical stress at the onset of cladding failure is predicted on the basis of a fracture criterion that is applicable for both large and small cracks. The proposed model is used to assess the critical size and weight of rock fall required to cause cladding failure as a function of the impact location, loading condition, and the geometry of rock fall, as well as the initial defect size, hydrogen content, and fracture toughness of the cladding at the time of rock fall. The result indicates that the critical rock radii for cladding failure are well within the size range of potential rock falls estimated based on the joint spacings of rocks at the repository site. These theoretical predictions remain to be verified by laboratory and field data.

A Decision Theory Perspective on the Disposal of High‐Level Radioactive Waste

In this paper the problem of high‐level nuclear waste disposal is viewed as a five‐stage, cascaded decision problem. The first four of these decisions having essentially been made, the work of recent years has been focused on the fifth stage, which concerns specifics of the repository design. The probabilistic performance assessment (PPA) work is viewed as the outcome prediction for this stage, and the site characterization work as the information gathering option. This brief examination of the proposed Yucca Mountain repository through a decision analysis framework resulted in three conclusions: (1) A decision theory approach to the process of selecting and characterizing Yucca Mountain would enhance public understanding of the issues and solutions to high‐level waste management; (2) engineered systems are an attractive alternative to offset uncertainties in the containment capability of the natural setting and should receive greater emphasis in the design of the repository; and (3) a strategy of “waste management” should be adopted, as opposed to “waste disposal,” as it allows for incremental confirmation and confidence building of a permanent solution to the high‐level waste problem.

Effects of Perched Water on Thermally Driven Moisture Flow at the Proposed Yucca Mountain Repository for High-Level Waste

Numerical modeling was conducted to identify potential perched-water sites and examine the effects of perched water on thermally driven moisture flow at the proposed Yucca Mountain repository for high-level nuclear waste. It is demonstrated that perched-water zones may occur at two horizons on the up-dip side of faults such as the Ghost Dance Fault (GDF): in nonwelded volcanic strata [such as the Paintbrush Tuff nonwelded (PTn) stratigraphic unit], where juxtaposition of welded strata against nonwelded may constitute a barrier to lateral flow within the nonwelded strata; and in fractured horizons of underlying welded units [such as the Topopah Spring welded (TSw) unit] because of focused infiltration fed by overlying perched zones. The potential perched zones (PPZs) may contain perched water (which would flow freely into a well or opening) if infiltration rates are high enough. At lower infiltration rates, the PPZs contain only capillary-held water at relatively high saturations. Areas of the proposed repository that lie below PPZs are likely to experience relatively high percolation flux even if the PPZ contains only capillary-held water at high saturation. As a result, PPZs that contain only capillary-held water may be as important to repository performance as those that contain perched water. Thermal loading from emplaced waste in the repository is not likely to have an effect on PPZs located an adequate distance above the repository (such as in the PTn). As a result, such PPZs may be considered as permanent features of the environment. On the other hand, PPZs close to the repository depth (such as those that may occur in the TSw rock unit) would experience an initial period of spatial growth and increased saturation following waste emplacement. Thereafter, drying would begin at the repository horizon with perched-zone growth simultaneously above and below the repository. As a result, after the initial period of expansion, PPZs close to the repository horizon would undergo a period of net dissipation followed by a period of net rejuvenation. The duration of perched-zone dissipation decreases with increasing infiltration rate from the ground surface and varies spatially, depending on distance from the controlling fault. Areas of the repository close to, and on the up-dip side of, faults such as the GDF are likely to experience relatively long periods of wetness and high percolation flux.

Nuclear Waste Package Corrosion Behavior in the Proposed Yucca Mountain Repository

AbstractThe corrosion performance of spent nuclear fuel waste packages is becoming increasingly important in establishing the viability of the proposed Yucca Mountain repository system. Current package concepts propose the use of a 2 cm thick nickel-base superalloy (Alloy 22) shell as the main barrier to prevent corrosion penetration over many thousands of years. The expected package service conditions, as well as their variability and uncertainty, are discussed. The electrochemical conditions known to be responsible for passive behavior and its breakdown in Alloy 22 and similar alloys are examined in the light of the predicted repository environment. Durability prediction approaches and their conclusions are considered. Efforts to determine the relative impact of localized modes of failure and uniform passive dissolution on package durability are reviewed, along with open issues in need of resolution and alternative package designs. The basic question of the validity of extrapolating corrosion behavior over many times the duration of the present base of experience is addressed.

Scientific Bases for Cladding Credit as a Barrier to Radionuclide Release at the Proposed Yucca Mountain Repository+

AbstractThe performance of Zircaloy nuclear fuel cladding in the environment of the proposed Yucca Mountain (YM) high-level waste (HLW) repository is evaluated. Because the uniform aqueous corrosion/oxidation rate is extremely slow, this evaluation focuses on mechanical failure, localized corrosion, stress corrosion cracking (SCC), hydrogen embrittlement, and initial failure. Mechanical failure is expected to result from (1) disruptive events such as rockfalls from seismicity, faulting, and igneous activities, (2) creep, and (3) splitting by oxidation of the spent fuel (SF) matrix. Effects of chloride ions and radiolysis are evaluated in localized corrosion and SCC. Embrittlement can be caused by delayed-hydride cracking (DHC) and hydride reorientation. Among these cladding failure modes, rockfalls, other disruptive events and initial defects can be important to performance prior to the container failure by corrosion. Confirmatory tests are required to evaluate the susceptibility to splitting by secondary mineral formation, localized corrosion, SCC, and hydride reorientation. Reliable temperature calculations with backfilling are also required. After breach, cladding may still assure slow release of radionuclides through perforations because localized failures may limit the exposure of the SF matrix and may provide high mass-transfer resistance. Failures prior to the repository emplacement arising from reactor operation, pool storage, dry storage, and transportation are also considered.

A decision theory perspective on the disposal of high-level radioactive waste.

In this paper the problem of high-level nuclear waste disposal is viewed as a five-stage, cascaded decision problem. The first four of these decisions having essentially been made, the work of recent years has been focused on the fifth stage, which concerns specifics of the repository design. The probabilistic performance assessment (PPA) work is viewed as the outcome prediction for this stage, and the site characterization work as the information gathering option. This brief examination of the proposed Yucca Mountain repository through a decision analysis framework resulted in three conclusions: (1) A decision theory approach to the process of selecting and characterizing Yucca Mountain would enhance public understanding of the issues and solutions to high-level waste management; (2) engineered systems are an attractive alternative to offset uncertainties in the containment capability of the natural setting and should receive greater emphasis in the design of the repository; and (3) a strategy of "waste management" should be adopted, as opposed to "waste disposal," as it allows for incremental confirmation and confidence building of a permanent solution to the high-level waste problem.

Impact of Ionic Strength on Colloid Mobility in Saturated And Unsaturated Porous Media

AbstractA model system consisting of well-characterized quartz sand, negatively charged carboxyl-modified latex microspheres, and synthetic J- 13 groundwater was used as a first-step in defining limiting conditions for colloid transport at the proposed Yucca Mountain Repository for radioactive waste. Colloid (280-nm diameter) and tracer (bromide) transport was determined using column methods under steady-state flow conditions for both saturated and unsaturated moisture conditions. Experiments were conducted at three water contents (13%, 70%, and 100% saturated) and four ionic strengths (deionized water, and 0.1 x, 1x, and 10x the ionic strength of J-13 groundwater, which has an ionic strength of 0.0116). Colloid mobility in the 13% saturated columns was appreciably less than in the 70% and 100% saturated columns, whereas colloid mobility in the 70% and 100% saturated columns were quite similar. Ionic strength had a significant impact on colloid retention, becoming more pronounced in the drier systems. Colloids were mobile in deionized water and the 0.1x J-13 groundwater. Retention was approximately 20% in lx J-13 groundwater and 100% in the 10x J-13 groundwater. Compared with the 70% and 100% saturated conditions, colloid mobility decreased for the 10% saturated condition, with greater than 50% retention for the 1x J-13 groundwater system. As observed for the 70% and 100% saturated conditions, colloids in the 10x J-13 solution were essentially immobile. Colloid mobility is greatly affected by the ionic strength of the soil solution and this effect is moisture saturation dependent.