Waste Isolation Research Articles

Seismic monitoring of radioactive waste repositories

Safe isolation of high-level radioactive waste (HLW) and spent nuclear fuel (SF) from the biosphere is an important concern for humanity. Many countries consider monitoring the evolving conditions of a HLW/SF repository over a long period noninvasively, in such a way as to not compromise the sealing of the repository. Remote active seismic methods are possible options. We performed extensive seismic measurements at the Mont Terri Rock Laboratory, located within a clay formation that is a potential host rock in the Swiss HLW/SF program. A scaled-down HLW/SF repository was simulated as a 13-m-long microtunnel of 1 m diameter. The dry sand-filled microtunnel was progressively water saturated and slightly overpressured. The monitoring system consisted of a pressure seismic source and a hydrophone cable located in two boreholes drilled perpendicular to the microtunnel and eight geophones equally distributed around the periphery of the microtunnel. Nonlinear anisotropic tomography of crosshole traveltimes allowed the gross velocity structure of the host rock to be determined, but due to the experiment geometry, the hydrophone traveltimes were inadequate for monitoring changes within the repository. In contrast, analysis of the geophone seismic traces enabled the saturation process to be characterized. An integrated interpretation of the geophone data in combination with extensive numerical modeling suggested that injection of the water resulted in significant self sealing of the microtunnel’s excavation damage/disturbed zone (EDZ) and substantial increases in the P-wave velocity of the microtunnel sand and EDZ. On the basis of our investigations, we concluded that seismic monitoring of a HLW/SF repository is feasible, as long as the sensors are installed as close to the repository as regulations allow.

An Experimental Study of Diffusivity of Technetium‐99 in Hanford Vadose Zone Sediments

One of the methods being considered at the Hanford site in the state of Washington for safely disposing of low‐level radioactive wastes (LLW) is to encase the waste in concrete and entomb the packages in the Hanford vadose zone sediments. The current plan for waste isolation consists of stacking low‐level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages with concrete. Any failure of the concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow or diffusion and move into the surrounding subsurface sediments. It is therefore necessary to conduct an assessment of the performance of the concrete encasement structure and the surrounding soil's ability to retard radionuclide migration. Because of their anionic nature in aqueous solutions, the radionuclides, 99Tc and 129I were identified as long‐term dose contributors in LLW. The leachability or diffusion of these radionuclide species must be measured to assess the long‐term performance of waste grouts when contacted with vadose zone pore water or groundwater. To supplement the previously obtained data, a set of experiments were conducted using 99Tc‐spiked concrete (with 0 or 4% metallic Fe additions) in contact with Hanford soil at extremely low moisture content (~1% by mass). The 99Tc diffusion profiles in the soil half‐cells were measured after a time lapse of ~1.9 yr and the diffusion coefficient was calculated to be ~1.0 × 10−9 cm2 s−1.

Development of a U.S. Nuclear Waste Repository Research Program in Salt

With the United States rethinking its strategy for the management and disposal of defense high-level radioactive waste and civilian used nuclear fuel (UNF), it is an opportune time to evaluate the near-term and long-term options and requirements for the U.S. geologic repository program. In this paper, we outline a research program investigating the behavior of salt when subjected to thermal loads like those that would be present in a high-level-waste (HLW) repository. This program builds upon the knowledge base developed as a result of previous repository program efforts and the successful licensing and operation of the Waste Isolation Pilot Project Transuranic waste repository. We present a preliminary evaluation of a conceptual repository design that, in principle, exploits the positive attributes of salt as a disposal medium while balancing heat management issues against other considerations such as efficiency of disposal operations and cost. The coupled thermal-mechanical behavior of the intact and crushed salt, which influences and is influenced by the liberation and movement of water present in the salt and hydrous minerals, will ultimately control the thermal and hydrochemical conditions in the repository and at the waste package. To address key scientific issues, we advocate a combination of laboratory-scale investigations, a thermal test in the field for a configuration that replicates a small portion of our conceptual repository design, and numerical simulations conducted to develop a validated model that can be used for future repository design or performance assessment purposes. Accompanying this testing program would be a broader set of investigations that we advocate be conducted in the context of an iterative and adaptive process for systematically reducing uncertainties as we build a science-based safety case for HLW disposal in salt.

Calibrating transmissivities from piezometric heads with the gradual deformation method: An application to the Culebra Dolomite unit at the Waste Isolation Pilot Plant (WIPP), New Mexico, USA

Summary The gradual deformation method (GDM) is a geostatistical parameterization technique introduced in reservoir engineering for compelling reservoir models to respect production history. It makes it possible to adjust the spatial distribution of the hydraulic properties populating the reservoir model so that the flow responses simulated for the modified model closely replicate the production data collected on the field. The work presented in this paper investigates the potential of the GDM to solve inverse problems in hydrogeology, or subsurface hydrology, and produce heterogeneous transmissivity fields conditional to both transmissivity and piezometric head measurements within a stochastic context. It focuses on the calibration of the Culebra Dolomite unit, which lies approximately 450 m above the repository horizon of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Results show that the proposed version of the GDM, with a local deformation in each sub-domain neighboring a well, is efficient and competitive compared to other geostatistical-based inverse methods such as the pilot point method.

Self-propagating high-temperature synthesis and characteristics of cermet matrices for isolation of wastes with long-lived radionuclides

The possibility of preparing by self-propagating high-temperature synthesis (SHS) metal-ceramic (cermet) matrices with simulated wastes of REE-actinide fraction and Tc was examined. The specimens consist of oxide crystalline phases, glass, and melts. In the aluminate composite, the component (Sm) simulating the REE-actinide fraction is in the garnet and glass phases, and in the titanate composite, in the pyrochlore, titanosilicate of perrierite structure, and glass phases. Rhenium (Tc simulator) is incorporated in alloy phases. To evaluate the prospects for radioactive waste immobilization by SHS, it is necessary to synthesize matrices containing actinide isotopes (Am) and Tc and to study their structure and isolation properties.

The Enriched Xenon Observatory: EXO-200 and Ba+ tagging

The Enriched Xenon Observatory (EXO) is a proposed ton-scale double beta decay experiment with a tentative design sensitivity to the Majorana mass of ∼10 meV. The first phase of EXO is EXO-200, which uses 200 kg of Xe enriched to 80% in 136Xe to search for neutrinoless double beta decay. EXO-200 is a liquid Xe time projection chamber with the ability to detect both scintillation and ionization signals. The detector is constructed from ultra-low background materials and is currently installed at the Waste Isolation Pilot Plant, a salt mine with a 1600 meter water equivalent overburden. The projected 2 year sensitivity for EXO-200 is T1/20ν>6.4×1025 y at 90% confidence level. Looking toward a ton-scale EXO, one unique feature of the experiment is the proposal to identify the barium daughter produced by 136Xe double beta decay on an event-by-event basis. This technique will allow for the elimination of all backgrounds other than the background from the two-neutrino double beta decay spectrum. The EXO Collaboration is exploring a number of options to implement Ba-daughter tagging in the next generation EXO experiment.

Phase composition and radiation stability of matrices for isolation of REE-actinide waste

Phase composition and radiation stability of matrices for isolation of REE-actinide waste

Current status of the AMS facility at the Tono Geoscience Center of the Japan Atomic Energy Agency

The JAEA-AMS-TONO system is routinely used for 14C measurements at Tono Geoscience Center, Japan Atomic Energy Agency (JAEA) and applied to neotectonics and hydrogeology, in support of research on geosphere stability applicable to the long-term isolation of high-level radioactive waste. 10Be AMS has been developed for geochronological studies to estimate sedimentation rates and exposure age of basement rocks, incorporating a gas ionization detector with a large-volume gas absorber cell. Test measurements on 14C and 10Be reference materials show the system’s performance and suitability for application in the geosciences.

Radionuclide Interaction with Clays in Dilute and Heavily Compacted Systems: A Critical Review

Given the unique properties of clays (i.e., low permeability and high ion sorption/exchange capacity), clays or clay formations have been proposed either as an engineered material or as a geologic medium for nuclear waste isolation and disposal. A credible evaluation of such disposal systems relies on the ability to predict the behavior of these materials under a wide range of thermal-hydrological-mechanical-chemical (THMc) conditions. Current model couplings between THM and chemical processes are simplistic and limited in scope. This review focuses on the uptake of radionuclides onto clay materials as controlled by mineral composition, structure, and texture (e.g., pore size distribution), and emphasizes the connections between sorption chemistry and mechanical compaction. Variable uptake behavior of an array of elements has been observed on various clays as a function of increasing compaction due to changes in pore size and structure, hydration energy, and overlapping electric double layers. The causes for this variability are divided between "internal" (based on the fundamental structure and composition of the clay minerals) and "external" (caused by a force external to the clay). New techniques need to be developed to exploit known variations in clay mineralogy to separate internal from external effects.

Geologic isolation of nuclear waste at high latitudes: the role of ice sheets

Geologic isolation of nuclear waste at high latitudes: the role of ice sheets

Influence of Carbonate on Uranium Solubility in the WIPP

ABSTRACTIn the performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP), the solubility of uranium (VI) was conservatively set at 10-3 M for all expected WIPP conditions, including the potential and likely effects of carbonate complexation [1]. Under WIPP-relevant conditions, long-term experiments were performed to establish the uranium (VI) solubility limits in WIPP-simulated brine over a broad range of pCH+ values [7.5-12.5] and to evaluate the contribution of carbonate complexation and hydrolysis to uranium (VI) speciation. Data obtained in carbonate-free ERDA-6 brine, a simulated WIPP brine, were reported earlier [2]. In the absence of carbonate, uranium solubility approached 10-7 M at the expected pCH+ in the WIPP (~ 9.5). In the presence of a significant amount of carbonate (millimole levels), recent experimental results showed that uranium (VI) concentrations will not exceed 10-4M. This measured solubility limit is an order of magnitude lower than the uranium solubility value currently used in the WIPP PA [3]. A small effect of borate complexation was found in the pCH+ range [7.5-10]. At pCH+ ≥ 10, hydrolysis overwhelmed carbonate effects, and no amphoteric effect was observed.

Lessons Learned from the Yucca Mountain Nuclear Waste Repository Project The Engineered Barrier System

ABSTRACTNuclear waste isolation programs both inside and outside the United States have provided evidence that there are many geologic options for a repository, but virtually all of them rely to some degree on an engineered barrier system (EBS) to isolate and/or retard the migration of radionuclides to the biosphere. At Yucca Mountain, the design of the EBS was unexpectedly challenging because of uncertainties in quantitatively determining the local environment of the EBS particularly during the thermal pulse. The EBS design for the Yucca Mountain site evolved from a thin-walled, limited-lifetime, corrosion-resistant canister through a corrosion-allowance canister, to the present design, which may have a lifetime of more than 106years. The EBS proposed for the Yucca Mountain repository has many individual sub-barriers, beginning with the spent fuel and waste, the cladding of the spent fuel, the geometry of the package, etc. The anticipated modes of degradation of engineering materials, including corrosion of the fuel, of the canister, and of the drip shield proposed specifically for the Yucca Mountain project, and the consequences of the materials degradation on the performance of the repository are presented. The roles of conservative modeling and simplifying assumptions for radionuclide mobilization and transport in the EBS on characterization of the source term are addressed.

Plutonium in the WIPP environment: its detection, distribution and behavior

The Waste Isolation Pilot Plant (WIPP) is the only operating deep underground geologic nuclear repository in the United States. It is located in southeastern New Mexico, approximately 655 m (2150 ft) below the surface of the Earth in a bedded Permian evaporite salt formation. This mined geologic repository is designed for the safe disposal of transuranic (TRU) wastes generated from the US defense program. Aerosol and soil samples have been collected near the WIPP site to investigate the sources of plutonium in the WIPP environment since the late 1990s, well before WIPP received its first shipment. Activities of (238)Pu, (239+240)Pu and (241)Am were determined by alpha spectrometry following a series of chemical separations. The concentrations of Al and U were determined in a separate set of samples by inductively coupled plasma mass spectrometry. The annual airborne concentrations of (239+240)Pu during the period from 1998 to 2010 show no systematic interannual variations. However, monthly (239+240)Pu particulate concentrations show a typical seasonal variation with a maximum in spring, the time when strong and gusty winds frequently give rise to blowing dust. Resuspension of soil particles containing weapons fallout is considered to be the predominant source of plutonium in the WIPP area. Further, this work characterizes the source, temporal variation and its distribution with depth in a soil profile to evaluate the importance of transport mechanisms affecting the fate of these radionuclides in the WIPP environment. The mean (137)Cs/(239+240)Pu, (241)Am/(239+240)Pu activity ratio and (240)Pu/(239)Pu atom ratio observed in the WIPP samples are consistent with the source being largely global fallout. There is no evidence of any release from the WIPP contributing to radionuclide concentrations in the environment.

Analysis of the FEBEX multi‐barrier system including thermoplasticity of unsaturated bentonite

SUMMARYDeep geological repository involving a multibarrier system constitutes one of the most promising options for isolating high‐level radioactive waste from the human environment. To certify the efficiency of waste isolation, it is essential to understand the behaviour of confining geomaterial under a variety of environmental conditions. To this end, results from a near‐to‐real experiment, the full‐scale engineered barriers in situ experiment, are studied by means of a thermo–hydro–mechanical finite element approach, including a consistent thermoplastic constitutive model for unsaturated soils. Laboratory tests are simulated to calibrate model parameters. The results of the numerical simulations are compared with sensor measurements and show the ability of the model to reproduce the main behavioural features of the system. The influence of the hysteretic and temperature‐dependent retention of water on the mechanical response is exhibited. Finally, those results are interpreted in the light of thermoplasticity of unsaturated soils, which reveals the highly coupled and non‐linear characters of the processes encountered. Copyright © 2011 John Wiley & Sons, Ltd.

Assessment study for multi-barrier system used in radioactive borate waste isolation based on Monte Carlo simulations

Radioactive waste generated from the nuclear applications should be properly isolated by a suitable containment system such as, multi-barrier container. The present study aims to evaluate the isolation capacity of a new multi-barrier container made from cement and clay and including borate waste materials. These wastes were spiked by (137)Cs and (60)Co radionuclides to simulate that waste generated from the primary cooling circuit of pressurized water reactors. Leaching of both radionuclides in ground water was followed and calculated during ten years. Monte Carlo (MCNP5) simulations computed the photon flux distribution of the multi-barrier container, including radioactive borate waste of specific activity 11.22KBq/g and 4.18KBq/g for (137)Cs and (60)Co, respectively, at different periods of 0, 15.1, 30.2 and 302 years. The average total flux for 100cm radius of spherical cell was 0.192photon/cm(2) at initial time and 2.73×10(-4)photon/cm(2) after 302 years. Maximum waste activity keeping the surface radiation dose within the permissible level was calculated and found to be 56KBq/g with attenuation factors of 0.73cm(-1) and 0.6cm(-1) for cement and clay, respectively. The average total flux was 1.37×10(-3)photon/cm(2) after 302 years. Monte Carlo simulations revealed that the proposed multi-barrier container is safe enough during transportation, evacuation or rearrangement in the disposal site for more than 300 years.

Environmental monitoring of radioactive and non-radioactive constituents in the vicinity of WIP

Abstract The Waste Isolation Pilot Plant (WIPP), a US Department of Energy (DOE) facility, is a deep geologic transuranic waste disposal site designed for the safe disposal of transuranic (TRU) wastes generated from the US defense program. Monitoring is a key component of the development and operation of any nuclear repository and is important to the WIPP performance assessment. Initial concerns over the release of radioactive and chemical contaminants from the WIPP led to various monitoring programs, including the independent, academic-based WIPP environmental monitoring (WIPP-EM) program conducted by the New Mexico State University (NMSU) Carlsbad Environmental Monitoring and Research Center (CEMRC) located in Carlsbad, NM. The mission of CEMRC is to develop and implement an independent health and environmental monitoring program in the vicinity of WIPP and make the results easily accessible to the public and all interested parties. Under the WIPP-EM program constituents monitored include: (1) selected radionuclides, elements, and ions of interest in air, soil, vegetation, drinking water, surface water and sediment from within a 100-mile radius of WIPP as well as in the air exiting the WIPP exhaust shaft, and (2) internally deposited radionuclides in the citizenry living within a 100-mile radius of WIPP. This article presents an evaluation of more than tens years of environmental monitoring data that informed the public that there is no evidence of increases in radiological contaminants in the region that could be attributed to releases from the WIPP. Such an extensive monitoring program and constant public engagement is an ideal model for all nuclear waste repositories anywhere in the world.

How to build a better sepulcher: Lessons from New Mexico’s Waste Isolation Pilot Plant

New Mexico’s Waste Isolation Pilot Plant took three decades, several lawsuits, myriad battles with the Department of Energy, and more than a few political twists and turns before becoming the first and only operational geologic radioactive waste repository in the world. The story of how Carlsbad, New Mexico, became the center of a national drama is an object lesson in how the United States and countries around the world can improve on this lengthy, contentious, and incredibly vital repository approval process. As more and more nations build nuclear programs and accumulate radioactive waste, there is no time to lose in initiating plans for safe disposal sites that can last—essentially—forever.

Physicochemical controls on initiation and evolution of desiccation cracks in sand–bentonite mixtures: X-ray CT imaging and stochastic modeling

The shrink-swell behavior of active clays in response to changes in physicochemical conditions creates great challenges for construction of geotechnical barriers for hazardous waste isolation, and is of significant importance for management of agricultural and natural resources. Initiation and evolution of desiccation cracks in active clays are strongly dependent on physicochemical initial and boundary conditions. To investigate effects of bentonite content (20, 40, 60%), pore fluid chemistry (0.05 and 0.5 M NaCl) and drying rates (40 and 60 °C) on cracking behavior, well-controlled dehydration experiments were conducted and X-ray Computed Tomography (CT) was applied to visualize and quantify geometrical features of evolving crack networks. A stochastic model based on the Fokker-Plank equation was adopted to describe the evolution of crack aperture distributions (CAD) and to assess the impact of physicochemical factors on cracking behavior. Analyses of crack porosity and crack specific surface area showed that both clay content and temperature had larger impact on cracking than pore fluid concentration. More cracks formed at high bentonite contents (40 and 60%) and at high drying rate (60 °C). The drift, diffusion and source terms derived from stochastic analysis indicated that evaporative demand had greater influence on the dynamics of the CAD than solution chemistry.

Monitoring of gross alpha, gross beta and actinides activities in exhaust air released from the waste isolation pilot plant

The simultaneous measurements of gross alpha and beta activities is one of the simplest radioanalytical technique used as a method for screening samples of both high and low activities of alpha and beta emitting radionuclides in environmental and bioassay samples. Such measurements are of great interest from both a radiological, waste disposal viewpoint, and to establish a trend of radioactivity based on long term monitoring. At the WIPP (Waste Isolation Pilot Plant) site, unfiltered exhaust air from the underground repository is the most important effluent. As part of its monitoring program, the particulates from WIPP exhaust air are collected everyday at a location typically called the Fixed Air Sampler (FAS) site or Station A, this site is located at the release point for aerosol effluents from the underground to the environment. The measurements of gross alpha and beta activity on air filter samples were performed using an ultra low level counter, PIC-MPC 9604-α/β, from Protean Instrument Corporation. The high sensitivity of the gross alpha and beta instrument enables detection of low value activity from the air filters. In 2009, the values of gross alpha and beta activity concentrations ranged from <MDC (≈0.1 mBq/m(3)) to 1.03 mBq/m(3) and <MDC (≈0.2 mBq/m(3)) to 15.8 mBq/m(3); the density varied from <MDC (≈0.9 Bq/g) to 63.5 Bq/g and <MDC (≈1.7 Bq/g) to 114 Bq/g.

Modeling of the pseudo-colloids migration for multi-member decay chains with a arbitrary flux boundary condition in a fractured porous medium

A deep geologic disposal is the prime option for the long-term isolation of high-level radioactive waste (HLW) in many countries. For deeply located repositories, a radionuclide released from a failed waste container moves through the engineered and natural barriers before it reaches a biosphere. The pseudo-colloid which a radionuclide is adsorbed on a moving natural humic or fulvic colloid can be generated in a fractured porous medium. The size of a colloid is in general in the order of a hundred nanometer so that its migration velocity in a fracture is higher than that of a radionuclide due to the hydrochromatic effect. A large colloid cannot diffuse easily into a surrounding rock. Also, there are many kinds of actinides and these actinides have decay chains. In this analysis, the canonical form solution is derived for a pseudo-colloid and a solute in a fracture and a surrounding rock with a realistic inlet boundary condition with multi-member decay chains. It is used the Fortran based a computational code which uses a special subroutine for the inversion of Laplace transform. Consequently, the role of the pseudo-colloid in the fractured porous medium is important and also their decay chains aren’t neglected in the performance assessment of the HLW.