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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Use of generic and site-specific underground research laboratories to support siting, design, and safety assessment - 15417

Under the auspices of the International Atomic Energy Agency (IAEA), nationally developed underground research laboratories (URLs) and associated research institutions are being offered for use by other nations. These facilities form an Underground Research Facilities (URF) Network for training in and demonstration of waste disposal technologies and the sharing of knowledge and experience related to geologic repository development, research, and engineering. In order to achieve its objectives, the URF Network regularly sponsors workshops and training events related to the knowledge base that is transferable between existing URL programs and to nations with an interest in developing a new URL. Toward that end, a three-day workshop was held in Albuquerque, New Mexico in October 2014 to identify and compare the understanding of URF Network members on the place and role of generic and site-specific URLs in the overall, iterative geological disposal development process, including repository siting, design, and post-closure safety assessment. Attention was also given to the need for URLs in salt host rock, since there is currently no such facility in operation in the world. This paper describes some of the outcomes of the workshop in the context of the general timeline for repository development and the evolving role of research, development, and demonstration (RD and D) activities in URLs. Of particular interest is the relationship between technology maturity, as measured by technology readiness level (TRL), and the stages or milestones in the repository development timeline. The long history of RD and D in worldwide URLs (> 40 years) can be argued to be the primary reason that nuclear waste disposal systems are currently at a high level of maturity. One of the key recommendations from the workshop was that an international, easily accessible knowledge base should soon be established to avoid the loss of important experiences and information obtained from existing and closed URLs. It was also recognized that a new generic URL in salt host rock would provide a number of benefits, but one should not be initiated until a national program specifically decides to site its repository in salt. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Increasing the Fissile Mass Loading of High-Level Waste Glass Canisters to Greater Than 2,500 g/m 3 at the Savannah River Site

To eliminate future fissile mass loading constraints for the Savannah River Site H-Canyon Facility and Liquid Waste system, the Savannah River National Laboratory recommends a repository evaluation of a uranium fissile mass loading in glass at 7,144 g/m 3 in addition to the existing International Atomic Energy Agency safeguards and security limit of 2,500 g/m 3 total plutonium. This recommended increase above the authorized 2,500 g/m 3 fissile mass loading limit is based on concentrations of uranium and plutonium shown to produce an acceptable glass waste form rather than projections of the maximum fissile mass loading in future sludge batches. An authorized fissile mass loading limit greater than 2,500 g/m 3 will increase facility flexibility, reduce the number of high-level waste canisters produced, and potentially avoid an increase in the Department of Energy Environmental Management mission life without compromising safety or glass product quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

An underground laboratory in the context of salt disposal research - 15549

Evaluating the basis of need for an underground research laboratory for salt science and engineering is imperative because of the significant commitment of time and money involved. Decades of salt repository studies, numerous experiments, and sophisticated modeling capabilities underpin the scientific database that supports safe disposal of nuclear waste in salt. The safety case for disposal of non-heat-generating waste such as transuranic waste interred at the Waste Isolation Pilot Plant is robust, with the only long-term releases to the environment projected to be by way of human intrusion. The scientific evidence also favors safe disposal of heat-generating waste. Technical evaluations for disposal of heat-generating waste in salt experienced a rather long hiatus in the United States subsequent to certification of the Waste Isolation Pilot Plant and issuance of the Nuclear Waste Policy Act Amendment that ended salt disposal research for the civilian program. Similar salt repository research in Germany was delayed by a ten-year moratorium that ended in 2010. In collaboration with German peers, the United States Department of Energy has reviewed and evaluated thermally driven processes in salt disposal and identified key technical areas in which to prioritize resources. The goal for disposal research in salt is to provide sufficient technical information to license a repository successfully. The necessity or utility of a salt underground laboratory is to be evaluated in the context of an overall research agenda that supports a license application. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Site selection in Germany - Start of the evaluation of the site selection law - 15561

Since 2011 Germany is pursuing a phase out strategy concerning the use of nuclear power for electricity production. This decision was strongly influenced by the Fukushima event. In 2013 the federal government announced that they also had achieved an agreement with the Federal States in Germany on a law to restart the site selection for a repository for spent fuel and high active heat producing waste from scratch. The consequence of this law is a delay of at least two decades to start operation of a final disposal site and additional costs of at least EUR 2.7 billion. The new law was passed in July 2013. At first a 34-member commission had been installed in April 2014 to evaluate the Site Selection Law and to develop basic principles for site selection, including safety requirements and selection criteria for rock formations. The commission includes representatives from the parliament, academia, civil society organizations, industry, the environmental organizations and trade unions and should forward its recommendations after a delay in starting the evaluation process now in 2016. The first author of this paper is member of the evaluation commission. The present law has to be reviewed within this time span. The site selection then might start after the next federal election in 2017 probably based on a new site selection law. This new site selection law will take into account or even follow the recommendations of the evaluation group. A new repository site should be according to the present law determined till 2031 and for this site the more detailed site investigation will take place followed by a detailed safety analysis, before the erection of the repository can start. Based on the present procedural steps it seems to be rather unlikely to determine a repository site till 2031. As shown in this publication, there will be a delay of at least 20 years compared to the schedule given in the site selection law until a repository site can be determined. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Impact of Dutch nuclear fuel cycle scenarios on radionuclide inventory intended for disposal - 15201

One of the main subjects of Dutch five-year research program into geological disposal of radioactive waste, OPERA, is the post-closure radiological safety assessment. Reliable estimates of the radionuclide inventory and matrix composition are an important input for the safety assessment. The amounts and categories of waste foreseen to be disposed in 2130 the Netherlands have been compiled by the Dutch waste management organisation COVRA. Additionally, NRG has made a characterisation of the waste (radionuclides and matrix). This reference inventory is based on the Dutch base nuclear energy scenario of operating the Dutch single reactor, the Borssele NPP, until its presently scheduled shut-down in 2033. However, changes in the presently adopted nuclear fuel cycle strategy in the Netherlands may impact both the quantities of generated radioactive waste and its composition, and therefore the source term of the radiological safety assessment. To get a grip on this uncertainty a set of alternative future fuel cycle scenarios in the Netherlands, which are in compliance with scenarios formulated in the Dutch vision documents, has been considered and evaluated. The scenarios have been analysed with the computer code DANESS, 'Dynamic Analysis of Nuclear Energy System Strategies'. The result of the DANESS analyses concerns the amount and type of the radioactive waste to be finally disposed in a deep geological repository. This also includes the inventories of the individual radionuclides that are relevant for the post-closure safety assessment to be performed in a subsequent stage of the OPERA program. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Summary of Consolidated Interim Storage Advantages and Disadvantages from an Integrated Systems Perspective from Prior Reports and Studies

The question of whether centralized storage of civilian spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. Centralized storage facilities were included as a potential component of the U.S. spent fuel management system in the Nuclear Waste Policy Act of 1982 (NWPA), but these facilities were not identified as being essential. When the NWPA was passed, it was expected that a permanent repository would be available around the turn of the century to meet the commitments that were made to start accepting waste by that time. For a number of reasons, a permanent repository was not available at that stage. This report summarizes the advantages and disadvantages of including an ISF as part of an integrated waste management system that have been identified in numerous analyses of the impacts of a consolidated interim storage facility (ISF), previously discussed as monitored retrievable storage, the form of ISF authorized in the NWPA) by the U.S. Department of Energy (DOE) and independent groups. Storing SNF before disposal is a functional requirement for an integrated waste management system. As described below, the studies reviewed in this report have identified a range of benefits that could be obtained by adding a consolidated ISF to the waste management system, including (1) earlier acceptance of fuel by the federal government, (2) reduction in the number of unintended long-term storage sites, (3) added system flexibility and opportunity for better integration, and (4) near-term development and demonstration of institutional and technical infrastructures for large-scale management of SNF. At the same time, these studies have identified potential drawbacks to deployment of an ISF, including (1) potential adverse impacts on development of a repository, (2) additional transportation of SNF, and (3) a large upfront economic investment to establish the consolidated storage capability. In general, the studies that have considered the positive and negative impacts of consolidated storage most comprehensively have concluded that in light of these considerations, the benefits of including an ISF as part of an integrated waste management system can outweigh any real or perceived disadvantages.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

International Collaboration Activities on Disposal in Argillite R&D: DECOVALEX19 Task C – Geochemical and Reactive Transport Modeling

International collaborations on nuclear waste disposal R&D are an integral part of the Spent Fuel Waste Science and Technology (SFWST) campaign within the DOE Fuel Cycle and Technology (FCT) program. These partnerships with international repository R&D programs provide key opportunities to participate in experiments developing laboratory/field data (underground research laboratories (URL)) of engineered barrier system (EBS) interactions (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field). The results of these experiments are used in the evaluation of coupled processes and their representation via state-of-the-art simulation approaches to evaluate repository performance. During the thermal heating period, increases in temperature from radionuclide decay in the spent fuel (SF) waste canisters will increase temperature in the surrounding EBS driving chemical and transport processes in the near- and far-field domains of the repository. URL heater-tests for extended periods of times (e.g., years) provide key information and data on thermal effects affecting engineered barriers in response to temperature and water saturation levels. Groundwater interactions with cementitious barriers are also important to in-drift chemistry and EBS performance during post-closure. Descriptions of the various URL experiments for various disposal design concepts according to the host country repository program and relevance to the US program is given elsewhere (Birkholzer et al.,2017;Jové Col& et al., 2016). The DECOVALEX-2019 Task C involves collaboration with the GREET (Groundwater REcovery Experiment in Tunnel) at the Mizunami URL, Japan ,which targets the development of monitoring methodologies of groundwater in granitic rock with applications to THMC simulations (Iwatsuki et al., 2005;Iwatsuki et al.,2015,2017). Some of the goals of GREET is to conduct a facility-scale geochemical characterization study of short- and long-term effects of tunnel excavation activities, impacts on groundwater flow and transport, and influences on groundwater chemistry (Iwatsuki et al.,2015). The data obtained from these URL activities is then used in the development and evaluation of THC models to support post-closure safety and performance assessments of the repository environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Assessments and Options for Removal and Immobilization of Volatile Radionuclides from the Processing of Used Nuclear Fuel

Nuclear fuel reprocessing is being considered in the US as part of an effort to help meet the nation’s energy needs. Various options are available to meet these needs (DOE 2010). The recent Fuel Options Study indicated that fuel cycles that utilize continuous recycling are among the “most promising” (Wigeland et al. 2014). After the valuable components have been extracted from the irradiated UO 2 during reprocessing, the remaining waste must be immobilized and sent to a repository. Some of the radionuclides in the fuel are short-lived isotopes that decay to innocuous levels soon after reactor discharge, e.g., 127 Xe (t 1/2 = 36.4 d). However, the preponderance of radionuclides requires long-term, geological storage. In this report, we assess options for capturing and immobilizing radionuclides expected to volatilize during aqueous reprocessing of used nuclear fuel. This document is an update to a previous report prepared in 2011 that was only released as a draft (FCR&D-SWF-2011-000305).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Bentonite buffer under high temperature: laboratory experiments and coupled process modeling

Abstract. Bentonite buffer in a geological repository will be simultaneously heated from decaying radioactive waste and hydrated from the surrounding host rock, triggering complex and coupled THMC (thermal–hydrological–mechanical–chemical) processes. Understanding the THMC behavior of bentonite-based engineered barrier system (EBS) is key to the evaluation and prediction of its long-term performance. Studies on the THMC process have been focused on conditions under 100 ∘C, as most design concepts impose a thermal limit of 100 ∘C in bentonite. Recently, studies under high-temperature conditions have been conducted to evaluate the possibility of raising the thermal limit and expanding the data/knowledge base to increase the confidence level. In this abstract, we present a series of bench-scale laboratory experiments at high temperatures (up to 200 ∘C) and the corresponding modeling work. Two sets of column tests were conducted, and each set consisted of two test columns: a control column undergoing only hydration (non-heated) and an experiment column experiencing both heating and hydration (heated). During the experiment, frequent X-ray computed tomography (CT) images were collected to provide a 3D visualization of the density distribution and present the spatiotemporal evolution of (1) hydration/dehydration, (2) clay swelling/shrinkage, (3) displacement, and (4) mineral precipitation. The two sets of tests differ with respect to several experimental conditions, such as bentonite type, compacted density and water content, water chemistry, and hydration pressure, but the important difference is that the first set used bentonite powder with a dry density of 1.28 g cm−3, whereas the second set used granulated bentonite (mixture of pellets and powder) with a dry density of 1.45–1.5 g cm−3. In both sets of experiments, a comprehensive post-dismantling characterization of bentonite samples was carried out after the column tests had been running for 1.5 years. Comparing non-heated and heated columns, the temperature gradient led to lower degree of homogenization of bentonite after bentonite became fully saturated; comparing the first and second sets, granulated and powdered bentonite exhibited drastically different hydration behavior. A THM model with a 2D axisymmetric grid system was used to interpret the data from the first set of tests. The model considers the combined impact of saturation, fluid pressure, and porosity change due to swelling/compression on the spatiotemporal distribution of bulk density and movement of the thermocouple modules. Observations from the tests help us understand the early perturbation of bentonite buffer under high temperature, and data from these tests improve the calibration of key constitutive hydrological and mechanical models and, therefore, enhance the modeling capability with respect to calculating the long-term evolution of bentonite buffer.

Zheng, Liange↗
DOE OSTI2015

Field Test to Evaluate Deep Borehole Disposal.

The U.S. Department of Energy (DOE) has embarked on the Deep Borehole Field Test (DBFT), which will investigate whether conditions suitable for disposal of radioactive waste can be found at a depth of up to 5 km in the earth’s crust. As planned, the DBFT will demonstrate drilling and construction of two boreholes, one for initial scientific characterization, and the other at a larger diameter such as could be appropriate for waste disposal (the DBFT will not involve radioactive waste). A wide range of geoscience activities is planned for the Characterization Borehole, and an engineering demonstration of test package emplacement and retrieval is planned for the larger Field Test Borehole. Characterization activities will focus on measurements and samples that are important for evaluating the long-term isolation capability of the Deep Borehole Disposal (DBD) concept. Engineering demonstration activities will focus on providing data to evaluate the concept’s operational safety and practicality. Procurement of a scientifically acceptable DBFT site and a site management contractor is now underway. The concept of deep borehole disposal (DBD) for radioactive wastes is not new. It was considered by the National Academy of Science (NAS 1957) for liquid waste, studied in the 1980’s in the U.S. (Woodward–Clyde 1983), and has been evaluated by European waste disposal R&D programs in the past few decades (for example, Grundfelt and Crawford 2014; Grundfelt 2010). Deep injection of wastewater including hazardous wastes is ongoing in the U.S. and regulated by the Environmental Protection Agency (EPA 2001). The DBFT is being conducted with a view to use the DBD concept for future disposal of smaller-quantity, DOE-managed wastes from nuclear weapons production (i.e., Cs/Sr capsules and granular solid wastes). However, the concept may also have broader applicability for nations that have a need to dispose of limited amounts of spent fuel from nuclear power reactors. For such nations the cost for disposing of volumetrically limited waste streams could be lower than mined geologic repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling the Fundamental Radiation Chemistry of the Organic Diluent, and the Effect of Metal Ion Complexation on the Radiochemical Behavior of Active Compounds

All used nuclear fuel (UNF) reprocessing technologies must operate efficiently in the presence of an intense, multi-component (predominantly alpha, beta, and gamma) radiation field. Consequently, radiation-induced degradation of reprocessing systems is of concern, as it negatively impacts process performance over time due to the destruction of both active compounds (ligands, phase modifiers, holdback agents, etc.) and the formation of degradation products. Reprocessing solvent system radiolysis has been linked to changes in separation efficiency and physical properties of solvent mixtures, solvent-recycle longevity, crud formation, and other unexpected outcomes that impact the efficient recovery of valuable materials (e.g., the actinides) and the volume of hazardous radioactive waste for final disposal, i.e., in a geological repository. Consequently, a fundamental understanding of radiolytic processes and their effects on reprocessing solvent system performance is critical for: (i) the cost-effective development and innovation of separation technologies; (ii) the design and implementation of predictive radiation chemical models for process monitoring and lifetimes; and (iii) potentially the ability to exploit radiolytic phenomenon to our benefit, e.g., strategic radiolysis of active molecules to liberate specific degradation products that aid subsequent process stages. Despite extensive investigation into the radiolytic behavior of active solvent system compounds, little attention has been given to understanding (i) the radiation chemical behavior and modification of the organic diluent and (ii) the effect of metal ion complexation on the radiochemical behavior of active compounds.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

Panel Session 62: Progress on Deep Repository Programs Around the World

This panel session focused on the progress of deep repository programs, worldwide. While efforts on a deep geological repository for SNF/UNF and HLW in the USA were terminated in 2010, several other repository programs continue to make considerable progress and deserve to be recognized. Panelists were invited from national, multi-national, and international repository programs and organizations to report upon the status and future as well as the perceived reasons underlying the success and the failure of components of their respective repository program. This panel provided a forum for interested parties to present, exchange, and take advantage of, as appropriate, repository progress, challenges, concepts/designs and other lessons learned around the world, including site characterization, site selection, research and development, licensing, construction, operation, and public acceptance and participation. Presentations proposed include: 1 - Cigeo (French Deep Geological Repository Project) update 2018 (Daniel Delort, Andra (France)); 2 - Progress on Deep Geological Repository Programs Around the World (Bruce McKirdy, Radioactive Waste Management (United Kingdom)); 3 - Geological disposal facility in Belgium (Didier Leonard, ONDRAF/NIRAS (Belgium))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Temperature response and brine availability to heated boreholes in bedded salt

Abstract There is a growing need for disposal of high‐level nuclear waste. To reduce uncertainty associated with brine availability to repository excavations in salt formations, a collaboration between Sandia, Los Alamos, and Lawrence Berkeley National Laboratories is performing a series of borehole‐scale coupled process tests. Here, we report on the first round of the Brine Availability Test in Salt (BATS) project, a “shakedown” experiment called Phase 1s. Experimental testing included placing a resistive heater, a 260‐W radiative heater, and a 750‐W radiative heater within previously drilled horizontal boreholes at the Waste Isolation Pilot Plant (WIPP) while monitoring temperature and water inflow. The experiments successfully achieved the targeted temperature of 120 °C when using the 750‐W radiative heater. Simulations using FEHM (Finite Element Heat and Mass transfer code) and TOUGH‐FLAC (Transport Of Unsaturated Groundwater and Heat–Fast Lagrangian Analysis of Continua) were able to accurately predict the coupled thermo–hydro–mechanical–chemical response of salt, matching the observed temperature and brine production. Due to the extremely low permeability of salt, these systems take many years to reach steady state when perturbed by mining activities. Long‐term numerical simulations are used to develop the initial pressure and saturation conditions. The inclusion of a damaged rock zone with higher permeability around the borehole also affects the saturation and pressure distributions and plays an important role in dissipating the potential for thermal pressurization. Knowledge gained from this round of experimentation and simulation will be used to conduct the next BATS project experiment in new boreholes at WIPP.

58 GEOSCIENCES↗
DOE OSTI2018

THM-Coupled Modeling of Barrier Integrity for HLW Disposal in Bedded Salt - 18289

Within the framework of the R and D project KOSINA in Germany, numerical studies analyzing the long-term integrity of a bedded salt barrier under the thermal loading of heat generating radioactive waste have been performed using different modeling approaches for various repository concepts and two geological structures. The THM-coupled calculations show an initially steep increase of temperatures in the vicinity of the HLW canisters which reduces an uplift above the repository. This in turn leads to a temporally and spatially limited reduction of horizontal stresses at the top of the salt, allowing for pressure-driven percolation of groundwater into the upper part of the hydraulic barrier. However, at any given time an undisturbed salt barrier of over 300 m remains, shielding the repository from groundwater inflow. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Effects of Gamma Radiation on Cementitious Materials in Repository Environment - 16300

Cementitious materials are widely used in the management of radioactive waste and have to perform as required for the period of service life of the installation. In order to assess the long term durability of cement paste in a deep borehole for radioactive wastes, specimens were cast in laboratory and exposed to a gamma radiation field. The effects of the radiation in the specimens were evaluated by X-Ray Diffraction, as changes observed in the mineralogy and microstructure of the material after irradiation doses varying between 0 (without radiation) and 10 MGy. It was observed that the specimens exposed to higher doses of gamma radiation (up to 3 MGy) presented differences between surface and core, when compared with the samples with no radiation or exposed to low doses. These results are important to model the behavior of cementitious materials under the conditions expected in a borehole repository and to provide data for the safety assessment of the installation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Effect of Iodide on Radiolytic Hydrogen Peroxide Generation

This report fulfills the milestone (M4SF-17PN010501041 Report on Radiolysis Modeling for the Defense Repository) to discuss continued integration of the PNNL Radiolysis Model and the ANL Mixed Potential Model. This work concerns the development of an integrated Radiolysis Model (RM) for evaluating defense waste materials (oxide and metal) degradation and radionuclide mobilization. Within an anoxic repository environment, primary oxidizing species (e.g., hydrogen peroxide (H 2 O 2 ), OH• radicals, as well as chlorate and other oxidizing species depending on the disposal environment) will be generated at the surface of the nuclear waste forms as a function of their specific activity. RM development has included expansion of chemical environments considered to encompass species for various disposal environments. PNNL has been coordinating this effort with ANL on the integration of the radiolysis work with the fuel degradation model. In this study, we demonstrate and approximate possible effects of iodide on H 2 O 2 generation. As has been shown these are conditions for which H 2 O 2 generation is reduced. We find that the presence of the iodide ion reduces the steady-state H 2 O 2 concentration, but not to the same degree as bromide at micro-molar concentrations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Experiments and Simulation of a Borehole in Salt to Understand Heat, Brine, and Vapor Migration - 19192

Disposition of heat-generating nuclear waste (HGNW) remains a continuing technical and sociopolitical challenge. Numerous concepts for HGNW management have been proposed and examined internationally, including an extensive focus on geologic disposal. One proposed geologic material is salt because of its low permeability and viscoplastic deformation that causes self-repair of damage done to the salt by waste emplacement activities. Evaluating the safety and technical challenges of storing HGNW in a salt repository is an ongoing process involving experiments and supporting numerical simulation. Currently an experiment is underway at the Waste Isolation Pilot Plant (WIPP) to explore how the presence of a heat generating source affects phenomena such as brine migration, vapor transport, and mechanical changes to the bedded salt. A sub-horizontal heated borehole test is in progress in the underground at the WIPP that includes a centrally located 10.2 cm diameter borehole with an adjustable heater surrounded with smaller diameter boreholes instrumented with thermocouples. The central borehole contains an inflatable packer, heating block, brine sampler, and constantly flowing nitrogen gas circulation system. Air-injection tests performed in the central borehole provide pressure measurements that are used to constrain permeability of the system. The steady-state temperatures, as well as the rise and fall of temperature when the heater is cycled on and off, have been measured for up to 60 days. In the borehole, dry nitrogen gas circulation evaporates water and outflows to a desiccant container where water mass is measured daily during the experiment to quantify vapor removal. Although this test is a 'shake-down' for a planned second round of fresh borehole testing, we have gathered a rich dataset. These data allow us to build simulations using the Finite Element Heat and Mass transfer code (FEHM) to evaluate the experimental results, determine field-scale parameters, and identify code improvements to reproduce important physical processes that may not be accounted for at present. A 3-D numerical mesh, built using LaGrit software (lagrit.lanl.gov; Miller et al., 2007), includes increasing resolution around the central borehole. Modeling of the experiment allows for determining the local thermal conductivity and permeability of any damaged bedded salt around the borehole, where damage from drilling may change the permeability, porosity and saturation conditions, by parameter testing and inverse methods. Additionally, assumptions about brine and vapor flow and transport are being tested by comparing measured and simulated results. The combination of experimental data and model results provide additional data to help support the safety case for safe and effective HGNW disposition in bedded salt formations. Initial results from this experiment show that water flow into the borehole is within previous experimental results. Further, we have found that the design of the heater block is restricting energy to flow into the rock salt. Thus, this test has proven useful in design of the next generation experiment where infrared heating may be used to bypass issues caused by air gaps located around the current stainless steel block heater. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

2020 NEPA Polychlorinated Biphenyl (PCB) Risk Assessment

The Waste Isolation Pilot Plant (WIPP) facility is a U.S. Department of Energy (DOE) operating repository 654 m below the surface in a thick salt formation in southeastern New Mexico. The DOE disposes transuranic (TRU) waste produced from atomic energy defense activities at the WIPP facility. A portion of the waste shipped to the WIPP facility contains TRU radionuclides co-mingled with polychlorinated biphenyls (PCBs), which fall under U.S. Environmental Protection Agency (EPA) regulations implementing the Toxic Substances Control Act (TSCA). This report documents the risks of PCBs co-mingled with TRU waste (hereafter designated as PCB/TRU waste) designated for disposal at the WIPP facility. This analysis is input to the National Environmental Policy Act (NEPA) assessment by the DOE Carlsbad Field Office (CBFO) for the proposed increase of the WIPP facility disposal area to include additional waste panels (but not to increase the legislated WIPP volume). This analysis is not a compliance calculation to support a certification renewal nor does it support a planned change request (PCR) or planned change notice (PCN) to be submitted to the EPA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗