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DOE OSTI2024

Drifting Acoustic Measurements around C-Power's SeaRay WEC

The repository contains underwater noise measurements and associated metadata collected around C-Power's SeaRay wave energy converter on July 15, 2024 and July 16, 2024 while it was deployed at the U.S. Navy's Wave Energy Test Site (WETS) in Kaneohe, HI. Measurements were obtained using Drifting Acoustic Instrumentation SYstems (DAISYs). DAISYs consist of a surface expression connected to a hydrophone recording package by a tether. Both elements are instrumented to provide metadata (e.g., position, orientation, and depth). Information about how to build DAISYs is available at https://www.pmec.us/research-projects/daisy. The repository's primary content is a compressed archive (.zip format), containing multiple MATLAB binary data files (.mat format). The structure of each file is included in the repository as a Word document (Data Description MHK-DR.docx). Each file contains time series information for a single DAISY deployment (file naming convention: WETS_DAISY_[Drift #].mat) consisting of processed hydrophone data and associated metadata. During these measurements, C-Power's SeaRay was located at approximately 21.48112 N, 157.74451 W.

16 TIDAL AND WAVE POWER↗
DOE OSTI2018

FRAMES Metadata Reporting Templates for Ecohydrological Observations, version 1.1

FRAMES is a a set of Excel metadata files and package-level descriptive metadata that are designed to facilitate and improve capture of desired metadata for ecohydrological observations. The metadata are bundled with data files into a data package and submitted to a data repository (e.g. the NGEE Tropics Data Repository) via a web form. FRAMES standardizes reporting of diverse ecohydrological and biogeochemical data for synthesis across a range of spatiotemporal scales and incorporates many best data science practices. This version of FRAMES supports observations for primarily automated measurements collected by permanently located sensors, including sap flow (tree water use), leaf surface temperature, soil water content, dendrometry (stem diameter growth increment), and solar radiation. Version 1.1 extend the controlled vocabulary and incorporates functionality to facilitate programmatic use of data and FRAMES metadata (R code available at NGEE Tropics Data Repository).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Analcime-wairakite formation during experimental cement-bentonite alteration at 200–300 °C

Engineered barrier system materials in nuclear waste repositories may undergo hydrothermal alteration in response to groundwater saturation and heating events over their long operational timescales. Hydrothermal interactions between engineered materials (e.g., bentonite buffers, cements, waste canister materials) and the host rock environment may drive alteration processes that affect the advantageous properties of some barrier materials. However, such alteration will also promote the formation of zeolites, hydrothermally formed minerals that may themselves act to isolate radionuclides. Understanding the environmental conditions that control radionuclide-sorbing properties of zeolites, such as Si/Al ratio or Na content, will be valuable for assessing the changing properties of engineered barrier materials in the case of an in situ heating event. Here, we present experimental work characterizing the formation of zeolites during hydrothermal interactions between generic clay and cement barrier materials, with a focus on the analcime-wairakite zeolite series, which has known radionuclide sorption and exchange properties. We present the results of hydrothermal experiments combining uncured ordinary Portland cement powder with steel and Wyoming bentonite in the presence of Opalinus clay and a synthesized Opalinus clay groundwater to simulate water-saturated conditions in an argillaceous rock repository. The experiments were conducted isothermally at 200 or 300 °C for 8 to 24 weeks. In contrast to analogous studies that did not include cement reactants, we observed the formation of analcime-group minerals in all experiments. The addition of cement resulted in lower Si/Al ratios in the zeolites compared to similar studies that did not include cement. The presence of cement as a reactant was interpreted to promote analcime formation at 200 °C. At 300 °C, we observed higher calcium and silica concentrations in the aqueous solutions as well as increased wairakite formation and decreased analcime formation compared to the experiments at 200 °C. These results show a fully realized analcime-wairakite solid solution that falls between Si/Al = 2 (ideal) and a trend of analcime minerals that have increasing Si/Al ratios with increasing Na/(Na+Ca). Our results show predictable relationships between the Si/Al ratio and the analcime-wairakite content under hydrothermal conditions and illustrate that repository material interactions may promote the formation of zeolites in the analcime-wairakite solid-solution series during heating events in the subsurface.

58 GEOSCIENCES↗
DOE OSTI2012

Benchmarking a new closed-form thermal analysis technique against a traditional lumped parameter, finite-difference method

A benchmarking effort was conducted to determine the accuracy of a new analytic generic geology thermal repository model developed at LLNL relative to a more traditional, numerical, lumped parameter technique. The fast-running analytical thermal transport model assumes uniform thermal properties throughout a homogenous storage medium. Arrays of time-dependent heat sources are included geometrically as arrays of line segments and points. The solver uses a source-based linear superposition of closed form analytical functions from each contributing point or line to arrive at an estimate of the thermal evolution of a generic geologic repository. Temperature rise throughout the storage medium is computed as a linear superposition of temperature rises. It is modeled using the MathCAD mathematical engine and is parameterized to allow myriad gridded repository geometries and geologic characteristics [4]. It was anticipated that the accuracy and utility of the temperature field calculated with the LLNL analytical model would provide an accurate 'birds-eye' view in regions that are many tunnel radii away from actual storage units; i.e., at distances where tunnels and individual storage units could realistically be approximated as physical lines or points. However, geometrically explicit storage units, waste packages, tunnel walls and close-in rock are not included in the MathCAD model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Investigation of Coupled THMC Processes and Reactive Transport: FY14 Progress

The multi-barrier system in a nuclear waste repository typically consists of the natural barrier system (NBS), which includes the repository host rock and its surrounding subsurface environment, and the engineered barrier system (EBS). The EBS represents the man-made, engineered materials placed within a repository, including the waste form, waste canisters, buffer materials, backfill, and seals (OECD 2003).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Materials for Consideration in Standardized Canister Design Activities.

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to establish corrosion rates and component lifetimes. Finally, it is unlikely that the aluminum-based neutron absorber materials that are commonly used in existing DPCs would survive for 10,000 years in disposal environments, because the aluminum will act as a sacrificial anode for the steel. We recommend additional testing of borated and Gd-bearing stainless steels, to establish general and localized corrosion resistance in repository-relevant environmental conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Evaluation of Used Fuel Disposition in Clay-Bearing Rock

Radioactive waste disposal in shale/argillite rock formations has been widely considered given its desirable isolation properties (low permeability), geochemically reduced conditions, anomalous groundwater pressures, and widespread geologic occurrence. Clay/shale rock formations are characterized by their high content of clay minerals such as smectites and illites where diffusive transport and chemisorption phenomena predominate. These, in addition to low permeability, are key attributes of shale to impede radionuclide mobility. Shale host-media has been comprehensively studied in international nuclear waste repository programs as part of underground research laboratories (URLs) programs in Switzerland, France, Belgium, and Japan. These investigations, in some cases a decade or more long, have produced a large but fundamental body of information spanning from site characterization data (geological, hydrogeological, geochemical, geomechanical) to controlled experiments on the engineered barrier system (EBS) (barrier clay and seals materials). Evaluation of nuclear waste disposal in shale formations in the USA was conducted in the late 70’s and mid 80’s. Most of these studies evaluated the potential for shale to host a nuclear waste repository but not at the programmatic level of URLs in international repository programs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Identification of Process Suitable Diluent

The Sigma Team for Minor Actinide Separation (STMAS) was formed within the USDOE Fuel Cycle Research and Development (FCRD) program in order to develop more efficient methods for the separation of americium and other minor actinides (MA) from used nuclear fuel. The development of processes for MA separations is driven by the potential benefits; reduced long-term radiotoxicty of waste placed in a geologic repository, reduced timeframe of waste storage, reduced repository heat load, the possibility of increased repository capacity, and increased utilization of energy potential of used nuclear fuel. The research conducted within the STMAS framework is focused upon the realization of significant simplifications to aqueous recycle processes proposed for MA separations. This report describes the research efforts focused upon the identification of a process suitable diluent for a flowsheet concept for the separation of MA which is based upon the dithiophosphinic acid (DPAH) extractants previously developed at the Idaho National Laboratory (INL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Investigation of Reactive Transport and Coupled THMC Processes in the EBS

Geological repositories for disposal of high-level nuclear waste generally rely on a multibarrier system to isolate radioactive waste from the biosphere. The multi-barrier system typically consists of the natural barrier system (NBS), which includes the repository host rock and its surrounding subsurface environment, and the engineered barrier system (EBS). The EBS represents the man-made, engineered materials placed within a repository, including the waste form, waste canisters, buffer materials, backfill, and seals (OECD, 2003). The EBS plays a significant role in the containment and long-term retardation of radionuclide release.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

DR Argillite Disposal R&D at LBNL

Within the Natural Barrier System (NBS) group of the Used Fuel Disposition (UFD) Campaign at the Department of Energy’s (DOE) Office of Nuclear Energy, LBNL’s research activities have focused on understanding and modeling EDZ evolution and the associated coupled processes and impacts of high temperature on parameters and processes relevant to performance of a clay repository to establish the technical base for the maximum allowable temperature. This report documents results from some of these activities. These activities address key Features, Events, and Processes (FEPs), which have been ranked in importance from medium to high, as listed in Table 7 of the Used Fuel Disposition Campaign Disposal Research and Development Roadmap (FCR&D-USED-2011-000065 REV0) (Nutt, 2011). Specifically, they address FEP 2.2.01, Excavation Disturbed Zone, for clay/shale, by investigating how coupled processes affect EDZ evolution; FEP 2.2.05, Flow and Transport Pathways; and FEP 2.2.08, Hydrologic Processes, and FEP 2.2.07, Mechanical Processes, and FEP 2.2.09, Chemical Process—Transport, by studying near-field coupled THMC processes in clay/shale repositories. The activities documented in this report also address a number of research topics identified in Research & Development (R&D) Plan for Used Fuel Disposition Campaign (UFDC) Natural System Evaluation and Tool Development (Wang 2011), including Topics S3, Disposal system modeling – Natural System; P1, Development of discrete fracture network (DFN) model; P14, Technical basis for thermal loading limits; and P15 Modeling of disturbed rock zone (DRZ) evolution (clay repository).

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

The On-line Waste Library (OWL): Usage and Inventory Status Report

The Waste Form Disposal Options Evaluation Report (SNL 2014) evaluated disposal of both Commercial Spent Nuclear Fuel (CSNF) and DOE-managed HLW and Spent Nuclear Fuel (DHLW and DSNF) in the variety of disposal concepts being evaluated within the Used Fuel Disposition Campaign. That work covered a comprehensive inventory and a wide range of disposal concepts. The primary goal of this work is to evaluate the information needs for analyzing disposal solely of a subset of those wastes in a Defense Repository (DRep; i.e., those wastes that are either defense related, or managed by DOE but are not commercial in origin). A potential DRep also appears to be safe in the range of geologic mined repository concepts, but may have different concepts and features because of the very different inventory of waste that would be included. The focus of this status report is to cover the progress made in FY16 toward: (1) developing a preliminary DRep included inventory for engineering/design analyses; (2) assessing the major differences of this included inventory relative to that in other analyzed repository systems and the potential impacts to disposal concepts; (3) designing and developing an on-line waste library (OWL) to manage the information of all those wastes and their waste forms (including CSNF if needed); and (4) constraining post-closure waste form degradation performance for safety assessments of a DRep. In addition, some continuing work is reported on identifying potential candidate waste types/forms to be added to the full list from SNL (2014 – see Table C-1) which also may be added to the OWL in the future. The status for each of these aspects is reported herein.

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

International Collaboration Activities on Engineered Barrier Systems

The Used Fuel Disposition Campaign (UFDC) within the DOE Fuel Cycle Technologies (FCT) program has been engaging in international collaborations between repository R&D programs for high-level waste (HLW) disposal to leverage on gathered knowledge and laboratory/field data of near- and far-field processes from experiments at underground research laboratories (URL). Heater test experiments at URLs provide a unique opportunity to mimetically study the thermal effects of heat-generating nuclear waste in subsurface repository environments. Various configurations of these experiments have been carried out at various URLs according to the disposal design concepts of the hosting country repository program. The FEBEX (Full-scale Engineered Barrier Experiment in Crystalline Host Rock) project is a large-scale heater test experiment originated by the Spanish radioactive waste management agency (Empresa Nacional de Residuos Radiactivos S.A. – ENRESA) at the Grimsel Test Site (GTS) URL in Switzerland. The project was subsequently managed by CIEMAT. FEBEX-DP is a concerted effort of various international partners working on the evaluation of sensor data and characterization of samples obtained during the course of this field test and subsequent dismantling. The main purpose of these field-scale experiments is to evaluate feasibility for creation of an engineered barrier system (EBS) with a horizontal configuration according to the Spanish concept of deep geological disposal of high-level radioactive waste in crystalline rock. Another key aspect of this project is to improve the knowledge of coupled processes such as thermal-hydro-mechanical (THM) and thermal-hydro-chemical (THC) operating in the near-field environment. The focus of these is on model development and validation of predictions through model implementation in computational tools to simulate coupled THM and THC processes.

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

International Approaches for Nuclear Waste Disposal in Geological Formations: Geological Challenges in Radioactive Waste Isolation—Fifth Worldwide Review

The overall objective of the Fifth Worldwide Review (WWR-5) is to document the current state-of-the-art of major developments in a number of nations throughout the World pursuing geological disposal programs, and to summarize challenging problems and experience that have been obtained in siting, preparing and reviewing cases for the operational and long-term safety of proposed and operating nuclear waste repositories. The scope of the Review is to address current specific technical issues and challenges in safety case development along with the interplay of technical feasibility, siting, engineering design issues, and operational and post-closure safety. In particular, the chapters included in the report present the following types of information: the current status of the deep geological repository programs for high level nuclear waste and low- and intermediate level nuclear waste in each country, concepts of siting and radioactive waste and spent nuclear fuel management in different countries (with the emphasis of nuclear waste disposal under different climatic conditions and different geological formations), progress in repository site selection and site characterization, technology development, buffer/backfill materials studies and testing, support activities, programs, and projects, international cooperation, and future plans, as well as regulatory issues and transboundary problems.

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

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock: FY17 Progress. Predecisional Draft

The focus of research within the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign is on repository-induced interactions that may affect the key safety characteristics of EBS bentonite and an argillaceous rock. These include thermal-hydrologicalmechanical- chemical (THMC) process interactions that occur as a result of repository construction and waste emplacement. Some of the key questions addressed in this report include the development of fracturing in the excavation damaged zone (EDZ) and THMC effects on the near-field argillaceous rock and buffer materials and petrophysical characteristics, particularly the impacts of temperature rise caused by waste heat. This report documents the following research activities. Section 2 presents THM model developments and validation, including modeling of underground heater experiments at Mont Terri and Bure underground research laboratories (URLs). The heater experiments modeled are the Mont Terri FE (Full-scale Emplacement) Experiment, conducted as part of the Mont Terri Project, and the TED in heater test conducted in Callovo-Oxfordian claystone (COx) at the Meuse/Haute-Marne (MHM) underground research laboratory in France. The modeling of the TED heater test is one of the Tasks of the DEvelopment of COupled Models and their VAlidation against EXperiments (DECOVALEX)-2019 project. Section 3 presents the development and application of thermal-hydrological-mechanical-chemical (THMC) modeling to evaluate EBS bentonite and argillite rock responses under different temperatures (100 °C and 200 °C). Model results are presented to help to understand the impact of high temperatures on the properties and behavior of bentonite and argillite rock. Eventually the process model will support a robust GDSA model for repository performance assessments. Section 4 presents coupled THMC modeling for an in situ test conducted at Grimsel underground laboratory in Switzerland in the Full-Scale Engineered Barrier Experiment Dismantling Project (FEBEX-DP). The data collected in the test after almost two decades of heating and two dismantling events provide a unique opportunity of validating coupled THMC models and enhancing our understanding of coupled THMC process in EBS bentonite. Section 5 presents a planned large in-situ test, “HotBENT,” at Grimsel Test Site, Switzerland. In this test, bentonite backfilled EBS in granite will be heated up to 200 °C, where the most relevant features of future emplacement conditions can be adequately reproduced. Lawrence Berkeley National Laboratory (LBNL) has very actively participated in the project since the very beginning and have conducted scoping calculations in FY17 to facilitate the final design of the experiment. Section 6 presents present LBNL’s activities for modeling gas migration in clay related to Task A of the international DECOVALEX-2019 project. This is an international collaborative activity in which DOE and LBNL gain access to unique laboratory and field data of gas migration that are studied with numerical modeling to better understand the processes, to improve numerical models that could eventually be applied in the performance assessment for nuclear waste disposal in clay host rocks and bentonite backfill. Section 7 summarizes the main research accomplishments for FY17 and proposes future work activities.

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

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

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

International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review

An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Pressurized Slot Testing to Determine Thermo-Mechanical Properties of Lithophysal Tuff at Yucca Mountain Nevada.

The study described in this report involves heated and unheated pressurized slot testing to determine thermo-mechanical properties of the Tptpll (Tertiary, Paintbrush, Topopah Spring Tuff Formation, crystal poor, lower lithophysal) and Tptpul (upper lithophysal) lithostratigraphic units at Yucca Mountain, Nevada. A large volume fraction of the proposed repository at Yucca Mountain may reside in the Tptpll lithostratigraphic unit. This unit is characterized by voids, or lithophysae, which range in size from centimeters to meters, making a field program an effective method of measuring bulk thermal-mechanical rock properties (thermal expansion, rock mass modulus, compressive strength, time-dependent deformation) over a range of temperature and rock conditions. The field tests outlined in this report provide data for the determination of thermo-mechanical properties of this unit. Rock-mass response data collected during this field test will reduce the uncertainty in key thermal-mechanical modeling parameters (rock-mass modulus, strength and thermal expansion) for the Tptpll lithostratigraphic unit, and provide a basis for understanding thermal-mechanical behavior of this unit. The measurements will be used to evaluate numerical models of the thermal-mechanical response of the repository. These numerical models are then used to predict pre- and post-closure repository response. ACKNOWLEDGEMENTS The authors would like to thank David Bronowski, Ronnie Taylor, Ray E. Finley, Cliff Howard, Michael Schuhen (all SNL) and Fred Homuth (LANL) for their work in the planning and implementation of the tests described in this report. This is a reprint of SAND2004-2703, which was originally printed in July 2004. At that time, it was printed for a restricted audience. It has now been approved for unlimited release.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modeling Radionuclide Releases from TRISO Particles by Simultaneous Diffusion Through and Corrosion of the Silicon Carbide Barrier Layer

TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗