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At least 883 records · Page 49

Corrosion of Steel During Long-Term Exposure to Evolving Cementitious Environments

SRS is proceeding with closure of the H-Area Tank Farm (HTF) and F-Area Tank Farm (FTF). Closure consists of removing the bulk waste, heel removal, and filling the tank with tailored grout formulations. The long-term performance of the concrete and steel materials of construction and the closure grout is an important consideration for Performance Assessment (PA) of closed tanks in FTF and HTF. Over PA timeframes of hundreds, thousands, to tens of thousands of years, the chemical and physical properties of the concrete and steel materials will slowly degrade due to environmental exposure and material aging. In the interim these materials will provide a barrier to the leaching of radionuclides into the soil. Savannah River Remediation (SRR) is updating relevant portions of these analysis inputs to support an imminent revision to the HTF PA and a future update to the FTF PA. This analysis provides an update to the previous PA inputs for steel corrosion. The Central Scenario for the PA analysis includes three postulated modeling cases: 1) Realistic Case, 2) Compliance Case, and 3) Pessimistic Case. This analysis also included a Fast Flow Path Case. This latter case considered the circumstance where the initial condition of the concrete was a completely degraded state and that grout shrinkage exposed the steel to the soil environment immediately. In effect, the steel was unprotected by the concrete and grout. The cases have various degrees of conservatism considered. Chemical, physical and tank configuration parameters were investigated to understand their effects on the predicted time to release of the contaminants. The assessment reviewed the initial tank and steel configuration, service life degradation of the steel, and potential corrosion mechanisms associated with degradation of the concrete materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

An Economics-by-Design Approach Applied to a Heat Pipe Microreactor Concept

Microreactors present a potential paradigm shift in the nuclear industry. Emphasis thus far has been on large-scale multi-billion-dollar projects that cater solely to grid electricity market. These projects can be challenging to finance and execute. On the other hand, microreactors are intended to target a wide variety of smaller niche markets and are expected to be factory-fabricated and more readily deployable. While diseconomies of scale for microreactors may tend to raise their costs per energy output (MWh) relative to large nuclear plants, offsetting gains can be expected from standardization, simplification, passive safety, lower radionuclide inventories, factory fabrication, fast installation, and low financing costs. To adequately assess these contributions, designers should have a different perspective on cost drivers than for large nuclear plants and can utilize novel approaches for systematic cost reduction. To account for these important aspects of microreactors, this report proposes an economics-by-design approach that places economic considerations at the center of the design process. The methodology builds on existing frameworks such as design-to-cost and value engineering, expanding them to new markets (beyond the grid), new attributes (beyond costs alone), and introducing the approach at earlier points in the design cycle. Design parameters and technical specifications are systematically evaluated until costs meet market entry points, while also providing the high-priority performance attributes of the particular use case. Determining first-order estimates for different components early in the process enables designers to focus R&D efforts on the biggest overall cost contributors and components with the most cost uncertainty. The analysis is always guided by market needs and threshold prices. In addition to microreactors, the approach is expected to be useful for other classes of nuclear reactors as well. The analysis was applied to a concept found in the open literature (the Design A heat-pipe reactor). A comprehensive bottom-up estimate was generated by leveraging a new microreactor-specific code of accounts and a range of cost equations. The initial estimate for levelized cost of electricity (LCOE) unsurprisingly exceeded market ranges since the use case had prioritized technological readiness over economic considerations in design choices. An alternate concept was then proposed, with various assumptions/targets made to reduce the largest cost contributors. Changes in the neutron spectrum, the power output, and building structures were found to make even the first-of-a-kind of this modified concept competitive with diesel generation in some remote communities. Learning rate (LR) assumptions indicated cost reductions achieved from sequential unit deployments could expand the range of competitiveness to include additional markets as deployments proceed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Mechanistic Source Term Considerations for Advanced Non-LWRs (Revision 1)

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (MSR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, hindering the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How a mechanistic source term is used within the Licensing Modernization Project (LMP) is also provided. Lastly, the characteristics of non-LWR mechanistic source terms are examined. This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps. This document is an update of SAND 2020-6730. An additional chapter is included as well as edits to original content.

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System Model Calculations for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington

Two lined trenches (Trench 31 and 34) located within the 200 West Low-Level Burial Ground (LLBG) area of the U.S. Department of Energy’s (DOE) Hanford Site are designated for permanent disposal of low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW). In accordance with DOE O 435.1, Radioactive Waste Management, radioactive waste shall be managed and disposed in a manner that is protective of worker, public health and safety, and the environment. DOE O 435.1 requires a site-specific radiological performance assessment (PA) that includes calculations of potential releases and subsequent doses to members of the public for a period of 1,000 years after closure of a low-level waste (LLW) disposal facility. The purpose of this environmental calculation file (ECF) is to document the various dose calculations performed to support DOE’s LLBG PA. The dose assessments are performed to evaluate the potential exposure of an all-pathway representative person to radionuclide contaminants of potential concern (COPCs) that may be released from the LLBG Trenches 31 and 34 to the point of calculation (POC) located at the outer edge of a 100 m buffer zone surrounding the trench boundary. This ECF uses inputs from other ECFs and their associated model package reports (MPRs) supporting the LLBG PA. The five major objectives of this ECF are to present the results of the following: (1) Deterministic dose assessments for both groundwater and atmospheric exposure scenarios; (2) Radon-222 (Rn-222) flux emanating at the surface from the disposed wastes; (3) Uncertainty analyses of the groundwater pathway annual dose due to uncertainty in input parameters; (4) Sensitivity analyses of the input parameters that could potentially impact the fate and transport of the contaminants and doses; (5) Doses arising from inadvertent intruder exposure scenarios.

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Argonne National Laboratory Site Environmental Report (CY2020)

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2020. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 Version 4.1 computer code, was used in preparing this report.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Generic FEPs Catalogue and Salt Knowledge Archive

This report describes the development of a comprehensive catalogue of generic features, events, and processes (FEPs) that are potentially important for the post-closure performance of a repository for high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in salt (halite) host rock. The FEPs and other supporting information have been entered into a “SaltFEP” Database. The generic salt repository FEPs include consideration of relevant FEPs from a number of U.S., Dutch, German, and international FEP lists and should be a suitable starting point for any repository program in salt host rock. The salt FEP catalogue and database employ a FEP classification matrix approach that is based on the concept that a FEP is typically a process or event acting upon or within a feature. The FEP matrix provides a two-dimensional structure consisting of a Features/Components axis that defines the “rows” and a Processes/Events axis that defines the “columns” of the matrix. The design of the FEP classification matrix is consistent with repository performance assessment – the Features/Components axis is organized vertically to generally correspond to the direction of potential radionuclide migration (from the waste to the biosphere) and the Processes/Events axis is designed to represent the common two-way couplings between thermal processes and other processes (such as thermal-mechanical or thermal-hydrologic processes). Related FEPs can be easily identified – related FEPs will typically be grouped in a single matrix cell or aligned along a common row (Feature/Component) or column (Process/Event). The online SaltFEP database can be downloaded from www.saltfep.org. It contains the FEP matrix, the FEPs, and the associated processes for each FEP. It provides a starting point to create and document site-specific individual FEPs. Furthermore, the FEP matrix is connected to the Salt Knowledge Archive (SKA), a database of about 20,000 references and documents representing the historical knowledge on radioactive disposal in salt. This work is the result of an ongoing collaboration between researchers in the U.S., the Netherlands, and Germany, and supports the NEA Salt Club Mandate. It builds upon prior work which is documented.

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Physical Sciences Vistas Issue 3 2021

In this issue, highlights of our contributions to mission operations include descriptions of the following. • How exact and scrupulous planning in concert with sophisticated science enabled a first-ever direct measurement of a radionuclide with a half-life as short as six days. The work, which included operations at the Los Alamos Neutron Science Centers (LANSCE) Weapons Neutron Research Facility and the Isotope Production Facility’s hot cell facility, is a boon for both astrophysics and weapons science. • An introduction to Christie Davis and her role in ensuring the Lab’s execution of simultaneous excellence. • A look at progress by AOT’s Target and Experimental Support Team in reclaiming long-dormant space in a radiological controlled area for new work improving target systems for the Lujan Center. • A story showcasing a cross-organizational effort to minimize the directorate’s legacy and environmental footprint. Staff from across the Lab joined together to ensure the safe and efficient disposal of Rocky Flats legacy waste stored in a transportainer at the Target Fabrication Facility. • How the Lab’s Utilities and Infrastructure experts collaborated with stakeholders in our directorate to safety execute preventive maintenance on the LANSCE mesa’s large electrical substation. The Safe Conduct of Research principles provided a common framework for the planning and process. • A description of the resources called upon and the procedures undertaken by MPA to inventory the division’s time-sensitive chemicals, including the careful and conscientious response of an alert team member when something seemed amiss.

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Silicon-Equivalent Neutron Dose Estimations

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within the various nuclear materials. Under long-term radiation exposure, some components may show signs of material damage such as hardening, cracking, discoloration, etc. resulting from the dose received from nearby radioactive sources. Radiation dose quantities are a useful means to correlate material damage or degradation to radiation interactions. The degradation is often dependent on how incident radiation interacts and deposits energy into the material. This dependence implies that radiation dose should be properly defined for the material effect of interest. Generic dose quantities, such as total energy deposited, may not necessarily correlate well for degradation of any given material.

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Evaluation of Attila and MCNP computational methods for dose and exposure estimation

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within various nuclear materials. Historically, dose calculations use MCNP (Monte Carlo N-Particle) transport code as the primary particle transport tool without a secondary computational tool to validate the results from the MCNP simulations [1]. The goal of this study is to make an independent check of the Monte Carlo solution from MCNP6 Version 6.2.1 with the discrete ordinates solution from Attila 10.2.0 Beta 3. As an example problem for this study, water-filled, stainless-steel vessels, modeled with an unstructured mesh (UM) with both MCNP and Attila [2], are exposed to 252Cf and 60Co point sources. This report also includes a discussion of the limitations of unstructured mesh in a MCNP calculation.

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Molten Salt Reactor Fundamental Safety Function PIRT

The phenomenon identification and ranking table (PIRT) process was employed to evaluate the current capability to assess the ability of liquid-salt–fueled molten salt reactors to achieve their fundamental safety functions (FSFs). The PIRT process provides a structured mechanism to elicit and document expert opinions on the most important phenomena and the corresponding level of knowledge with regard to achieving the FSFs. The PIRT panel included reactor developers, accident progression evaluation tool developers, US Department of Energy (DOE) national laboratory technical staff, university researchers, and US Nuclear Regulatory Commission (NRC) staff. The information-gathering process concentrated on phenomena related to accidents in which the fuel salt—including any in the cover gas—has been released from the first barrier layer. This information-gathering process included identifying all potential accidents that could result in the release of a substantial amount of radionuclides that could breach or bypass the first barrier. The PIRT did not identify any previously unrecognized systemic vulnerabilities. However, the elicitation process identified several areas with low levels of knowledge and significant potential impact on accident progression. These areas either require additional research to improve the state of knowledge or additional design conservatism to accommodate the remaining unknowns.

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Development of an Activation Analysis Methodology to Support the Disposal of the High Flux Isotope Reactor Metal Pool Waste

During operation and maintenance of the High Flux Isotope Reactor (HFIR), metal waste from reactor components is generated. Due to space limitations in the pool fuel storage, it is necessary to remove the waste and send it to off-site disposal. Most of this waste was in the reactor and is radioactive through neutron activation. Prior to shipping and disposal at a waste disposal facility, the radiological activity of the waste must be determined. Therefore, a conservative methodology to characterize the radiological activity of the waste must be performed. The methodology used was based on being able to determine a conservative overall activity and radionuclide inventory based on radiological survey data and neutron activation products for the materials present.

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Testing to Evaluate Processes Expected to Occur during MSR Salt Spill Accidents

Obtaining a license for a new nuclear reactor requires the identification and assessment of the potential consequences of specified accident scenarios, which are achieved using accident progression models. Those models need to be parameterized and validated using experimental data, but existing experimental data addressing processes relevant to molten salt reactor accidents are sparse. Specifically, experimental data that quantify the sensitives of important processes to the initial conditions of the spill, the ambient environment, and the containment features are needed to parameterize individual process models. Integrated experiments that simulate accident scenarios are also needed to provide data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation. The overarching objectives of this work are to develop the methods for simulating the targeted processes, to determine the effectiveness of the methods in producing the data required for model development, to generate data that can be used to parameterize individual process models, and to provide key insights into the behavior of spilled molten salt that should be considered in models. Experimental methods were designed to quantify aspects of individual processes expected to occur during or after a molten salt spill accident that will affect the fate of spilled molten salt and the radionuclides within. These processes include 1) molten salt spreading and heat transfer, 2) molten salt flowing and freezing in tubing, 3) stainless steel corrosion kinetics in molten salt, and 4) molten salt splashing and aerosol generation. The initial tests described in this report were conducted using eutectic FLiNaK to demonstrate the test methods, the data that are generated, and the analyses of the data to derive values needed for modeling. The primary variables that were tested include initial salt temperature and the presence of volatile surrogate fission products (e.g., cesium and iodine). The developed methods are shown to be effective in quantifying the desired processes and can be applied to study more complex salt compositions of interest to molten salt reactor developers, a wide range of environmental conditions of interest to modelers, and additional variables relevant to salt spill accidents. The developed methods and insights gained from laboratory tests can also be incorporated in future large-scale integrated tests used to simulate molten salt spill accidents.

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Biokinetic and Dosimetric Models [Slides]

Presentation describes internal dosimetry as an intersection of disciplines involving Physiology, Anatomy, Physics, Mathematics and Computer Science, which some people call "The art of Internal Dosimetry". The basic components for internal dose calculations are: Biokinetic (metabolic) models describing the intake, distribution, retention and excretion of radionuclides in the body, Dosimetric models describing the interaction of radiation within the several body organs and tissues, and System of dose limitation.

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Organic Iodide Sorption from Dilute Gas Streams

Reprocessing used nuclear fuel releases volatile radionuclides, including 129 iodine (I), into the off-gas of a processing plant. Volatile radioiodine could be present in several forms, depending on the chemistry of the process used and the off-gas stream. Inorganic I 2 is expected to be the predominant I species in the dissolver off-gas (DOG), with minor organic iodides present. The bulk of the I is expected to volatilize into the DOG in parts-per-million-level (ppm) concentrations. In contrast, in the vessel off-gas (VOG), most of the volatile I is expected to be found as organic iodides, such as CH 3 I, C 4 H 9 I, and C 12 H 25 I. These species are expected to be present in parts-per-billion-level (ppb) concentrations but require abatement, even at low expected concentrations, to meet regulatory emissions limits in the United States. Historically, studies of I abatement by Ag-functionalized sorbents have focused on inorganic I in the DOG, but in the last few years, more research attention has been given to organic iodides, especially longer chain species, such as C 4 H 9 I, and C 12 H 25 I. This report has three main goals: (1) to present new data generated at Oak Ridge National Laboratory (ORNL) in FY21 on the sorption behavior of organic iodides on AgZ, (2) to summarize and synthesize organic iodide data produced by ORNL and Idaho National Laboratory (INL) over the last 4 years to answer questions on organic iodides behavior outlined in the 2018 joint test plan (Jubin et al. 2018), and (3) to propose a VOG abatement system design that can provide the capture efficiencies required to meet I emission limits. The 2021 ORNL experimental campaign tested the effects of organic iodide speciation and concentration in the off-gas, superficial velocity of the off-gas, and effects of aging on AgZ sorbent capacity. These studies found that the sorption rate of organic iodides by AgZ depends on the hydrocarbon chain length and the concentration in the off-gas. Higher molecular weight organic iodides adsorb to AgZ more slowly than I. At a concentration of 50 ppm concentration in the off-gas, CH 3 I loads 8% slower, C 4 H 9 I loads 20% slower, and C 12 H 25 I loads 40% slower than I. The lowest concentration loading rates calculated were in 5 ppm organic iodide gas streams in which AgZ gained on average 0.14 mg I/g sorbent/hour in the bench scale test system. Thus, longer sorbent beds might be needed to accommodate slower loading rates onto AgZ in lower concentration gas streams. Although sorption rate varies as a function of hydrocarbon chain length, the saturation concentration of the sorbent for these I-bearing species does not vary. Aging AgZ in a humid air stream for 9 months drops the overall sorbent capacity by ~35% for CH 3 I, ~50% for C 4 H 9 I, and ~40% for C 12 H 25 I. This results in a saturation capacity between 35 and 70 mg I/g sorbent for the aged AgZ. In conjunction with recent data produced by INL, these data are used to estimate the mass transfer zone (MTZ) and decontamination factor (DF) for sorbent beds of AgZ. Sorption tests performed with iodide gas concentrations of about 1 ppm and higher at a superficial gas velocity of 10 m/min, indicate that MTZ depths for these conditions tend to range between about 8-20 cm. Tests performed at lower concentrations between 50-90 ppb and at gas superficial velocities of 1, 10, and 20 m/min indicate that the MTZ depth increases with increasing superficial gas velocity. The 20 m/min test indicates that the MTZ for those conditions was at least 22 cm, and doubling the superficial gas velocity from 10 to 20 m/min could roughly double or triple the MTZ depth. Doubling and tripling the bounding MTZ depth of 20 cm for the body of MTZ estimates made at with 10 m/min superficial velocity would extend the MTZ for a superficial velocity of 20 m/min to 40-60 cm. This bounding limit applies to all of the organic iodides that have been tested. These results also indicate that the sorption rate-limiting step is not sensitive to the superficial gas velocity; otherwise the MTZ depth would not have increased approximately in proportion to the increase in the gas superficial velocity. This further suggests that the rate limiting step is not associated with mass transfer of the sorbate to the sorbent surface, or mass transfer of the reaction byproducts from the sorbent surface, but is associated with sorption or chemical reactions on the sorbent surface or in sorbent pores. Deep bed testing at INL has established DFs of >2,000 for I, CH 3 I, and C 4 H 9 I under a range of conditions (Soelberg et al. 2021, Bruffey et al. 2019). DF does not seem to be affected by the concentration of the organic iodide in the gas stream over the range of 1 to 50 ppm. Thus, if the MTZ is accommodated in sorbent bed design for the DOG and VOG, then regulatory DFs will be met. To meet the third objective outlined in this report, these experimental data were used to update an engineering evaluation of the VOG first completed in 2016. The updated VOG design can be found in an accompanying document (Welty et al., 2021; INL- LTD-21-64587). This report finds that the VOG will decrease in both size and complexity, relative to previous designs, and will still meet regulatory requirements for all iodine forms.

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Combined Technologies for In Situ Remediation of Tc-99 and U in Subsurface Sediments

In this study, combinations of chemical remedies were tested in bench-scale batch experiments to evaluate a two-step reduction-sequestration approach to effectively stabilize high concentrations of inorganic contaminant mixtures. Bench tests simulated contaminant and geochemical conditions of a perched aquifer located within the Central Plateau at the Hanford Site, located in southeastern Washington State (USA). Pairwise combinations of a reductant [e.g., zero valent iron, sulfur modified iron (SMI), or calcium polysulfide] and a sequestering agent [e.g., calcite, apatite, or dilute alkaline solution (e.g., NaOH)] were evaluated for immobilization and stabilization of technetium (Tc) (50,000 pCi/L), uranium (U) (150 mg/L), and nitrate (NO 3 ) (200 mg/L) in high ionic strength groundwater. The results of these batch studies demonstrated that reduction by SMI and sequestration in apatite or calcite are the most effective combination for these contaminant mixtures and conditions. Aqueous concentrations of Tc and U decreased by 95.6% ± 2.5% and 101.1% ± 5.2%, respectively, with SMI-apatite and 98.3% ± 0.0% and 101.2% ± 5.2%, respectively, with SMI-calcite. Sequential extractions showed that sequestered contaminants had limited capacity for re-oxidation; in fact, less than 10% of immobilized Tc and U was recovered by selective extraction of mineral phases most susceptible to oxidation. In addition, X-ray absorption near edge structure analysis of the sediment samples treated with SMI-calcite showed the presence of only U(IV), while both U(IV) and U(VI) were present in the SMI apatite combination [ratio of 0.43 U(IV):0.59 U(VI)].This study describes preliminary results that a two-step approach for stabilizing contaminant mixtures of long-lived radionuclides can be effective at reducing contaminant fluxes to groundwater from vadose and perched water zones.

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Nuclear Forensics National Frameworks [Slides]

Nuclear forensics is the examination of nuclear or other radioactive material, or of evidence contaminated with radionuclides, in the context of legal proceedings under international or national law related to nuclear security. The goal of forensic science is to discover linkages among people, places, materials, and events. Nuclear forensics is an essential component of national response plans to events involving nuclear or other radioactive materials out of regulatory control and informs prevention, detection, and response.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Environmental Protection Agency Drinking Water Protective Action Guides Implementation Recommendations

This report documents the analysis conducted by Sandia National Laboratories on the effect of various alterations to published methodologies to calculate Derived Response Levels for Environmental Protection Agency Drinking Water Protective Action Guides. Specifically, this study sought to assess and provide recommendations on calculation of the Derived Response Level accounting for decay during the consumption period, assess the impact of decay on laboratory Minimal Detectable Concentration in water samples as compared to the Derived Response Level, make a recommendation on the calculation of Derived Response Level consistent with existing Public Protection Methods, and make a recommendation on the use of six age groups versus eight age groups based on available dose coefficients for calculation of the Derived Response Level. The authors analyzed these various factors using nominal radionuclide mixes from four scenarios and compared calculation of the Derived Response Level accounting for decay and no decay and then compared those results to the laboratory Minimal Detectable Concentrations. The authors concluded that decay should be included in the calculation of the Derived Response Level, existing Public Protection Methods should be employed to calculate the Derived Response Level, six age groups should be used versus eight, and the use of both decay and Public Protection Methods result in little to no concern for water samples meeting the Minimum Detectable Concentration requirements. The results of this study may be used in further developing and implementing a method for the Environmental Protection Agency Water Derived Response Level calculation in the Federal Radiological Monitoring and Assessment Center Assessment Manual.

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E-Area Low Level Waste Generator Inventory Uncertainty Estimation

This report describes the method used to estimate the uncertainty and bias in the low-level waste inventory, as reported by Waste Generators, for waste disposed in the E-Area Low-Level Waste Facility at SRS. The waste cut uncertainty and bias is summed to calculate the current total uncertainty and bias in each disposal unit in the E-Area Low-Level Waste Facility. This information is used to estimate the future uncertainty and bias when all disposal units are closed. For most waste cuts, there are two areas of inventory uncertainty and bias: in the isotopic characterization of the waste stream and in the measurement of total waste cut activity. Waste stream characterization is accomplished through a combination of process knowledge and analytical measurements. This characterization is documented by a technical baseline that describes the relative quantity of each isotope present in the waste stream. The total activity of each individual waste cut is usually measured directly or is calculated through a ratio of activity per weight or activity per volume. The waste cut total activity is multiplied by the waste stream isotopic fraction to calculate the total isotopic activity present in the waste cut. The waste cut uncertainty and bias is calculated by analyzing the uncertainties and biases in the waste stream characterization and waste cut measurements and adding the uncertainties in quadrature and multiplying the reported isotopic activities by the bias factors. The current isotopic inventory, uncertainty, and bias of a disposal unit is the sum of the currently disposed individual waste cut isotopic activities, uncertainties, and biases. The future uncertainty and bias is estimated based on the average “worst-case” (largest) uncertainties and biases of currently closed and operational disposal units. This report details the method, data sources, and calculation techniques used to calculate these values. It also presents some of the difficulties in working with the reported data and gives examples of the calculations and analysis. The software used to carryout the analysis is included and the location of the complete individual analyses is given. After discussion with waste generators, examination of multiple waste characterization documents, and a review of SRS Manual 1S, the reported ELLWF radionuclide activities are clearly conservative and likely represent reasonable upper bounds for the true activities. However, quantifying the bias and uncertainty in each waste cut requires detailed analysis of the documentation for each waste cut measurement and waste stream characterization.

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