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DOE OSTI2017

AGR Equipment and Materials Disposition

Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWXT equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

AGR Equipment and Materials Disposition

Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWX Technologies, Inc. (BWXT) equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository. Representatives for the United States Department of Energy may intervene in intended disposition of specific equipment or materials. Revision 3 on document

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

AGR Equipment and Materials Disposition

Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWX Technologies, Inc. (BWXT) equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository. Representatives for the United States Department of Energy may intervene in intended disposition of specific equipment or materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Elevated-temperature cyclic properties of advanced manufactured materials

Microreactor developers at the 2019 GAIN Microreactors workshop expressed a need for advanced manufacturing to fabricate microreactor components. Powder metallurgy (PM) hot isostatic pressing (HIP) is more mature than other advanced manufacturing techniques for the following reasons: 1. a code case exists for Section III, Division 1, Subsection NB components of light water reactors to be manufactured using PM HIP Alloy 316L, and 2. tensile and creep properties have been shown to be equivalent to wrought material for Type 316 stainless steel and Grade 91. Thus, PM HIP is more readily deployable in the near term. Elevated-temperature cyclic properties of PM-HIP materials, however, have not been established. The scope of testing for Code qualification could be significantly reduced compared to a new material for the following reasons: 1. only a limited amount of time consuming creep testing would be required, and 2. testing of fatigue and creep-fatigue behavior could be reduced if it can be demonstrated that the properties are comparable to wrought material. The objective of this work package is to develop an understanding of the relationship of advanced materials processing on the material characteristics as they relate to elevated-temperature component design and construction. This will be achieved by the following: • Elevated-temperature fatigue and creep-fatigue testing of PM HIP and wrought Alloy 316L. • Fracture and microstructural characterization of the PM HIP and wrought Alloy 316L. • Analysis of the elevated-temperature fatigue and creep-fatigue properties of PM HIP Alloy 316L. • Draft a report evaluating the elevated-temperature fatigue and creep-fatigue properties of PM HIP Alloy 316L. • Pursue procurement of Alloy 316H and Grade 91 manufactured using PM HIP. • Initiate scoping studies of the procured Alloy 316H or Grade 91 manufactured using PM HIP contingent upon successful material procurement.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Nuclear Power Fault Diagnostics and Preventative Maintenance Optimization

The nuclear industry is beginning to see reactors shut down—even after their operating licenses have been extended—because they are not economically competitive with other energy sources. These early closures happen primarily due to economic reasons, despite excellent safety records. Therefore, it is imperative to reduce costs in order to prevent these early closures. One of the contributors to these economic reasons is the large operations and maintenance costs. This paper showcases recent research on advanced fault diagnostics techniques and preventative maintenance optimization (PMO) for reducing NPP maintenance costs. Specifically, it focuses on the feedwater and condensate system (FWCS) for both pressurized- and boiling-water reactor (BWR) systems. The computerized maintenance management system (CMMS), which contains the plant’s digital record of all corrective maintenance (CM) and preventative maintenance (PM) work orders, provided the ground truth for locating potential faults and labeling the process data as either healthy or faulted. Various feature extraction techniques were used to further differentiate the faulted data from the healthy data. Through a cross-validation procedure, support vectors machines were used to label other test sets of process data as either healthy or faulted. With relatively few faults identified in the BWR system, the potential for PMO opens up, since an unnecessary amount of PM leads to inflated maintenance costs. The steps for PMO are summarized, from component health determinations to recommendations for action. An example of PMO assessment is presented for condensate pumps, condensate booster pumps, and the respective motors that drive them.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Nuclear Power Fault Diagnostics and Preventative Maintenance Optimization NPIC presentation

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Robotic Understanding of Spatial Relationships Using Neural-Logic Learning

Understanding spatial relations of objects is critical in many robotic applications such as grasping, manipulation, and obstacle avoidance. Humans can simply reason object's spatial relations from a glimpse of a scene based on prior knowledge of spatial constraints. The proposed method enables a robot to comprehend spatial relationships among objects from RGB-D data. This paper proposed a neural-logic learning framework to learn and reason spatial relations from raw data by following logic rules on spatial constraints. The neural-logic network consists of three blocks: grounding block, spatial logic block, and inference block. The grounding block extracts high-level features from the raw sensory data. The spatial logic blocks can predicate fundamental spatial relations by training a neural network with spatial constraints. The inference block can infer complex spatial relations based on the predicated fundamental spatial relations. Simulations and robotic experiments evaluated the performance of the proposed method.

Wang, Dali↗

DOE OSTI2022

Plasma Processing for In-Situ Field Emission Mitigation of Superconducting Radiofrequency (SRF) Cryomodules

Field emission (FE) is one of the main limiting factors of superconducting radio-frequency (SRF) cavities operating in accelerators and it occurs whenever contaminants, like dust, metal flakes or even absorbates, are present on the surface of the cavity high electric field region. Field emission reduces the maximum achievable accelerating field and generates free electrons that may interact with the beam, damage or activate the beamline. One practical method that can be used to mitigate this problem is in-situ plasma cleaning, or plasma processing. The development of a processing that can be applied in-situ is extremely advantageous, since it enables the recovery of the cryomodule performance without the need of disassembling the whole cryomodule, which is an extremely expensive and time-consuming process. On the other hand, plasma processing only requires the cryomodule warm-up to room-temperature and the subsequent processing of the contaminated cavities. The entire process is reasonably quick and involves a limited number of personnel. For these reasons we would like to advocate for continuing to invest in the R&D of plasma processing to optimize its applicability in cryomodules and for extending the technique to other frequency ranges and cavities geometries.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

An Experimental Investigation of Students’ Learning Effects When Using a Simplified Nuclear Simulator

Securing enough data has been a main challenge in human reliability analysis (HRA). Many researchers and institutes have made a lot of efforts for collecting HRA data to produce reasonable human error probabilities (HEPs) as well as reduce the uncertainty of HRA quantification. Representatively, U.S. Nuclear Regulatory Commission (U.S. NRC), Korea Atomic Energy Research Institute (KAERI) and Idaho National Laboratory (INL) have led lots of empirical research regarding the HRA data collection. The U.S. NRC and KAERI have mainly carried out full-scope simulator research collecting HRA data through experiments using full-scope simulators with actual operators. In contrast, INL has experimentally collected the data using simplified simulators and student operators. INL has proposed the Simplified Human Error Experimental Program (SHEEP) framework to complement full-scope data collection efforts by suggesting a way to infer full-scope data based on experimental data collected from students operating simplified simulators, specifically the Rancor Microworld Simulator (Rancor) and the Compact Nuclear Simulator (CNS). The aim of the SHEEP framework is to lower the entry point for collecting useful HRA data by securing large sample sizes at a reasonable amount of cost and labor while also guaranteeing a high degree of freedom when designing experiments. The authors’ previous research investigated whether data collected from the SHEEP framework could support a representative full-scope study. Besides, human performance differences between professional and student operators when using Rancor and CNS have been analyzed to understand the lack of fidelity of the simplified simulators and student operators within the SHEEP study. As a follow up research, this study experimentally investigates students’ learning effects and the performance trends over a certain period when using Rancor. This study aims to find out 1) how much training or education is required to collect HRA data from non-experts (i.e., students) when using Rancor and 2) how much differences there are in human performance measures between students and professional operators. In this study, a longitudinal experiment is developed. The four experiment trials with two weeks interval are carried out for sixteen undergraduate students majoring nuclear engineering at Chosun University. Totally four scenarios randomly selected from ten Rancor scenarios are used in each experiment trial. Four human performance measurements (i.e., workload, situation awareness, time and error) are considered in the experiment. Lastly, the trend of students’ performance is compared with operator data having been collected from the previous experiment.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2023

Simulation-Based Recovery Action Analysis Using the EMRALD Dynamic Risk Assessment Tool

A recovery action is defined as the action that prevents deviant conditions from producing unwanted effects. It generally indicates a kind of countermeasure performed in response to a failure of human action. The recovery actions especially play an important role in complex systems like nuclear power plants (NPPs), which consist of highly sophisticated controllers to ensure that desired performance and safety must be achieved and maintained. This is because a combination of human error and its recovery failure may be able to cause a catastrophic effect on a system. Analyzing recovery actions has been a critical part of HRA, which is a technique to evaluate human errors and provide human error probabilities (HEPs) for application in probabilistic safety assessment (PSA). If recovery actions are not adequately analyzed and applied to PSA models, the PSA results may be under-estimated or be not able to reasonably account for the failure of human actions in the context of PSA. For this reason, some regulatory documents such as ASME/ANS RA-Sb-2013 by the American Society for Mechanical Engineers and the American Nuclear Society and NUREG-1792 by U.S. Nuclear Regulatory Commission have emphasized the importance of recovery analysis within the HRA. A couple of existing HRA methods, such as the Technique for Human Error-Rate Prediction (THERP), the Cause-Based Decision Tree (CBDT), and the Korean Standard HRA (K-HRA), have respectively suggested their own approaches to the HRA recovery analysis. However, there are a couple of limitations to treating recovery actions using only the current HRA methods available. The biggest limitation is that the existing recovery analysis does not explicitly consider a variety of recovery action types and recovery sequences as they occur in actual NPPs. To handle the limitations of existing recovery analysis, this study proposes a simulation-based recovery analysis method using the Event Modeling Risk Assessment Using Linked Diagram (EMRALD) software. The EMRALD software is a dynamic simulation tool for PSA. It supports realistic and dynamic modeling of human actions as they would be performed at NPPs. It is also favorable to simultaneously model the specific moment at which an action is performed, the time it takes to perform the action, and the failure probability of that action. In this paper, a detailed methodology for modeling recovery actions in the simulation platform is proposed with a couple of examples. Then, outputs from the simulation are discussed as reviewing if this novel approach can complement the challenges of existing recovery analyses.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2023

Uncertainty-aware photovoltaic generation estimation through fusion of physics with harmonics information using Bayesian neural networks

We develop an aggregate photovoltaic generation estimation methodology that uses diverse inputs and can reason on its current input-dependent predictive uncertainty. Named PVPHEst, for PhotoVoltaic Physics- & Harmonics-driven Estimator, the resulting tool is intelligently weighing and fusing information carried by the output of physics models, harmonics, and line sensors, using Bayesian neural networks and related techniques aimed at solving machine learning problems with intrinsic uncertainty quantification. As each of the three input classes carries heterogeneous information that only sheds light on one facet of the estimation problem but its value can diminish in the face of diverse grid phenomena, PV-PHEst with its estimation and uncertainty reasoning capabilities perform a nontrivial and potentially mission-critical task of value to grid operators.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2023

A Data Processing Pipeline for Adversarial Socio-Technical Network Analysis

With the rapid adoption of emerging technologies, there is a need to catalog and model sociotechnical interdependencies that have been historically used to influence the operation of Critical Infrastructure networks including the impacts of mergers and acquisitions, hostile takeovers, and foreign investment. Our research intends to address this need with two primary contributions. First, we have developed a data curation and processing pipeline to generate sociotechnical networks extracted from a variety of data sources including SEC filings and infrastructure asset databases. The pipeline, implemented in Apache Airflow, extracts and normalizes the representation of entities and relations, specified within ontologies. Our intent is to provide an extensible, machine-actionable approach to quickly communicate such models, reproduce previous results, and adapt them to new, unanticipated situations. Second, networks produced by our pipeline enable the development of graph-theoretic metrics that consider the properties of network components in addition to its topology. Metadata associated with network components---whether semantic, temporal, or geospatial---affects the alignment of generated networks with assumptions underlying complexity metrics. Validation of generated networks relative to component types defined by an ontology, may allow the research community to adapt metrics to the semantics of the domains being studied. Generated networks may be processed as knowledge, dynamic, or spatial graphs and enables a variety of analyses including automated reasoning and measures of network complexity. Automated reasoning views extracted entities and relations as a knowledge graph; this enables application of inference rules that represent historically-attested adversarial business methods and applies that behavior to a specific geographic context. Measures of network complexity, including degree distribution, reachability analyses, temporal analysis, and community detection can be adapted to indicate adversarial organizational influence.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Comparison of Socio-Technical Threat Models

Given the adoption of emerging technologies and the increasing complexity of managing such systems with a lifecycle much shorter than that of critical infrastructure systems, there is a practical need to be able to analyze sociotechnical dependencies and their associated evolving risks. Threat models based on social influence techniques can be used to implement adversarial tactics analogous to the cyber kill chain and attested to within the MITRE ATT&CK for ICS framework including Initial Access, Persistence, Collection, and Impact. Furthermore, as with cyber disruptions, the impact of social influence threat models can have an asymmetric impact that is not spatially-localized. Finally, unlike cyber attacks with a reasonably short duration (ransomware takes days to months), social influence based attacks have the potential to persist for much longer as they are based on long-term strategic infrastructure investments within the private sector. Given the increased importance of electric vehicle charging stations as a long-term, strategic infrastructure investment within the Energy and Transportation Sectors, we provide initial results that compare the impact of a Loss of Availability (T0826) realized through cyber and social influence based threat models. The analysis employs techniques from automated reasoning and measures of network complexity to understand evolving dominance of EV payment and charging networks within geographic region of interest. Within this context, we compare the impact of a loss of availability due to ransomware versus that of loss of support due to a merger and acquisition. Results across several different metro areas will be provided.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2023

Can We Lower the Voltage of a Silicon Module and Still Make It Just as Efficient

For thousands of years, scientists have studied the universe for many different reasons, but one reason, in particular, comes down even to the smallest materials in the universe. The universe is made of thousands of different particles. These cosmic particles are hugely important to figuring out how the universe works. But to figure out how these cosmic particles work and their purposes, there has to be something that can collect data and information on them.

Mason, Eddie↗

DOE OSTI2011

RAMA Surveillance Capsule and Component Activation Analyses

This paper presents the calculated-to-measured ratios associated with the application of the RAMA Fluence Methodology software to light water reactor surveillance capsule and reactor component activation evaluations. Comparisons to measurements are performed for pressurized water reactor and boiling water reactor surveillance capsule activity specimens from seventeen operating light water reactors. Comparisons to measurements are also performed for samples removed from the core shroud, top guide, and jet pump brace pads from two reactors. In conclusion: The flexible geometry modeling capabilities provided by RAMA, combined with the detailed representation of operating reactor history and anisotropic scattering detail, produces accurate predictions of the fast neutron fluence and neutron activation for BWR and PWR surveillance capsule geometries. This allows best estimate RPV fluence to be determined without the need for multiplicative bias corrections. The three-dimensional modeling capability in RAMA provides an accurate estimate of the fast neutron fluence for regions far removed from the core mid-plane elevation. The comparisons to activation measurements for various core components indicate that the RAMA predictions are reasonable, and notably conservative (i.e., C/M ratios are consistently greater than unity). It should be noted that in the current evaluations, the top and bottom fuel regions are represented by six inch height nodes. As a result, the leakage-induced decrease in power near the upper and lower edges of the core are not well represented in the current models. More precise predictions of fluence for components that lie above and below the core boundaries could be obtained if the upper and lower fuel nodes were subdivided into multiple axial regions with assigned powers that reflect the neutron leakage at the top and bottom of the core. This use of additional axial sub-meshing at the top and bottom of the core is analogous to the use of pin-wise meshing in peripheral bundles to accurately represent radial leakage effects. The representation of thermal neutron fluence and activations are found to be reasonably accurate and consistently conservative, as demonstrated by comparison to the reactor component thermal neutron reaction activation measurements. Further improvement in the comparisons to measurements could be achieved by exploring the impact of enhanced sub-meshing of the geometry near the components of interest. The mesh densities utilized in the current evaluation are consistent with the mesh requirements for high energy neutron transport. The substantially shorter transport lengths for thermal neutrons relative to high energy neutrons suggests that localized regions of finer meshing are needed in the vicinity of those reactor components requiring thermal neutron fluence evaluations. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

German experiences in local fatigue monitoring

The ageing management of nuclear power plants (NPP) has gained an increasing importance in the last years. The reasons are mainly due to the international context of extending period of plants operation. Moreover, new scientific discoveries, such as the corrosive influence of the medium on the fatigue process (environmentally assisted fatigue - EAF) play an important role and influence the code development (ASME, EAF code cases). The fatigue damage process takes a central position in ageing mechanisms of components. It must be ensured through appropriate evidence that facilities are being operated under allowable boundary conditions. In the design phase of NPP, fatigue analyses are still based on theoretical considerations and empirical values, which are summarized in the design transient catalogue, necessary for licensing. These analyses aim at proving the admissibility of the loads in terms of stress and fatigue usage. These analyses will also provide the fatigue-relevant positions in the NPP and give a basis for future design improvements and optimization of operating modes. The design transients are in practice conservatively correlated with the real transients occurring during operation. Uncertainties reveal very conservative assumptions regarding forecast temperatures, temperature gradients and frequencies of events. During operation of the plant, it has to be recurrently proved, that the plant is being operated under designed boundary conditions. Moreover, operating signals are constantly acquired to enable a fatigue evaluation. For example, in Germany fatigue evaluation is based on decades of experience and regulatory requirements. The rule KTA 3201.4 [1] establishes the rules for qualified fatigue monitoring. The rule DIN 25475-3 [2] on fatigue monitoring systems is available in draft version. Experience shows that some significant differences occur between the design transients and the real occurred transients during plant operation. The reasons for it are the various manual control options and also different operating modes. It is clear that showing the covering of real loads by design loads, requires a relatively complex and well-qualified detection process. The difficulty of this task is increased due to the lack of data or incomplete information and the exclusive reliance on existing operation plant data. The strategy of employing local fatigue monitoring is a straightforward solution enabling the direct measurement of loads on the fatigue-sensitive zones. Nowadays a direct derivation of the complete stress tensor at the fatigue-relevant locations is enabled thanks to the recorded local loads and combination with finite element (FE) analyses. So, additionally to the recorded temperature curves, a representation of the time evolution of the six stress components for each monitored component is possible. This allows the application of the simplified elasto-plastic fatigue check according to design codes. The fatigue level can be realistically analyzed with a suitable cycle-counting method. Furthermore, the knowledge of the time evolution of the stresses and strains enables to take into account an environmental factor to include the corrosive fluid influence in the calculations. Without local recording, it is impossible to calculate realistic fatigue usage. AREVA offers the AREVA fatigue concept (AFC) and the new fatigue monitoring system integrated (FAMOSi), necessary tools to monitor local fatigue and to provide realistic assessment. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Opportunities for Cost Effective Disposal of Radioactively Contaminated Solid Waste on the Oak Ridge Reservation, Oak Ridge, TN - 13045

Department of Energy (DOE) requirements for the release of non-real property, including solid waste, containing low levels of residual radioactive materials are specified in DOE Order 458.1 and associated guidance. Authorized limits have been approved under the requirements of DOE Order 5400.5, predecessor to DOE Order 458.1, to permit disposal of solid waste containing low levels of residual radioactive materials at solid waste landfills located within the DOE Oak Ridge Reservation (ORR). Specifically, volumetric concentration limits for disposal of solid waste at Industrial Landfill V and at Construction/Demolition Landfill VII were established in 2003 and 2007, respectively, based on the requirements in effect at that time, which included: an evaluation to ensure that radiation doses to the public would not exceed 25 mrem/year and would be as low as reasonably achievable (ALARA), with a goal of a few mrem/year or less (in fact, these authorized limits actually were derived to meet a dose constraint of 1 mrem/year); an evaluation of compliance with groundwater protection requirements; and reasonable assurance that the proposed disposal is not likely to result in a future requirement for remediation of the landfill. Prior to approval as DOE authorized limits, these volumetric concentration limits were coordinated with the Tennessee Department of Environment and Conservation (TDEC) and documented in a Memorandum of Understanding (MOU) between the TDEC Division of Radiological Health and the TDEC Division of Solid Waste Management. These limits apply to the disposal of soil and debris waste generated from construction, maintenance, environmental restoration, and decontamination and decommissioning (D and D) activities on the DOE Oak Ridge Reservation. The approved site-specific authorized limits were incorporated in the URS/CH2M Oak Ridge LLC (UCOR) waste profile system that authorizes disposal of special wastes at either of the RCRA Subtitle D landfills. However, a recent DOE assessment found that implementation of the site-specific authorized limits for volumetrically contaminated waste was potentially limited due in part to confusion regarding the applicability of volumetric concentration limits and/or surface activity limits to specific wastes. This paper describes recent efforts to update the authorized limits for Industrial Landfill V and Construction/Demolition Landfill VII and to improve the procedures for implementation of these criteria. The approved authorized limits have been evaluated and confirmed to meet the current requirements of DOE Order 458.1, which superseded DOE Order 5400.5 in February 2011. In addition, volumetric concentration limits have been developed for additional radionuclides, and site-specific authorized limits for wastes with surface contamination have been developed. Implementing procedures have been revised to clarify the applicability of volumetric concentration limits and surface activity limits, and to allow the use of non-destructive waste characterization methods. These changes have been designed to promote improved utilization of available disposal capacity of the onsite disposal facilities within the DOE Oak Ridge Reservation. In addition, these changes serve to bring the waste acceptance requirements at these DOE onsite landfills into greater consistency with the requirements for commercial/ public landfills under the TDEC Bulk Survey for Release (BSFR) program, including two public RCRA Subtitle D landfills in close proximity to the DOE Oak Ridge Reservation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗