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At least 1,621 records · Page 90
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Computational study of critical mass curves for various ratios of {sup 238}Pu to {sup 239}Pu

A MCNP study for a water-reflected sphere of plutonium metal water pseudo-solution was performed for a various ratios of {sup 238}Pu to {sup 239}Pu to develop minimum critical mass curves. The {sup 238}Pu atom percent (at. %) was varied from 35 to 100 percent. It was demonstrated that at {sup 238}Pu ratios of ≥ 65 at. % the high absorption cross section of {sup 238}Pu in the thermal energy region dominates the system reactivity for water moderated systems resulting in a minimum critical mass in the metal rather than solution regime. However, due to the lack of available benchmark experiments to validated use of the {sup 238}Pu cross-section data for quantities above gram-level, the models fell outside the area of applicability (AOA) of the Los Alamos National Laboratory's (LANL) Nuclear Criticality Safety Division's (NCSD) MCNP validation report. An additional of margin of subcriticality (MOS) is needed for {sup 238}Pu systems. This additional MOS will be dependent on the amount and significance of the {sup 238}Pu, which varies in the MCNP study. Whisper is a computational software package designed to assist with validation studies for MCNP. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application model. Therefore, Whisper 1.0.0 was used on the curve minima in the solution regime for the 65 at. % {sup 238}Pu curve to estimate an appropriate additional margin of subcriticality and create a subcritical curve based on the suggested upper subcritical limits (USL)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

SISTec: Mathematical Calibration of Large Clearance Monitors - 18265

One major path for clearance of large amount of waste arising from the decommission of nuclear power plants is based on measurement with large clearance monitors (also known as free release measurement facilities). These devices utilize multi-detector plastic scintillator arrays (large clearance monitors can consist of up to 24 plastic scintillators) to offer both, extremely low minimum detectable activities (MDA) and short measurement times for waste packages weighing up to one metric ton. However, a major problem with these devices is to obtain nuclide specific efficiencies, especially for geometrically complex large waste packages. Normal procedure for estimating the efficiency is to device dummy waste packages and attach point sources with well-known activities to unfavourable positions in these packages. In accordance with this procedure, calibration factors can be derived in a radiological conservative way. Devising such test dummies is not always possible and the procedure is prone to generating large measurement uncertainties. In this publication, we propose a new method to derive calibration factors by means of the Los Alamos National Laboratory (LANL) simulation tool Monte Carlo N-Particle (MCNP) and a statistical-phenomenological model for the detector characterization. In a multi-step approach, first the detector response for a distribution of activity in a waste package is calculated and subsequently the detectors are mathematically characterized. Applying state of the art statistical and optimizing procedures, parameters for the detector response are derived in a self-consistent way. This also allows to rigorously propagate the uncertainties to the final calculated calibration factors according to the Guide to the Expression of Uncertainty in Measurement (GUM supplement 1). To validate the derived parameters for the detector characterization a minimally invasive experimental setup was devised and installed into the large clearance monitor. It consisted of an aluminum swivel arm which could position point sources to the relevant coordinates. The so collected measurement data not only allowed to investigate the precision of the calculated calibration factors, but also to probe the location of concentrated activity sources in the waste package. Furthermore, data on measurements with dummy waste packages were collected to test the reproducibility of well-known calibration factors. We present the software tool SIStec, which will provide an intuitive user-interface to generate and simulate waste package models and execute the above-mentioned method. Using SIStec, different scenarios for the distribution of activity in the waste package can be probed. This will also allow to optimize the clearance process on various levels. For example, measurement time for pre-measurements can be reduced in exchange for unfavourable calibration factors, which would still allow to meet the regulator limits for clearance. (authors)

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

Overview of IAEA MODARIA II Program Working Group 1 on Assessment and Decision Making of Existing Exposure Situations for NORM and Nuclear Legacy Sites - 18308

The International Atomic Energy Agency (IAEA) launched the Modelling and Data for Radiological Impact Assessments (MODARIA) II Programme in October. As part of the IAEA effort to enhance the capabilities of its Member States in assessing radiological impacts of radionuclides released to the environment from sites that were not operated in accordance with current standards, and to ensure protection of the public and the environment, the MODARIA II Programme aims to: (1) test the performance of models developed for predicting the transfer of radionuclides in the environment and assessing radiological impacts to the public and the environment; (2) develop and improve models for particular environments and, where appropriate, to agree on data sets that are generally applicable for use in testing environmental transfer models; and (3) provide an international forum for the exchange of experiences, ideas and research information. The Programme will run for 3 years from 2016 to 2019. The Programme comprises seven working groups covering a wide range of topics (https://wwwns.iaea.org/projects/modaria/modaria2.asp?s=8 and l=129). MODARIA II Working Group 1 (WG1) is focused on dose assessment and decision making regarding existing exposure situations due to naturally occurring radioactive material (NORM) and nuclear legacy sites. WG1 was formed as a combination of two MODARIA I WGs: WG1 focused on application of decision-aiding tools in remediation; and WG3 focused on risk assessment for NORM and radioactively contaminated legacy sites (https://wwwns.iaea.org/projects/modaria/default.asp?s=8 and l=129). It was recognized that the decision-making needs will inform the scope of risk assessments, and the risk insights and structured decision analyses will aid in informed decision-making. Therefore, MODARIA II WG1 is focused on developing methods and toolsets for risk-informed decision-making for NORM and nuclear legacy sites, including demonstration of decision-making processes and tools through case studies. MODARIA II WG1 has 91 registered participants from 33 Member States and IAEA (as of November 2017), some 30 of whom are actively engaged in various collaborative research and development activities within WG1. Members of WG1 participated in the First Technical Meeting of MODARIA II, which was held in Vienna, Austria from 31 October to 4 November 2016. The group then held its First Interim Meeting in Brussels, Belgium from 26 to 30 June 2017, which included a field trip to the legacy Tessenderlo Chemie (TCH) Phosphate Processing Site. At this meeting, the TCH Site and the Pridneprovsky Chemical Plant (PChP) in Ukraine were selected for case studies. The Los Alamos National Laboratory (LANL) Material Disposal Area B in the USA and the Beaverlodge Uranium Mill and Mill Site in northern Canada were selected as examples for sharing of good practices and lessons learned in risk-informed decision-making for environmental remediation and land conveyance. At the Second Technical Meeting held in Vienna on 30 October to 3 November 2017, an additional NORM site for tin processing in Brazil was added to the case studies, with another NORM site in Yunnan, China included for consideration at the Second Interim Meeting of WG1. The Second Interim Meeting will be held in in early to mid-May, 2018. The group has also decided to prepare an Interim Report by October 2018 to document the WG1 discussions and analyses for the period of 2016-2018, which will be finalized at the conclusion of MODARIA II in October 2019. Initial assignments have been discussed for IM1 participants to collect data, conduct modelling and analyses, and prepare draft sections for the Interim Report. This paper will provide an overview of the work plans for WG1. It will also highlight some near-term activities that the group is currently working on. (authors)

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

Hydrologic Model Analyses for Decision Support at the Los Alamos National Laboratory Site - 18363

Modeling hydrologic flow and transport in the subsurface presents a number of challenges including large datasets that contain limited information about the properties of the subsurface and models with a large number of parameters characterizing the properties of the subsurface and the ambient hydrologic conditions (e.g., groundwater recharge). This, in turn, leads to substantial uncertainty in predictions about quantities of interest (e.g., the concentration of a contaminant at points of compliance). We present a number of methods and tools that have recently been developed to address these problems. The model calibration tools are capable of assimilating large datasets (for example, on the order of 1,000,000 model calibration targets) into numerical models, calibrating models with a large number of model parameters (on the order of 1,000,000), or both. These tools leverage recently developed methods in numerical linear algebra such as matrix-free methods and randomized matrix sketching. The uncertainty quantification tools are highly-parallel and enable us to robustly quantify the uncertainties in model parameters (and therefore uncertainty in quantities of interest). Having these capabilities in model analysis tools enables more informed decision making related to remediation of groundwater contaminants. The model analyses that we present are related to the Los Alamos National Laboratory (LANL) chromium site. The analyses are performed by MADS (Model Analysis and Decision Support; http://mads.lanl.gov). (authors)

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

Transient Unsaturated Zone Modeling of Radioactive Waste Pits at Los Alamos National Laboratory - 87544 - 18432

Waste disposals at Material Disposal Area G (MDA G) at Los Alamos National Laboratory (LANL) occur primarily in a series of long, rectangular pits into which layers of waste have been emplaced beginning in the late 1950's. MDA G is situated in north-central New Mexico on top of Mesita del Buey, a narrow 3 km long mesa in the Parajito Plateau. The mesa is composed primarily of volcanic tuff from two principle volcanic eruptions in the Jemez Mountains which deposited the Otowi and Tshirege Members of the Bandelier Tuff. The Cerros del Rio Basalt underlay the tuff units and extends down for approximately 200 meters. The water table lies within the basalts at around 300 m below the surface of the mesa. The local climate is temperate and semiarid, averaging around 34 cm of precipitation per year. Ambient subsurface moisture conditions are dry, with water content typically below 10%. During disposal operations, open pits are subject to enhanced infiltration compared to the final covered configuration due to the unvegetated disturbed surface and precipitation events. Additionally, water added to the pits during disposal operations for tasks like waste compaction and dust suppression introduces additional moisture to the system during disposal operations that could potentially provide a pathway for contaminants in the waste zone to migrate through the unsaturated zone. Modeling was conducted using the HYDRUS-2D/3D software to evaluate the spatial extent of saturation changes and fluxes across geologic layer interfaces in the unsaturated zone throughout the life cycle of a waste pit. A general 2-dimensional framework was developed that included a cross section of a waste pit and the underlying geologic units. Several geological layerings were developed to represent the variation across the mesa at Area G. The model was run stochastically for hundreds of realizations to understand how variations in hydraulic parameters (e.g., porosity, saturated hydraulic conductivity, Van Genuchten parameters, etc.) and operational parameters (e.g., pit open period duration, pit depth, initial saturation of the waste zone, etc.) affect the extent of saturation changes and fluxes in the deep unsaturated zone. Distributions for these parameters were developed based on available site data. Results of the unsaturated zone modeling indicate that a significant amount of water can reach the lower unsaturated zone under some circumstances, particularly when the amount of water introduced during pit operations is high. A statistical study of modeling results was performed to develop a regression that describes the variations of water flux in time at the geological layer interfaces. These statistical model results will be used to evaluate solute transport in a future effort. (authors)

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

Investigation and Recovery Following the Drum Event at INL's ARP-V Facility - 19517

Contractors working under the Department of Energy's (DoE's) Idaho Field Office have been exhuming and processing buried legacy waste from early weapons development and productions operations since 1984. Since operations began, nearly 400,000 drums have passed through this process. On April 11, 2018, repackaged drums were staged in an airlock (radiological buffer area) awaiting transfer to another facility for characterization. Late in the evening after operations had ended, four of the drums over pressurized causing an energetic displacement of their lids. A significant fraction of drum contents was ejected and dispersed throughout the airlock. The event activated the fire detection system in the facility initiating an emergency response from the site fire department, radiological protection, and operations. Initial intervention by fire department personnel was unsuccessful at quenching the reaction, so the responders egressed to a safe distance and monitored the situation. After determining the event had terminated and it was safe to enter, the facility was assessed revealing significant contamination in the airlock from the ejected waste. The area surrounding the facility was surveyed and it was determined that no contamination was released outside the facility. The Fluor Idaho, LLC (Fluor Idaho) Program Manager commissioned an investigation and recovery effort, which focused on three aspects: (1) technical investigation, (2) event and causal analysis, and (3) facility recovery. A team of highly qualified individuals from around the DOE Complex was assembled to support the technical investigation. The team included representatives from several Fluor Idaho, Fluor Corporate, Battelle Energy Alliance, LLC (INL), LANL, SRNL, South West Research Institute, DOE Carlsbad Field Office, and Defense Nuclear Facility Safety Board, and DOE-Headquarters. This team addressed the reaction mechanism of the event by guiding the sampling activities and coordinating analysis and testing by the analytical laboratories. Independent and experienced investigators were brought in to conduct a rigorous event and cause analysis. The Causal Analysis Team pursued direct, root, and contributing causes. Interviews were conducted, and documents were reviewed to develop an event and causal analysis. The Accelerated Retrieval Project V Operations team mobilized to decontaminate and recover the facility. Significant progress has been made in this effort, area and airborne contamination levels have been significantly reduced allowing limited activities to resume. In addition, this group supported sample collection efforts as directed by the Technical Investigation Team. This paper reviews key findings and milestones from the technical investigation, the event and causal analysis, and facility recovery operations. From these elements of the investigation, lessons learned are identified that would benefit other organizations involved with legacy waste management and disposition. Waste management is common to all sites in the DOE Complex and the events and recovery efforts described in this paper may provide beneficial insights to these operations. (authors)

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

Offsite Source Recovery Program; High Activity Beta Gamma Removal, Transport and Disposition - 19663

The Department of Energy's (DOE) National Nuclear Security Administration's (NNSA) Office of Radiological Security's (ORS) Off-Site Source Recovery Program (OSRP) has been tasked with the removal and disposition of radioactive sealed sources in the interest of national security and public health and safety. It is the mission of the program to recover radioactive sources to prevent them from falling into the hands of individuals who would use this material for malicious intent. OSRP is a collaborative effort involving Los Alamos National Laboratory (LANL), Idaho National Laboratory (INL), and Lawrence Livermore National Laboratory (LLNL). At the inception of the program, Pu-239 was the only isotope acceptable for recovery. As a wider security scope developed, OSRP expanded its mission to include the recovery of other sources that posed a potential risk to national security, health, and safety. Starting in 2004, OSRP began recovering cesium and cobalt devices, largely from medical, research, and industrial facilities. These disused higher-activity devices present a burden to licensees in terms of security, and demand for OSRP's assistance in removing them has been high. Today, OSRP removes around 40 high-activity devices per year in support of domestic demand as well as ORS's Cesium Irradiator Replacement Project (CIRP). This represents an average of over 50,000 Ci of Cs-137 and Co-60 removed from the private sector each year for permanent disposition. A major component of this effort is the use of the proper shipping container to transport the devices. As a way to increase transportation capabilities ORS has developed two new Type B containers. The most recently completed of these containers, the 435-B, is currently certified to transport Gammators and Gammacell-1000, -3000, and -40 irradiators in an overpack configuration. This provides greater capacity and flexibility to safely and effectively transport these devices, for which acceptable shipping configurations had been previously limited. (authors)

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

Distribution Development for the RDX Regional Model at Los Alamos National Laboratory - 20385

Representing uncertainty in model inputs often means finding a balance between uncertainty and physical reality. Developing wide distributions may seem conservative in principle, but this approach may lead to unrealistic model results. Characterizing the current state of knowledge of stochastic inputs presents many challenges, especially if the data available are limited or have limited relevance to the site. Relationships among these inputs may also be important to represent but are typically complex or difficult to define. Often special adjustments must be made to account for reduced credibility in particular data. If parameters are strongly related to one another, a correlation structure may be developed for input to the model. Other techniques such as regression models may be used to incorporate relationships between the information available and the desired parameters. The process of developing distributions must consider details of the model in terms of what the distribution is meant to represent. This paper uses the example of a probabilistic fate and transport model for hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX) in the regional aquifer at Los Alamos National Laboratory (LANL). For many parameters, a single draw is applied to all space and time over which the model is run, for a single iteration. This simplification is often made for many reasons, and can often be beneficial, but also adds additional complexity in the distribution development process. Defining the distributional goals as they relate to the modeling process is an important step, which should take place prior to evaluation of the data. Usually, the distributions developed are meant to characterize the average value of the parameter over the spatial and temporal domain of the model. Distribution development requires consideration of many sources of information on the parameter where available, ideally from multiple references. Examples of different sources include data from different references but also from different conditions, measurement methods, or experimental types. Depending on these conditions and the reliability or relevance of particular references, different sources of data may each contribute valuable information but have varying relevance to the site. In these cases, weighting data unequally is a useful way to incorporate this information. As an example, aqueous dispersivity data are available for a variety of rock types. Only a few values are available for the desired rock type, and this is not enough to develop a distribution. Therefore, dispersivity values from other rock types are included in distribution development but are down-weighted such that the best data have the most influence on the distribution developed. In another case, K{sub d} distributions in the model are meant to represent a known composition of multiple soil types. Data from these materials are weighted accordingly to develop a distribution for the weighted average K{sub d} across soil types. Other cases include varying reliability of different sources, and weighting data according to the confidence in these sources. In some cases, input parameters are correlated with one another. An example is advective porosity, which is positively correlated with total porosity and must be less than total porosity. Paired data with both parameters must exist to discern the relationship between the two parameters and if it is necessary to build a correlation structure into the model. In general, correlated parameters may be represented in the model by a multivariate distribution, or perhaps more desirably, capturing the correlation within the developed distributions which may be treated as independent from one another. In the example of porosity, this can be done by transforming advective porosity into a proportion of total porosity which may be drawn independently from the distribution of total porosity. This paper explores how complex data and correlations can be incorporated to meet the distributional goals of the model, using the RDX regional model as a detailed example. (authors)

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

Bayesian Approach to Estimation of Water Table Elevations Using Historical Rasters as Prior Information 2019 - 20430

In cases of complex but only partially known geology and a lack of spatial control in observation well locations, water table elevation estimation is very challenging. In some cases, auxiliary information, such as observations of the movement of tracers, operation of injection and extraction wells, and calibration of groundwater models against the historical elevation data, can be combined with expert judgement to estimate flow directions in areas of sparse data and to aid in the production of more reliable contour maps (and associated rasters) than could be produced by relying on sparse well elevation data alone. Given a historical sequence of these raster maps, the question arises how to automate, to the extent possible, the process of producing new raster maps to reflect data from previous times, the current data and the operation of expert judgement. One solution is to adopt a Bayesian point of view and to regard the historical well elevation data, auxiliary information and historical raster maps as prior information. The well elevations for water table wells, as well as those for injection/extraction wells and the data associated with other relevant variables, can be viewed as predictors for the raster surface. From this prior information, we can, conditional on the values of these predictors for a new time period, compute an expected value map and a standard deviation map for the new raster. These then can be taken to specify a prior predictive distribution for the pixels in the new raster map. Then we condition the pixels, corresponding to water level observation wells within the raster, on the observed values in those wells (which in general will differ from the regression estimate) for the new time period. Given the smoothness of the water table surface, we then smooth the surface of deviations from the mean surface, based on the variograms of the historical rasters, and add this smoothed surface to the regression mean surface. The error structure of the produced raster map is defined by the regression error structure and the error due to smoothing based on the estimated variograms. This methodology has been developed and is being further refined for groundwater monitoring and remediation at LANL. It is a very flexible method that can also be applied with a variety of other predictors applied to model the water level wells in the area of interest over the historical record. The smoothness of the spatial process and its possible evolution over time can then be estimated from the residuals from this regression. This can be augmented by expert hydrogeological opinion based on site topography and hydrogeology. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Verification of Special Nuclear Material (SNM) using Active Well Coincidence Counter (AWCC) and a High Purity Germanium (HPGe) Detector

SNM verification can be accomplished via two methods i.e. destructive analysis (DA) or nondestructive analysis (NDA). The DA modes are much more accurate compared to NDA modes; nevertheless, the sample's physical integrity is destroyed in the process. Techniques such as ICP-MS, ID-ICP-MS, TI-MS are few examples of DA, where a bulk material is ground and mixed with a solution (some form of acid) to acquire a homogenous sample. In contrast, NDA modes utilize signals emitted by the sample such as gamma, neutrons, light, or heat to analyze samples under investigation. HPGe detectors are some of the most frequently used gamma detectors to assay SNM. However, in a high activity scenario it provides a large uncertainty and sometimes becomes unfeasible to measure low energy gamma rays. Meanwhile, neutron detectors are not affected by the gamma background, therefore they can be used in addition to the gamma detectors to verify total fissile mass ({sup 235}U + {sup 239}Pu) of SNM. The AWCC is one of the thermal neutron NDA systems developed at the Los Alamos National Laboratory (LANL). It consists of 42 {sup 3}He detectors to measure gross neutrons (singles), coincidence neutrons (doubles), higher order multiplicity counting (triples, quadruples, etc.). It can effectively operate in both active and passive neutron interrogation modes. It can utilize AmLi, AmBe, or {sup 252}Cf active interrogation neutron sources. It is used to verify declared fissile mass ({sup 235}U or {sup 239}Pu) present in SNM. Objectives: Calibrate HPGe and AWCC for lightly irradiated (with enough {sup 137}Cs gamma background) ∼93% enriched HEU fuels; Verify fissile mass ({sup 235}U +{sup 239}Pu) content using AWCC and HPGe; Model and benchmark AWCC in MCNP6. Neutron coincidence and multiplicity counting: Two or more neutrons are coincident if they are detected by the system within the specified gate window; In AWCC: Gate window is 128 μsec; Induced fission in fissile material such as {sup 235}U and {sup 239}Pu - emission of time correlated neutrons - Singles, Doubles, Triples, or Quadruples; No induced fission in non-fissile materials- no correlated neutrons - only singles. SNM can be verified with high accuracy and precision using gamma and neutron NDA instruments such as HPGe and AWCC. Interaction with thermal neutrons. Non-fissile atoms - do not emit correlated neutrons. Fissile atoms - emit correlated neutrons during induced fission events. The time correlated neutrons can be detected by using coincidence or multiplicity counters such as AWCC. Likewise, {sup 235}U produces 186 keV gamma peak that can be measured by a HPGe gamma detector. Counts under 186 keV gamma peak provide information about {sup 235}U content. Combining AWCC and HPGe results can provide better understanding of special nuclear material under investigation.

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

Refreshing Waste Tracking Software for the 2020's - 20059

The Waste Compliance and Tracking System (WCATS) is a software application designed to manage waste operations from cradle to grave. The system provides the support needed for characterization, generation, processing, and shipment of all waste generated or stored at Los Alamos National Laboratory (LANL). Waste is tracked with task records for containers in real-time via both desktop and mobile device applications. The software allows application administrators to configure sites with appropriate limits for Pu-239 equivalent Curies (PE-Ci), fissile gram equivalents (FGEs), volume limits, as well as custom limits for site-specific technical safety requirements (TSRs). WCATS was developed based on a need at Los Alamos National Laboratory for a modern, customizable waste tracking system that could support many different waste types. The application is built using Oracle{sup R} database and Oracle{sup R} ADF Swing application framework. The technologies and software frameworks selected for the system were finalized in 2006. Although Oracle{sup R} Database has proven to be robust, the user interface and application framework interface with the database have been deprecated and need to be replaced. The technologies selected for modernization follow the open-source model, which allows the WCATS team to more easily recruit developers, improve the speed of development, and ease the deployment and adoption of the system. The tools that support the re-engineering effort are Vue.js{sup R}, a JavaScript{sup R} framework, and Spring Boot{sup R}, an API (application programming interface) development tool written in Java{sup R}. An API allows for simplified communication from the database to the user interface of the application. In addition, legacy Crystal Reports{sup R} will be replaced with reports developed with the open source Pentaho{sup TM}f Report Designer community edition. Because the application uses HTML and JavaScript{sup R}, it builds upon tools that have been used for many years in web development. These tools have large community support and their proven track record will ensure that enhancements to WCATS can continue for years to come. (authors)

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

Savannah River Site H-Canyon Advancing Technologies for Remote Inspections - 20345

In 2017, the DOE Environmental Management Office of Technology Development (DOE-EM TD) sponsored the H-Canyon Advanced Technology Demonstration (ATD) to demonstrate to DOE facilities the value of using new commercial-off-the-shelf (COTS) and near-ready technologies to solve difficult problems and enhance worker safety. The DOE Savannah River Site (SRS) H-Canyon Air Exhaust Tunnel (HCAEX) inspection task was identified as representative of the hazardous, human denied environments which could benefit from advanced technologies. The HCAEX underground concrete tunnel is visually inspected biannually using a camera mounted on a remotely operated vehicle (ROV) designed and built by SRNL. While tunnel images have provided valuable visual information, it is desirable to have a higher order of understanding of the environment to support a more thorough structural integrity (SI) analysis and for long term planning purposes. As part of the ATD, the Concrete Integrated Product Team (CIPT) was formed to identify and evaluate available sensors and methods mature enough to remotely obtain tunnel concrete characterization data of high value and with a high probability of success. The team included SMEs and H-Canyon stakeholders in the field of concrete, nondestructive examination (NDE), structural integrity, sensors and remote systems from SRNL, SRNS, LANL, DOE-SR and the Army Corps of Engineering. The CIPT completed an in-depth identification of customer concrete inspection needs and potential technology solutions. Sensors and methods were evaluated on performance, data usefulness, cost and the feasibility of a successful deployment given the unique tunnel access challenges and environment. Two technologies were identified as promising by the CIPT for near term demonstration and evaluation: Lidar (Light Detection and Ranging) 3-dimensional (3D) mapping and remote robotic deployment of NDE instrumentation. Laser spectroscopy to characterize tunnel surface chemical changes was also of interest, but presently cost prohibitive. This paper will include a discussion of the two efforts underway to evaluate and implement the CIPT recommendations. First, the status of the November 2019 deployment of Lidar at a single location into the tunnel is presented. This initial deployment provided the team a learning curve and lessons learned on the challenges of tunnel deployment to include remote operation and data collection, stabilization of the sensor in high air flow (∼30 mph), ability to achieve a tolerance accuracy of 0.25-inches, and the probability to identify change in tunnel dimensions over time. Secondly, a discussion on the development of the Robotic Arm Concrete Inspection Test Bed capable of deploying NDE instruments to examine custom concrete forms will be presented. Concrete forms simulating the rough concrete surfaces, strength, composition and potential structural defects that can be found at our DOE EM facilities have been designed and built for the test bed. Two state-of-the art concrete NDE instruments have been identified as having potential to work on rough concrete walls, they are being tested and characterized as to their ability to provide desired structural integrity data to include wall thickness and defect identification on the developed test beams. Lastly, lessons learned, and the path forward will be presented. (authors)

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

Robust Method to Identify Groundwater Affected By Redox Conditions at Los Alamos National Laboratory Legacy Cleanup Site - 20497

To facilitate the mission of the Department of Energy's (DoE's) Environmental Management Program at Los Alamos National Laboratory (LANL), a significant number of groundwater monitoring wells with depths ranging from 152-396 meter were completed for characterization and monitoring. Groundwater at Los Alamos tends to be oxygen saturated with very low concentrations of organic matter. Drilling fluids can introduce residual carbon causing an increase in microbial activity and local reducing conditions around the well, potentially impacting representativeness of groundwater data quality for redox sensitive constituents. We assert that the redox state of the wells can robustly be identified by aqueous solution concentrations of iron and manganese. An automated review process has been built into a computer-based data management system enabling an efficient screening process for identifying reducing conditions that are not caught using standard data-validation protocol, and ensures that important data quality issues are thoroughly identified and addressed in an efficient and timely manner. (authors)

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

Analysis of Hydrology and Interim Measure Performance in the Chromium Plume at Los Alamos National Laboratory - 20498

Hexavalent chromium, Cr(VI), is present in the regional aquifer beneath Los Alamos National Laboratory (LANL) at concentrations greater than the New Mexico groundwater standard (50 ppb). The source of the Cr is blowdown from a power plant where potassium dichromate was used as a corrosion inhibitor from 1956 to 1972. Blowdown comingled with effluent from another source and migrated approximately 3 miles down a canyon, primarily as surface flow, until reaching an alluvial infiltration zone. The infiltrating water moved vertically through a thick (approximately 900 ft), complex stratigraphic sequence of unsaturated tuffs, basalt, and unconsolidated alluvial deposits. Horizontal transport also took place within spatially limited perched groundwater horizons in the vadose zone (VZ). Breakthrough at the regional aquifer water table is thought to have occurred in multiple hydraulic windows or 'drip points' located approximately 3 miles from the initial release site at the power plant. The regional aquifer is highly heterogeneous and anisotropic, and displays unconfined behavior in an upper zone and confined behavior at depth. The Cr(VI) contamination has so far remained confined to the upper approximately 25 m of the aquifer. The ambient hydraulic gradient in the regional aquifer is toward the east/southeast, but the water table is extremely flat in the Cr plume area compared to surrounding areas. There is some local variability in flow direction, potentially related to effects of seasonal pumping from nearby Los Alamos county water-supply wells, which draw from the regional aquifer several hundred feet below the water table. An interim measure (IM) is currently operating at the site to achieve hydraulic control of plume migration using a series of extraction wells, ion-exchange treatment, and injection along the downgradient portion of the plume. The migration of Cr contamination in the aquifer has been previously evaluated using a calibrated probabilistic numerical model, which has also been used to design the spatial configuration of extraction and injection wells for the IM, and to identify locations for monitoring wells. The model is built using the Finite Element Heat and Mass (FEHM) transfer code, and the Model Analysis and Decision Support (MADS) software is used for analyses. To date, the portion of the IM system that has operated in a continuous manner has been very effective at hydraulic control. Modifications to the flow field associated with the influence of IM pumping and injection are evaluated here, with modeling constraints to be provided by recent observations of tracers injected into both monitoring wells and IM injection wells, as well as opportunistic observations of geochemical signatures of treated ion-exchange effluent injected into IM injection wells. The tracer and geochemical data provide insights into modified flow directions within the Cr plume area that cannot be otherwise obtained. Results of the modeling work are being used to inform adaptive management of IM operations and will be a valuable tool for development and refinement of long-term remediation strategies. (authors)

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

Understanding Manufacturing Scale-Up of Plasma Sprayed Zr Barrier Coatings on U-10Mo Fuels

The United States government is committed to further strengthening nuclear security and nonproliferation in order to reduce the threat of terrorists acquiring nuclear material. To meet this important mission, the US Department of Energy/National Nuclear Security Agency's Office of Material Management and Minimization (M{sup 3}) Reactor Conversion Program minimizes the use of highly enriched uranium (HEU) in civilian applications by working with governments and facilities around the world to convert research reactors to the use of non-weapons-usable low-enriched uranium (LEU) fuels. Each reactor converted not only eliminates the need for HEU at civilian sites, but also reduces the amount of HEU being manufactured, stored, and transported, where it is at its most vulnerable. In instances where suitable LEU fuels do not exist for particular reactors to convert, the M{sup 3} Reactor Conversion Program (CONVERT) contributes to the development of new LEU fuels. Currently, the M{sup 3} Reactor Conversion Program is committed to developing a high-density U-Mo LEU fuel that will be able to convert high performance research reactors while establishing an efficient and economic fabrication capability to make the fuel. To that end, Los Alamos National Laboratory (LANL) has been tasked with developing a method to affix a Zirconium fuel diffusion barrier via plasma spraying that will prevent a reaction between the U-Mo fuel and the aluminum cladding. This report discusses the transitions from development of the plasma spray deposition process to a scientific understanding of pseudo-production conditions and quantification of coating layer properties, characterization, reproducibility, etc. Understanding quality assurance and its applications to experiment work is non-trivial; statistical analysis of various measurement techniques not only improve quality, but will allow for scale up of manufacturing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Characterizing the Benefits of Seismic Isolation of Nuclear Structures in Terms of Reduced Risk and Cost

Seismic isolation is a mature technology that has been implemented in many civil structures, including buildings, bridges, liquid natural gas tanks, and off shore oil platforms, both in the United States and other countries, to mitigate the damaging effects of earthquakes. Seismic isolation has also been used in nuclear structures in France and South Africa, but not yet in the United States: neither in the Department of Energy facilities nor in commercial nuclear power plants (NPPs). This is primarily due to a lack of guidelines, codes and standards for the analysis, design and construction specific to seismically isolated nuclear structures. However, seismic isolation of nuclear structures has seen increased research interest in the recent years and the recently published national consensus standard, America Society of Civil Engineers (ASCE) Standard 4-16 'Seismic analysis of safety related nuclear structures', now incorporates language and commentary (Chapter 12) for seismically isolating surface or near-surface-mounted nuclear facilities, including NPPs. Seismic isolation substantially reduces horizontal seismic loads (demands) on structures, systems, and components. Reduction in demand results in four potential benefits: (1) economic: reduction in capital cost, (2) increased safety: reduction in the mean annual frequency of unacceptable performance, (3) insurance: protection against increases in the known seismic hazard after construction by minimizing the effort to re-qualify and re-certify structures, systems and components, and (4) recertification: the opportunity to certify an existing NPP design for a region of higher seismic hazard. Item (2) above, an increase in safety (reduction of seismic risk), was explored in Huang et al., wherein it was demonstrated that the implementation of seismic isolation reduced seismic risk in nuclear power plant structures. Studies that assess item (1) have not been performed although the use of isolation may lead to large reductions in the capital cost of safety-related nuclear facilities. Funding provided by Nuclear Safety Research and Development in the Department of Energy (DOE) allowed the authors to develop a framework for assessing the economic benefits and reductions in seismic risk afforded by the use of seismic (base) isolation. The framework includes probabilistic risk assessment and estimation of overnight capital costs (OCC's) of a sample generic DOE nuclear facility (GDNF). The project is documented in detail in Bolisetti et al. and Yu et al., and key results are presented here. The GDNF considered in this study is based on a facility constructed at the Idaho National Laboratory. The GDNF is a two-story reinforced concrete building with an embedded basemat and is founded on soft rock at the boundary of Site Class B and Site Class C (B/C boundary) as characterized by ASCE/SEI Standard 7. The GDNF is assumed to handle radiological materials, and its structure, systems and components (SSC's) are designed to effectively confine these materials in the event of an internal failure. The GDNF is populated with components that are generic to safety-related nuclear structures. A single system, which is intended to confine a loss of materials-at-risk (MAR) to the interior of the structure, is chosen for this study. Accordingly, a hypothetical event tree and fault tree are assumed for risk assessment. The isolation system for the GDNF is designed per Chapter 12 of ASCE/SEI 4-16 but is not optimized. The isolation system consists of 38 lead rubber bearings installed in one horizontal plane below the basemat. Details of the GDNF, including the SSC's, event tree and fault tree, and isolation system, are presented in the project report and Yu et al.. This study includes seismic risk assessment and cost estimation for four cases: (1) the GDNF is located at a site of low to moderate seismic hazard and is constructed on a conventional (non-isolated) foundation; (2) the GDNF is located at the same site as case 1, but is seismically isolated; (3) the GDNF is located at a site of high seismic hazard and is constructed on a conventional foundation; and (4) the GDNF is located at the same site of high seismic hazard as case 3, but is seismically isolated. In this study, the DOE site at Idaho National Laboratory (INL) is chosen as the low to moderate seismic hazard site for cases 1 and 2, and the Los Alamos National Laboratory (LANL) is chosen as the high seismic hazard site for cases 3 and 4.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Current space environment measurements made by Los Alamos National Laboratory - 25614

Los Alamos National Laboratory has a long history in the measurement of the space environment to meet scientific and national security needs. This proceeding provides an overview of the current measurements collected by a subset of the Los Alamos built instruments and their capability to contribute to the understanding of complex environment where satellites operate. Provided is an overview of sensors on the GPS constellation, NASA's Van Allen Probes, and the LANL-GEO instruments at geostationary orbit. Discussion is aimed at the scientist not well versed in space physics or space radiation detection but familiar with laboratory radiation detection. Data examples are provided as well as brief discussion of the features in the presented data and their relation to spacecraft hazards. (authors)

54 ENVIRONMENTAL SCIENCES↗