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At least 1,639 records · Page 91
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Testing of ENDF71x: A new ACE-formatted neutron data library based on ENDF/B-VII.1

The ENDF71x library [1] is the most thoroughly tested set of ACE-format data tables ever released by the Nuclear Data Team at Los Alamos National Laboratory (LANL). It is based on ENDF/B-VII. 1, the most recently released set of evaluated nuclear data files produced by the US Cross Section Evaluation Working Group (CSEWG). A variety of techniques were used to test and verify the ENDF7 1x library before its public release. These include the use of automated checking codes written by members of the Nuclear Data Team, visual inspections of key neutron data, MCNP6 calculations designed to test data for every included combination of isotope and temperature as comprehensively as possible, and direct comparisons between ENDF71x and previous ACE library releases. Visual inspection of some of the most important neutron data revealed energy balance problems and unphysical discontinuities in the cross sections for some nuclides. Doppler broadening of the total cross sections with increasing temperature was found to be qualitatively correct. Test calculations performed using MCNP prompted two modifications to the MCNP6 source code and also exposed bad secondary neutron yields for {sup 231,233}Pa that are present in both ENDF/B-VII.1 and ENDF/B-VII.0. A comparison of ENDF71x with its predecessor ACE library, ENDF70, showed that dramatic changes have been made in the neutron cross section data for a number of isotopes between ENDF/B-VII.0 and ENDF/B-VII.1. Based on the results of these verification tests and the validation tests performed by Kahler, et al. [2], the ENDF71x library is recommended for use in all Monte Carlo applications. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

A Mobile Automated Tomographic Gamma Scanning System - 13231

Canberra Industries have recently designed and built a new automated Tomographic Gamma Scanning (TGS) system for mobile deployment. The TGS technique combines high-resolution gamma spectroscopy with low spatial resolution 3-dimensional image reconstruction to provide increased accuracy over traditional approaches for the assay of non-uniform source distributions in low-to medium-density, non-heterogeneous matrices. Originally pioneered by R. Estep at Los Alamos National Laboratory (LANL), the TGS method has been further developed and commercialized by Canberra Industries in recent years. The present system advances the state of the art on several fronts: it is designed to be housed in a standard cargo transport container for ease of transport, allowing waste characterization at multiple facilities under the purview of a single operator. Conveyor feed, drum rotator, and detector and collimator positioning mechanisms operated by programmable logic control (PLC) allow automated batch mode operation. The variable geometry settings can accommodate a wide range of waste packaging, including but not limited to standard 220 liter drums, 380 liter overpack drums, and smaller 20 liter cans. A 20 mCi Eu-152 transmission source provides attenuation corrections for drum matrices up to 1 g/cm{sup 3} in TGS mode; the system can be operated in Segmented Gamma Scanning (SGS) mode to measure higher density drums. To support TGS assays at higher densities, the source shield is sufficient to house an alternate Co-60 transmission source of higher activity, up to 250 mCi. An automated shutter and attenuator assembly is provided for operating the system with a dual intensity transmission source. The system's 1500 kg capacity rotator turntable can handle heavy containers such as concrete lined 380 liter overpack drums. Finally, data acquisition utilizes Canberra's Broad Energy Germanium (BEGE) detector and Lynx MCA, with 32 k channels, providing better than 0.1 keV/channel resolution to support both isotopic analysis with the MGA/MGAU software and a wide 3 MeV dynamic range. The calibration and verification of the system is discussed, and quantitative results are presented for a variety of drum types and matrices. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Data and Model-Driven Decision Support for Environmental Management of a Chromium Plume at Los Alamos National Laboratory - 13264

A series of site investigations and decision-support analyses have been performed related to a chromium plume in the regional aquifer beneath the Los Alamos National Laboratory (LANL). Based on the collected data and site information, alternative conceptual and numerical models representing governing subsurface processes with different complexity and resolution have been developed. The current conceptual model is supported by multiple lines of evidence based on comprehensive analyses of the available data and modeling results. The model is applied for decision-support analyses related to estimation of contaminant- arrival locations and chromium mass flux reaching the regional aquifer, and to optimization of a site monitoring-well network. Plume characterization is a challenging and non-unique problem because multiple models and contamination scenarios are consistent with the site data and conceptual knowledge. To solve this complex problem, an advanced methodology based on model calibration and uncertainty quantification has been developed within the computational framework MADS (http://mads.lanl.gov). This work implements high-performance computing and novel, efficient and robust model analysis techniques for optimization and uncertainty quantification (ABAGUS, Squads, multi-try (multi-start) techniques), which allow for solving problems with large degrees of freedom. (authors)

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

The SAVY 4000 Container Storage Program at Los Alamos National Laboratory - 17313

Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in the U.S. Defense Nuclear Safety Board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based storage container design, the SAVY-4000{sup R}. The SAVY-4000{sup R} is the first vented general use nuclear material container demonstrated to meet the requirements of DOE M 441.1-1, Nuclear Material Packaging Manual. The SAVY-4000{sup R} is an innovative and creative design demonstrated by the fact that it can be opened and closed in a few seconds without torque wrenches or other tools; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas, yet retains 99.97% of plutonium particulates, and prevents release of material even in a 12 foot drop. Finally, it has been fire then drop tested and will reduce the risk to the public in the event of an earthquake/fire scenario. This will allow major nuclear facilities to credit the container towards source term Material at Risk (MAR) reduction. The container was approved for nuclear material storage in the TA-55 Plutonium Facility on March 15, 2011, and there are currently over 2000 SAVY-4000{sup R} containers available for use within the facility. A Safety Analysis Report was developed to demonstrate Manual 441.1-1 compliance and was approved April, 2014. The first four SAVY-4000{sup R} containers were packaged with plutonium on August 2, 2011. Key aspects of the SAVY-4000{sup R} vented storage container design will be discussed which include design description, qualification and testing, surveillance strategy, and the design life extension program as enhanced by surveillance activities with the intent to extend well beyond the current five year design life. (authors)

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

Self-Absorption Corrections for Plutonium Measurements With the Tomographic Gamma Scanner - 17325

The Tomographic Gamma Scanner (TGS) is a Non-Destructive Assay (NDA) instrument that combines high-resolution gamma spectroscopy with low spatial resolution three-dimensional image reconstruction in order to reduce bias and improve precision for measurements of drummed waste. The TGS technique is especially suited to measure containers with non-uniform source distributions and low-to me-dium-density heterogeneous matrices, compared with traditional NDA methods. Canberra recently built, installed, and commissioned a TGS system, intended primarily for the measurement of plutonium (Pu) bearing waste, at a European nu-clear fuel reprocessing facility. Pu assays using the TGS technique present technical challenges distinct from those typically encountered in traditional LLW applications. Among these is the need to apply corrections for self-absorption effects that occur on scales smaller than the TGS spatial resolution, and can compromise the assay result if not included in the TGS imaging methodology. We present a brief review of the Self Absorption Correction (SAC) technique, and discuss the complications involved with its application to the TGS methodology with examples from our recent experience. This work continues a Canberra history of further developing and refining a technique to solve emerging problems, having commercialized the original methodology by partnering with the research institution that pioneered it. The TGS and the SAC methodologies were originally developed by Los Alamos National Laboratory (LANL) and John Fleissner respectively, and this work illustrates how Canberra has implemented an advancement of these techniques in a new TGS system. (authors)

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

Simulated Impacts of Water Introduced into Pits 37 and 38 at Technical Area 54, Area G, Los Alamos National Laboratory - 17386

Water is used in the disposal of bulk waste generated by cleanup activities at LANL, including large quantities of waste retrieved from Material Disposal Area (MDA) B. This water is used to wash material from the metal bins in which the material is transported to Area G, to decontaminate the containers so they can be returned to service, to decontaminate the trucks used to transport the bins, and for general dust suppression during waste placement. The introduction of this water into pits 37 and 38 at Area G during disposal of MDA B waste in those pits may affect the rates at which water percolates through, and leaches contaminants from, the waste in the disposal pits, and thus may hasten the transport of contamination to the regional aquifer below the disposal facility. A special analysis was conducted to evaluate the potential impacts of the water used during bulk waste disposal on the long-term performance of the disposal facility. (authors)

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

Diversity and the citizens' advisory board: aligning disparate views for better results - 15469

The Northern New Mexico Citizens' Advisory Board (NNMCAB) is a group of -citizen experts,- i.e., experts on being citizens who take the time and focus their attention to grow more informed about nuclear science, nuclear waste remediation, and environmental monitoring issues. The Board digests complex information associated with these topics in light of the Los Alamos National Laboratory's (LANL) and Department of Energy's (DOE) activities in Northern New Mexico and discusses and conveys this information clearly to a diverse regional stakeholder population that includes a 400+ year old Hispanic culture, a 1,000+ year old Native American pueblo culture, a highly educated scientific and technical community of federal workers and their families, rural agricultural communities, creative/artistic communities, and political and peace activists. Our Board has either representation from each of these diverse New Mexican sectors or enjoys public comment from these sectors. Projects that involve such stakeholder-expert exchanges include remediation of an underground chromium plume, continued investigation and remediation associated with a radiation leak at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, the transport of Transuranic Waste off the Los Alamos 'Hill,- the ultimate disposition of -33 Shafts,- currently in place at Los Alamos National Laboratory, and the distilled voice of the citizen for continued and sufficient federal funding each year for responsible clean-up and monitoring of nuclear activities. Through the NNMCAB's communication venue of regular public meetings, mis-characterization is prevented, public input is incorporated into the official record, and recommendations to the DOE are promulgated. Scientific and technical experts join the bi-monthly NNMCAB meetings and monthly committee meetings to present and discuss current topics. The availability of these experts and their deeper reference materials is key to the interchange, as is the respectful and patient dialogue that exists between board members and the subject matter experts, both in the public forums and between meetings. Meetings are professionally organized and orchestrated to maximize the value of the experts' and board members' participation time. The professional format of the meetings ensures maximum learning and relationship-building which fosters effective channels for public dialogue. (authors)

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

Testing and Versatility of the SAVY 4000 Nuclear Material Container - 15634

The SAVY 4000 container was designed by Nuclear Filter Technology, Inc. to meet DOE M441.1-1 requirements for safe storage of nuclear material, and it is currently being used at Los Alamos National Laboratory, Livermore National Laboratory, Idaho National Laboratory and Nevada National Security Site. The 316L stainless steel container can be manufactured in a range of sizes from 1 quart to 10 gallon capacity, it can be opened and closed easily by hand, it is sealed with a Viton{sup R} O-ring, and the filter retains particulates and resists liquid water entry while allowing hydrogen gas to escape. This innovative design has proven to provide benefits and uses beyond its original intent. For example, the container was tested in a simulated facility design basis accident involving high temperatures, fuel fire, and subsequent drop, and the damage ratio was measured. This has allowed the LANL plutonium facility to take credit for a material-at-risk reduction of approximately two orders of magnitude for material stored in the SAVY 4000 container. In addition, the water resistant filter allows the nuclear facility to credit the container as a safety feature to mitigate the risk of a criticality accident in the event of facility sprinkler system activation. The container has also been approved to transfer small quantities of solid nuclear material by hand carry between Los Alamos facilities, and it has been approved and used as a waste container for transuranic waste destined for the Waste Isolation Pilot Plant. The 1 quart capacity container and the future 2 quart design both fit inside the 9977 shipping container, and this has facilitated the transport of nuclear material from Idaho National Laboratory to Nevada Nuclear Security Site, obviated the need for repackaging the material for storage upon arrival. Finally, Los Alamos has recently contracted with Nuclear Filter Technology, Inc. to perform testing to certify the SAVY 4000 container as a DOT Type A liquid transport container. This paper will provide an overview of the various testing regimes and describe the resulting versatility of the container for nuclear material containment. (authors)

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

ZEM: Integrated Framework for Real-Time Data and Model Analyses for Robust Environmental Management Decision Making - 16438

Typically environmental management (EM) problems require analysis of large and complex data sets originating from concurrent data streams with different data collection frequencies and pedigree. These big data sets require on-the-fly analysis, assimilation and integration into a series of models with different complexity for various types of model-based analyses. This is needed to provide fast iterative model analyses based on the latest available data to guide decision-making. Furthermore, the data and model are associated with uncertainties. The uncertainties are probabilistic (e.g. measurement errors) and non-probabilistic (unknowns, e.g. alternative conceptual models characterizing site conditions). To address all of these issues, we have developed an integrated framework for real-time data and model analyses for environmental decision-making called ZEM. The framework allows for seamless and on-the-fly integration of data and modeling results for robust and scientifically-defensible decision-making applying advanced decision analyses tools such as Bayesian-Information-Gap Decision Theory (BIG-DT). The framework also includes advanced methods for optimization that are capable of dealing with a large number of unknown model parameters, and surrogate (reduced order) modeling capabilities based on support vector regression techniques. The framework is predominantly coded in Julia, a state-of-the-art high-performance programing language (http://julialang.org) and incorporates the MADS computational tools (Model Analysis and Decision Support; http://mads.lanl.gov). We demonstrate the application of ZEM for data and model analyses at the LANL Chromium site. The ZEM framework will be released as open-source code and can be applied to any environmental management site. (authors)

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

Scientific Investigations on the Chemical Reactivity of Los Alamos Remediated Nitrate Salt Wastes - 16540

The February 14, 2014 release of radioactivity at the Waste Isolation Pilot Plant (WIPP) was the result of runaway thermal reactions that occurred in a TRU waste drum containing nitrate salts remediated using an organic absorbent. Experimental and modeling studies performed at LANL indicate that mixtures of metal nitrate salts (oxidizer) with sWheat Scoop{sup R} organic kitty litter (fuel) create the potential for exothermic chemical reactions. The use of sWheat{sup R} absorbent in the processing of nitrate salt wastes can be pinpointed as the critical processing decision that led to the failure of the breached drum in the WIPP repository. The most plausible scenario is that a production of heat, either from low-level chemical reactions or the growth of natural microbes, in concert with mixed metal nitrate salts, bismuth lined glovebox gloves and/or lead nitrates when combined with the sWheat{sup R} organic kitty litter, generated a series of exothermic reactions that heated and pressurized the drum resulting in the venting of high-temperature gases and radioactive material into the room. Observations and modeling studies indicate that the probability of thermal runaway in another drum has decreased with time since the WIPP event. However, the complexity of the mixtures, ambiguity in procedures such as those used for neutralization, the heterogeneity of the drum contents, and the difficulty of sampling leads to an irreducible level of uncertainty that mandates the exercise of caution in managing RNS wastes. The situation also requires that this waste stream continue to be stored in a temperature-controlled environment and monitored until the waste can be remediated to remove the hazard permanently. We recommend that proposals to treat nitrate salts with organic absorbents of any kind be accompanied by experimental studies to confirm that similarly sensitive mixtures are not inadvertently created over the range of conditions (mixture quantities, moisture content, etc.) anticipated during processing and storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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↗