SEARCH · Engineering Papers

Results for “radioactive waste processing and disposal”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 253 records · Page 14

All sources NASA NTRS DOE OSTI

DOE OSTI2024

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada

This Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) is for Corrective Action Unit (CAU) 577, “Area 5 Chromium Containing Waste Disposal Cells,” which is located in Area 5 of the Nevada National Security Site (NNSS). CAU 577 includes five low-level waste cells at the Area 5 Radioactive Waste Management Site (RWMS) where buried waste received from Nuclear Fuel Services, Inc. (NFS), was subsequently determined to contain chromium that exceeded the Toxicity Characteristic Leaching Procedure regulatory limit, which would require the waste to carry hazardous waste code D007. To resolve issues related to this alleged violation of hazardous waste regulations, a Settlement Agreement was executed between the Nevada Division of Environmental Protection (NDEP) and the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office on April 25, 2019, which identified agreed-upon mitigation measures. The Settlement Agreement required that the chromium-containing waste received from NFS would be addressed following the closure process laid out in the Federal Facility Agreement and Consent Order (FFACO).

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2023

(Project 18-15502) Reducing Uncertainty in Radionuclide Transport Prediction Using Multiple Environmental Tracers (Final Report)

In order to successfully site and design a nuclear waste disposal facility, DOE scientists are required to show safe containment of the radioactive waste for up to 1 million years. A significant hurdle to accurately and convincingly demonstrating disposal safety is predicting the fate of radioactive elements once they enter the groundwater system surrounding the repository. These predictions are often made with computer models, which contain accurate physics of groundwater movement and chemical reactions that occur during groundwater flow. As computer power increases, the physics and chemistry of these computer simulators can become more and more realistic and the physically based error decreases. However, the ability of these computer models to provide accurate predictions in a specific place, over long time periods, requires the scientists and engineers to know the subsurface properties of the Earth that control groundwater movement. In particular, groundwater scientists and engineers need to know the groundwater fluid velocity, which can change over time and strongly vary with location within the groundwater system. Because the Earth’s subsurface cannot be directly seen, and can only be sampled at drilling locations, the properties of the groundwater system are never known completely, and computer models of groundwater transport will always have some amount of uncertainty. This project’s principal goal was to use chemicals and isotope “tracers”, which have been introduced to the groundwater system by natural processes over long time periods, to help inform computer models of the groundwater velocity and subsurface properties. The goal was to calculate how much better predictions of groundwater transport were when these “tracers” were used to inform the computer models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2026

Reduced-Order CFD Modeling to Support Waste Loading Optimization in Hanford WTP Vitrification

The U.S. DOE Hanford Site stores over 56 million gallons of radioactive liquid tank waste that must be treated and immobilized for long-term disposal The Waste Treatment and Immobilization Plant (WTP) will vitrify this waste by feeding it into Joule-heated melters, where it is incorporated into a stable borosilicate glass Computational fluid dynamics (CFD) simulations of glass melters can provide insight into the maximum achievable waste loading under varying melter operating conditions Fully resolved VOF multiphase simulations were used as the reference model to capture bubble-driven convection in the melter, including bubble formation, rise behavior, and induced glass melt circulation Effective bubble column diameter and rise velocity were extracted from the resolved simulations, compared with empirical correlations, and refit across relevant viscosity and gas flow rate conditions Explicit gas–liquid interface tracking was replaced with a single-phase momentum source term model, enabling faster steady-state CFD simulations while preserving the dominant hydrodynamic effects of bubbling New empirical correlations were developed for effective bubble column diameter and bubble rise velocity by fitting resolved simulation data across expected melter viscosity and gas flow rate ranges, providing improved inputs for the momentum source term model compared with existing literature correlations The momentum source term model reduced fluid-domain mesh size by 89% and achieved an 8.4× computational speedup relative to resolved bubbling simulations The validated momentum source term approach enables prediction of process-relevant heat transfer behavior in the integrated melter model, including heat transfer from the molten glass to the cold cap, plenum, refractory walls, and surrounding structural regions under varying melter operating conditions

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

DOE OSTI2023

Review of Mercury Sequestration in Cementitious Waste Forms

Decontaminated salt solution from Tank 50 at the Savannah River Site (SRS) is known to contain dissolved mercury species which are present as hydroxides and oxides. These species are reported in analytical measurements as the organic cation [CH 3 Hg] + and the inorganic cation Hg 2+ . The salt solution is mixed with cementitious reagents, blast-furnace slag (BFS) and thermally beneficiated Class F fly ash in the Saltstone Processing Facility and pumped into engineered Saltstone Disposal Units, (SDUs), to produce a solidified waste form. The solidified waste form, saltstone, encapsulates and chemically stabilizes the mercury and other contaminants, sequestering them from release into the environment. With the discovery of methylmercury in the salt solution, testing was performed to determine the extent to which saltstone prepared with methylmercury (MeHg) retains mercury in the Toxicity Characteristic Leaching Procedure (TCLP). Results from this testing indicated that the amount of mercury extracted remained below the regulatory limit, and that the saltstone formulation with a greater proportion of BFS performed slightly better at retaining Hg in the solid.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

GDSA Framework Development and Process Model Integration FY2023

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

PFAS Waste Management for Industrial, Water, and Wastewater Treatment Facilities

Per- and polyfluoroalkyl substances (PFAS), including perfluorooctanesulfonic acid (PFOS) and perfluorooctanoic acid (PFOA), are widely used in industrial and consumer products due to their unique water and oil repelling properties. These slow-degrading and bio-accumulating compounds, often called "forever chemicals," are not typically removed by conventional industrial and municipal water/wastewater (W/WW) treatment. This tip of the month provides comprehensive strategies for managing PFAS-contaminated waste streams. Understanding the various types of PFAS waste streams is the first step in managing them. Primary PFAS waste includes contaminated process W/WW and solid materials directly from the production and processing of PFAS-containing materials. These might come from manufacturing facilities, firefighting training areas, or other industrial operations where PFAS compounds are used or were historically present. Secondary PFAS waste results from treatment and disposal mechanisms of primary PFAS waste streams, including spent activated carbon, used ion exchange resins, membrane reject streams, PFAS-laden backwash water, landfill leachate, incineration residuals, etc. TABLE 1 lists examples of the most common PFAS waste streams in different industrial facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Physical Sciences Vistas Issue 3 2021

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

2020 LANL Contributions to the BATS Test in WIPP

The DOE Office of Nuclear Energy (DOE-NE) repository research and development program seeks to provide a sound technical basis for multiple viable disposal options, increase confidence in the robustness of generic disposal concepts, and develop science and engineering tools needed to support disposal concept implementation. Sandia, Los Alamos, and Lawrence Berkeley National labs are conducting research into salt, which includes the Brine Availability Test in Salt (BATS) field test at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management facility. BATS is leveraging the existing infrastructure associated with the WIPP to advance science informing generic disposal concepts. BATS is an ongoing heated borehole experiment being conducted in the WIPP underground. The goal of this experiment is to reduce the uncertainty associated with spent nuclear fuel disposition in geologic salt formations. The BATS experiments is designed to accomplish this by increasing our understanding of brine migration in salt, assessing damaged zones from mining and drilling, simulating a post-closure environment, confirming salt properties, and providing data for model validation. Phased testing for generic salt repository R&D was proposed in 2015. Subsequently, borehole thermal testing was developed as a first step in this process. This document summarizes the work conducted on the BATS experiment during fiscal year 2020 with an emphasis on the contributions by Los Alamos National Lab.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Crucible Melter Simulation in Nek5000

Nuclear tank waste at the Hanford site is slated for vitrification in large scale refractory-lined melters to transform it into a stable borosilicate waste form suitable for long-term storage or disposal. Molten glass corrodes the refractory lining over time at a rate correlated to the velocity of molten glass against the refractory-lined wall. To properly design a melter and plan for maintenance, it is necessary to accurately model the glass flow and find the correlation between flow velocity and corrosion rate. This modeling was started in STAR-CCM+ CFD software and is being continued in Nek5000 for its fast-running and quick turnaround code. This poster explains the basic process of reconstructing the geometry, fluid properties, and heating of test melters in Nek5000. The geometry is imported from a mesh file and boundary conditions assigned based on surface IDs. The fluid properties are set in the .par and .usr case files. The heating of the actual crucibles in joule heating with electrodes but simulated in Nek5000 with a volumetric heat source term. This volumetric heat source is fit to the actual heating profile with piecewise polynomials and exponentials. Basic results and verification methods are explained in the poster, as well as further work that must be done to complete modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Canonsburg, Pennsylvania, UMTRCA Title I Disposal Site 35 Years Postclosure (1985-2020) - 20288

The DOE Office of Legacy Management (LM) Canonsburg, Pennsylvania, Disposal Site is a former uranium ore-processing site located in the borough of Canonsburg and an important part of US history. For 46 years (1911-1957), the site processed ore, for vanadium, radium and then uranium. Soils and groundwater beneath the site were contaminated by the milling operations. The 15-hectare (37-acre) site is owned by the US government and managed by LM, which serves as the federal land manager and steward of cultural, historical, and natural resources at Cold War legacy sites that have been successfully cleaned up to remedial standards. In 1978 Congress directed DOE to remediate 22 inactive uranium-ore processing sites under the Uranium Mill Tailings Radiation Control Act (UMTRCA), in accordance with standards promulgated by the US Environmental Protection Agency in Title 40 CFR Part 192 (40 (CFR 192). Subpart B of 40 CFR 192 regulates the cleanup of contaminated groundwater at the processing sites. The radioactive materials were encapsulated in US Nuclear Regulatory Commission (NRC)-approved disposal cells, and the NRC general license for UMTRCA Title I sites is established in 10 CFR 40.27. Canonsburg was the first site to be remediated under UMTRCA. From 1983 to 1985, the Canonsburg site underwent cleanup under this US law that provides for the safe and environmentally sound disposal, long-term stabilization, and control of uranium mill tailings in a manner that minimizes or eliminates radiation health hazards to the public. The site is unique in that it is one of only two UMTRCA Title I sites in the eastern half of the United States, the other site being the Burrell, Pennsylvania, Title I Disposal Site, a vicinity property to the Canonsburg site. It has been 35 years since the disposal cell at the Canonsburg site was completed. This paper discusses the many efforts that have been taken to maintain the site and presents some of the lessons learned concerning the long-term care of the site. The discussion includes but is not limited to the following: (1) Preserving site history so the community can better appreciate and understand the site. (2) Community outreach efforts, and the importance of maintaining transparency of environmental monitoring results with the community. (3) Pursuing reuse opportunities that offer the potential for improving the community. (4) Remaining engaged with the local community. (5) Remaining proactive in the long-term care of the site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Evolving radon diffusion through earthen barriers at uranium waste disposal sites

Field measurements of Rn-222 fluxes from the tops and bottoms of compacted clay radon barriers were used to calculate effective Rn diffusion coefficients (D Rn ) at four uranium waste disposal sites in the western United States to assess cover performance after more than 20 years of service. Values of D Rn ranged from 7.4 × 10 -7 to 6.0 × 10 -9 m 2 /s, averaging 1.42 × 10 -7 . Water saturation (S W ) from soil cores indicated that there was relatively little control of D Rn by S W , especially at higher moisture levels, in contrast to estimates from most steady-state diffusion models. Further, this is attributed to preferential pathways intrinsic to construction of the barriers or to natural process that have developed over time including desiccation cracks, root channels, and insect burrows in the engineered earthen barriers. A modification to some models in which fast and slow pathway D Rn values are partitioned appears to give a good representation of the data; 4% of the fast pathway was needed to fit the data regression. For locations with high S w and highest D Rn (and fluxes) at each site, the proportion of fast pathway ranged from 1.7% to 34%, but for many locations with lower fluxes, little if any fast pathway was needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 An Integrated Regional Approach To Risk Management of Industrial Systems, S. Chakraborty and R. Stratton; 9 Annual Technical Meeting of the NRC Cooperative Severe Accident Research Program, E. G. Silver; ACCIDENT ANALYSIS: 13 Deterministic Severe Accident Criteria as Severe Accident Design Criteria and Policy for the New Production Reactor-Heavy Water Reactor, P. T. Rhoads; 20 Proposed Deterministic Severe Accident Criteria for the Heavy Water Reactor-New Production Reactor Containment, K. D. Bergeron, S. E. Slezak, and C. E. Leach; 33 Quasi-Static Core Liquid Level Depression and Long-Term Core Uncovery During a PWR LOCA, Y. Kukita, R. R. Schultz, H. Nakamura, and J. Katayama; 49 LIRA: An Advanced Containment System to Minimize the Accidental Radioactivity Releases, A. Turricchia; 63 Errata to “A Review of Hydrogen Production During Melt/Water Interaction in LWRs,” Vol. 33, No. 4; CONTROL AND INSTRUMENTATION: 64 The Nuplex 80+™ Advanced Control Complex from ABB Combustion Engineering, F. Ridolfo, D. Harmon, and K. Scarola; DESIGN FEATURES: 76 A Scheme for Passive Isolation of the Containment of a Reactor, A. K. Ghosh, V. V. Raj, and K. Kakodkar; 84 Utility Requirements for Safety in the Passive Advanced Light-Water Reactor, T. U. Marston, W. H. Layman, and G. Bockhold, Jr.; WASTE AND SPENT FUEL MANAGEMENT: 91 Activities Related to Waste and Spent Fuel Management, Compiled by M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 103 Testing Deficiencies in Auxiliary Systems Feedwater Systems, J. D. Kueck; 110 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 113 Selected Safety-Related Events, Compiled by G. A. Murphy; 115 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 133 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 145 Reports, Standards, and Safety Guides, D. S. Queener; 150 Proposed Rule Changes as of Sept. 30, 1992; ANNOUNCEMENTS: 75 Harvard Short Course on Risk: Science, Assessment, and Management; 102 International Conference on Reactor Physics and Reactor Computations; 102 The Fourth International Conference on Nuclear Waste Reprocessing and Waste Management [RECOD ’94]; 132 International Symposium on the Radiological Exposure of the Population of the European Community from Radioactivity in the Mediterranean Sea [MARINA-MED]; 162 International High-Level Radioactive Waste Conference ‘Technology Integration—Above and Below”; 154 The Authors; 158 Indexes to Nuclear Safety, Volume 33.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Vadose Zone Model for B Plant Area for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The scope of this ECF is to document the development and results of the B Plant Area vadose zone model. The B Plant Area model encompasses the region around B Plant (221-B) in the western part of 200 East Area. The plant was used to separate plutonium from irradiated fuel using the bismuth-phosphate process from 1945 to 1956, and for strontium and cesium recovery from 1968 to 1985.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Vadose Zone Model for SALDS for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037,Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed "dry" in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗