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DOE OSTI2017

Annual Status Report (FY2016) Performance Assessment for the Environmental Restoration Disposal Facility

DOE O 435.1, Radioactive Waste Management, and DOE M 435.1-1, Radioactive Waste Management Manual, require that a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Annual summaries of low-level waste (LLW) disposal operations must be prepared with respect to the conclusions and recommendations of the PA and CA, and a determination of the need to revise the PA or CA must be made. The annual summary requirement provides a structured approach for demonstrating the continued adequacy of the PA and CA in demonstrating a reasonable expectation that the performance objectives will be met. This annual summary addresses only the status of the Environmental Restoration Disposal Facility (ERDF) PA (CP-60089, Performance Assessment for the Environmental Restoration Disposal Facility, Hanford Site, Washington, formerly WCH-520 Rev. 1)1. The CA for ERDF is supported by DOE/RL-2016-62, Annual Status Report (FY 2016): Composite Analysis of Low Level Waste Disposal in the Central Plateau at the Hanford Site. The ERDF PA portion of the CA document is found in Section 3.1.4, and the ERDF operations portion is found in Section 3.3.3.2 of that document.

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DOE OSTI2015

Information on the Advanced Plant Experiment (APEX) Test Facility

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, "Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing," Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

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DOE OSTI2017

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

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DOE OSTI2017

Effluent Management Facility Evaporator Bottom-Waste Streams Formulation and Waste Form Qualification Testing

This report describes the results from grout formulation and cementitious waste form qualification testing performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates three grout formulations proposed for wide-range treatment of different waste stream compositions expected for the Hanford Effluent Management Facility (EMF) evaporator bottom waste. This work supports the technical development need for alternative disposition paths for the EMF evaporator bottom wastes and future direct feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form qualification testing. The work contained within this report relates to waste form development and testing, and does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY 2017 and future waste form development efforts. The provided results and data should be used by (1) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (2) decision makers interested in off-site waste form disposal, and (3) the U.S. Department of Energy, their Hanford Site contractors and stakeholders as they assess the IDF PA program at the Hanford Site. The results reported help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottom waste, and improve the technical defensibility of long-term waste form risk estimates.

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DOE OSTI2017

Impact of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities

Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The SRS liquid waste contractor requested an assessment of the impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this impact had not been previously evaluated. A literature review revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion. For the proprietary coating systems applied to the DWPF concrete, glycolic acid was deemed compatible since the coatings were resistant to more aggressive chemistries than glycolic acid. Additionally similar coating resins showed acceptable resistance to glycolic acid.

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DOE OSTI2018

Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility – Fiscal Years 2015-2017

The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA -54), Area G disposal facility.

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DOE OSTI2018

Effluent Management Facility Evaporator Bottoms: Waste Streams Formulation and Waste Form Qualification Testing

This report describes the results from grout formulation and cementitious waste form testing performed by Pacific Northwest National Laboratory for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates grout formulations proposed for encapsulating a wide range of compositions predicted to be present in evaporator bottoms wastes from the Hanford Effluent Management Facility (EMF). This work supports the technical development need for alternative treatment and disposition paths for the EMF evaporator bottoms waste and future direct-feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form testing. The work contained within this report relates to waste form development and testing, but does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains information useful for future PA updates [beyond fiscal year (FY) 2017] and future waste form development efforts. These analytical results can be used by (i) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (ii) decision makers interested in off-site waste form disposal, and (iii) the U.S. Department of Energy, their Hanford Site contractors, and stakeholders as they continue to assess the IDF PA program at the Hanford Site. The reported results help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottoms waste, and improve the technical defensibility of long-term waste form risk estimates.

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DOE OSTI2017

A Critical Review of Ion Exchange in Nuclear Waste Glasses to Support the Immobilized Low-Activity Waste Integrated Disposal Facility Rate Model

There are over 50 million gallons of waste in underground tanks at the Hanford site in Washington State which was generated as a byproduct of nuclear materials fabrication for United States strategic defensive purposes. The waste, also known as high-level waste (HLW), is comprised of both radioactive and non-radioactive contaminants. To deal with the tank waste, the United States Department of Energy (DOE) plans to separate out the liquid, or supernatant, and salt-cake portions and treat them as low-activity waste (LAW). The LAW and HLW will be vitrified in separate facilities at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The LAW will be disposed of at the on-site Integrated Disposal Facility (IDF). The main species of concern in the immobilized low-activity waste (ILAW) include 99 Tc, 129 I, and U and other non-radioactive contaminants. Ultimately, it must be demonstrated that the contaminants released from the waste form pose minimal risk to humans and the environment. To aid in this demonstration, work is currently being done to expand the understanding of long-term glass waste form behavior, especially with regards to exposure to aqueous environments.

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DOE OSTI2019

Frit Recommendation For Sludge Batch 9 Coupled Operation With The Salt Waste Processing Facility

Prior to the processing of Sludge Batch 9 (SB9) with Frit 803 at the Defense Waste Processing Facility (DWPF), assessments and experimental glass work were completed to support SB9 coupled processing with streams from the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU). This previous work did not address coupled processing with streams from the Salt Waste Processing Facility (SWPF) because the Product Composition Control System (PCCS) models limited the TiO2 concentration in glass to 2 weight percent (wt.%). These high activity streams from SWPF include monosodium titanate (MST) and sludge solids from the Sludge Solids Receipt Tank (SSRT) as well as Cscontaining strip effluent (SE). The SWPF streams are expected to reach TiO2 concentrations in glass greater than 2 wt.% based on an MST strike of 0.4 g MST/L of salt solution. Therefore, revisions to the models contained in PCCS were completed to allow the evaluation of glasses containing greater than 2 wt.% TiO2, but less than 6 wt.%. Since the initial SB9 qualification for ARP-MCU coupled operation, additional Tank 40 analyses have been completed, which served as the basis for an updated projection of the SB9 blend composition in Tank 40. This projection will be used to support SB9 coupled operation with SWPF in 2019. Savannah River Remediation (SRR) has requested that the Savannah River National Laboratory (SRNL) select a second frit to optimize SB9 coupled operation with SWPF. The selected frit would provide operational flexibility by allowing an acceptable waste loading (WL) to be targeted while maintaining an operating window of at least ± 4 percentage points around this target. The objective of this study was to select a second frit based on a paper-study utilizing Measurement Acceptance Region (MAR) assessments with DWPF PCCS. To support these assessments, SRNL developed a new method for rigorous and directed optimization of frit compositions, which allowed for rapid screening and an assessment of a large array of candidate frits, and the ability to simultaneously satisfy the following single and double MST strike operating scenarios.

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DOE OSTI2018

Thermal evaluations for mixed organic and nitrate salt drums in interim storage at the WCS Federal Waste Facility

Waste Control Specialists LLC (WCS) operates the Federal Waste Facility (FWF) in Texas that is licensed to process and store certain types of mixed low-level radioactive waste packages. The specific waste currently housed at WCS includes some of the transuranic (TRU) waste that originated at the DOE Los Alamos National Laboratory (LANL) and are destined for disposal at the DOE Waste Isolation Pilot Plant (WIPP) facility. The TRU waste within each Standard Waste Box (SWB) was packaged in 55-gallon drums at LANL. Some of the drums contain mixed organic and nitrate salt contents suspected of experiencing a chemical reaction. The drums (up to four drums in a SWB) were then overpacked in a SWB prior to shipment to WCS. In the most common configuration, WCS placed the SWB containing the drums on metal pallets and stored two SWBs into Modular Concrete Canisters (MCCs) in a double stacked array. The MCCs were placed in a staggered array at a segregated area of the FWF and covered with sand to mitigate the heating that occurred when stored above ground on waste pads. The temporary interment is also intended to provide additional containment of the suspect reactive waste to mitigate release of radioactivity in an incident such as that which occurred at WIPP for one such drum. Void space of each MCC was filled with 1-inch pea gravel to facilitate retrievability. Thermocouple probes were placed within each MCC to monitor heat generation from the SWBs and the MCC lids were closed. The primary objective of this analysis was to estimate the heat generation of the SWB drum contents when the SWB package temperature inside the MCC and the FWF ambient temperature were monitored.

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DOE OSTI2019

HALEU Decontamination Investigations for EBR-II Recovered Uranium. HALEU Drip Casting Results in the Fuel Conditioning Facility Cathode Processor

Many of the new advanced reactor designs being proposed by the nuclear private sector require uranium fuel enriched between 5 and 20%, otherwise known as High Assay Low Enriched Uranium (HALEU), to support their designs. There are currently no commercial facilities based in the United States which have the capability to produce this material, and as a result the nuclear industry has requested support from the Department of Energy (DOE) to meet this need. One approach being considered by DOE to respond to this demand is to recover highly enriched uranium from used nuclear fuels and downblend it to enrichments less than 20 % U-235. DOE’s past research related to liquid metal fast breeder reactors resulted in the irradiation of more than 3 metric tons (MT) of nuclear fuel composed of highly enriched metallic uranium which resides at the Idaho National Laboratory (INL). This inventory of used nuclear fuel is currently being conditioned using an electrometallurgical treatment process (EMT) to facilitate its eventual disposition to a geologic repository. As part of this treatment process, the highly enriched component of this used fuel is separated, recovered, downblended to <20% U-235, and stored as a metallic uranium ingot. The majority of this fuel was irradiated in the Experimental Breeder Reactor II (EBR-II) over the course of its 30 years of operation. This sodium cooled fast reactor utilized a core composed of highly enriched, sodium-bonded, metallic fuel to sustain or “drive” the reactor. This EBR-II driver fuel is currently being treated for sodium neutralization in the Fuel Conditioning Facility (FCF) at the Materials and Fuels Complex (MFC) and the metallic uranium recovered as a component of this process has been proposed as a potential source of supply of high assay low enriched uranium. The irradiation of the metallic uranium fuel results in the creation of fission products and some minor actinides which are not compatible with many reuse scenario’s, and introduces significant challenges associated with fabrication of new fuel elements using this material. However, it is believed that the current EMT process can be slightly enhanced to address these concerns and result in the recovery of a product that can be considered as a source of supply for HALEU applications. Based on this, DOE requested the INL research methods to enhance the EMT process to produce a recovered uranium material that has a lower contaminant level and lower radiological dose rate and could be utilized in glovebox based fuel fabrication scenarios. The report that follows documents the progress toward meeting this goal.

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DOE OSTI2018

Special Analysis: Impact of Updated GSA Flow Model on E-Area Low-Level Waste Facility Groundwater Performance

In March 2018, the Savannah River National Laboratory (SRNL) notified Solid Waste Management (SWM) of new information that could potentially impact groundwater (GW) disposal limits and possibly require temporary protective measures (Crowley 2018). GW flow directions in the E-Area Low-Level Waste Facility (LLWF) Slit Trenches (STs), Engineered Trenches (ETs), and the Low Activity Waste Vault (LAWV) have notably changed in the new 2018 General Separations Area (GSA) flow model (Flach 2018a). The 2018 GSA flow model had been updated using 20 years of new hydrologic field data and model calibration methods that employed mathematical optimization software. It was later determined that the primary reason for the change in GW flow directions was due to the influence of low permeability caps that had been placed over the Old Burial Grounds and the Low Level Radioactive Waste Disposal Facility (LLRWDF). Updated flow directions in the model produce a higher degree of plume overlap for disposal units (DU’s) in the southeastern portion of E-Area than had been predicted in the 2008 E-Area LLWF Performance Assessment (PA). SWM and SRNL outlined the scope for this Special Analysis (SA) and proposed a set of interim measures (IM’s) to protect trench operations from exceeding Performance Objectives (PO’s) during the preparation of this SA, both of which were approved by the SWM Performance Assessment Review Committee (PARC) (Mooneyhan 2018).

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DOE OSTI2019

Selection of Glasses in Support of the Sludge Batch 9 Sulfate Solubility Limit for Coupled Operation with the Salt Waste Processing Facility

For Sludge Batches 3 through 8, the sulfate (SO 4 2- )solubility limit for the Defense Waste Processing Facility (DWPF) has been determined by laboratory-scale crucible testing. In preparation for Sludge Batch 9 (SB9) processing, a comparison demonstrated that minor compositional differences existed between the Sludge Batch 8 (SB8)-Frit 803 and SB9-Frit 803 glass composition regions for sludge-only and coupled operations with the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU). Thus, no experimental testing was necessary and the SB8 SO 4 2- limit of 0.65 wt.% was recommended for SB9 processing. This previous SB9 assessment did not address coupled processing with streams from the Salt Waste Processing Facility (SWPF), which is anticipated to start operations in November 2019. These high activity streams will include monosodium titanate (MST) and sludge solids from the Sludge Solids Receipt Tank (SSRT) as well as Cs-containing strip effluent (SE). The incorporation of these SWPF streams is expected to reach TiO 2 concentrations in glass greater than 2 wt.% based on single and double MST strike scenarios. In support of coupled operation with SWPF, the DWPF Product Composition Control System was revised and allows a TiO 2 concentration of up to 6 wt.% (minus measurement uncertainty) in glass. For the SB8-Frit 803 SO 4 2- solubility study, the maximum TiO 2 concentration evaluated was 0.94 wt.%. Since the influence of increased TiO 2 on the retention of SO 4 2- in the SB9 glass region is unknown, a test matrix of ten glass compositions has been proposed to determine whether the 0.65 wt.% SO 4 2- limit is still appropriate for SB9 coupled operation with SWPF. This report documents the development of a test matrix that will support this study.

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DOE OSTI2018

Effluent Management Facility Evaporator Bottoms: Waste Streams Formulation and Waste Form Qualification Testing

This report describes the results from grout formulation and cementitious waste form testing performed by Pacific Northwest National Laboratory for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates grout formulations proposed for encapsulating a wide range of compositions predicted to be present in evaporator bottoms wastes from the Hanford Effluent Management Facility (EMF). This work supports the technical development need for alternative treatment and disposition paths for the EMF evaporator bottoms waste and future direct-feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form testing. The work contained within this report relates to waste form development and testing, but does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains information useful for future PA updates [beyond fiscal year (FY) 2017] and future waste form development efforts. These analytical results can be used by (i) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (ii) decision makers interested in off-site waste form disposal, and (iii) the U.S. Department of Energy, their Hanford Site contractors, and stakeholders as they continue to assess the IDF PA program at the Hanford Site. The reported results help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottoms waste, and improve the technical defensibility of long-term waste form risk estimates.

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DOE OSTI2020

Influence of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities – SCC Testing

Glycolic acid is being evaluated as an alternate reductant in the preparation of High Level Waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to waste facilities. The SRS liquid waste contractor requested an assessment of the influence of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this concern had not been previously evaluated. The influence of glycolic acid on the occurrence of stress corrosion cracking (SCC) had not been evaluated specifically in the previous testing. The present testing used electrochemical and coupon immersion testing to assess this influence. Additionally, the primary guidance on corrosion for operating DWPF with a formate-based flowsheet is based on testing that utilized test solutions with a small number of constituents. Recent testing used more complex solutions based on flowsheet development for DWPF. Additional electrochemical tests were also conducted to assess the influence of this difference on test results.

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DOE OSTI2020

Default Radioactive and Hazardous Waste Management (RHWM) Dose Conversion Factors (DCFs) and Weighting Factors and plutonium-239 equivalent Curie (PE-Ci) methodology for acceptance into the Waste Storage Facilities

DOE-STD-5506-2007 (Ref. 1) is the standard that provides the methodology for preparing hazard analysis and accident analysis for transuranic (TRU) waste facilities. The Waste Storage Facilities hazard and accident analysis for the DSA is developed in accordance with DOE-STD- 5506-2007. Two related issues need to be addressed when implementing DOE-STD-5506-2007 regarding the calculation of consequences from TRU waste releases in accident analysis: Consequences determined in accident analysis must be modeled using inhalation Dose Conversion Factors (DCFs) that are consistent with those from the International Commission on Radiological Protection Publication 72 (ICRP-72, Ref. 2).; The use of a statistical Material at Risk (MAR) representing the distribution of radiological loading of TRU waste containers based on the use of a plutonium-239 equivalent Curie (PE-Ci) methodology. The PE-Ci methodology simplifies accident analysis by allowing one to model releases based on the Curie value of one standard nuclide (e.g., Pu-239) instead of needing to model releases for every potential nuclide of interest. This is accomplished by normalizing the Curie value of all nuclides to a single Curie value of a standard nuclide by using a Weighting Factor that is a ratio of the DCFs associated with the nuclides and that of the standard nuclide. By using the ICRP-72 DCFs to calculate PE-Ci values used in the statistical MAR, the accident analysis that uses that statistical MAR will be consistent with the consequence methodology identified in DOE-STD- 5506-2007.

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DOE OSTI2020

Modeling the Destruction of Glycolate in the Defense Waste Processing Facility (DWPF) Recycle Stream and Concentration Factors for Glycolate in the 2H Evaporator

Two models were developed to predict maximum glycolate concentrations in the Savannah River Site (SRS) Concentration, Storage, and Transfer Facility (CSTF) from implementation of the Nitric-Glycolic flowsheet at the Defense Waste Processing Facility (DWPF). One model describes the kinetics of glycolate destruction via chemical oxidation with sodium permanganate. This model conservatively predicts glycolate concentration delivered to the CSTF with a high probability the actual glycolate concentration is lower than predicted. The second model describes the potential concentration of said residual glycolate within the 242-16H (i.e., “2H”) Evaporator system.

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DOE OSTI2020

Sandia National Laboratories Annual Site Environmental Report, 2019: Tonopah Test Range, Nevada and Kaua'i Test Facility, Hawai'i

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the United States Department of Energy (DOE) National Nuclear Security Administration. The National Nuclear Security Administration’s Sandia Field Office administers the contract and oversees contractor operations at Sandia National Laboratories, Tonopah Test Range (SNL/TTR) in Nevada and Sandia National Laboratories, Kaua‘i Test Facility (SNL/KTF) in Hawai‘i. Activities at SNL/TTR are conducted in support of DOE weapons programs and have operated at the site since 1957. SNL/KTF has operated as a rocket preparation launching and tracking facility since 1962. DOE and its management and operating contractor are committed to safeguarding the environment, assessing sustainability practices, and ensuring the validity and accuracy of the monitoring data presented in this Annual Site Environmental Report. This report summarizes the environmental protection, restoration, and monitoring programs in place at SNL/TTR and SNL/KTF during calendar year 2019. Environmental topics include air quality, ecology, environmental restoration, oil storage, site sustainability, terrestrial surveillance, waste management, water quality, and implementation of the National Environmental Policy Act. This report is prepared in accordance with and as required by DOE O 231.1B, Admin Change 1, Environment, Safety, and Health Reporting, and has been approved for public distribution.

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