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At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Solvent Hold Tank Sample Results for MCU-20-32-33-34 (Quarterly Report, September 2020)

In late FY13, MCU implemented the Next Generation Solvent (NGS) flow sheet. Facility personnel added a non-radioactive, NGS “cocktail” containing the new Extractant (MaxCalix) and a new Suppressor (TiDG) to the SHT heel to implement the NGS flow sheet. The resulting “blend” solvent (“NGS blend solvent”) is essentially NGS with residual amounts of calix[4]arene-bis(tert-octylbenzo-crown-6) (BOBCalixC6) and trioctylamine (TOA). For process monitoring, SHT samples are sent to Savannah River National Laboratory (SRNL) to examine solvent composition changes over time. With the exception of Isopar™ L which is regularly added to the SHT due to its high vapor pressure, this report shows the cumulative chemical composition data, including impurities like mercury, of the September 2020 (MCU-20-32-33-34) solvent SHT sample. A summary report of the September SHT solvent sample was issued earlier. This report examines the cumulative results from these and several past monthly reports. These samples are intended to verify that the solvent is within the specified composition range. A baseline “scratch” solvent - a scratch solvent is a preparation of all 6 solvent components (this includes the old extractant BoBCalix and the old suppressor TrioctylAmine that remained when solvent was converted to Next Generation Solvent [NGS]) of the composition that approximates the blend of cocktail and heel solvent - was prepared in the lab (September 2018) and used for comparison and evaluation. The results from the analyses are presented in this document. This report is the last characterization analysis of the SHT solvent. MCU entered a lay-up state and about 128 gallons of the SHT solvent were pumped into 55 gallons drums for future disposal.

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DOE OSTI2021

Cesium Batch Contact Equilibrium Testing of Crystalline Silicotitanate (CST) Sieve Cuts in SRS Average Simulant and Examination of CST Samples Aged in Caustic and Simulant in Support of Tank Closure Cesium Removal 1A (TCCR-1A)

Batch contact testing to determine cesium equilibrium loading on Crystalline Silicotitanate ion exchange media in Savannah River Site Average Waste Simulant at 25 °C indicated that smaller diameter particles isolated by sieving pretreated CST media may load slightly higher (<10%) amounts of cesium, though the differences are within analytical uncertainty. In addition, ion exchange media sub-samples stored in 2-4 M NaOH and caustic simulant solutions for ~2.5 years were examined by optical microscopy and the 3 M NaOH sample was also analyzed to determine whether changes in the particle size distribution occurred during storage. No visual indications of particle attrition or agglomeration were observed for any sample. Particle size analysis indicated that a slight decrease occurred in the average particle diameter following contact with 3 M NaOH (541 µm average diameter versus 566 µm for the pretreated CST prior to contact). A small increase (from 0 to <0.5 wt. %) in the number of particles ranging from 271 and 322 µm was also observed for the CST sample contacted with 3 M NaOH relative to a sample of the original pretreated material. However, this small change could be due to sub-sampling differences or analytical uncertainty. It does not appear that small particles are formed to a significant degree during CST caustic contact or that small particles which do form (presumably from attrition of larger particles during pretreatment) load significantly more cesium than the bulk material. Minimal other negative consequences were observed associated with CST extended caustic or simulant contact, except for the tendency for more concentrated salt solutions to form some salt crystals which deposit on the media over time.

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DOE OSTI2022

Characterization Results for the March 2022 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the March 2022 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed on a single sample replicate. A summary of the analytical results for the 3H evaporator overhead sample includes the following. The measured cesium-137 activity in the 3H evaporator overhead sample is 4.59E+02 dpm/mL which is above the ETP WAC limit of 3.28E+02 dpm/mL. The strontium-90 activity in the 3H evaporator overhead sample is <2.95E+01 dpm/mL. This is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample is <1.92E-02 dpm/mL. This is below the ETP WAC limit of 1.00E+00 dpm/mL.

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DOE OSTI2022

Characterization Results for the April 2022 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the April 2022 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC.

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DOE OSTI2023

Trinity 78th Anniversary: re-discovered and newly edited unclassified footage shows use of tanks to collect radioactive samples immediately after Trinity test (Rev. 1)

On July 16, 1945, at 5:29 a.m., a team of Los Alamos scientists successfully detonated the world’s first atomic bomb at a remote section of south-central New Mexico desert, then known as the Alamogordo Bombing Range. This full-scale test, dubbed “Trinity” by the Lab’s first Director J. Robert Oppenheimer, proved the viability of the implosion-type bomb known as The Gadget and ushered in the Atomic Age. Three weeks after Trinity, two Los-Alamos created atomic weapons – Little Boy and Fat Man – were released above Japan, helping to end World War II and changing the course of history forever.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Characterization Results for the March 2023 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the March 2023 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC.

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DOE OSTI2023

Analysis of H-Canyon Process Tanks in Preparation of Consolidation and Blending for HALEU Fuels

High-Assay Low- Enriched Uranium (HALEU) fuels are being developed to support the replacement of Highly Enriched Uranium (HEU) fuels used in U.S. High-Performance Research Reactors (USHPRR) as well as advanced nuclear power reactor designs. The projected demand for HALEU far exceeds the supply and studies are underway to assess various options to partially mitigate the potential short supply. The H Canyon facility at the Savannah River Site (SRS) Low Enriched Uranium (LEU) containing 4.95% U-235 from the reprocessing of highly enriched foreign and domestic research reactor fuel for the Tennessee Valley Authority’s (TVA) commercial power reactor market for several decades. The production of LEU at the H-Canyon facility can be readily transitioned to produce 19.75% HALEU solutions from the current separated inventory of purified HEU solutions in H-Canyon storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Characterization Results for the September 2023 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents the average characterization results for the September 2023 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed in duplicate, and all three measured radionuclide concentrations were within ETP WAC limits. A summary of the analytical results for this 3H evaporator overhead sample includes the following: The measured cesium-137 activity in the 3H evaporator overhead sample averaged 5.70E+01 dpm/mL, which is below the ETP WAC limit of 1.30E+03 dpm/mL. As seen through historical results, Cs-137 continues to present a decreasing activity since 2020. The strontium-90 activity in the 3H evaporator overhead sample averaged <2.16E00 dpm/mL, which is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample averaged <9.08E-02 dpm/mL, which is below the ETP WAC limit of 1.00E+00 dpm/mL

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DOE OSTI2024

Zirconium Sludge Criticality Calculations in Large Process Tanks

The Savannah River Site’s H-Canyon facility has been tasked with accelerating the disposition of used research reactor fuels. • Future missions will involve dissolution of “Non-Aluminum” clad fuels – not easily dissolved – Nitric acid is not capable of dissolving stainless steel, Zirconium, and Hastelloy, clad fuels by itself • The electrolytic dissolver has been (re) selected as a disposition path for these fuels. 2

Devine, Nathan P.↗
DOE OSTI2024

Sampling and Analysis Plan for LSL2 Underground Storage Tank Evaluation of PFAS in Soil: Protection of Groundwater

This SAP gathers all pertinent information relevant to the sampling and analysis of soil at the removal site for a former 10,000-gallon UST and reference sites in the vicinity of the excavation site. Data Quality Objectives establishes the process and steps for the acquisition of data. Data Flow Diagram establishes the decisions and next steps. Quality Assurance Project Plan establishes the quality requirements for data collection, including planning, implementation, and assessment of sampling, field measurements, and laboratory analysis. The work is being completed for DOE-SC Pacific Northwest Site Office.

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DOE OSTI2024

Nitrate and Nitrite at Hanford – From Tanks to Natural Attenuation

Washington River Protection Solutions, LLC (WRPS) contracted Pacific Northwest National Laboratory (PNNL) in support of their maturation testing and analysis to develop and deploy an ambient temperature solidification/stabilization process for low-activity waste (LAW) (referred to as “grout” in this report). The intent of this effort is to support a sample-and-send regulatory and processing strategy in the Hanford 200 West Area, specifically considering grout waste forms disposal at Hanford in the Integrated Disposal Facility (IDF). Currently the disposal of grouted LAW is not an option because the IDF is not permitted to receive this waste form. The material presented in this report is intended to help inform the U.S. Department of Energy (DOE), the site operating contractors, regulatory agencies, and stakeholders of the future implications of IDF disposal of a grouted LAW waste form, specifically for the behavior of nitrate (NO 3 - ) and nitrite (NO 2 - ) at the Hanford. The only calculation of nitrate release from a hypothetical grouted LAW inventory in the IDF showed that the overall release was slightly above the compliance limit of 45 mg/L nitrate and 3.32 mg/L nitrate after 1000 years (Asmussen et al. 2019). As such, uncertainty around the impact of nitrate and nitrite release from the IDF was identified in the recent National Academies of Sciences study of Hanford supplemental LAW by both the national laboratory team and Hanford stakeholders (Bates et al. 2023). However, these calculations (and others for the IDF involving secondary waste grout) may be conservative as no nitrate and nitrite retention in the grout waste form was assumed, nor any attenuation in the subsurface. There is evidence from the literature of processes that can attenuate nitrate migration both in the waste form and in the Hanford subsurface. If found to be technically defensible and likely to occur in the IDF and associated subsurface, then these processes should be included in modeling of the IDF to represent the behavior of nitrate and nitrite more accurately. Doing so would remove unnecessary conservatism in the modeling projections and reduce the uncertainty in assessments of facility compliance of a grouted LAW waste form in the IDF.

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