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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Sample Tank 202

An Extraction, Scrub, and Strip (ESS) test was performed on a sample from Tank-202. The purpose of testing this sample is to determine if the solvent is displaying the correct extraction and stripping behavior with cesium. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 14.2, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, the strip cesium distribution ratios fell within acceptable ranges.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Unlocking America's Abundant Marine Energy Resources

NREL Marine Energy One-Pager: A resource designed for use during NREL campus visits - especially with high-profile guests - as well as at events in Washington, D.C., when engaging with the new administration, and for displays at WPTO or lab booths. The U.S. holds vast untapped marine energy resources - wave, tidal, river, and ocean currents - that can strengthen grid resilience, support coastal communities, and advance energy independence. NREL leads innovation in this space through advanced modeling tools, patented technologies, and world-class testing facilities. NREL's simulation platforms, like OpenFAST and SAM, help reduce development time and risk. Patented devices such as PKelp and FlexWEC demonstrate flexible, resilient approaches to energy capture. At Flatirons Campus, NREL offers motion simulation, structural testing, wave tanks, and megawatt-scale microgrid emulation via the ARIES platform. These capabilities help developers refine and validate devices before in-water trials. With upcoming open-water testing at PacWave and strong partnerships, NREL is accelerating the path to commercialization - positioning the U.S. as a global leader in marine energy innovation.

17 WIND ENERGY↗
DOE OSTI2020

Investigation of Thermolytic Hydrogen Generation Rate in Tank 44 Dissolved Saltcake Samples

Saltcake core samples collected from Tank 44 in 2006 were dissolved to provide material for HGR measurements applicable to F-Area dissolved saltcake material. Additionally, characterization was performed on the Tank 44 saltcake material. The following are key results from the Tank 44 saltcake characterization. The Tank 44 Upper Saltcake Composite, corresponding to the 171 to 285 inch tank level, contained by mass approximately 69% sodium nitrate, 11% sodium carbonate, 8% sodium nitrite, smaller amounts of other salts and components, and 9% unquantified (which includes water, water of hydration, oxygen/hydrogen content of oxides and hydroxides, and uncertainty). The Tank 44 Lower Saltcake Composite, corresponding to the 76 to 114 inch tank level, contained by mass approximately 49% sodium carbonate, 18% sodium nitrate, smaller amounts of other salts, at least 8% sludge, and 9% unquantified (see above). The dissolved saltcake contained free hydroxide less than quantifiable (<0.01 M) due to the limited quantity of material that could be removed from the Shielded Cells based on the sample radioactivity. Measurement by pH paper provided an approximate pH of 12. The following are key results from the Tank 44 HGR testing. During boiling at 106.7 °C, HGR for Tank 44 dissolved saltcake without added glycolate was 7.2×10 -8 ft 3 h -1 ga l-1 . During boiling at 106.9 °C, HGR for Tank 44 dissolved saltcake with 1000 mg/L of added glycolate was 8.2×10 -8 ft 3 h -1 gal -1 . ∙ For the test without added glycolate, the first several HGR measurements at 70, 85, and 100 °C gave indication of the release of dissolved hydrogen and should not be used to represent the sustained thermolytic HGR for those temperatures. The measurements at boiling are the best representation of thermolysis in this testing. Carbon dioxide was observed at concentrations up to 6 vol% in the flow-system offgas for the test at boiling. ∙ Methane generation was observed at 100 °C and boiling. Methane concentration in the total gas generated during testing remained well below the lower flammability limit for methane in air. The addition of 1000 mg/L of glycolate did not have a significant impact on the hydrogen generation rates measured during this testing. The low hydroxide concentration in the Tank 44 dissolved saltcake likely influenced the relatively low thermolytic HGR and high carbon dioxide release observations in this testing. Based on the observation that methane was generated or released upon heating SRS radioactive Tank 44 waste samples to 100 °C and above, we recommend gaining a greater understanding of the cause and mechanism of its generation. First, the applicable literature should be reviewed to reveal the thermolytic methane generation mechanisms of possible methane generating species in the SRS CSTF. If warranted, a plan should be developed for simulant tests with methylated siloxanes and other applicable compounds in order to gain a better mechanistic understanding of methane generation in the SRS CSTF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Investigation of Thermolytic Hydrogen Generation Rate in Tank 44 Dissolved Saltcake Samples

Saltcake core samples collected from Tank 44 in 2006 were dissolved to provide material for HGR measurements applicable to F-Area dissolved saltcake material. Additionally, characterization was performed on the Tank 44 saltcake material. The following are key results from the Tank 44 saltcake characterization. The Tank 44 Upper Saltcake Composite, corresponding to the 171 to 285 inch tank level, contained by mass approximately 69% sodium nitrate, 11% sodium carbonate, 8% sodium nitrite, smaller amounts of other salts and components, and 9% unquantified (which includes water, water of hydration, oxygen/hydrogen content of oxides and hydroxides, and uncertainty). The Tank 44 Lower Saltcake Composite, corresponding to the 76 to 114 inch tank level, contained by mass approximately 49% sodium carbonate, 18% sodium nitrate, smaller amounts of other salts, at least 8% sludge, and 9% unquantified (see above). The dissolved saltcake contained free hydroxide less than quantifiable (<0.01 M) due to the limited quantity of material that could be removed from the Shielded Cells based on the sample radioactivity. Measurement by pH paper provided an approximate pH of 12. The following are key results from the Tank 44 HGR testing. During boiling at 106.7 °C, HGR for Tank 44 dissolved saltcake without added glycolate was 7.2×10 -8 ft 3 h -1 gal -1 . During boiling at 106.9 °C, HGR for Tank 44 dissolved saltcake with 1000 mg/L of added glycolate was 8.2×10 -8 ft 3 h -1 gal -1 . For the test without added glycolate, the first several HGR measurements at 70, 85, and 100 °C gave indication of the release of dissolved hydrogen and should not be used to represent the sustained thermolytic HGR for those temperatures. The measurements at boiling are the best representation of thermolysis in this testing. Carbon dioxide was observed at concentrations up to 6 vol% in the flow-system offgas for the test at boiling. Methane generation was observed at 100 °C and boiling. Methane concentration in the total gas generated during testing remained well below the lower flammability limit for methane in air. The addition of 1000 mg/L of glycolate did not have a significant impact on the hydrogen generation rates measured during this testing. The low hydroxide concentration in the Tank 44 dissolved saltcake likely influenced the relatively low thermolytic HGR and high carbon dioxide release observations in this testing. Based on the observation that methane was generated or released upon heating SRS radioactive Tank 44 waste samples to 100 °C and above, we recommend gaining a greater understanding of the cause and mechanism of its generation. First, the applicable literature should be reviewed to reveal the thermolytic methane generation mechanisms of possible methane generating species in the SRS CSTF. If warranted, a plan should be developed for simulant tests with methylated siloxanes and other applicable compounds in order to gain a better mechanistic understanding of methane generation in the SRS CSTF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Power capture and power take-off load of a self-balanced dual-flap oscillating surge wave energy converter

Wave energy converters are an important part of future renewable energy infrastructure. Predicting their power matrix, capture width ratio, and power take-off loads at a targeted site is required for performance assessment before deployment. Because their testing is very expensive, numerical modeling and simulations play a significant role in those assessments. Linear potential flow theory has limited accuracy under large amplitude wave forcing. More accurate predictions can be obtained by using higher-fidelity models, which are computationally expensive. We present a framework for multi-fidelity numerical simulations to determine the hydrodynamic response, wave capture capability, and power take-off load of a full-scale dual-flap oscillating surge wave energy converter. This design exploits out-of-phase motion by setting the distance between the flaps to half the wavelength of the most occurring wave. The simulations are validated using a 1:10 model experiments in a wave tank. Based on these validations, it was determined that Euler simulations provide an acceptable prediction with 90% reduction in computational time with only 11% error. Utilizing Euler simulations at full-scale, the results demonstrate that the annual electrical energy output is 1.79 GWh under regular wave conditions. Here, one significant improvement over single-flap designs is the capture width ratio which exceeds unity.

16 TIDAL AND WAVE POWER↗
DOE OSTI2020

Full-Scale Implementation of Propane Biosparge System for In-Situ Remediation of 1,4-Dioxane - 20174

1,4-Dioxane is a common co-contaminant with chlorinated solvents but is not readily remediated via similar treatment approaches (e.g., sorption, reductive dechlorination). However, 1,4-dioxane can be cometabolically biodegraded in the presence of alkane gases and oxygen. At Vandenberg Air Force Base in California, USA, historical use of chlorinated solvents resulted in 1,4-dioxane in groundwater across three vertical groundwater zones. Between April 2013 and December 2016, laboratory testing and various field demonstrations were conducted by us, and others. The results from these tests showed in-situ propane biosparging as a promising technology for reduction of both chlorinated solvent and 1,4-dioxane concentrations, with up to 99 percent reduction in groundwater concentrations. Stable isotope probing (SIP) was also used in 2015 to verify that biodegradation was a (destructive) mechanism occurring in the subsurface.). The success of the propane biosparge demonstrations, and confirmation of the biodegradation mechanism via SIP, has led to the full-scale implementation of a propane biosparge treatment system at Site 24 at Vandenberg Air Force Base. The treatment area is a relatively small footprint (e.g., less than 61 meters (200 feet) long), with 83 new wells installed simultaneously across the three groundwater zones. Ultimately 93 wells will be connected to an above-ground treatment system the includes an air compressor, a propane tank, controls to ensure safe operating conditions, and nutrient amendment elements. This full-scale system is currently known to be among the first of its kind. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Pacific Marine Energy Center Laser Doppler Velocimetry Commissioning

First commissioning data for the new laser doppler velocimetry (LDV) system that will be used at the Tyler Flume at the University of Washington. The LDV system can measure three components of velocity at a point. For this dataset the three components were operated in non-coincident mode and data were acquired at the center of the empty facility. Comparisons of freestream turbulence were made with a Vectrino slightly upstream of the LDV measurement location.

16 TIDAL AND WAVE POWER↗
DOE OSTI2020

Testing of Crystalline Silicotitanate to Support Tank-Side Cesium Removal System Operations - 20458

The direct feed of waste from the Hanford tank farms to the Low-Activity Waste Facility at the Hanford Waste Treatment and Immobilization Plant requires an intermediate treatment step that prepares the waste for vitrification. Washington River Protection Solutions (WRPS) selected a near-tank treatment system known as the Tank-Side Cesium Removal (TSCR) system to perform the required intermediate treatment functions. The approach in the TSCR system is to filter waste supernatant with a sintered metal dead-end filter and then use a series of ion exchange columns to remove cesium; this system has several similarities to the Tank Closure Cesium Removal system that has been deployed on the Savannah River Site. The ion exchange media proposed for use is crystalline silicotitanate (CST), which is a non-elutable media with a high affinity for cesium. Because the media is non-elutable, loaded columns will be blown down with compressed air, moved into interim storage, and replaced with new columns when cesium capacity is reached. The assumed extent and rate of drying that could be achieved in the TSCR column geometry lacked confirmation by experimental data. In addition, once the columns are loaded with cesium, they generate flammable gases (primarily hydrogen) via radiolysis acting on any resident liquids, i.e., moisture, remaining in the media bed. An assessment of available information on CST concluded that gas generation data that bounds expected operational conditions were needed to support the TSCR system safety basis and planned operations. Working with WRPS, Pacific Northwest National Laboratory (PNNL) designed and conducted testing with CST media to address the need for (a) representative in-column drying data, and (b) bounding flammable gas generation data. Drying testing was performed using a full-height column with a diameter of approximately 2 inches with injected air at temperatures of 18 deg. C and 30 deg. C. Gas generation testing was conducted in an engineered bunker that allowed simultaneous irradiation of up to eight samples at a time. These two experimental approaches used at PNNL for the CST testing are summarized and the major outcomes are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The simultaneous removal of technetium and iodine from Hanford tank waste

The simultaneous removal of radionuclides technetium-99 and iodine-129 from an actual decontaminated Hanford tank waste sample (a mixture of decontaminated waste from tanks 241-AP-105 and 241-AP-107) was demonstrated for the first time in this work. A series of commercially available ion exchange resins were evaluated in batch contact tests in the tank waste, and all showed removal of both Tc and I. The highest Tc removal was observed for Purolite A530e while the highest iodine removal was observed for ResinTech SIR-110-MP. Batch tests in simulated tank waste with these two resins showed that the SIR-110-HP-MP had consistently higher K d for both pertechnetate and iodide and much higher K d than previous works on Tc removal from Hanford waste. As such, the SIR-110-MP was evaluated in a dual -column (lead/lag) test processing 5.2L of the tank waste mixture showing 60% breakthrough of Tc on the lead column and no significant breakthrough on the lag after 625 bed volumes (BV, 6 mL size) while significant iodine breakthrough (>50%) occurred after 28 BV. The limited iodine uptake was attributed to the column conditions generating mass transfer limitations. A fraction of the Tc and I was not captured by the resin (<10%) in either the batch tests or column tests. The iodine fraction was identified to be an iodide, likely organo-iodide. The fraction of the Tc was identified as a non-pertechnetate species, which is the first time non-pertechnetate has been identified in AP-105 and AP-107 tanks, although the exact species is still unknown.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

CST R9120-B 30x60 Media Pretreatment and Digestion Standard Development to Support Tank Closure Cesium Removal 1A (TCCR-1A) Operations

In support of continued operations of Savannah River Remediation’s (SRR) ion exchange process to remove radioactive cesium from Savannah River Site (SRS) dissolved saltcake, referred to as the Tank Closure Cesium Removal 1A process (TCCR-1A; containing the second set of at-tank ion exchange columns), SRNL was tasked with: 1) validating the field protocol planned for pretreating the crystalline silicotitanate (CST) ion exchange media to be used in the columns, 2) preparing CST media for use in in-tank batch contact testing, and 3) developing a digestion standard to verify complete CST dissolution prior to characterization following each in-tank batch contact test. A procedure (SRNL L29 Manual, ITS-0229) was developed previously in support of the TCCR Demonstration (which contained the first set of ion exchange columns) based on evaluations of a different CST production lot with a larger particle size range to document the protocol for pretreating CST utilizing conditions similar to what will be performed in the field.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Summary of Analytical Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3 and Modeling Results for Cs Loading on CST

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. Some results from analysis of those samples have been previously reported. This report documents the remaining analyses of the in-tank batch contact samples as well as the results of ZAM (Zheng, Anthony, Miller) isotherm modeling performed for comparison to the measured results. Results of the additional analyses include analysis of the loading of other radionuclides besides 137 Cs on the crystalline silicotitanate (CST) contained within the in-tank batch contact test samples. Results from those analyses revealed the next highest contributor to the activity on the CST was 90 Sr with an average loading of 2.76E+08 dpm/g CST compared to 3.56E+10 dpm/g CST for the 137Cs. Isotopes of plutonium were also detected on the samples. ZAM modeling was performed using the measured composition of the Tank 10H Batch 3 qualification samples. The modeling predicted a maximum Cs loading approximately 2.2x higher than the measured result. This is a slightly lower ratio (expected/measured) compared to what was observed for the prior TCCR in-tank batch contact testing performed for Batches 1A and 2 where the ZAM results were 2.7-2.8x higher than the measured values.

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

Analysis of Respirator Cartridge Performance Testing on the 702-AZ Primary Exhauster for the Hanford AY/AZ Tank Farms

Washington River Protection Solutions (WRPS) conducted tests on two types of chemical cartridges for use in air purifying respirators to determine the period of time that the cartridges would provide adequate performance to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the 702-AZ primary exhauster for the Hanford AY/AZ tank farms. The Occupational Safety and Health Administration (OSHA) identifies cartridge testing as a valid approach for establishing cartridge change schedules. Testing is commonly applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Testing was conducted from August 26–28, 2016, on a slipstream from the 702-AZ exhauster, under static conditions fed to a respirator cartridge test stand developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose respirator cartridges, SCOTT 7422-SD1 and SCOTT 7422-SC1(SCOTT Safety, Monroe, North Carolina) were assessed on separate days. Sample media (sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridge, and were subsequently analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and change-out frequency.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Tank 50 Simulant Grout and Toxicity Characteristic Leaching Procedure (TCLP) Results for Methyl Mercury Waste Acceptance Criteria (WAC) Limit

Testing of simulated Tank 50 salt solution grout, spiked with methyl mercury (MeHg) and ionic mercury, used the Toxicity Characteristic Leaching Procedure (TCLP) to evaluate leaching performance with respect to Hg and the Waste Acceptance Criteria (WAC) limits for the Saltstone Production Facility (SPF). This report describes the test parameters for simulant preparation and confirmatory analysis, grout preparation, TCLP leach testing and subsequent leachate analyses. These scoping tests were designed to provide an initial evaluation of the effect of increased MeHg and ionic Hg on the TCLP and were requested by Savannah River Remediation LLC (SRR) through a Technical Assistance Request (TAR). It is expected that this preliminary testing will likely lead to future, more detailed, testing involving the mercury (Hg) WAC and Hg limits with respect to the Tank 50 TCLP for mercury. The TCLP concentrations for total Hg have approached 0.06 mg/L in recent years for Tank 50 radioactive solutions containing up to 105 mg/L of Hg in the form of ionic (Hg(I/II), elemental (Hg(0)) and organo-mercury such as MeHg. In discussions with SRR personnel, it was decided for this non-baseline work to target Hg spike concentrations, using predominately MeHg, in the range of 150 to 350 mg/L total Hg to investigate corresponding Hg TCLP leaching behaviors.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

TEAMER: Pitching Foil Crossflow Turbine Efficiency Data

This dataset documents the efficiency testing of a pitching foil crossflow turbine, conducted at the University of New Hampshire's (UNH) Chase Ocean Engineering Laboratory tow tank facility. The tests explored various pitch phases and amplitudes, ranging from 0 to 18 degrees, across different flow speeds and turbine RPMs. Specifically, testing was performed at pitch amplitudes of 0,3,6,9,12,15 and 18 degrees. The results indicate that turbine efficiency improves with increased pitch amplitude, reaching optimal performance at 12 degrees. Modifications were made to the test frame by UNH to enhance the stiffness of the support structure. The testing specifically mapped the efficiency of a single quadrant of the Ocean Renewable Power Company (ORPC) Autonomous Turbine Generator Unit (ATGU), covering the full range of allowable pitching amplitudes and phases. The dataset provides both raw and processed data, including drag and torque measurements, and contains Python scripts used for data processing and visualization. This research was supported by funding from TEAMER RFTS 1 (Request for Technical Support).

16 TIDAL AND WAVE POWER↗
DOE OSTI2021

End-of-Life Evaluation of Compressed Natural Gas Vehicle Fuel Tanks

This study evaluated the structural integrity of Type III and Type IV CNG fuel tanks from the Los Angeles (LA) County Metropolitan Transportation Authority to characterize the fuel tank conditions after experiencing a full service life of 15 years in transit bus application. The data produced provide insight about the condition of the CNG fuel tanks at the conclusion of their defined EOL and potential risks of continued operation of the CNG vehicle without replacement of the expired fuel tanks. In addition to physical testing performed on the fuel tanks, a nondestructive evaluation of modal acoustic emission (MAE) was utilized to assess the structural integrity of the tanks. The CNG fuel tanks were evaluated in their received condition and after experiencing artificial damage. The MAE results were then compared to visual inspection results of the procedures defined in the Compressed Gas Association’s (CGA’s) C-6.2 and C-6.4 standards (Compressed Gas Association 2013; 2012) to better understand the effectiveness of the visual inspection and potential safety risks of continued use of the tanks.

03 NATURAL GAS↗
DOE OSTI2021

Residual Insoluble Solids Expected from Saltcake Dissolution

A review of previous SRS saltcake dissolution and characterization studies was performed to inform the analysis of the amount of residual insoluble solids that may result from saltcake dissolution. A realistic value covering most cases for the maximum volume of a settled layer of residual insoluble solids that would result from dissolution of a volume of saltcake was calculated as 0.08 mL/mL (8 vol%). Applying factors to account for dissolution test uncertainty and differences between in-tank settling and laboratory testing, the recommended value for a settled layer of residual insoluble solids that would result from dissolution of a volume of saltcake is 0.16 mL/mL (16 vol%). This should apply to dissolution of saltcake with typical minimal amounts of entrained sludge but may not apply to saltcake with significant sludge. As it is composed of primarily non-radioactive isotopes, the precipitated aluminum hydroxide material does not contribute to the heat load.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Performance of Oak Ridge National Laboratory spallation neutron source proton power upgrade cavities and cryomodule production

The Proton Power Upgrade initiative at Oak Ridge National Lab’s Spallation Neutron Source aims to greatly boost the capability of proton beam power. This upgrade involves the integration of seven additional cryomodules, each housing four six-cell high-beta ( β = 0.81 ) superconducting radio frequency cavities. These cavities, manufactured and processed by Research Instruments in Germany, underwent meticulous treatment, including electropolishing as the bulk and final vital step. Upon delivery to Jefferson Lab, a total of 28 cavities for seven cryomodules, along with an additional four cavities for a spare cryomodule, underwent thorough vertical qualification tests, meeting the required specifications. Following cavity tanking and subsequent rf testing, the assembly of eight cryomodules was successfully executed, with all 32 cavities demonstrating compliance with acceptance criteria. The transportation of the cryomodule to SNS for high-power testing in the tunnel yielded exceptional results. Notably, the performance surpassed specifications, in terms of quality factor and accelerating gradient. Published by the American Physical Society 2024

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Analysis of Respirator Cartridge Performance Testing on the 702-AZ Primary Exhauster for the Hanford AY/AZ Tank Farms during a Waste-Disturbing Event

Washington River Protection Solutions (WRPS) conducted tests using two types of chemical cartridges for use in air-purifying respirators (APR) to determine the period of time that the cartridges would provide adequate performance1 for APRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the 702-AZ Primary Exhauster for the Hanford AY-AZ tank farms. Unlike prior cartridge testing on the 702-AZ Primary Exhauster, the recent tests were performed during a waste-disturbing event. The Occupational Safety and Health Administration (OSHA) identifies cartridge testing as a valid approach for establishing a cartridge change schedules. Testing is commonly applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate the cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Testing was conducted from February 10–11, 2017, on a slipstream from the 702-AZ exhauster fed to a respirator cartridge test stand developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose respirator cartridges, SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina), were assessed on separate days. Sample media (sorbent tubes) and canisters (e.g., Summa) sampling were used to collect samples of the vapor stream entering and exiting the respirator cartridge and were subsequently analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗