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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Exploring Sustainability in Scientific Software through Code Quality & Test Coverage Metrics

Context: Scientific open-source software (SciOSS) plays a foundational role in research and engineering, yet its long-term sustainability has often been overlooked and remains a significant concern. Objective: This study investigates the long-term sustainability of SciOSS through code and test quality metrics. Method: We analyze CASS Software Portfolio projects, classifying them by sustainability and comparing their code structure, test coverage, and links between code quality and testing across the dataset. Results: Sustainable projects show higher, more consistent test coverage and clearer code-test correlations, while unsustainable ones show weaker patterns. Overall, test coverage is low in scientific software, and high complexity and coupling reduce testability. Conclusion: In this study, we present a practical, data-driven approach for assessing sustainability in scientific software, offering a foundation for evaluating long-term software health and supporting future efforts in quality assurance and sustainability monitoring.

Md mushfiqur rahman, Sheikh [University of Tenness↗
DOE OSTI2023

Quantification and prediction of error in MIMO tests

Dynamic testing of systems is most realistic of real-world conditions when multiple input and multiple output (MIMO) techniques are used. To replicate measured environmental conditions, a series of desired outputs (responses) on the system must be realized by inverting the frequency response functions (FRF) matrix for input estimation. System identification is dependent on the number of inputs and outputs, which affect the type of solutions. In input estimation, the auto spectra of the determined inputs are affected by both the auto spectra and cross spectra of the desired outputs. This research evaluates errors and investigates two sources in an experimental setting. The first significant error source is test-to-test variability in FRFs from system identification tests, quantified by magnitude variability from long-duration tests. The second significant error source is the realization of time histories from frequency domain inputs. The spectral content of the realized random process deviates from the desired content and is also quantified. The structure is a linear three-story frame fixed at the base; two inputs are provided uniaxially with the use of suspended electrodynamic shakers. The proposed model for error sources was found to be effective in predicting errors in a SISO (single-input single-output) and a square (2-input 2-output) MIMO test.

Yousef, Odey↗
DOE OSTI2022

High throughput embryonic zebrafish test with automated dechorionation to evaluate nanomaterial toxicity

Engineered nanomaterials pose occupational health and environmental concerns as they possess unique physical and chemical properties that can contribute to toxicity. High throughput toxicity screening methods are needed to address the increasing number of nanomaterials in production. Here we used a zebrafish photomotor response (PMR) test to evaluate a set of fifteen nanomaterials with military relevance. Automated dechorionation of zebrafish embryos was used to enhance nanomaterials bioavailability. Optimal PMR activity in zebrafish embryos was found at 30–31 hours post-fertilization (hpf). Behavioral and toxicological responses were measured at 30 and 120 hpf; behavioral responses were found for thirteen of the fifteen nanomaterials and acute toxicity (LC50) levels for nine of the fifteen nanomaterials below the maximum test concentration of 500 μg/ml. Physico-chemical characterization of the nanomaterials detected endotoxin and bacterial contamination in two of the tested samples, which may have contributed to observed toxicity and reinforces the need for physical and chemical characterization of nanomaterials use in toxicity testing. The zebrafish PMR test, together with automated dechorionation, provides an initial rapid assessment of the behavioral effects and toxicity of engineered nanomaterials that can be followed up by physico-chemical characterization if toxicity is detected, reducing the amount of time and monetary constraints of physico-chemical testing.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Tundra SGS – Critical Importance of Site-Specific Injectivity Testing for Carbon Dioxide Injection Well Simulation Calibration

Paper presented at 16th International Conference on Greenhouse Gas Control Technologies (GHGT-16), October 23–27, 2022, Lyon, France. Brine injectivity tests were conducted on both the Broom Creek (BC) and Deadwood (DW) storage formations in one of the cored stratigraphic test wells prior to finalized simulations. The injectivity tests entailed injection of brine into the formations, followed by shutting in the wells. The shut-in duration represents the fall-off period which was used to carry out pressure transient analyses of the tests. The analysis of injectivity tests indicated that core-indicated and modeled permeability distributions should be conservatively multiplied by 2.5× and 5× in the BC and DW geologic models, respectively, prior to numerical simulations of CO 2 injection using Computer Modelling Group’s (CMG’s) GEM equation-of-state reservoir simulator. The scaler permeability multipliers allowed the fluid flow simulations to fully represent well test-measured injection well rates in the respective formations.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2020

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

14 MeV DT Neutron Test Facility at the Sandia Ion Beam Laboratory

A recently completed LDRD project has provided a new facility at Sandia for testing effects of energetic neutrons on electronic components. 14 MeV neutrons are produced with a deuterium ion beam onto a thin-film tritide target. The goal of the project was to increase the neutron fluence to levels needed for radiation effects testing and qualification. This goal was achieved through two technical advances. First, a new multi-layer target concept was developed to reduce the rate of tritium loss from the target by isotope exchange, thereby reducing tritium usage and increasing target lifetime. The second advance was the construction of a new test chamber designed to maximize neutron flux at test locations. Together, these increased the available neutron fluence by several orders of magnitude. This new capability is being used in tests for Sandia nuclear weapon programs, evaluation of commercial parts such as highly-scaled CMOS SRAM lCs, and tests of new devices under development at Sandia such as lll-V HBTs, gallium nitride high-voltage diodes, and for fundamental studies of physical mechanisms of device failure.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Sierra/SD - Verification Test Manual - 4.56

This document presents tests from the Sierra Structural Mechanics verification test suite. Each of these tests is run nightly with the Sierra/SD code suite and the results of the test checked versus the correct analytic result. For each of the tests presented in this document the test setup, derivation of the analytic solution, and comparison of the Sierra/SD code results to the analytic solution is provided. This document can be used to confirm that a given code capability is verified or referenced as a compilation of example problems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Swab Tensile Testing Results and Procedures - PA11 Rev. 10 Swabs

This tensile test is to mimic the worst-case scenario of the swab being caught on an obstruction when being pulled out of the nasopharyngeal space. It is to determine how much tensile force the swab can withstand without breaking. All swabs were provided by HP and were the PA11 material and Rev 10 design. All of the swabs were tested using the protocol outlined in Section 2 (Tensile Testing Procedure). Not all of the swabs were pre-treated using the Autoclave. For those swabs that were pre-treated using the Autoclave, the tensile tests were performed within 6 hours of completion of the Autoclave pre-treatment, with the exception of the “Aged” swabs. For the “Aged” swabs, the tensile tests were performed 7 days after the Autoclave pre -treatment. The “Aged” swabs that were not pre-treated using the Autoclave, were tested 7 days after those that were not “Aged.”

36 MATERIALS SCIENCE↗
DOE OSTI2020

Markforged Continuous Fiber Composite Material Testing

Continuous fiber composite 3D printing promises to greatly expand the design space of polymer additive manufacturing (AM). A series of mechanical tests were performed to attempt to fully characterize the composite materials built from feedstock provided by the Markforged company. These tests included tensile, compression, Charpy impact, three point bending, coefficient of thermal expansion, and calorimetry tests. Specimens were prepared on the Mark Two 3D printer at LANL during the spring and summer of 2019. These samples were then shipped to New Mexico Institute of Mining and Technology’s Thermo-Mechanical Lab for testing during the summer and fall of 2019. Test data are reported and analyzed for each of the tests.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Sodium Exposure Tests of A709 Steel (FY2020 Progress Report)

Sodium exposure tests were conducted on the first commercial heat, H58776 of A709 steel. Six different processing and heat treatment conditions were investigated, including three processing conditions, namely, Argon-Oxygen-Decarburization (AOD), Electro-slag Remelting (ESR), and Electro-slag Remelting with subsequent homogenization (HOMO), and two heat treatment conditions, i.e. solution annealing at 1100°C (SA1100°C) or 1150°C (SA1150°C). Specimens of AOD-SA1100°C, AOD-SA1150°C, ESR-SA1100°C, ESR-SA1150°C, HOMO-SA1100°C, and HOMO-SA1150°C were tested in sodium at 650°C and for times up to 10,152 h. Specimens in the AOD-SA1100°C and AOD-SA1150°C conditions were also tested in sodium at 550 and 600°C. Parallel thermal aging experiments were carried out for comparison with sodium exposure tests to separate the thermal and sodium effects. It was found that A709 H58776 generally showed a weight loss after sodium exposures at 650°C except a few data points of the ESR-SA1100°C specimens. The corrosion rate showed large variations in the initial stage and decreased with increasing exposure time. Long-term exposure tests are needed to determine its steady-state corrosion rate. The AOD-SA1100°C and HOMO-SA1100°C specimens showed similar tensile responses after thermal or sodium exposures at 650°C. The yield stress increased, the ultimate tensile strength decreased, and the uniform and total elongations were reduced after thermal or sodium exposures. A longer exposure time had a minimal effect on the tensile strength but further reduced the tensile ductility. The sodium-exposed specimens had higher tensile strength than the thermally-aged specimens. In contrast, the ESR-SA1100°C specimens showed an increase in the yield stress and the ultimate tensile strength after thermal or sodium exposures at 650°C and ~10,000 h. The AOD-SA1150°C and ESR-SA1150°C specimens showed similar tensile behavior after thermal or sodium exposures at 650°C.Thermal or sodium exposures at 650°C increased the yield stress, decreased the ultimate tensile strength, and reduced uniform and total elongations. There were minimal differences in tensile properties between the sodium-exposed and thermally-aged specimens after an ~3000 h exposure. After an exposure for ~10,000 h the sodium-exposed specimens showed higher yield stress and ultimate tensile strength and lower uniform and total elongations than the thermally-aged specimens. For the HOMO-SA1150°C specimens, thermal or sodium exposures at 650°C significantly increased the yield stress but had a minimal effect on the ultimate tensile strength. Both uniform and total elongations were significantly reduced at an exposure time of ~3,000 h, and they were further reduced after an exposure of ~10,000 h. The sodium-exposed specimens had lower tensile ductility than the thermally-aged specimens. Before exposure, the stress-strain curves of the specimens in all six processing-heat treatment conditions showed serrations indicating the dynamic aging effect at 650°C. Thermal aging or sodium exposures at 650°C after ~10,000 h completely removed the serrations of the flow curves in A709 H58776, indicating strong precipitation during thermal or sodium exposures. Longer-term sodium exposure tests are needed to determine the steady-state corrosion rates in sodium for A709 H58776 and fully understand the influence of sodium exposures on its mechanical performance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Transient Testing of Nuclear Fuels Performed in the Original Operation of TREAT (2nd ed.)

This report provides summary information on experiments performed in the Transient Reactor Test Facility (TREAT) during its original operation from 1959-1994, and more specifically, before the TREAT core was upgraded at the end of the 1980s. The tests performed since about 1970 are summarized in considerable detail. Earlier tests are noted by their designated series and by some parameters that help to identify and distinguish each series from the others. The detailed summary descriptions are intended to help inform future researchers regarding the large, historical, empirical data base generated from that 35-year span during which hundreds of TREAT experiments were performed to investigate and demonstrate the transient behavior of a wide variety of nuclear fuels during severe off-normal and accident conditions -- using direct nuclear heating of the test samples. That empirical basis is foundational and valuable for guiding future qualification and safety evaluation of improved fuel designs and for future transient test planning and performance in TREAT. Since Edition 1 of this report, descriptions of the following test series have been added: D, E, EOS, F, H, J, R, RX, and S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Surrogate Assembly 30 cm Drop Test

The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. While obtaining data on the actual fuel is not a direct requirement, it provides definitive information which aids in quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. The 30 cm drop test with the full-scale surrogate assembly conducted in May 2020 was the last step needed for quantifying the strains on the surrogate assembly rods under NCT. The full-scale surrogate assembly used in the 2020 30 cm drop test was built using a new 17x17 Pressurized Water Reactor (PWR) Westinghouse skeleton filled with the copper rods and 3 zircaloy rods from the full-scale surrogate assembly used in the Multi-Modal Transportation Test (MMTT). Felt pads were attached to the surrogate assembly bottom prior to the 30 cm drop to adequately represent the effects of the impact limiters and the cask. Note that felt "programming material" has been used extensively in past drop tests and is known to be a good material for programming a desired shock pulse. The felt pad configuration was determined during a previous series of tests reported in. The acceleration pulses observed on the surrogate assembly during the test were in good agreement with the expected pulses. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would if it was dropped in the cask with the impact limiters.

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

Report on FY 2020 creep, fatigue and creep fatigue testing of Alloy 709 base metal at ORNL

The testing activities and research in support of ASME Code qualification of Alloy 709, an advanced austenitic steel, are being carried out at Oak Ridge National Laboratory (ORNL), Argonne National Laboratory, and Idaho National Laboratory. This report summarizes the status and results of FY 2020 planned testing at ORNL. Uniaxial tensile tests on the electroslag remelt (ESR) solution-annealed plate (heat number 58776-3RBC) with additional heat treatment were performed for the baseline mechanical properties evaluation for ASME code qualification of Alloy 709. The tensile properties were found to meet ASME SA-213 specifications and were comparable to the Nippon Steel NF709 data generated for an ASME Section I Code Case for seamless tubing. ORNL was tasked to carry out a subset of the Code Case testing for creep rupture. Creep rupture data from 18 tests were generated on a solution-annealed ESR plate with a solution annealing temperature of 1100°C. There are total of 19 intermediate and long-term creep rupture tests of Alloy 709 ongoing at ORNL. A preliminary fatigue design curve at 760°C was developed for Alloy 709, and the results show that the fatigue design curve of Alloy 709 is comparable to that of Alloy 800H at 760°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sensitive Resources Assessment and Forest Analysis for the Proposed Versatile Test Reactor, Oak Ridge, Tennessee

The US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) is a leading institution in advanced materials, supercomputing, neutrons, and nuclear science. As a research laboratory managed by UT-Battelle, LLC for DOE, ORNL has national priorities in energy, security, and scientific discovery that necessitate facility improvements and expansions. DOE is also committed to environmental stewardship. The laboratory is located on the ~32,000-acre (~13,000-ha) Oak Ridge Reservation (ORR), much of which is categorized as a National Environmental Research Park (NERP) and a state Wildlife Management Area. DOE works with the Tennessee Wildlife Resources Agency (TWRA), Tennessee Department of Environment and Conservation (TDEC), US Fish and Wildlife Service (USFWS), US Department of Agriculture, and other agencies to serve as an effective steward of the ORR. Accordingly, project managers must conform to environmental regulations, agreements, and policies at the federal, state, and institutional levels. Per 40 CFR (Code of Federal Regulations) 1508.14, potential effects on research and science education also represent potential effects of federal actions on the NERP, and impacts on, e.g., deer harvest, must be considered on the Oak Ridge Wildlife Management Area when other aspects of the human environment are affected. The United States currently has no fast neutron testing capability to support advanced nuclear research and development. The proposed Versatile Test Reactor (VTR) will take advantage of current investments by the US government and private industry in nuclear reactors to expedite the design and construction process, using proven technology to create a world-class scientific infrastructure. The VTR will take advantage of fast neutrons provided by this proven technology, along with a capability to rapidly insert, conduct, and remove state-of-the-art experiments. An advantage of the VTR is that it can support future innovations in experimental capabilities without modifying the facility. The VTR will support progress in a variety of science and technology areas, including testing and qualification of advanced reactor fuels; testing and qualification of innovative structural materials; testing of innovative components and instruments; validation of advanced modeling and simulation tools; and versatility for future technical missions. Through proven technology, the VTR can take advantage of existing reactor designs and operating experience to reduce the risk, cost, and time for design and construction. The top available resources of DOE laboratories, industry, and universities will be used to expedite reactor design and construction toward developing the scientific infrastructure that affords a strong testing capability that can be sustained over many years. This report summarizes current knowledge of natural and cultural resources primarily within the VTR construction area. At the time of this report, the proposed VTR site design includes a construction area of ~150.4 acres (~69.9 ha), which contains an ~51.3-acre (~20.8-ha) operations area, located within forested natural areas of the ORR. The primary goal of the work presented here was to evaluate potential effects on sensitive resources that might result from development and construction activities associated with VTR. In addition to on-the-ground surveys during spring and summer 2020 by the ORNL Natural Resources Management Program and Aquatic Ecology Group staff, this report makes use of historical (pre-1995) and contemporary (1995 to present) data from additional confirmed sources (e.g., TDEC). Likewise, forest conditions were compiled from a 2011 forest inventory and supplemented with limited ground observations in 2020. The individuals who obtained and compiled the data presented here are familiar with and routinely assess sensitive resources on the ORR. Anyone who references this report must consider that the timing of surveys did not permit a complete delineation of the resources that will be affected. If the VTR project proceeds, additional surveys will be required to account for the seasonal patterns of various threatened and endangered species. Data deficiencies and potential resources that likely went undetected are indicated where possible. Accordingly, this report should facilitate more environmentally sound decisions during planning and 2 development of the VTR site, provide a foundation for further assessment of sensitive and cultural resources, and help project managers better address regulatory guidance and DOE policies on sustainable development in compliance with, for example, the US Endangered Species Act (ESA), Migratory Bird Treaty Act (MBTA), Tennessee Rare Plant Protection and Conservation Act of 1985, Tennessee Nongame and Endangered or Threatened Wildlife Species Conservation Act of 1974, several federal and state regulations regarding aquatic resource protection, and site-specific policies as outlined in various ORR management plans developed by ORNL and TWRA for DOE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Coupon Container Pressure Test

A total of six containers were tested to verify the Maximum Allowed Working Pressure (MAWP) rating of 100 psi. Three Inner Container Pressure Test Assemblies (1006759387-AA) and three Outer Container Pressure Test Assemblies (1006729168-AA) were tested. The container serial numbering is the drawing number followed by the serial number (i.e.,1006759387-###). The test procedure was performed as follows: 1. Pressurize to 50 psi and hold for 5 minutes. 2. Increase pressure to 100 psi and hold for 5 minutes. 3. Increase pressure to 150 psi and hold for a minimum of 20 minutes. A container is considered to pass if the pressure drops less than 1% at the end of the 20 minutes. All six containers passed the test.

42 ENGINEERING↗
DOE OSTI2020

Improvements to CTF Code Verification and Unit Testing (FY2020)

In 2010, the U.S. Department of Energy created its first Energy Innovation Hub, which focuses on improving Light Water Reactors (LWRs) through Modeling and Simulation. This hub, named the Consortium for the Advanced Simulation of LWRs (CASL), attempts to characterize and understand LWR behavior under normal operating conditions and use any gained insights to improve their efficiency. In collaboration with North Carolina State University (NCSU), CASL has worked extensively on the thermal-hydraulic subchannel code Coolant Boiling in Rod Arrays—Three Field (COBRA-TF). The NCSU/CASL version of COBRA-TF has been rebranded as CTF. This document focuses on code verification test problems that ensure CTF converges to the correct answer for the intended application. The suite of code verification tests are mapped to the underlying conservation equations of CTF, and significant gaps are addressed. Convergence behavior and numerical errors are quantified for each of the tests. Tests that converge at the correct rate to the corresponding analytic solution are incorporated into the CTF automated regression suite. A new verification utility is created for this purpose, which enables code verification by generalizing the process. For problems that do not behave correctly, the results are reported but the problem is not included in the regression suite. In addition to verification studies, this document also quantifies the existing tests of constitutive models. A few existing gaps are addressed by adding new unit tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

PMK 1000 cone-socket ventilator fittings force testing to ISO 5356-1:2015 Standard

This memo summarizes the results of verification tests on a cone-socket fitting to the ISO 5356-1:2015 standard, annexes B and F. The components were found to meet requirements holding 50N tensile force and more than 35 N-cm of torque for 10 seconds without separating at 35°C with a fresh socket. Humidity control was unavailable; consequently, humidity was significantly lower than the values in the ISO standard. Several repeat tests of the fitting with force, torque, and time excursions give us confidence that the fittings perform suitably to the ISO standard. Force overload tests suggest that the samples will disengage under tension at around 69 N, while torque overload tests under a force of 50 N will fail at 79 N-cm of torque. If repeated tests are performed on the same component, the socket can deform and lead to the connection failing the standard requirements. However, such failures only occur with repeated testing, and the gripping of such units can be modified so that they pass the standard. This observation demonstrates that socket gripping is responsible for those failures and would not be a problem in application.

42 ENGINEERING↗
DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) Cartridge Performance Testing on Hanford Tanks SX-101 and SX-104 (Vol.1)

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 performance1 to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the headspaces of Hanford tanks SX-101 and SX-104. The Occupational Safety and Health Administration (OSHA) considers cartridge testing a valid approach for establishing cartridge change schedules.2 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 ensure 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 June 16-17, 2017, and June 23-24, 2017, using headspace vapors from Hanford tank SX-104 and Hanford tank SX-101, respectively. Headspace vapors under static conditions were 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 cartridges, and subsequently were 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↗